ML041760477

Univ of VA Reactor Decommissioning Project, License R-66 Termination Request, Volume 1, Narrative

ML041760477
Person / Time
Site:	University of Virginia
Issue date:	06/30/2004
From:	Benneche P University of Virginia
To:	Hughes D Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041760477 (294)

Text

The University of Virginia Reactor Decommissioning Project

__ CH2MH ILL License R-66 Termination Request Volume 1 Narrative

-Request Letter

-UVA Decommissioning Plan Performance Summary

-Final Status Survey Report

-Master Final Status Survey Plan and Addenda 001 through 008 June 2004

THE ORIGINAL VERSION OF THIS DOCUMENT IS BEING SENT TO THE NRC DOCUMENT CONTROL DESK COPIES HAVE BEEN SENT TO DAN HUGHES & STEVEN HOLMES OF THE U.S.N.R.C.

UNIVERSITY OF VIRGINIA NUCLEAR REACTOR FACILITY U.S. MAIL ADDRESS STREET ADDRESS P.O. Box 400322 675 Old Reservoir Road Charlottesville, VA Charlottesville, VA 22903 229044322 Telephone: 434-982-5440 Faxr 434-982-5473 June 18, 2004 Docket 50-62 Document Control Desk United States Nuclear Regulatory Commission Mail Stop 03-H3 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2783 Attention: Mr. Daniel E. Hughes, Project Manager Operating Reactor Improvements Program, Mail Stop 012-G13

Subject:

University of Virginia License Termination Request and Transmittal of the University of Virginia Decommissioning Plan Performance Summary, April 2004 and the Final Status Survey Report-- Evaluation of Radiological Results Relative to Termination of NRC License R-66, University of Virginia, Charlottesville, Virginia, May 2004

References:

1. Amendment No. 26 to Amended Facility Operating License No. R-66 for the University of Virginia Research Reactor, Docket 50-62

2. Transmittal R. U. Mulder to D. E. Hughes, "Transmittal of the University of Virginia Reactor Decommissioning Project Master Final Status Plan, UVA-FS-002, Rev 0, March 2003" dated April 4, 2003

3. Transmittal R. U. Mulder to D. E. Hughes, " Transmittal of University of Virginia Reactor Decommissioning Project Groundwater Report and Final Survey Status Addenda: Underground Waste Tank Excavation; Reactor Facility Piping; Pond Sediments; Interior Structure Surfaces; Exterior Soil and Paved Surfaces; Exterior Structure Surfaces; and Special Soils Areas" dated June 18, 2003

4. Transmittal P. E. Benneche to D. E. Hughes, "Transmittal of University of Virginia Reactor Decommnissioning Project Continuing Characterization Summaries: Reactor Pool Soil Areas; Reactor Pool Interior; Plant Stack; Laboratory Rooms M008 and M005; Demineralizer Room; Hot Cell Room; Heat Exchanger Room; Source Room; Beamport Area; Impacted Drain Lines and Piping; Mezzanine Crawl Space; and Pond Sediment" dated June 29, 2003

June 18, 2004 Page 2

Dear Mr. Hughes,

The University of Virginia requests that License R-66 for the University of Virginia Research Reactor be terminated. We have performed the activities described in Amendment 26 (Reference 1), which approves the Decommissioning Plan for the University of Virginia Research Reactor.

The performance details are contained in the "University of Virginia Decommnissioning Plan PerformanceSumnary, April 2004." The many documents that implemented the Decommissioning Plan under the direction of the Project Management Plan are maintained at the University for your inspection.

Reference 2 established the protocols for the Final Status Survey, while Reference 3 customizes those protocols to evaluate the characterization results documented in Reference

4. The Final Status Surveys were performed to implement those customized Final Status Survey protocols. Results of the Final Status Survey measurements are summarized in "Final Status Survey Report- Evaluation of RadiologicalResults Relative to Terminationof NRC License R-66, University of Virginia, Charlottesville, Virginia, May 2004. The raw data is maintained in hard copy at the University for your inspection.

Please note that Reference 3 transmitted the results of the groundwater investigation required by Amendment 26.

We are pleased to transmit for your information the Decommissioning Performance Summary and the Final Status Survey Report, prepared for the University of Virginia by CH2M HILL and its subcontractor, Safety and Ecology Corporation. These reports have been reviewed and approved by the Reactor Decommissioning Committee. The reports are:

the "University of Virginia Decommissioning Plan Performance Summary, April 2004" and the "Final Status Survey Report- Evaluation of Radiological Results Relative to Termination of NRC License R-66, University of Virginia, Charlottesville, Virginia, May 2004."

We believe this transmittal completes the transmittals required by the University of Virginia Decommissioning Plan in order for the Commission to terminate License R-66. If you have any questions please contact me at 434-982-5440.

I declare under penalty of perjury that tie forgoing is trite and correct.

Sincerely, Date: J ,2 o/o If Paul Benneche Reactor Director University of Virginia

June 18, 2004 Page 3 N1

Enclosures:

1. University of Virginia Decormmissioning Plan Performance Summary, April 2004

2. Final Status Survey Report-Evaluation of Radiological Results Relative to Termination of NRC License R-66, University of Virginia, Charlottesville, Virginia, May 2004 cc: Ralph Allen, Chair, Univ. of Va. Reactor Decommissioning Cormmittee Daniel Hughes, USNRC Stephen Holmes, USNRC

University of Virginia Decommissioning Plan Performance Summary Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 9189 South Jamaica Street Englewood, CO 80112 April 2004

University of Virginia I ..

Decommissioning Plan Performance Summary Prepared for University of Virginia Reactor Facility Decommissioning Project Approved By:

-RFa)JQ &/ V C Date: APLQ 5°0*DDt LWAR Technical Director Approved By: Date: -6, °i WVA Reactor Decommissioning Committee Prepared by

. is CH21UHILL 9189 South Jamaica Street Englewood, CO 80112

. April 2004 DECOESS1ONNGPM PMERFORMN4CE 2

University of Virginia Decommissioning Plan Performance Summary Prepared for University of Virginia Reactor Facility Decommissioning Project Approved By: Date: /e I9' Project Engineer Approved By: Date: f/.0/4; Project Manager Prepared by CH2MHILL 9189 South Jamaica Street Englewood, CO 80112 April 2004 DECOMMISSIONING PLAN PERFORMANCE 3

Summary The University of Virginia has completed the decommissioning of the University of Virginia Research Reactor (License R-66). The University of Virginia has determined that the remaining residual radioactivity results in a total effective dose equivalent that does not exceed the 25 mrem per year site release limit. This determination was made by comparing the Final Status Survey results to the Derived Concentration Guideline Levels (DCGLs) that were presented in the Nuclear Regulatory Commission (NRC)-approved "University of Virginia Reactor Decommissioning Plan" (Decommissioning Plan) approved by Amendment No. 26 to Amended Facility Operating License No. R-66 for the University of Virginia Research Reactor, Docket 50-62. Each DCGL is a radionuclide-specific value that correlates acceptable annual dose for a particular site and dose pathway with a concentration (pCi/g) or surface activity (dpm/100 cm2 ) value. The DCGL value is an average level for the survey unit or sample.

The DCGLs for the University of Virginia Reactor (UVAR) project were chosen from the NRC's DandD computer code list of "default screening values" that were developed by using dose-based models with conservative input parameters. Because of the conservatism inherent in the model inputs, the NRC has determined that use of these "default screening values" eliminates the requirement to perform separate As Low As Reasonably Achievable (ALARA) evaluations of the DCGL values. For certain situations, excursions above a DCGL level are allowable for small surface areas, provided the average level for the survey unit is not exceeded. However, the "default screening values" represent upper limits, and there are no allowances for small areas of activity exceeding those values.

For practical field application, a gross DCGL may be developed for surface activity when there are multiple radionuclides. This DCGL5 TOSS takes the radionuclide mixture into account, essentially providing a means for demonstrating that the combination of contaminants on a surface satisfies the Unity Rule (sum of fractions) for the individual dose-based DCGLs of the radionuclides in the mixture. In the case of soil, a surrogate DCGL is often developed. The DCGLswrogatw enables measurement of a specific radionuclide and the inference from that measurement of the levels of other contaminants that may not be as easily measured-again with the objective of satisfying the Unity Rule.

The approved Decommissioning Plan was implemented essentially as written. Proposed changes and clarifications were evaluated and approved by the Reactor Decommissioning Committee. One proposed change "to revise the DPP scanning levels for impacted areas to values consistent with the MARSSIM" (December 5,2003, Benneche to Hughes, Requestfor Approval of Final Status Survey Plan Coverage Consistent with MARSSIM Requirementsfor thte University of Virginia Reactor [License R-66J) required approval from the NRC. That change was approved (March 31, 2004, Hughes to Benneche, Approval of Final Status Survey Plan Coverage Chlange for License No. R University of Virginia [TAC NO. MA37371).

There are two visible changes to the facility that differ from the conditions envisioned in the Decommissioning Plan: (1) after continuing characterization surveys of the Reactor Room ventilation system indicated that surface contamination values were less than DCGL ,

therefore, the system did not need to be removed as radioactive waste, the Reactor Room DECOMMISSIONING PLAN PERFORMANCE 4

ventilation system was left in place; and (2) the Reactor Pool Co-60 Irradiation Facility source remains stored at the facility.

In summary, the UVAR facility has met the criteria for the unrestricted termination of Reactor License R-66.

Introduction Decommissioning was controlled through the use of a series of project plans:

• Project Management Plan

• Environmental Safety and Health Plan

• Quality Assurance Project Plan

• Radiation Protection Plan

• Environmental Compliance Plan

• Waste Management Plan.

These plans invoked the controls necessary to implement the Decommissioning Plan. For instance, training of personnel commensurate with their tasks, use of Radiation Work Permits, ALARA, radiological protection, general site training, respiratory protection training and use, approved supplier list, quality assurance audits, stop work authority, procedural controls, document production and approval, instrument calibration and controls, sampling and chain-of-custody protocols, and record production and retention are all controlled through the Project Management Plan control system. These plans allowed the physical work to be performed safely using a set of controls that established consistent management and worker expectations.

Decommissioning Plan field work started with mobilization of CH2M HILL Constructors, Inc. (CCI) to the facility in April 2002. The subcontractors, Safety and Ecology Corporation (SEC) and Bartlett Services, Inc. (Bartlett), quickly followed. WMG and Underwater Construction Company mobilized in August 2002 for pool component removal. The project was also supported by Penhall Corporation (concrete cutting), Parham Construction Co.

(heavy crane operation and earth-moving), and NLB Corporation (water jet cutting). The physical work was completed on May 29,2003 and Bartlett demobilized on May 30,2003.

SEC remained on site to perform the Final Status Survey and then demobilized on August 15, 2003.

Field Activities The characterization, which had not been completed at the time of the Decommissioning Plan approval, was performed and delivered to the NRC in July (July 29,2003, Benneche to Hughes, Transmittalof University of VirginiaReactor DecommissioningProject Continuing CharacterizationSummaries: ReactorPool Soil Areas; Reactor Pool Interior;PlantStack; Lnboratory Rooms M008 and M005; DemineralizerRoom; Hot Cell Room; Heat Exchanger Room; Source Room; Beamport Area; Impacted Drain Lines and Piping; Mezzanine Crawvl Space; and Pond Sediment).

A summary of the results of the characterization and in-process surveys of the UVAR facility follow. For the purposes of the Decommissioning Plan, the UVAR facility consisted DECOMMISSONING PLAN PERFORMANCE 5

of three levels of offices and laboratories (reactor building) attached to the 61-foot-high (ceiling to the bottom of the reactor pool) reactor room that contained the 27-foot-deep reactor pool embedded in the hillside. This building was surrounded by land and pond areas protected by a fence and gate.

In the discussions that follow, distances and areas listed in feet or square feet followed by (meters) or (square meters) are "footprint" or floor area values unless otherwise stated.

Values listed only in meters or square meters are values associated with the radiological survey or sampling area descriptions. Because all sampling plans were established to be consistent with the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) methodology of performing Final Status Surveys, no conversion to English measurement units were made.

• The 102,000 square feet (9,390 m2 ) of facility external to the building were screened and

-2,520 square feet (232 m2 ) were found to be potentially contaminated above the radiological release limits

• The 22,000 square feet (2,030 m2 ) of building interior floor space contained

-11,346 square feet (1,050 m2 ) of as-found free release area and -10,654 square feet (980 m2 ) of as-found area potentially contaminated above the radiological release limits

• No structural demolition of the building was required or performed since the facility was expected to be refurbished as engineering offices and laboratories after decommissioning.

All areas or items in the area that were contaminated above the release limits were either decontaminated to below the release limits or physically removed and processed as radioactive waste. Waste minimization on a cost-to-ultimate-disposal basis was implemented. A few pieces of slightly radioactive equipment (mineral irradiation facility

[MIF], transfer casks, etc.), desired by other facilities, were physically transferred to other licensed research reactors for their use Waste and Material Removals Ten radioactive waste shipments were made. One shipment was made to Barnwell, SC in August 2002. The other nine waste shipments were made to Envirocare with the last shipment being made on May 29,2003. These 10 shipments comprised 270,127 pounds of waste shipped for disposal (260,832 pounds to Envirocare and 9,295 pounds to Barnwell).

Six 55-gallon drums containing 3,511 pounds of radioactively-contaminated soils and asphalt, manifested for disposal at Envirocare, have been staged at UVA for shipment as part of a future UVA shipment. One B-25 container of the unused portions of soil and pond sediment samples remains at the site and will be disposed after the independent verification is completed.

Over 1,000,000 pounds of soil were surveyed, confirmed to not be radioactively contaminated, and were reused on site. Over 600,000 pounds of non-soil materials were surveyed and contaminated items were included in the Envirocare shipments. Non-soil materials surveyed and confirmed to not be radioactively contaminated included 331,271 pounds of material released for reuse, recycle, surplus sales, or disposal as sanitary waste (construction rubble).

DECOMMISSIONING PLAN PERFORMANCE 6

Safety and ALARA Over 36,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> have been worked with no lost-time injuries or OSHA-recordable injuries or illnesses since the project began. Additionally, there were zero environmental releases from field activities with zero environmental Notices of Violation.

With all dosimetry for the decommissioning having been processed, the official total project dose is 702 mrem. This number represents a significant reduction from the Decommissioning Plan's schedule-adjusted prediction of 3,000 mrem.

Reactor Confinement Structure The polar crane was re-certified for use in decommissioning and remained operational at the completion of decommissioning activities. Loose items, the reactor control room, and the instrumentation room were size-reduced as necessary and removed to the bare walls of the confinement structure (Room 131). After the reactor pool had been emptied, the concrete floor was cleaned with a water jet cutting process. The floor drains were then inspected and decontaminated or removed as necessary. When all decommissioning activities that might benefit from ventilation system operation were completed, the reactor ventilation system and the building off-gas stack were surveyed. Because the indicated surface contamination values were less than DCGLgV, the ventilation systems were left in place, following review by the Reactor Decommissioning Committee. The ventilation openings required no special closures or monitoring.

Reactor and Pool The water in the pool was used to provide shielding during the segmenting and removal of the highly activated components in the pool. Using divers to perform the work achieved a dose reduction of 80 percent (from the originally planned in-air operation). The component segmentation process began with the placement of a cask liner in the reactor pool.

Segmentation was then performed underwater by divers using plasma arc cutting equipment. The liner was loaded underwater, the higher activity items were preferentially loaded nearest the center of the cask and the lower activity materiel (hardware, beam port nosepieces, etc.) loaded in the liner annulus to provide additional shielding. The activated components were shipped to the Barnwell disposal site in a CNS 8-120B shipping cask.

Because air sampling performed while segmentation was occurring demonstrated that airborne contaminants were not produced, a confinement structure was not required.

After the shipment to Barnwell, the remaining pool water was sampled and confirmed suitable for discharge to the sanitary sewer. The water was discharged from the pool through filters to a temporary surge tank, where a second pump pumped the filtered water directly to the sanitary sewer. Using this system, the pool discharge was about 63,000 gallons; about 35 gallons remained in the piping and pool to be processed by the routine liquid release pathway. When the structure surface survey of the empty pool determined that no "hot particles" were present, decontamination and cleaning of the pool surfaces began. Air sampling performed while cleaning with a water jet cutting process demonstrated that airborne contaminants were not produced. The confinement structure was used to control cleaning water over-spray. The cleaning water produced by the water jet cleaning was collected, large solids were settled out, and the remaining water was DECOMMISSIOMNG PLAN PERFORMANCE 7

evaporated off the concrete and epoxy fines. Both solids were dried and disposed of as low-level radioactive waste.

Once the pool surfaces had been cleaned to bare concrete, a structural evaluation was performed. No additional structural protection measures were required for the remainder of the decommissioning activities. Potential leakage paths were investigated. Concrete surface and interior core samples were evaluated for contamination activation. The only activated concrete was detected radially around the beam tubes through the pool wall. Soil sampling under the pool floor and horizontally through the pool walls was performed.

Because the structural analysis indicated that it was not necessary to maintain the pool structural integrity, and to allow access by the verification team, these sampling locations were not repaired. Results of sample analyses are described in the characterization summaries referenced earlier in this document.

The characterization results led to the removal of the entire west beam port tube liner in a 30-inch diameter cylinder of concrete. The east beam port tube liner was removed in similar fashion to a depth of about 24 inches from the interior face of the pool. All metal in the pool that had been in direct contact with pool water was removed except for the embedded pool gate guides and structural anchor plates, After all of these items were removed, which lowered background levels in the pool, the pool characterization survey was performed.

Pool gate guide coupons were removed to investigate the portions that were in contact with the pool wall concrete. Several small concrete surface contamination areas were decontaminated. The embedded flange on the heat exchanger suction, located immediately under the reactor core, was removed because it contained activation products. The remaining heat exchange and drain lines were cleaned and left in place. The "knee wall" at the top of the pool was cut off flush with the floor. After the Final Status Surveys of the pool interior were complete, industrial walkway grating was installed across the pool to eliminate fall hazards, yet maintain access for the verification team.

Remaining Rooms and Structures With the reactor room (Room 131) completed, approximately 8,000 square feet (740 m2 ) of building floor space remained to be decommissioned. This space included the primary heat exchanger room, the demineralizer system rooms, the liquid waste storage tanks area, the hot cell rooms, the source and instrument storage areas, the pneumatic rabbit room, and two laboratory rooms. These rooms were cleared to the bare walls of their reactor-associated components and the remaining contaminated items decontaminated or disposed as low-level radioactive waste.

For instance, the installed laboratory counters, sinks, and hoods that had contamination levels less than the DCGLu,,rrogate were left in place, while the potentially internally-contaminated rabbit transfer system was removed completely and processed as low-level radioactive waste. Contaminated surfaces were decontaminated or removed (exhaust blowers, filters and some ductwork). The contaminated laboratory hood exhaust ducting that penetrated the wall to the outside, had contamination levels less than the DCGLsurrogate and was left in place.

The cooling tower on the roof of the mezzanine level was characterized before removal by a crane to the parking area. Characterization results allowed remediation of the asbestos as DECOMMISSIONING PLAN PERFORMANCE 8

clean asbestos and remaining tower materials as clean construction debris. The Reactor Pool Co-60 Irradiation Facility source has decayed to about 900 curies and remains in the facility at this time. The hot cell lead-glass oil-filled window and manipulators were surveyed, found free from radiological contamination, and were removed for reuse by another company.

Underground Tanks and Vaults The outdoor spent fuel transfer tank was internally contaminated from previous transfer operations. It was enclosed in a ventilation containment to capture airborne contamination while being size-reduced by oxygen-acetylene torch cutting.. Although air sampling confirmed that respiratory protection was not required due.to airborne radiological levels, the confinement served to prevent contamination dispersion. The sand base for the tank was removed and processed as low-level waste. Subsequent concrete basemat screening indicated the basemat had contamination levels less than the DCGLO.

Two large underground liquid waste tanks and two smaller hot cell drain tanks were excavated, removed, and size-reduced for disposal as low-level waste. The liquid waste storage tanks were size-reduced outdoors in a similar confinement structure to that used for the spent fuel transfer tank. Some of their associated buried piping was removed as part of the removal operation of the tanks and enclosures. The remaining underground pipe sections were surveyed, found to be free from contamination, and were left in place. The block wall and gravel floor of the liquid waste tank blockhouse were found to be contaminated, and were removed and processed as low-level radioactive waste. The poured-concrete hot-cell tank vault was removed, surveyed, and found to be free from radiological contamination, allowing disposal as construction debris. The blockhouse and the vault structures were removed completely to bare soil.

The soils removed to uncover the tanks were surveyed and determined to be clean (i.e., no radioactive contamination present). That soil was staged for future replacement. Soil screening of the excavation confirmed the contaminated areas had been removed. The industrial hazards presented by the excavation opening and the stability of the adjacent roadway required mitigation. Soil sampling of the completed excavation area, in accordance with the Final Status Survey Plan, was performed. Then placement of the staged soil back into the excavation was performed until hazard mitigation was achieved.

Soil sampling, sample splitting, and placement of the samples under Chain-of-Custody controls was witnessed by the NRC site inspector. Results of these analyses are available to the verification team, as are the split samples that are stored at the reactor facility. The excavation was restored to a stable configuration and re-vegetated to maintain that stable configuration.

Outdoor Areas, Drains, and Sewers Storm drains, building drains, and the sanitary sewer line were surveyed and confirmed to be clean or to have radioactive contamination levels less than the DCGL S. The exterior piping access via well casings and manways remain in place for use by the verification team.

Initial characterization efforts had identified previously contaminated surface soil adjacent to the liquid waste storage tank blockhouse. These soils and the pond sediments were re-DECOMMISSIONING PLAN PERFORMANCE 9

characterized and found to not require remediation. The only "soil" remediation required was performed when the underground liquid waste storage tanks blockhouse floors were removed. These dirt and gravel floors were removed to a depth of nominally 1 foot and were packaged for shipment to Envirocare.

The other outdoor area remediated was the asphalt pad just outside the reactor room roll-up door. Four 55-gallon drums of Cs-137-contaminated asphalt were removed before the contamination values were less than DCGLross. This contamination is believed to have come from a contaminated storage cask that been stored at this location.

During the investigation of the reactor pool drain lines, a portion of the clay-tile footing drains was located. Perched upon a foot of coarse gravel fill, the footing drains functioned to maintain low external water pressures on the pool footing. All areas accessed indicated clean piping or less than DCGLZ,.. This pool footing drain combined with the reactor building roof drains and discharged on the pond hillside to the south of the reactor building. These flows did not enter the storm drainage system. All portions of the combined drains, exterior to the building, were determined to be not contaminated.

Final Status Survey The Final Status Survey Plan was transmitted to the NRC (Transmittal R. U. Mulder to D. E.

Hughes, "Transmittalof the University of VirginiaReactor Decommissioning ProjectMaster Final Status Survey Plan, UVA-FS-002, Rev 0, March 2003") on April 4,2003 and the Final Status Survey Addenda were transmitted (Transmittal R. U. Mulder to D. E. Hughes, "Transmittal of University of VirginiaReactor Decommissioning Project GroundwaterReport and Final Survey Status Addenda: UndergroundWaste Tank Excavation; ReactorFacilityPiping;Pond Sediments; InteriorStructure Surfaces; Exterior Soil and Paved Surfaces; Exterior Structure Surfaces; and Special Soils Areas") on June 18, 2003.

On December 12,2003, a Request for Additional Information (RAI) was received. Responses and clarifications were provided to the NRC (Transmittal P. E. Benneche to D. E. Hughes, "University of Virginia Master Final Status Survey Plan and Addenda 001-008 (TAC NO.

MB82331") on January 22,2004. One proposed change (December 5, 2003, Benneche to Hughes, Requpestfor Approval of Final Status Survey Plan Coverage Consistent wvith MARSSIM Requirementsfor tihe University of Virginia Reactor[License R-661) required approval from the NRC. The proposed change "to revise the DPP scanning levels for impacted areas to values consistent with MARSSIM requirements," was approved (March 31, 2004, Hughes to Benneche, Approval of Final Status Survey Plan Coverage ChangeforLicense No. R University of Virginia (TAC NO. MA3737]). While the changes and clarifications did not affect the validity of the field measurements, the Final Survey Plan and Addenda required revision. These documents were revised to reflect the RAI response reviews. Revision 1 of the Final Status Survey Plan and the seven Addenda are included in this submittal.

At the completion of the physical decommissioning for each final survey plan area, surveys were performed. Eight areas were identified as having elevated activity, decontaminated as necessary, re-characterized and the Final Status Survey for that area performed. The Final Status Survey results are contained in the document, "Final Status Survey Report-Evaluation of RadiologicalResults Relative to Ternmination ofNRC License R-66, University of Virginia, Charlottesville, Virginia, October2003" included in this submittal. This report confirms that DECOMMISSIOMNG PLAN PERFORUMACE 1 10

remaining residual radioactivity, distinguishable from background radiation, results in a total effective dose equivalent (to an average member of a critical group) that does not exceed the 25 mrem per year site release limit. The UVAR facility has met the criteria for unrestricted release and termination of the Reactor License (R-66).

DECOMMISSIONING PLAN PERFORMANCE 11

Final Status Survey Report EVALUATION OF RADIOLOGICAL RESULTS RELATIVE TO TERMINATION OF NRC LICENSE R-66 UNIVERSITY OF VIRGINIA CHARLOTTESVILLE, VIRGINIA Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 May 2004

Final Status Survey Report EVALUATION OF RADIOLOGICAL RESULTS RELATIVE TO TERMINATION OF NRC LICENSE R-66 UNIVERSITY OF VIRGINIA CHARLOTTESVILLE, VIRGINIA Prepared for University of Virginia Reactor Facility Decommissioning Project May 2004 Approved by: .' /na Reacto Dcommissioning Committee Chair Date Approved by., eQ S-26-04F UVAR Technical Director Date

Final Status Survey Report EVALUATION OF RADIOLOGICAL RESULTS RELATIVE TO TERMINATION OF NRC LICENSE R-66 UNIVERSITY OF VIRGINIA CHARLOTTESVILLE, VIRGINIA Prepared for University of Virginia Reactor Facility Decommissioning Project May 2004 Approved by: . -,2D-te

,Aadiol!ical Controls Supervisor Date Approved by: I aractnztin and Survey Supervisor Date Approved by: 7 4 < 5/fo Project Manager Date

Contents Acronyms ............................................................ viii Executive Summary ........................................................... ix 1.. Introduction .................................................................................................................................. 1-1

2. Facility Description ........................................................... 2-1

3. General Final Status Survey Approach ........................................................... 3-1 3.1 Radiological Contaminants and Criteria ........................................................... 3-4 3.2 Data Quality Objectives ........................................................... 3-6 3.3 Final Status Survey Tasks ........................................................... 3-7 3.4 Survey Instrumentation and Methods ........................................................... 3-7 3.4.1 Surface Beta Radioactivity Scan Surveys ................................................... 3-8 3.4.2 Gamma Surface Scans ................... ........................................ 3-8 3.4.3 Integrated Direct Surface Beta Radioactivity Measurements ................. 3-9 3.4.4 Smear Sample Collection and Analysis ..................................................... 3-9 3.4.5 Volumetric Sample Collection and Analysis ......................... ................. 3-10 3.5 Quality Assurance ........................................................... 3-10 3.5.1 Data Quality Controls ...................... ..................................... 3-10 3.6 Data Assessment and Evaluation ........................................................... 3-11 3.7 Background Determination and Reference Areas .................................................. 3-11

4. Final Status Survey ........................................................... 4-1 4.1 Background ........................................................... 4-1 4.1.1 General ........................................................... 4-1 4.1.2 Survey Activities and Results ............................................. .............. 4-1 4.2 Underground Waste Tank Excavation ............................................ ............... 43 4.2.1 Description ........................................................... 4-3 4.2.2 Survey Activities ........................................................... 4-4 4.2.3 Survey Results ............................................................ 4-6 4.2.4 Conclusion ........................................................... 4-9 4.3 Facility Piping ........................................................... 4-9 4.3.1 Description ........................................................... 4-9 4.3.2 Survey Activities ........................................................... 4-11 4.3.3 Survey Results ........................................................... 4-15 4.3.4 Conclusion ........................................................... 4-17 4.4 Pond Sediment ........................................................... 4-17 4.4.1 Description ........................................................... 4-17 4.4.2 Survey Activities ........................................................... 4-18 4.4.3 Survey Results ........................................................... 4-19 4.4.4 Conclusion ........................................................... 4-22 iv

4.5 Interior Structure Surfaces ........................ 4-22 4.5.1 Description ........................ 4-22 4.5.2 Survey Activities ........................ 4-26 4.5.3 Survey Results ........................ 4-29 4.5.4 Conclusion ........................ 4-35 4.6 Exterior Soils and Paved Areas ........................ 4-35 4.6.1 Description ........................ 4-35 4.6.2 Survey Activities ........................ 4-36 4.6.3 Survey Results ........................ 4-38 4.6.4 Conclusion ............................  :-

4.7 Exterior Structure Surfaces ........................ 4-42 4.7.1 Description ........................ 4-42 4.7.2 Survey Activities ........................ 4-42 4;7.3 Survey Results ........................ 4-48 4.7.4 Conclusion ........................ 4-48 4.8 Special Soils Areas ........................ .49 4.8.1 Description ........................ 4-49 4.8.2 Survey Activities ........................ 4-53 4.8.3 Survey Results ........................ 4-57 4.8.4 Conclusion ........................ 4-67 4.9 Facility Ventilation....................................................................................................4-68 4.9.1 Description ........................ 4-68 4.9.2 Survey Activities .................................................... 4-69 4.9.3 Survey Results .................................................... 4-73 4.9.4 Conclusion .................................................... 4-74

5. Quality Assurance .................................................... 5-1

6. Summary ..................................................... 6-1

7. Works Cited .................................................... 7-1 TABLES Table 3-1 UVAR Survey Areas and Classifications .................................................... 3-2 Table 3-2 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces ............................................. 3-5 Table 3-3 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil .................................................... 3-6 Table 3-4 Instrumentation for UVAR Final Status Survey .................................................... 3-8 Table 4-1 Concentration of Radionuclides in Background Soil ................................................. 4-3 Table 4-2 Results of Soil Sampling from Waste Tank Excavation ............................................. 4-8 Table 4-3 Analyses of Composite Samples from Waste Tank Excavation ............................... 4-9 Table 4-4 Facility Piping Survey Units ..................................................... 4-11 Table 4-5 Beta Scan Results for Facility Piping .................................................... 4-16 Table 4-6 Facility Piping Beta Activity Measurement Summary ............................................. 4-16 Table 4-7 Masslinn Swabs of Piping ..................................................... 4-17 Table 4-8 Results of Gamma Analyses of Systematic Samples ................................................ 4-22 Table 4-9 Survey Units for UVAR Building Interior Surfaces ................................................. 4-27

Table 4-10 Survey Results Forms for UVAR Building Interior Surfaces ................................ 4-29 Table 4-11 Results of Gamma Scans of UVAR Building Interior Surfaces ............................ 4-31 Table 4-12 Results of Floor Monitor Beta Scans of UVAR Building Interior Surfaces ......... 4-32 Table 4-13 Results of 43-68 Detector Beta Scans of UVAR Building Interior Surfaces ........ 4-32 Table 4-14 Summary of Beta Activity Measurements for UVAR Building Interior Surfaces ................................................................... 4-33 Table 4-15 Summary of Gamma Spectrometry Analysis for Soil Samples from Site Open Land Areas ................................................................... 4-40 Table 4-16 Concentrations of Radionuclides in Composite of Systematic Soil Samples ..... 4-41 Table 4-17 Summary of Beta Surface Activity Measurements on Exterior Paved Surfaces ....  ; 4-41 Table 4-18 Exterior Structure Surfaces Beta Activity Summary . 4-48 Table 4-19 Range of Gamma Scan Levels on Surfaces of Interior Soil . 4-63 Table 4-20 Results of Gamma Spectrometry of Mezzanine Crawl Space Soil Samples . 4-63 Table 4-21 Results of Gamma Spectrometry of Demineralizer Wall Excavation Soil Samples . 4-64 Table 4-22 Results of Gamma Spectrometry of Soil Samples from "M" and "B" Areas . 4-64 Table 4-23 Results of Gamma Spectrometry of Soil Samples from Beneath Reactor Pool . 4-65 Table 4-24 Results of Gamma Spectrometry of Surface Soil Samples from Beneath Reactor Room Floor . 4-66 Table 4-25 Results of Gamma Spectrometry of Soil Samples Representing 1-m Fill Intervals Beneath Reactor Room Floor . 4-66.

Table 4-26 Analyses of Composite Samples from Interior Soil Areas . 4-67 Table 4-27 Beta Scans of Facility Ventilation . 4-74 Table 4-28 Ventilation Beta Activity Measurement Summary . 4-74 Table 5-1 Normalized Absolute Difference for Duplicate Measurements . 5-2 FIGURES Figure 2-1 Map of Charlottesville Area Surrounding the UVAR Site . 2-1 Figure 2-2 Western Grounds of the University of Virginia . 2-2 Figure 2-3 UVAR Facility Site . 2-3 Figure 2-4 Building Plan Views . 2-4 Figure 4-1 Background Soil Locations . 4-2 Figure 4-2 WVAR Facility and Environs Indicating Location of Waste Tanks . 4-5 Figure 4-3 Waste Tank Survey Unit, Indicating Grid for Survey and Sample Locations . 4-7 Figure 4-4 Drains from CAVALIER Facility . 4-12 Figure 4-5 Ground Floor and Exterior Piping . 4-13 Figure 4-6 Reactor Room Piping . 4-14 Figure 4-7 University of Virginia Reactor Facility and Environs . 4-20 Figure 4-8 Plot Plan of Pond, Indicating Reference Grid and Sampling Location . 4-21 Figure 4-9 UVA Reactor First Floor Plan View . 4-23 Figure 4-10 UVA Reactor Mezzanine Floor Plan . 4-24 Figure 4-11 UVA Reactor Ground Floor Plan View . 4-24 Figure 4-12 University of Virginia Reactor Facility and Environs . 4-37 vi

Figure 4-13 Plot of Site, Indicating Reference Grid System, and Measurement Sampling Locations ............................................................. 4-39 Figure 4-14 WVA Reactor Floor Plan View Indicating Roof Areas ......................................... 4-44 Figure 4-15 UVA Reactor First Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ............................................................. 4-45 Figure 4-16 UVA Mezzanine Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ..................... 4-46 Figure 4-17 UVA Reactor Ground Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ..................... 4-47 Figure 4-18 UVA Mezzanine Floor Plan View Indicating the Location of the Mezzanine Crawl Space ...................... 4-50 Figure 4-19 UVA Reactor First Floor Plan View Indicating the Location of Soil Fill Around Reactor Pool................... 4.-51 Figure 4-20 UVA Reactor Ground Floor Plan View Indicating the Location of the Soils Beneath the Reactor Pool ..................... 4-52 Figure 4-21 Excavation of Pool Fill at Demineralizer Room Wall .......................................... 4-54 Figure 4-22 Excavation of Sub-floor Fill at "M" and "B" Core Locations Of Pipe Leaks ............................................................. 4-55 Figure 4-23 Mezzanine Crawl Space Floor, Indicating Soil Sampling Locations ................. 4-58 Figure 4-24 Soil Sampling Locations Beneath the North End of the Reactor Pool ............... 4-59 Figure 4-25 Soil Sampling Locations Beneath the South End of the Reactor Pool ............... 4-60 Figure 4-26 Locations of Surface Sampling of Fill Soil Beneath Reactor Room Floor .......... 4-61 Figure 4-27 Locations of Sub-surface Samples from Fill Beneath Reactor Room ................. 4-62 Figure 4-28 UVA Reactor First Floor Indicating Potentially Impacted Ventilation Systems ...  ; 4-70 Figure 4-29 UVA Mezzanine Floor Indicating Potentially Impacted Ventilation Systems .... 4-71 Figure 4-30 UVA Reactor Ground Floor Indicating Potentially Impacted Ventilation Systems...................................................................................................................... 4-72 APPENDICES Appendix A Master Final Status Survey Plan and Addenda ..................................... A-1 Appendix B Final Survey Data Sheets .............. ............................... B-1 Appendix C Quality Control Charts and Log Sheets ............................................. C-1 Ai

Acronyms ALARA As Low As Reasonably Achievable CAVALIER Cooperatively Assembled Virginia Low Intensity Educational Reactor cm centimeter cm2 square centimeters DCGL Derived Concentration Guideline Level dpm disintegrations per minute DQO Data Quality Oblective FSS Final Status Survey g gram km kilometer LBGR Lower Bound of the Gray Region mn meter m 2 square meters MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual mrem millirem NRC Nuclear Regulatory Commission Norn number of data points pCi picocurie QA Quality Assurance QC Quality Control SEC Safety and Ecology Corporation UVAR University of Virginia Reactor WRS Wilcoxon Rank Sum nii

Executive Summary Field activities to decommission the University of Virginia pool-type research reactor (Nuclear Regulatory Commission (NRC) License No. R-66) were conducted by CH2M HILL Constructors, Inc., assisted by several specialty subcontractors, beginning in April 2002.

Decommissioning criteria established for this project were the NRC's default screening guidelines for structure surfaces and soil, which provide a conservative approach to assure that future facility uses do not result in radiation doses to the public in excess of 25 mrem per year.

The final status surveys to demonstrate that these guidelines have been satisfied were performed by one of the CH2M HILL team subcontractors, Safety and Ecology Corporation.

Final surveys followed the recommendations of the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) and other NRC guidance. Co-60 and Cs137 were identified as primary contaminants present; however, because of the limited extent of impacted facility surfaces and media, assumptions as to other contaminants were based on limited radiological data. Conservative adjusted gross or surrogate guidelines were confirmed after the initial final survey data were evaluated.

The results of the final status surveys, documented in this report, demonstrate that the project decommissioning criteria have been satisfied and that the facility meets the requirements for termination of NRC License No. R-66.

ix

1.Introduction The University of Virginia operated a light-water cooled, moderated, and shielded pool-type nuclear research reactor beginning in June 1960. Reactor uses included radiation research, activation analysis, isotope production, neutron radiography, radiation damage studies, and training of Nuclear Engineering students. The reactor was initially commissioned to operate at a maximum power of one Megawatt (MW) thermal; it was

-upgraded to a power level of two MW in January 1971. Aluminum clad high-enriched uranium fuel was initially used; the reactor was converted to low-enriched uranium fuel in early 1994. The reactor operated under NRC License No. R-66.

In June 1998, the reactor was permanently shutdown, and the fuel was removed and shipped offsite between the shutdown date and mid 1999. Beginning in July 1999 GTS Duratek performed a radiological characterization of the reactor, the facility housing the reactor, and the surrounding fenced and gated land area, collectively referred to as the UVAR facility; results of that characterization are presented in a April 2000 Characterization Survey Report (Ref. 1). The University of Virginia submitted a Decommissioning Plan for the UVAR facility to the NRC in February 2000 (Ref. 2).

Beginning in March 2002, the University of Virginia contracted with CH2M HILL to conduct the decommissioning of the UVAR. CH2M HILL teamed with Waste Management Group, Inc. (WMG), Safety and Ecology Corporation (SEC), Bartlett Nuclear, Inc., and Parallax, Inc.

to accomplish this effort. This team conducted additional characterization surveys; surveyed and released or disposed of materials, depending on radiological conditions; and performed decontamination of components, where appropriate.

Following the removal or decontamination of surfaces and materials, a Final Status Survey of the facility was performed to demonstrate that the radiological conditions satisfy NRC-approved criteria for use without radiological restrictions and termination of License No.

R-66. This document describes the methodologies, results, and data evaluation for the Final Status Survey (FSS) of the UVAR facility.

1.1

2. Facility Description The UVAR facility is located on the Western grounds of the University of Virginia approximately 0.6 kilometers (km) west of the city limits of Charlottesville in Albemarle County Virginia. (Figure 2-1).

Figure 2-1 Map of Charlottesville Area Surrounding the UVAR Site 2-1

CH2MHILL The UVA Research Reactor and the decommissioned former CAVALIER facility, as well as offices for former faculty, students of the former Department of Nuclear Engineering, and the reactor staff, are housed in the facility. The CAVALIER Facility was decommissioned separately; a Final Status Survey was peirformed and license termination requested (see CAVALIER Final Status Report (Ref. 3) for details); the CAVALIER Facility area was added to the UVAR Facility and included in this Final Status Survey. To the north, east, and south of the facility (no closer than 0.5 km) there are city residential districts. The only access to the facility is by way of Old Reservoir Road (Figure 2-2).

Figure 2-2 Western Grounds of the University of Virginia Location 1: Aerospace Research Laboratory Location 12: Reactor Facility Location 2: Alderman Observatory Residence Area Location 14: Shelboume Hall Location 6: Hereford Residential College Location 15: Slaughter Recreation Location 7: High Energy Physics Laboratory Facility Location 9: McCormick Observatory' Location 16:

• Special Materials Location 10: National Radio Astronomy Observatory Handling Facility The land and facilities are the property of the University of Virginia, which is responsible for facility oversight and support. The UVAR facility site is depicted in Figure 2-3. The facility is located on the north side of a narrow valley with the land gradient falling north to south and west to east.

2

CH2MHILL Figure 2-3 UVAR Facility Site

,I.

i Figure 2-4 shows -the three levels of the UVAR facility.

The Reactor Confinement Room (Rm 131), which housed the former UVA Research Reactor, is located on the upper floor (first floor). This room contained the 9.8 m long by 3.7 m wide by 8.2 m deep reactor pool, associated operating equipment and systems, the operating controls, and some research/experimental equipment. This room is circular and has an elevated (-10 m) ceiling. In addition, the Instrument Shop (Rm 128), the Shipping Area (Rm 127), and multiple offices and other support facilities for staff and students are located on this building level.

On the Mezzanine level were located the Demineralizer (Rm M021), Mechanical Room (Rm M020), HP Laboratory (Rm M019), several partially contaminated laboratories (Rms M005

[Tc-99 contamination] and M008 [Ni-63 contamination]), and multiple offices and other support facilities for staff and students. A crawl space (MCS) is accessed from the stairwell on the Mezzanine level.

3

CH2MHILL Figure 2-4 Building Plan Views UVA Reactor First Floor Plan View UVA Reactor Mezzanine Floor Plan UVA Reactor Ground Floor Plan View The ground floor contained the Heat Exchanger (Rm G024), Rabbit Room (Rm G005), Beam port/Experimental area (Rm G020), Hot Cell (Rms G025, G026, and G027), Counting Room (Rrn G004), Woodworking and Machine Shop (Rm G008), Source Storage (Rmns G022, G018, 4

CH2MHILL and G007A), the former CAVALIER facility (Rm G007), and miscellaneous support facilities and areas.

There was a cooling tower located on the roof of the Mezzanine level, adjacent to the Reactor Confinement room; this facility provided cooling for the reactor secondary system water.

The 2030 m2 (interior floor space) UVAR facility building is situated on a 9390 m2 -fenced parcel of land. This land area included 2 sets of underground tanks for collection of potentially contaminated facility liquid wastes, a pond used for collection and holdup of facility discharges containing radioactive contamination, a tank used during fuel shipments at ground level at the front of the building, underground storm and sanitary sewer drainage systems, and miscellaneous larger materials and equipment with little or no potential for being radiologically impacted.

The UVAR building is of metal and concrete block construction with brick veneer. Floors are concrete slab. Internal walls are block and drywall.

Most impacted reactor and support systems and components were removed and disposed of as radioactive waste or surveyed and released for use without radiological restrictions.

The CH2M HILL decommissioning report records detailed information regarding the current facility status (Ref.4).

5

3. General Final Status Survey Approach This survey was performed in accordance with the guidelines and recommendations presented in the "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSIM),

NUREG-1575 (Ref. 5). Guidance provided in NUREG-1757, "ConsolidatedNMSS Decommissioning Guidance" (Ref. 6) was followed in the design, implementation, and evaluation of this final status survey. The process emphasizes the use of Data Quality Objectives and Data Quality Assessment, along with a quality assurance/quality control program. The graded approach concept was followed to assure that survey efforts were maximized in those areas having the greatest potential for residual contamination or the highest potential for adverse impacts of residual contamination.

For the purposes of guiding the degree and nature of FSS coverage, MARSSIM first classifies areas as impacted, i.e., areas that may have residual radioactivity from licensed activities, or non-imnpacted, i.e., areas that are considered unlikely to have residual radioactivity from licensed activities. Non-impacted areas do not require further evaluation. For impacted areas MARSSIM identifies three classifications of areas, according to contamination potential.

• Class 1 Areas: Impacted areas that, prior to remediation, are expected to have concentrations of residual radioactivity that exceed the guideline value.

• Class 2 Areas: Impacted areas that, prior to remediation, are not expected to have concentrations of residual radioactivity that exceed the guideline value.

• Class 3 Areas: Impacted areas that have a low probability, typically on the order of containing residual activity. Typically levels will not exceed 25-35% of the guideline value.

Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities were the bases for classification.

See Table 3-1 for area classifications and survey unit numbering.

This survey was performed in accordance with the Master Final Status Survey Plan and eight addenda to the plan (see Appendix. A). Each addendum provided the survey approach and requirements for each common area of survey. Section Four of this report follows the addenda sections for survey approach, data summary, assessment, and evaluation. The supporting documentation for this report is contained in Appendices B and C.

3-1

CH2MHILL Table 3-1 UVAR Survey Areas and Classifications Surface Area Survey Room or Area Surface Class (m2 ) or Unit Survey Length (m) No.

131 Reactor Room, West Floor 1 92 1 131 Reactor Room, East Floor 1 92 2 131 Reactor Room Lower Walls 1 103 3 Reactor Pool, South Floor and Walls 1 117 4 Reactor Pool, North Floor and Walls 1 117 5 M005/005A Floor and Lower Walls 1 65 6 M008 Floor and Lower Walls 1 89 7 M019 Floor and Lower Walls 1 107 8 M020 Floor and Lower Walls 1 104 9 M021/021A Floor, Walls, and Ceiling 1 158 10 Bio Shield Surfaces Wall 1 54 11 G005 Floor, Walls, and Ceiling 1 99 12 G007/GO07A Floor, Pit and Lower 1 167 13 Walls G018 Floor, Walls, and Ceiling 1 92 14 G020, West Floor and Lower Walls 1 55 15 G020, Center Floor and Lower Walls 1 67 16 G020, East Floor and Lower Walls 1 120 17 G022 Floor, Walls, and Ceiling 1 48 18 G024 Floor, Walls, and Ceiling .1 105 19 G025/G026/G027 Floor, Walls, and Ceiling 1 146 20 Pond Sediments 1 160 21 Waste Tank Area Soil 1 350 22 Reactor Stack Ductwork, stack, blowers 1 N/A 24 Ventilation System 1 Ductwork, stack, blowers 1 N/A 25 Ventilation System 2 Ductwork, stack, blowers 1 N/A 26 Heat Exchanger Piping Piping interior 1 13 m 27 Reactor Pool Drains Piping interior 1 20 m 28 Reactor room floor drains Piping interior 1 19 m 29 Sanitary sewer release path Piping interior 1 21 m 30 Drain to LWST Piping interior 1 11 m 31 Hot Cell Drain Piping interior 1 13 m 32 Reactor Drains to Pond Piping interior 1 36 m 33 Fill Around Reactor Pool Soil 1 1000 N/A Soil Beneath Reactor Pool Soil 1 140 N/A Reactor Pool cores "M" and "B" Soil 1 12 N/A areas Demineralizer room wall core Soil 1 <10 N/A Outside reactor room Roll-up Asphalt 1 <10 62 door Ventilation System 3 Ductwork, stack, blowers 1 N/A 61 3-2

CH2MHILL Table 31 (Continued) UVAR Survey Areas and Classifications Surface Area Survey Room or Area Surface Class (m2 ) or Unit Survey Length (m) No.

131 Reactor Room Upper Walls and Ceiling 2 690 34 127/128/130 Floor, Walls, and Ceiling 2 176 35 107/124/124A/124B Floor and Lower Walls 2 311 36 M005/005A. Upper Walls and Ceiling 2 50 37 M008 Upper Walls and Ceiling 2 56 38 M019 Upper Walls and Ceiling 2 72 39 M020 Upper Walls and Ceiling 2 76 40 M006/M014/M015/M030/

M031 Floor and Lower Walls 2 259 41 MCS (crawl space) Floor, Walls, and Ceiling 2 153 42 G004/GO05A Floor and Lower Walls 2 154 43 G006 Floor and Lower Walls 2 64 44 G007B/G008/G0O8A/G016

/G017/G019 Floor and Lower Walls 2 362 45 Stairwell 1 Floor and Lower Walls 2 119 46 Stairwell 2 Floor and Lower Walls 2 184 47 Reactor Confinement Roof All 2 214 48 Main Building Roof All 2 863 49 Outside Paved Areas All 2 2236 50 Outside soil areas Soil 2 6264 52 CAVALIER Facility Drain Piping Interior 2 22 m 51 Storm and sanitary drains Basins and piping 2 34 m 23 MCS Soil Floor Soil 2 43 53 G007/GO07A Upper Walls and Ceiling 3 104 54 G020 Upper Walls and Ceiling 3 437 55 107/124/124A/124B Upper Walls and Ceiling 3 220 56 M006/M014/M015/M030/ Upper Walls and Ceiling 3 192 57 M031 G004/GO05A Upper Walls and Ceiling 3 107 58 G006 Upper Walls and Ceiling 3 31 59 G007B/G008/G0O8A/G016 Upper Walls and Ceiling 3 280 60

/G017/G019 G002 All 3 71 63 Elevator All 3 21 64 Mezzanine Offices All 3 1190 65 First Floor Offices All 3 1934 66 Outside Exterior Walls Doors, vents 3 1362 67 N/A = Not Applicable 3-3

CH2MHILL 3.1 Radiological Contaminants and Criteria The GTS Duratek initial characterization survey and continuing characterization by the CH2MHILL team indicated that the radiological contamination present was generally low level and was limited to the fenced grounds. Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclides were Co-60 and Cs-137; smaller activities of fission and activation products, namely C-14, Fe-55, and Eu-152 were identified in some media. Ni-63 and Tc-99 were surface contaminants from research projects in labs M008 and M005, respectively. Elevated levels of uranium decay series nuclides were identified in the pond sediments; pool fill soil and surface soil of the facility grounds, however, these were considered to be of natural origin and not to have originated from licensed reactor operations.

The Decommissioning Plan established the criteria for residual radioactive material contamination on WAR facility surfaces and in facility soil. UVAR facility criteria also referred to as derived concentration guideline levels (DCGLs) were selected from the tables of NRC default screening values (refer to NUREG-1757, Ref. 6 and NUREG/CR-5512, Vol.3, Ref. 7). The screening values for total surface contamination for radionuclides anticipated at UVAR are listed in Table 3-2; guideline levels for removable activity are 10% of the values in that table. Screening values for anticipated contaminants in soil are listed in Table 3-3. These screening criteria were based on assuring that estimated doses to facility occupants and the public during future facility use do not exceed annual doses of 25 mrem; default screening criteria were based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exceeded.

The criteria described in this section are net (above background) concentrations and activity levels of radionuclides; appropriate adjustments for instrument background levels were made to survey data before comparing data to the respective criteria.

Use of default screening values as decommissioning guidelines does not allow for areas of elevated activity. Therefore, there are no area factors for small areas of contamination, and all surface activity levels and radionuclide concentrations in soil must satisfy those guideline levels and methodology of Appendix E of NUREG-1757 (Ref. 6). In addition, because of use of the conservative default screening values, further evaluations and actions, relative to demonstrating the final conditions satisfy ALARA, are not required.

34

CH2MHILL Table 3-2 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces Radionuclide Symbol Screening Value Source (dp /i00 cm 2 )

Hydrogen-3 (Tritium) H3 1.2E+08 NUREG-1757 Carbon-14 C14 3.7E+06 NUREG-1757 Sodium-22 Na 2 2 9.5E+06 NUREG-1757 Sulfur-35 S35 1.3E+07 NUREG-1757 I Chlorine-36 C136 5.OE+05 NUREG-1757 Manganese-54 Mn 5 4 3.2E+04 NUREG-1757 Iron-55 Fe5 5 4.5E+06 NUREG-1757 Cobalt-60 Co60 7.1E+03 NUREG-1757 Nickel-63 Ni63 1.8E+06 NUREG-1757 Strontium-90 Sr9 0 8.7E+03 NUREG-1757 Technetium-99 Tc99 1.3E+06 NUREG-1757 Iodine-129 I129 3.5E+04 NUREG-1757 Cesium-137 Cs13 7 2.8E+04 NUREG-1757 Europuium-152 Eu' 52 1.3E+04 NUREG/CR-5512, Vol. 3 Plutonium-238 Pu3 3.1E+01 NUREG/CR-5512, Vol. 3 Plutonium-239 PU239 2.8E+01 NUREG/CR-5512, Vol. 3 Plutonium-241 Pu241 1.4E+03 NUREG/CR-5512, Vol. 3 Americium-241 Am24 1 2.7E+01 NUREG/CR-5512, Vol. 3 Notes:

a Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume for screening purposes that 100 percent of the surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. Users may calculate site-specific levels using available data on the fraction of removable contamination and DandD version 2.

b Units are disintegrations per minute (dpm) per 100 square centimeters (dpm/100 cm 2). One dpm is equivalent to 0.0167 becquerel (Bq). Therefore, to convert to units of Bq/m 2 , multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSv/y (25 mrem/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies (see Part 20, Appendix B, Note 4).

3-5

CH2MHILL Table 3-3 Acceptable License Termination Screening Values of Comnmon Radionuclides for Surface Soil Radionuclide Symbol Screening Value Source (pCi/g)

Hydrogen-3 (Tritium) H3 1.1E+02 NUREG-1757 Carbon-14 C14 1.2E+01 NUREG-1757 Manganese-54 Mn 5 4 1.5E+01 NUREG-1757 Iron-55 Fe55 1.OE+04 NUREG-1757 Cobalt-60 Co60 3.8E+00 NUREG-1757 Nickel-63 Ni63 2.1E+03 NUREG-1757 Strontium-90 Sr90 1.7E+00 NUREG-1757 Technetium-99 Tc99 1.9E+01 NUREG-1757 Iodine-129 I129 5.0E-01 NUREG-1757 Cesium-137 Cs137 1.1E+01 NUREG-1757 52 Europium-152 Eu' 8.7E+00 NUREG/CR-5512, Vol. 3 Plutonium-238 PuZ 8 2.5E+00 NUREG/CR-5512, Vol. 3 Plutonium-239 Pu239 2.3E+00 NUREG/CR-5512, Vol. 3 Plutonium-241 Pu241 7.2E+00 NUREG/CR-5512, Vol. 3 Americium-241 Am24 ' 2.1E+00 NUREG/CR-5512, Vol. 3 Notes:

a These values represent superficial surface soil concentrations of individual radionuclides that would be deemed in compliance with the 25 mrem/y (0.25 mSv/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies; see Part 20, Appendix B, Note 4.

b Screening values are in units of (pCi/g) equivalent to 25 mrem/y (0.25 mSv/y). To convert from pCi/g to units of Becquerel per kilogram (Bq/kg), divide each by 0.027. These values were derived using DandD screening methodology (NUREG/CR-5512, Volume 3). They were derived based on selection of the 90th percentile of the output dose distributionfor each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at "Standard Man" or the mean of the distribution for an average human.

3.2 Data Quality Objectives The objective of the FSS was to demonstrate that the radiological conditions of the facility satisfy the decommissioning criteria established in the NRC-approved Decommissioning Plan. The Data Quality Objectives (DQOs) demonstrated at the 95%

confidence level that these criteria have been met. Decision errors were 5% for both Type I and Type II errors. Such a Type I (alpha) decision error provides a confidence level of 95% that the statistical tests will not determine that a surveyed area satisfies 36

CH2MHILL criteria when, in fact, it does not. The Type II (beta) decision error provides a confidence level of 95% that the statistical tests will not determine that a surveyed area does not satisfy criteria when, in fact, it does. Measurement sensitivities will enable quantification of contaminants at or below the DCGL values at the 95% confidence level.

Data quality indicators for precision, accuracy, representativeness, completeness, and comparability, have been established.

• Precision is determined by comparison of replicate values from field measurements and sample analyses; the objective is a relative percent difference of 20% or less at 50% of the guideline value.

• Accuracy is the degree of agreement with the true or known value; the objective for this parameter is +/- 20% at 50% of the guideline value.

• Representativeness and comparability do not have numeric values.

Performance is assured through selection and proper implementation of sampling and measurement techniques.

• Completeness refers to the portion of the data that meets acceptance criteria and is thus acceptable for statistical testing; the objective for this survey is 90%.

3.3 Final Status Survey Tasks

• Surface Beta Radioactivity Scan Surveys

• Gamma Surface Scans

• Integrated Direct Surface Beta Radioactivity Measurements

• Smear Sample Collection and Analysis

• Volumetric Sample Collection and Analysis 3.4 Survey Instrumentation and Methods Table 3-4 lists the instrumentation used for survey activities described in this FSS Report, along with nominal operating parameters and estimated detection sensitivities.

Instrument response was based on use of an average surface efficiency of 0.25 (per recommendations of ISO 7503) (Ref. 8). This conservatively low efficiency, based on Tc-99, underestimates the true detector response for the higher beta energies associated with Cs-137 and Co-60. Thus, for areas with no Tc-99 present, calculated quantities will be higher than those actually present.

Detection sensitivities were estimated, using the guidance in NUREG-1575 (MARSSIM) and NUREG-1507 (Ref. 9). Instrumentation and survey techniques were chosen to obtain detection sensitivities below the applicable DCGLs for both scanning and direct measurement, with the objective of achieving *25% of the DCGL. This objective was not achievable for gamma scans of land surfaces using Model 44-10 gamma scintillation detectors and piping scans using a Model 491-30 GM detector, but these survey activities were sufficient to assure identification of areas of elevated activity of a size and activity level that could adversely affect the average for the survey units.

3-7

CH2MHILL All instruments had current calibrations using NIST-traceable standards. Operational and background checks were performed at the beginning of each day of FSS activity and whenever there was reason to question instrument performance.

Table 3-4 Instrumentation for UVAR Final Status Survey 3.4.1 Surface Beta Radioactivity Scan Surveys Beta scanning of structure surfaces was performed to identify locations of residual surface activity. Gas-flow proportional detectors were used for beta scans. Floor monitors with 580 cm 2 detectors were used for floor and other larger accessible horizontal surfaces; hand-held 125 cm2 detectors were used for surfaces not assessable by the floor monitor. Scanning was performed with the detector within 0.5 cm of the surface. Scanning speed was no greater than 1 detector width per second. Audible signals were monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage was 100% for Class 1 surfaces, 25% for Class 2 surfaces, and 10% for Class 3 surfaces (Ref. 10). Coverage for Class 2 and Class 3 surfaces was biased towards areas considered by professional judgment to have highest potential for contamination.

3.4.2 Gamma Surface Scans Gamma scanning surfaces were performed on structure and land surfaces to identify locations of residual surface activity. NaI gamma scintillation detectors (2' x 2") were

CH2MHILL used for these scans. Scanning was performed by moving the detector in a serpentine pattern, while advancing at a rate of approximately 0.5 m per second. The distance between the detector and the surface was maintained within 5 cm. Audible signals were monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage was 100% for Class 1 surfaces, 25% for Class 2 surfaces, and 10% for Class 3 surfaces (Ref. 10). Coverage for Class 2 and Class 3 surfaces was biased towards areas considered by professional judgment to have highest potential for contamination.

3.4.3 Integrated Direct Surface Beta Radioactivity Measurements Measurements of surface beta radioactivity were performed using a Ludlum Model 43-68 handheld 125 cm 2 gas proportional detector coupled to a Ludlum Model 2221 ratemeter/scaler. Counts were integrated for a one-minute counting interval to obtain measurement sensitivity less than the DCGL. Two measurements were performed at each measurement location. The first of these was a surface measurement, performed in the typical manner (i.e., with the detector face uncovered); this measurement included contributions from beta particles emitted from the surface and interactions of ambient gamma photons with the detector. The second measurement was performed at the same location with the detector face covered by a layer of material. A piece of wood approximately 1.27 cm (1/2-inch) thick, which contained no significant beta-emitting component and which has sufficient density thickness to shield out the beta particles, but not reduce the gamma photon level. The detector response for this second measurement was representative of the contribution from gamma radiation only. The difference between measurements with an uncovered (unshielded) detector and covered (shielded) detector represented the level of beta activity, only, which was then compared with the surface contamination criterion. Instrument beta response factors (efficiencies) incorporate considerations for source efficiency, due to potential adverse surface conditions. As recommended in International Organization for Standardization (ISO) 7503 for beta emitters with maximum energies less than 0.4 megaelectron volts (MeV), a source efficiency factor of 0.25 was used in determining the effective total instrument efficiencies. Total efficiency factors for the various instruments used for direct beta measurement were within very close agreement. Effective total instrument efficiencies were used for converting count-rate data to activity units.

3.4.4 Smear Sample Collection and Analysis Smear samples for removable activity were collected by wiping a 5 cm (2-inch) diameter cloth disc over approximately 100 cm 2 (15.5 in2) areas of the surface, while applying moderate pressure. Smear samples were obtained at each location of direct surface activity measurement. Smear samples were counted on a Tennelec LB 5100 automatic gas proportional counter for alpha and beta radioactivity. The primary purpose of collecting smear samples was to provide data to confirm the assumption of the dose assessment that transferable radioactivity is less than 10% of total radioactivity. The DQO for smear samples was to insure that the dose model assumptions used to develop the criteria were appropriate.

3-9

CH2MHILL 3.4.5 Volumetric Sample Collection and Analysis Volumetric samples were collected as prescribed in the FSS plan. Samples were managed under chain-of-custody procedures and submitted to Severn Trent Laboratory, located in Earth City, Missouri and Eberline Services, Oak Ridge laboratory, Oak Ridge, Tennessee for gamma spectroscopy and 10 CFR Part 61 analyses for hard to detect nuclides.

3.5 Quality Assurance Instruments used for the Final Status Survey were maintained and calibrated to manufacturers' specifications to ensure that required traceability, sensitivity, accuracy, and precision of the equipment/instruments was maintained. The SEC laboratory located in Knoxville, Tennessee followed standard procedures per ANSI N323A-1997 and used National Institute of Standards and Technology (NISTI) traceable sources to calibrate the equipment/instruments.

Before and after daily use, instruments were quality control (QC) checked by comparing the instruments' response to designated radiation sources and to ambient background.

These performance checks were performed at a predetermined site reference location within the UVAR facility.

Instrument responses to the designated QC check sources were plotted on control charts and evaluated against the average established at the start of the field activities. A performance criterion of +20% of this average was used as an investigation action level.

No instruments were removed from service for not meeting operational requirements.

During QC checks, instruments used to obtain radiological data were inspected for physical damage, current calibration, and erroneous readings in accordance with applicable procedures and/or protocols. Instrumentation not in compliance with the specified requirements of calibration, inspection, or response check was removed from operation. If the instrument failed the QC response check, any data obtained to that point, but after the last successful QC check were considered invalid due to faulty instrumentation. No data were rejected during the FSS due to QC criteria.

3.5.1 Data Quality Controls Project data were recorded in a Project Data Logbook or on standard, preprinted data forms. Records were reviewed daily.

Sample chain-of-custody was maintained for volumetric samples.

Duplicate sampling and measurements were performed for 20% of the sampling/measurement locations.

The analytical laboratory performed laboratory spike and blank analyses. Relative percent differences were determined for the spike results and compared to a project performance criterion of +20%. Blank sample results were compared with a performance criterion of no detectable activity.

3-10

CH2MHILL 3.6 Data Assessment and Evaluation Data was reviewed to assure that the type, quantity, and quality were consistent with the Final Survey Plan and design assumptions. Data standard deviations were compared with the assumptions made in establishing the number of data points. Individual and average data values were compared with guideline values and proper survey area classifications were confirmed. Individual measurement data in excess of the guideline level for Class 2 areas and in excess of 25 % of the guideline for Class 3 areas prompted investigation. Patterns, anomalies, and deviations from design assumption and Plan requirements were identified. Need for investigation, reclassification, remediation, and/or resurvey was determined; a resolution was initiated and the data conversion and assessment process repeated for new data sets.

3.7 Background Determination and Reference Areas The UVAR Decommissioning Plan identified Ragged Mountain Reservoir as a comparable area to the UVAR facility to obtain background soil samples. However, further evaluation indicated that the soils at UVAR facility contained higher levels of naturally occurring radionuclides than those at Rugged Mountain Reservoir. Therefore, samples from the immediate vicinity of UVAR, but without a potential of being impacted by site operations, were obtained to determine background levels of radionuclides in the area.

Radionuclides, which are potential contaminants of concern due to licensed UVAR activities, are not naturally occurring in site soil or construction materials at concentrations greater than 10% of the project DCGLs. Therefore, no soil background reference areas were determined to be required for this Final Status Survey.

Static count backgrounds for area surveys were determined at the time of survey, utilizing a shielded/unshielded probe approach.

3-11

4. Final Status Survey

4.1 Background

4.1.1 General Radionuclides, which are potential contaminants of concern due to licensed UVAR activities, are not naturally occurring in site soil or construction materials at concentrations greater than 10% of the project DCGLs. Therefore reference areas for evaluation of soil contamination are not applicable to the UVAR Decommissioning Project. Survey designs were based on data requirements for the Sign Test, with evaluation of soil survey data to be based on comparison of specific radionuclides with default screening DCGLs or surrogate DCGLs, established for a soil area of interest.

Direct measurement gamma and beta background levels were noted to be variable and occasionally elevated throughout the facility and surrounding land areas. This was due to local geologic formations containing naturally occurring K-40, uranium, and thorium; a wide-variety of construction materials, also containing varying levels of naturally occurring radionuclides; and localized areas of elevated ambient radiation from stored radioactive sources regulated under other licenses. Because this background variability was not conducive to establishing background reference areas, it was decided that individual direct measurement locations would have unshielded and shielded gross activity determinations and the difference would be the basis for determining activity for comparison with established guideline levels. Since reference areas were not used, the statistical design of required data points was based on the Sign test.

4.1.2 Survey Activities and Results The UVAR Decommission Plan identified Ragged Mountain Reservoir as a potential offsite location for soil background determination. However, because of the reasons indicated in Section 4.1.1, it was decided that a soil reference area was not necessary, and, furthermore, that the geology of the Ragged Mountain Reservoir area was not comparable with the naturally occurring nuclides at the UVAR site. Although a soil reference area was not to be used in final status survey evaluation, it was necessary to establish the identity and activity range of the naturally occurring radionuclides in site soils and thus enable these radionuclides to be eliminated from consideration in the evaluation.

Streambed samples were obtained from two locations (#1 and #2) approximately 200-m and 250-m upstream from the pond. Rock face and fill samples were obtained by hand auger at two locations (#3 and #4) upgradient from the UVAR Building; three samples from each location represented the 0-15 cm, 15-45 cm and 45-75 cm depths. See Figure 4-1 for sample locations.

4-1

I-'

CH2MHILL Figure 4-1 Background Soil Locations 4

0D0 I E C

'N7 CE 9 ~' '.

Results of soil analyses are summarized in Table 4-1. Concentrations of C-14 and the uranium series are consistent with those typically present in background soils, while K40 and thorium series radionuclides are slightly (2 to 4 times) higher than typically noted in soil. It is believed that the elevated levels of these radionuclides are likely due to the bedrock underlying the site.

Positive concentrations (up to 1.8 pCi/g) of Eu-155 were also reported for these samples.

This radionuclide is not naturally occurring and is not typically identified as a contaminant of reactor origin, particularly considering that Eu-152 and Eu-154, which are of reactor origin, were not identified in these samples. Considering the cross-sections of activation and natural abundances of the stable elements from which these Eu-isotopes are produced, the Eu-155 level would be expected to be lower than of Eu-152 and Eu-154. Finally, the photon energies of Eu-155 are very close to X-ray and gamma photon energies present due to naturally occurring radionuclides. For these reasons, the Eu-155 by gamma spectrometry is considered a misidentification, and not a contaminant of reactor origin.

4-2

CH2MHILL Table 4-1 Concentration of Radionuclides in Background Soil Sample Activity (pCi/g)

Location C-14 K-40 U-Series Th-Series

• 1 6.5 + 2.55 22.6 +4.2 1.6 0.4(c) 1.6 + 0.3 (e) 2 2.99 +1.54 29.3 + 3.8 1.6 0.4(c) 1.9 + 0.4 (e) 3-1 N/D (a) 34.9 + 4.4 1.4 + 0.3 (d) 2.8 + 0.4 (b) 3-2 N/D (a) 42.8 + 5.2 1.4 + 0.3 (d) 2.9 + 0.4 (b) 3-3 N/D (a) 41.1 + 5.1 1.3 +/- 0.3 (d) 3.3 +0.5 (b) 4-1 N/D (a) 20.8 + 2.8 0.9 0.2 (d) 2.2 +/- 0.3 (b) 4-2 N/D (a) 32.3 + 4.2 1.8 _ 0.3 (d) 4.5 + 0.5 (b) 4-3 N/D (a) 38.7 + 4.8 1.5 +/- 0.3 (d) 4.0 + 0.5 (b)

Note a Not Determined Note b Based on Ac-228 measurement Note c Based on U-238 measurement Note d Based on Bi-214 measurement Note e Based on Th-232 measurement Variable gamma radiation levels were noted with 2" x 2" NaI detectors, used for surface scanning surveys. General background levels outside the building were approximately 8,000 cpm, but ranged up to 28,000 cpm in contact with rock outcroppings on the site. Such levels are not unexpected, considering the natural content of K-40 and thorium in site soil and rock. Gamma levels were also elevated and highly variable (up to 88,000 cpm) in the vicinity of the Hot Cell doors, due to the presence of radioactive source in storage in that facility.

Inside the building, gamma background levels generally ranged from about 6,000 to 12,000 cpm. Higher values (up to 30,000 cpm) were observed in portions of the structure, where materials such as concrete, ceramic tile, cinder block, and brick, containing naturally occurring radioactive materials, were present, where geometry was enhanced due to small rooms and at interfaces of two or more surfaces, and in areas adjacent to source storage.

Beta backgrounds on instruments used for scans and direct requirements also varied, depending on the specific detector, the surface material, and the ambient gamma levels from radioactive sources in the immediate area.

For direct beta measurements, adjustment for background was made by conducting unshielded and shielded measurements at each data point; the difference represented the surface activity level, conservatively over-estimating the true activity present by not correcting for naturally occurring activity in the surface.

4.2 Underground Waste Tank Excavation 4.2.1 Description Two sets of underground metal tanks, located southeast of the UVAR facility adjacent to the pond, were used for collection/holdup of liquid wastes, which were potentially contaminated with low concentrations of radioactive materials (Figure 4-2). Two of these tanks (HCfs) serviced the hot cell facility and two tanks (LWSTs) were used for collection of 4-3

CH2MHILL demineralizer regeneration liquids from the 2-MW UVA Reactor. Both of these tank sets were initially equipped for environmental discharge to the pond, provided the liquid met appropriate release criteria following dilution with pond water. However, the demineralizer regeneration liquid tanks were later replumbed to discharge directly into the pond spillway.

All tanks, associated piping, valves, pumps, etc., have been removed along with their concrete enclosures and foundations. The floor of the LWST enclosure contained small quantities of contaminated soil-like materials near the tanks which were removed. The resulting excavation was approximately 175 m2 in area and ranged up to approximately 3 m in depth; including the unexcavated soil edges. The area addressed by this survey was approximately 350 M2 .

Before excavation of the tanks, a sample of waste tank sludge, composed of resin fines and sediments, was collected from the demineralizer regeneration liquid tanks and found to contain 6,930 pCi/g of Co-60, 8,142 pCi/g of H-3, 1,110 pCi/g of Fe-55, and much smaller or non-detectable concentrations of multiple other radionuclides (refer to Final Status Survey Plan, Addendum 001, in Appendix A). Only four radionuclides, (Co-60, H-3, Mn-54, and Sr-90) were present at levels which would potentially be responsible for greater than 1% of the total dose from the mixture. These results are consistent with expected liquids from a pool-type reactor with minimal fuel leakage and stainless steel and aluminum components.

Co-60 at a DCGL gate of 3.4 pCi/g was established as the guideline for evaluation of excavation soils for compliance. In addition, because this guideline was based on a single sample, a composite of all systematic samples was analyzed for hard-to-detect radionuclides to confirm the absence of such contaminants in this survey unit.

During waste tank removal, a small area of contaminated soil was identified at the base of the waste tank blockhouse interior. Samples from this location identified only Co-60 and Cs-137 at detectable concentrations. Additional soil was removed to a depth of approximately 0.5 m from the interior of the blockhouse excavation, where these samples were obtained. Hand augured samples from three other locations did not contain detectable concentrations of facility related gamma-emitting radionuclides. Approximately 40 additional samples of excavated (non-impacted) soil were collected and analyzed by gamma spectrometry. No gamma-emitting radioactive contaminants were identified at detectable levels in these samples.

4.2.2 Survey Activities A 5-meter grid was established over the excavation area and extended to unexcavated soil surrounding the excavation. This grid was an extension of the reference grid established for the survey of the Pond Sediments, thus enabling the sampling locations to be related to the federal and/or state planar coordinate system. Figure 4-3 illustrates the reference grid system's relationship to the excavation. Based on facility use history and identification of contaminants of license origin in the soils of this area, the survey area is designated Class 1 for FSS planning and implementation purposes. The area of the excavation and surrounding soil is approximately 350 m2 ; this is within the MARSSIM-recommended area of 1,000 m2 for Class 1 open land survey units. Therefore, the area is a single survey unit.

44

CH2MHILL Figure 4-2 UVAR Facility and Environs Indicating Location of Waste Tanks Demineralizer Regeneration Tanks Gamma walkover surface scans were performed using a 2"X 2" Nal detector (Ludlum Model 4-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5 to 10 cm of the soil surface and moved from side to side in a serpentine pattern while noting any indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Scanning coverage was 100% of the soil surface.

Compliance with decommissioning requirements was demonstrated by comparison of results of FSS sample analyses with the Co-60 DCGLswrnogate. Because radionuclides identified as potential contaminants are not present in background soil at concentrations, which are significant fractions of the release guidelines, correction of FSS sample data for background levels was not required. Statistical testing of results utilized the Sign test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors are 0.05 (Type I and Type II).

The number of data points required for the Sign test was determined to be 17 (refer to Section 4.5 of Addendum 001). A triangular pattern, based on 17 samples and 350 m2 area, was used to determine sampling locations. The distance between samples was 5.0 m. A random start point of 7.9 m N and 11.2 m W for the pattern was based on survey unit dimensions and random numbers from the MARSSIM random number table. Because only 14 locations fell onto soil surfaces, an additional line of locations at -0.7 m N was added; the resulting number of sampling locations was 19. Figure 4-3 indicates the sampling locations.

4-5

CH2MHILL Surface (0 to 15 cm) soil samples of at least 500 g were collected at the 19 systematic sampling locations. If a sample could not be obtained from a pre-identified location, one was obtained from the nearest soil location available, and the deviation noted in the survey record. The licensee and the NRC Inspector witnessed the soil sampling and selected samples for confirmatory purposes. Requested samples were homogenized and split. This process was used to accelerate backfilling to maintain slope stability. Samples were assigned unique identification numbers and a chain of custody record and analytical request were prepared.

Samples were screened by on-site gamma spectrometry and then sent to an off-site commercial laboratory for individual gamma analyses. To confirm the absence of non-gamma emitting contaminants at significant levels, a composite sample, consisting of 20 grams from each of the individual systematic samples, was prepared and analyzed by the off-site laboratory for hard-to-detect (10 CFR Part 61) radionuclides.

4.2.3 Survey Results Detailed field survey data forms are included in Appendix B. (survey number tVAR-FS-449). Gamma scans ranged from 11,000 cpm to 31,000 cpm; for comparison, ambient gamma levels in the immediate vicinity of the excavation ranged from approximately 10,000 cpm to 15,000 cpm. Scan levels were higher within the excavation, where detector - source geometry was optimized, and at locations of exposed or near-surface bedrock and areas of rock fill; elevated gamma scan levels were not observed to be associated with possible contamination by materials from the tanks. An additional

("biased") soil sample was collected from the location of highest gamma scan level, near reference grid coordinate 5N, lOW.

446

C. CH2M;,IL FIGURE 4-3 WASTE TANK SURVEY UNIT, INDICATING THE GRID FOR SURVEY AND SAMPLE LOCATIONS 4-7

CH2MHILL Table 4-2 presents the results of gamma spectrometry analyses for the 19 systematic soil K> samples and 1 soil sample from the location of highest scan gamma level. None of the samples contained detectable levels of Co-60. Four samples contained Cs-137 activity above the detection sensitivity; the maximum Cs-137 concentration was 0.56 pCi/g. No other gamma emitters of license origin were present at detectable concentrations. All samples contained well below the Co-60 DCGLsiirrogate of 3.4 pCi/g. The average and standard deviation of Co-60 concentrations for the systematic samples is <0.17 pCi/g and 0.05 pCi/g respectively. Retrospective calculation of the relative shift yields approximately 65, which is much greater than the survey design value, thus indicating adequate data points were obtained.

The sample from the location of highest gross gamma scan results did not contain any detectable gamma radionuclides of facility origin. Analyses of the composite of 19 systematic samples are summarized in Table 4-3. None of the four radionuclides potentially contributing greater than 1% of the total dose from the mixture were present at detectable concentrations. Of the radionuclides of potential license origin only Am-241, Fe-55, Pu-238, and Pu-241 had levels of activity above their method detection sensitivities. Relative to their respective default DCGL values, the highest ratio of concentration to DCGL for these radionuclides with positive concentrations was 0.157 for Am-241. These results confirm the absence of significant levels of hard-to-detect contaminants.

Table 4-2 Results of Soil Sampling from Waste Tank Excavation Sample Location Co-60 Cs 137 Other (c)

North West (pCi/g) (pCi/g) Nuclides

-0.7 1.2 <0.22 <0.17 None Detected

-0.7 6.2 <0.19 0.46 None Detected

-0.7 11.2 <0.12 0.1 None Detected

-0.7 16.2 <0.15 <0.14 None Detected

-0.7 21.2 <0.26 0.56 None Detected 3.6 -1.3 <0.16 <0.14 None Detected 3.6 3.7 <0.17 <0.16 None Detected 3.6 8.7 <0.17 <0.14 None Detected 3.6 13.7 <0.11 <0.10 None Detected 3.6 18.7 <0.21 0.28 None Detected 3.6 23.7 <0.16 <0.31 None Detected 7.9 1.2 <0.14 <0.12 None Detected 7.9 6.2 <0.15 <0.14 None Detected 7.9 11.2 <0.17 <0.14 None Detected 7.9 16.2 <0.22 <0.21 None Detected 7.9 21.2 <0.12 <0.11 None Detected 12.2 8.7 <0.10 <0.23 None Detected 12.2 13.7 <0.21 <0.18 None Detected 12.2 18.7 <0.10 <0.11 None Detected 5 10 <0.17 <0.14 None Detected 48

CH2MHILL Table 4-3 Analyses of Composite Samples from Waste Tank Excavation ad .d Concentration Radionuclide (/9)

Am-241 0.33 +/- 0.15 Cm-244 <0.09 Co-57 <0.10 Co-60 <0.15 Cs-137 <0.12 Eu-152 <0.90 Eu-154 <0.37 Fe-55 1.99+/- 1.75 H-3 <3.74 1-129 <0.61 Mn-54 <0.12 Ni-63 <2.32 Pu-238 0.19 +/- 0.08 Pu-239 . <0.03 Pu-241 4.98 +/- 1.53 Sr-90 <0.65 Tc-99 <0.16 Zn-65 <0.30 4.2.4 Conclusion All individual systematic samples contain well below the Co-60 DCGLsursogate of 3.4 pCi/g, and no significant levels of other contaminants of license origin were'detected. Statistical testing is not required for data evaluation relative to the established guideline. These results demonstrate that the soils of the Waste Tank Excavation satisfy the established project decommissioning criteria.

4.3 Facility Piping 4.3.1 Description The bulk of known potentially contaminated piping was removed from the UVAR Facility during remediation activities, but sections of radiologically impacted piping previously associated with the reactor coolant system and various drains from the reactor facility remain.

This remaining piping is embedded in concrete or buried beneath concrete or asphalt paving and soil. The piping is generally of small diameter (2 in (5 cm) to 4 in (10 cm) ID); however there are several short sections of larger diameter. All or portions of the following impacted piping remain:

• Heat exchanger lines: Stainless Steel (SS), 6 in ID x 22 ft and 6 in ID x 32 ft.

• Reactor pool drains: SS, 2 in ID x 32 ft and 2 in ID x 36 ft.

• Reactor Room floor drains: Cast Iron (CI), 2 in ID x -160 ft (multiple sections).

4-9

CH2MHILL

• Ground floor drains to Pond standpipe: CI, 2 in ID x 40 ft and 4 in ID x 140 ft.

• Reactor Demineralizer drain to outside underground collection tanks: CI, 2 in IDx75ft.

• Hot Cell drain to outside underground collection tanks: Duriron with PVC repair, 2 in ID x 55 ft.

• Ground floor Bulk Access Facility drains to Pond hillside: CI, 2 in ID x 40 ft and terra cotta, 4 in ID x 80 ft.

• Sanitary sewer from liquid release point to sewer manway: 4 in CI by 40 ft.

• Drain lines from CAVALIER facility to Pond hillside.

Additional, non-radiologically impacted piping includes building and pool footing drains, storm drains, and the portion of the sanitary sewer drains and the portion of the sanitary sewer not on the liquid release path. These lines are located underneath the building floors and underneath the paved parking area between the building and the pond.

Figures 4-4 through 4-6 illustrate the locations of the reactor facility piping.

Visual (boroscope) inspection of the internal surfaces of reactor piping revealed breaks or blockages in the floor drain piping system beneath the Reactor Room floor. This inspection also identified accumulations of scale and loose debris, concentrated on the bottom surfaces of the piping. Visual inspection of the sanitary system piping was performed and the lines appeared clean and free from scale. Visual inspection of the storm drain system was not conducted.

Broken or damaged areas of piping were accessed, and contaminated pieces of pipe and soil were identified and removed. Dashed lines on Figure 4-6 denote piping removed from the reactor room. Hydrolazing of reactor piping internal surfaces was performed to remove scale and loose debris. Piping access points were created to enable the performance of the final status survey.

Preliminary scans, direct measurements, swabs, and water rinses of remaining piping were performed to identify the presence of contamination. Contaminated surfaces were removed or remediated.

Soil removed during excavation of the underground waste tanks, soil from the vicinity of piping breaks, debris collected from piping, and pieces of removed piping were analyzed on site by gamma spectroscopy. These analyses identified Co-60 as the primary potential contaminant in most of the remaining piping. Cs-137 was the major potential contaminant associated with the Hot Cell drain. Because piping did not contain sufficient activity levels to enable meaningful determinations of the contaminant mixture, particularly for the hard-to-detect radionuclides, it was assumed that the mixture in the reactor piping was the same as that reaching the waste tanks (refer to Master Final Status Survey Plan (FSSP) Addenda 001 and 002-Appendix A). This activity mixture is dominated by Co-60 (39%) and H-3 (46%). On a dose basis, Co-60 contributes 87.7% of the dose and Pu-241 contributes 12.2% of the total dose from this mixture.

4-10

CH2MHILL The Decommissioning Plan established the criteria for residual radioactive material contamination on UVAR facility surfaces. UVAR facility criteria, also referred to as derived concentration guideline levels (DCGLs), are selected from the table of NRC default screening values. The DCGLgOS for all radionuclides at the activity fractions present is 15,200 dpm/100 cm2 . Based on only beta emissions from Co-60, Mn-54 and Sr-90 being measurable (i.e., 41.4% of the radionuclides present will be detectable), the approach described in Appendix A of the Master Final Status Survey Plan yields a DCGLgToss of 7,390 dpm/100 cm 2 and an DCGLadjlsted gross of 6,320 dpm/100 cm2 . This latter value (6,320 dpm/100 cm 2 ) was used as the applicable total gross P criteria for all facility piping. Removable activity criteria are 10% of this value. This criteria represents a conservative approach for Hot Cell piping, in which the contaminant is more likely to be Cs-137 with a less restrictive guideline value.

4.3.2 Survey Activities Nine facility piping survey units were identified; Table 4-4 is a listing of those survey units.

Based on the facility use history and identification of contaminants of license origin in the remaining impacted piping, the reactor facility piping surfaces were designated Class 1. Storm drains, building and pool footing drains, the CAVALIER Facility drains, and the non-release path portion (west line) of the sanitary sewer system were designated Class 2.

Table 4-4 Facility Piping Survey Units Survey Unit Description Class 23 Sanitary Sewer, Storm Drain, French Drain 2 27 Heat Exchanger Piping 1 28 Reactor Pool Drains 1 29 Reactor Room floor Piping 1 30 Room G022 to Track Pit Drain 1 31 Demineralizer Piping 1 32 Hot Cell Drain 1 33 Reactor Header Drain 1 51 CAVALIER Piping 2 Compliance with decommissioning requirements was demonstrated by comparing the results of FSS with DCGL adjusted grossf 6,320 dpm/100 cm2 . Because of the variability in instrument background levels due to varying levels of naturally occurring radionuclides in soil, rock and building construction materials and piping materials, appropriate reference areas were not applicable. Instead, shielded and unshielded measurements were performed at the same locations and the difference compared to the contamination criteria. Statistical testing of results utilized the Sign Test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors are 0.05 (Type 1 and Type 2).

4-11

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Figure 4-4 Drains from CAVALIER Facility I

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c. ( CH2MLL Figure 4-6 Reactor Room Piping

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\. PIPING REMOVED FROM

.REACTOR ROOm TO HEAT EXCHANGER ROOM SUMP REMAINING PIPING SHOWN SOLID DOTTED LINE SHOWS REMOVED PIPING INDICATES SEPARATED LINE LOCATION I .

4.14

CH2MHILL The number of data points required for the Sign Test was determined to be 20 (refer to Section 4.5 of FSSP Addendum 002). Direct measurements were obtained at equally spaced intervals along the piping to assure a minimum of 20 data points. Although the relative shift (A/a) would be higher and the number of data points required would be lower for Cs-137 as the contaminant, for consistency the minimum number of data points (i.e., 20) remained the same for all piping survey units.

Scans and surface activity measurements of interior surfaces of 6 in (or larger) ID piping were performed using Ludlum Model 43-68 gas proportional detectors; Piping which was not accessible with this detector was surveyed using a Victoreen Model 491-30 GM detector. This latter detector has a 30 mg/cm 2 wall thickness and in an unshielded configuration had an effective field of view of slightly more than 100 cm 2 in a 2 in ID pipe. The overall diameter of the 491-30 detector assembly is approximately 2.9 cm, enabling access to most piping surfaces.

Detector response to Co-60 in piping was determined by cross calibration, using a section of contaminated piping containing a measured activity level. The detection sensitivity of the 491-30 detector and survey technique was estimated as 2,292 dpm/100 cm2 for static measurements and 4,643 dpm/100 cm 2 for scans (refer to Appendix A of FSSP Addendum 002). Although these sensitivities did not satisfy the target objective of 25% of the DCGL, both were less than the DCGL, providing a high level of confidence that any significant residual contamination would be identified.

Interior piping surfaces were scanned by passing the detector through the pipe. The rate of detector movement was approximately 1 detector width/sec for the gas proportional and pancake GM detectors and 2.5 to 3.0 cm/sec for the 491-30 GM detector. Model 2221 scaler/ratemeters used with the detectors was monitored for changes in audible signal and indication of elevated count rate, suggesting possible presence of radioactive contamination, was noted for further investigation. Scan coverage was 100% of the length of Class 1 piping and 25% of the length of Class 2 piping. One-minute static beta counts were performed at the designated systematic locations and at locations of elevated count rate identified by scans.

A Masslinn swab was passed through each pipe section to collect removable activity and scanned for activity using a 125 cm 2 gas proportional detector. Detector sensitivity for this technique was 500 dpm/100 cm 2 . A static one-minute beta measurement was performed at the location of maximum activity, or at a representative location, if elevated activity was not identified.

Following FSS activities, piping access points were covered to prevent recontamination and to allow for future NRC confirmatory actions.

4.3.3 Survey Results Table 4-5 presents the results of the beta scans. Scans utilizing a model 43-68 gas proportional detector ranged from 255 cpm to 960 cpm; those utilizing a model 491-30 GM detector ranged from 12 cpm to 110 cpm. Higher ambient levels were observed in piping embedded in concrete and in piping drains of terra cotta construction. No specific locations of elevated activity were identified.

4-15

CH2MHILL I _

Table 4-5 Beta Scan Results for Facility Piping Survey Survey Counts Per Minute Instrument Number Unit Location Minimum Maximum Set UVA-FS-74 23 Sewer, Storm & French Drains 17 61 15 (a)

UVA-FS-74 23 Sewer, Storm & French Drains 280 500 9 (b)

UVA-FS-72 27 Heat Exchanger Piping 300 960 9 (b)

UVA-FS-75 28 Reactor Pool Drain 20 74 15 (a)

UVA-FS-31 29 Reactor Room Floor Drains 280 475 10 (b)

UVA-FS-12 30 Room G022 to Track Pit Drain 50 110 14 (a)

UVA-FS-12 30 Room G022 to Track Pit Drain 255 280 9 (b)

UVA-FS-73 31 LWST Drain Lines 28 85 14 (a)

UVA-FS-04 32 Hot Cell drain 54 110 14 (a)

UVA-FS-77 33 Reactor header drain 12 61 15 (a)

UVA-FS-76 51 CAVALIER Drains 13 55 15 (a)

Notes:

(a) GM Detector (b) Gas Proportional Detector Total activity measurement results are summarized in Table 4-6. The maximum activity level was 3,152 dpm/100 cm 2 in the Heat Exchanger piping. All systematic measurements were below the DCGLadjstd g of 6,320 dpm/100 cm 2, thus statistical testing is not necessary to demonstrate compliance with the guidelines.

Table 4-6 Facility Piping Beta Activity Measurement Summary Survey Instrument Number of Beta Activity (dpm/100 cm2 )

Unit Type Measurements Minimum Maximum Mean Std Dev 23 491-30 20 -185 446 86 137 43-68 7 163 1491 724 427 27 43-68 19 208 3152 1681 1049 491-30 20 -185 338 55 149 28 43-68 2 252 310 281 41 29 491-30 20 -519 1556 604 478 30 491-30 30 -444 2593 37 846 43-68 7 163 1491 724 427 31 491-30 20 -46 400 132 140 32 491-30 20 -305 712 119 207 43-68 2 22 208 11 132 33 491-30 25 -123 262 73 27 491-30 19 -46 508 83 135 5 43-68 3 230 816 452 318 4-16

CH2MHILL From 19 to 37 data points were obtained for each survey unit as compared to the design number of 20. The maximum standard deviation of direct measurements was 1049 dpm/100 cm 2 in Survey Unit 30. The relative shift for this standard deviation is approximately 3.4. This is greater than the design basis relative shift of 1.4 for 20 data points; sufficient data points were therefore obtained for each survey unit. (he design number of data points includes an additional 20% for potential data losses and quality control purposes; therefore the actual number of data points are adequate to satisfy the design number of 20).

Masslinn swabs were passed through piping sections to determine if loose surface contamination was present. Beta scans were performed on each masslinn swab to determine the highest activity location on the swab. A static measurement was performed on each masslinn swab in the highest activity swab location to determine relationship to the loose surface activity inside the piping. No removable activity above criteria was identified (refer to Table 4-7).

Table 4-7 Masslinn Swabs of Piping unit Survey Location Beta Scans Static Counts (epm/100 cm2)

Number Counts per Minute Lowest Highest Sanitary Sewer, Bulk Access, 23 VA-FS-74 Storm & French Drains 280-500 69 630 27 VA-FS-72 Heat Exchanger Piping 190-270 140 181 28 VA-FS-75 Reactor Pool 2" Drain Line 270-420 128 389 29 VA-FS-31 Reactor Room Floor Piping 280-400 -66 84 Room G022 to the Track Pit 30 VA-FS-12 Drain 255-280 -83 -83 31 TA-FS-73 LWST Drain Lines 200-280 5 5 32 VA-FS-04 Hot Cell Drain Line 225-315 11 106 33 VA-FS-77 Reactor Header Drains 330430 246 374 51 VA-FS-76 CAVALIER Drains 250-400 117 416 During characterization and remediation activities, gamma scans and on-site gamma spectrometry of soil at piping access locations identified potential soil contamination as a consequence of piping breaks beneath the reactor room floor and the reactor pool. Sampling to demonstrate adequate remediation of these piping break locations is described in Section 4.8 of this report.

4.3.4 Conclusion Surveys demonstrate that remaining potentially impacted facility piping does not contain residual contamination in excess of established project guidelines.

4.4 Pond Sediment 4.4.1 Description Storm runoff from the adjacent land areas and overflow from the storm drain on the UVAR site were collected in a small pond, located to the south of the UVAR Building (see Figure 4-7). The pond covers a surface area of about 1,600 m2 , and ranges in depth from approximately 2 to 4 m.

K> The pond bottom is covered with sediments, ranging from a few centimeters to several meters 4-17

CH2MHILL thick. Figure 4-8 is a plot plan of the pond, indicating pertinent features. Some laboratory drains, floor drains, and other sources of non-sanitary wastewater with low potential for radiological or other hazardous constituents also routinely discharged to this pond. Two underground waste tanks serviced the Hot Cell, and two tanks were used for collection of demineralizer regeneration liquids from the reactor. Both of these sets of tanks were originally plumbed to allow the contents to be discharged to the pond, provided the liquid met appropriate release criteria following dilution with the pond water; the demineralizer regeneration tanks were later replumbed so they could be discharged directly into the pond spillway.

During facility operation, there were several intentional and unintentional discharges of low-level contaminated liquids to the pond occurred. Two of these occurred in laboratories M005 and M008, and involved contamination by Tc-99 and Ni-63, respectively. Reactor Pool water discharges to the pond were made in the 1960's. A break in the piping from the demineralizer regeneration tanks resulted in release of low-level contaminated liquids, containing primarily Cs-137 and Co-60, onto the bank of the pond. Because of this history, there was a potential for the sediments to be contaminated with facility-derived radionuclides. Pond sediments analyzed during the 1999 GTS Duratek characterization identified positive levels of Cs-137, Co-60, Eu-152, and Pu-241 in some samples; however, many analyses did not identify activity levels above the method detection limits.

Sampling during the CH2M HILL continuing characterization identified only Cs-137, Ni-63, and Pu-241 as contaminants of license origin (refer to FSSP Addendum 003 in Appendix A).

Based on these findings, Cs-137 will be used as a surrogate for all potential contaminants at a DCGL of 5.9 pCi/g.

In September 2002, the pond was drained and additional characterization was performed. This characterization was designed and implemented such that the data could be used for FSS purposes, if appropriate. Detailed field survey data forms are included in Appendix B (survey number UVAR-0245).

4.4.2 Survey Activities A 10-m reference grid was established over the pond area. This grid is shown in Figure 4-8.

Based on use history and previous characterization findings, the pond sediments were classified as Class 1; the area comprises one survey unit of the pond sediments to the high-water mark and the immediately adjacent north bank, where building drains discharged.

Gamma walkover surface scans were performed using a 2" x 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5-10 cm of the sediment surface and moved from side to side in a serpentine pattern while noting any indication of elevated count rate, which might indicate the presence of radioactive contamination. Surface gamma scans were performed over 100% of the pond sediment surface, the north bank, and the pond discharge stream for approximately 20 in downstream of the spillway. Locations of elevated gamma radiation were noted for further investigation.

Surface (0-15 cm) sediment samples of approximately 500 g were collected at 16 systematic and 18 judgmental sampling locations. In soft sediments, sediment columns were obtained by K.. driving PVC pipe to refusal, capping and removing the pipe. In more resistant sediments, f

4-18

CH2MHILL boreholes were augured through the sediment s to the underlying soil using a 2-in diameter bucket auger. Some locations required a combination of both methods. The resulting samples were obtained from depths of 15 to 45 cm, 45 to 75 cm, and 75 to 105 cm. where thickness of sediment allowed. If a sample could not be obtained from a pre-identified location, one was obtained from the nearest sediment location available; the survey/sampling record noted this situation. A total of 92 samples were obtained. Duplicate samples were collected at 4 locations.

Samples were assigned unique identification numbers and a chain of custody record and analytical request were prepared.

Boreholes were gamma logged at 30 cm intervals from the surface to the bottom of the borehole; where necessary to maintain a borehole open, thin-walled PVC piping was inserted into the borehole as the auger was advanced.

Sample cores were scanned for gamma and beta activity. All samples were analyzed in the on-site laboratory by gamma spectrometry. Based on the results of surface scans, borehole logging, sample core scans, and on-site analyses, samples from 6 locations were sent to an off-site commercial laboratory for gamma spectrometry and analysis for hard-to-detect (10 CFR Part 61) radionuclides. Results of these analyses were used to develop a DCGLuogate for Cs-137. All FSS samples were analyzed for gamma emitters and results compared with the Cs-137 DCGLgate.

4.4.3 Survey Results Gamma scans of pond sediments surveyed from 9,000 to 24,000 cpm, the highest levels were near the spillway and locations where facility drains discharged to the pond. Samples from these locations of elevated surface gamma levels were analyzed to establish the contaminant mix for the sediments. No elevated gamma levels were identified in the creek, downstreamrof the pond discharge.

Analyses of systematic samples, summarized in Table 4-8, identified only Cs-137 as a gamma-emitting radionuclide of potential license origin; the highest concentration was 1.75 pCi/g in the sample near the location where the former underground waste tanks had discharged into the pond. All results were below the Cs-137 surrogate DCGL of 5.9 pCi/g and therefore the sample results demonstrate the established project criteria are satisfied without need for further statistical evaluation. The average Cs-137 level is 0.39 pCi/g with a standard deviation of OA1 pCi/g. These results yield a retrospective relative shift of 13.1; this is higher than the design relative shift, indicating that adequate systematic data points for evaluation were obtained.

As described in the Final Status Survey Plan for Pond Sediments (Addendum 003), analyses of judgmental samples for the purpose of determining the contaminant mix, clearly identified only Cs-137 (maximum 3.46 pCi/g), Ni-63 (maximum 22.9 pCi/g), and Pu-241 (15.8 pCi/g).

4-19

CH2MHILL Figure 4-7 University of Virginia Reactor Facility and Environs ACA Roo

• FENCE Removed during decommissioning 4-20

(  : .C.IHILL Figure 4-8 Plot Plan of Pond, Indicating Reference Grid and Sampling Location.

-- uas +~___--

. ~SITE N o POND SAMPLE ,

N.OE SCALE ,,

sO 6Nsf0 No 0 E 104METER 2>x* '\ t

.40 N. 0 E

£ ,'  %

30;0£s N.* ,( ,', .' 0 -- TCSDSTS

. / ,S J/ ROW/ tJ. }':*

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Property Gr.ld=

O. N. O E6O S N

- 42

CH2MHILL No individual radionuclide concentration in these samples was above its specific default

) screening DCGL and none of the samples yield a sum of fractions value greater than 1 (one).

Table 4-8 Results of Gamma Analyses of Systematic Samples Sample Location Concentr tion (pCi/g)

Sample Cs-137 Co-60 Other Radionuclides 4.9 S 7.3 W 0.25 +/- 0.15 <0.21 None Detected 4.9 S 17.8 W 0.31 i 0.10 <0.11 None Detected

49. S 28.3 W 0.23 +/- 0.16 <0.16 None Detected 4.9 S 38.8 W <0.15 <0.16 None Detected 4.9 S 49.3 W <0.17 <0.17 None Detected 13.9 S 12.5 W 0.26 +/- 0.09 <0.09 None Detected 13.9 S 23.0 W 0.16 +/- 0.09 <0.12 None Detected 13.9 S 33.5 W 0.89 +/- 0.21 <0.16 None Detected 4.1 N 2.0 W 0.89 +/- 0.22 <0.15 None Detected 4.1 N 12.5 W 1.74 +/- 0.36 <0.20 None Detected 4.1 N 23.0 W 0.41 +/- 0.11 <0.15 None Detected 4.1 N 33.5 W <0.15 <0.15 None Detected 4.1 N 45.0 W 0.47 +/- 0.17 <0.14 None Detected 4.1 N 55.5 W <0.22 <0.18 None Detected 4.1 N 66.0 W 0.58 +/- 0.20 <0.17 None Detected 13.1 N 49.3 W 0.18 +/- 0.19 <0.16 None Detected Drainage Creek #1 <0.16 <0.14 None Detected Drainage Creek #2 <0.14 <0.15 None Detected Drainage Creek #3 <0.09 <0.14 None Detected 4.4.4 Conclusion Results of direct survey and sampling demonstrate that contaminants of license origin in WAR pond sediments satisfy the Cs-137 DCGLurrogate. Judgmental characterization samples did not contain significant levels of hard-to-detect radionuclides. On the basis of these results the radiological status of the pond sediments satisfies established project decommissioning criteria.

4.5 Interior Structure Surfaces 4.5.1 Description The three-story UVAR building housed the WVA Research Reactor and the CAVALIER facility, as well as offices for the reactor staff and faculty and students of the Department of Nuclear Engineering, miscellaneous laboratories, and other support facilities for the reactors and Department of Nuclear Engineering.

Figures 4-9 through 4-11 show the three levels of the UVAR facility. The upper level has approximately 620 m2 of floor area. The Reactor Confinement Room (Rm 131), which housed 4-22

CH2MHILL the former UVA Research Reactor, is located on the upper floor (first floor). This room contained the 9.8 m long by 3.7 m wide by 8.2 m deep reactor pool, associated operating equipment and systems, the operating controls, and some research/experimental equipment.

This room is circular and has an elevated (-10 m) ceiling. In addition, the Instrument Shop (Rm 128), Shipping Area (Rm 127), and multiple offices and other support facilities for staff and students are located on this building level.

On the approximately 670 m2 Mezzanine level were located the Demineralizer (Rm M021),

Mechanical Room (Rm M020), HP Laboratory (Rm M019), several partially contaminated laboratories (Rms M005 [Tc-99 contamination] and M008 [Ni-63 contamination]), and multiple offices and other support facilities for staff and students. A crawl space (MCS) is accessed from the stairwell on the Mezzanine level.

The 740 m2 ground floor contained the Heat Exchanger (Rm G024), Rabbit Room (Rm G005),

Beam port/Experimental area (Rm G020), Hot Cell (Rms G025, G026, and G027), Counting Room (Rm G004), Woodworking and Machine Shop (Rm G008), Source Storage (Rms G022, G018, and G007A), the former CAVALIER facility (Rm G007), and miscellaneous support facilities and areas.

The WVAR building is of concrete block construction with brick veneer. Floors are concrete slab.

Internal walls are block and drywall. Most offices, hallways, and small laboratories have a dropped ceiling of acoustical tile, and tile floors.

In preparation for implementing the Final Status Survey, impacted reactor and support systems and components were removed and disposed of as radioactive waste or surveyed and released for use without radiological restrictions. Contaminated facility surfaces and materials were removed or decontaminated. Major actions of this nature are described below.

Figure 4-9 UVA Reactor First Floor Plan View 4-23

CH2MHILL Figure 4-10 UVA Reactor Mezzanine Floor Plan Figure 4-11 UVA Reactor Ground Floor Plan View OmuI

. n.

The neutron beamport area is located on the ground floor level, lower west side of the reactor pool in the biological shield through to the south west wall, approximately 2 meters off the pool floor. Two beam ports were located in this area. The north beam port (referred to as the hot beamport or No.1 beam port) was radioactively activated by the reactor's neutron beam and has been removed along with the activated concrete surrounding the port. The south beamport (referred to as the cold beam port or No. 2 beamport) was not used as extensively and only the first 18-inches of beam tube was removed from the pool wall with a portion of surrounding concrete. Post-remediation surveys identified residual contamination exceeding the release guidelines and additional concrete removal was performed.

Demnineralizer area, rooms M021 and M021A, are located on the mezzanine floor level, northeast corner of the UVAR facility. The room contains painted block and poured concrete walls, painted precast concrete panel ceiling and poured concrete floor. The area contained the reactor pool demineralizer system (resin and charcoal vessels, pumps and associated piping and 4-24

CH2MHILL equipment) and the associated regenerating equipment. All piping, motors, pumps, resin and KHi charcoal vessels and associated materials were removed during D&D operations.

The heat exchanger area, room G024, is located on the ground floor level, northeast corner of the UVAR facility. The room contains painted block and poured concrete walls, painted precast concrete panel ceiling and poured concrete floor. The area contained the reactor pool heat exchanger, primary and secondary side pumps and associated piping and equipment. All piping, motors, pumps, heat exchanger and associated materials were removed during D&D operations. A section of the primary suction and return piping was left in place inside the floor and wall. A valve-gallery sump was located along the middle of the south wall. It contained the valves for the reactor pool drains, heat exchanger primary suction piping drain, and a connection to the reactor room floor drain header for discharge to the pond. The sump is including in this survey; drain piping is addressed separately in Section 4.3.

The hot cell area rooms, G025, G026 and G027, are located on the ground floor level, southeast corner of the UVAR facility. The room contained painted, poured concrete walls, ceiling, and floor. The area contained the hot cell interior, manipulator arms, lead glass window and mechanical hoist. Both manipulator arms and the lead glass window were removed during D&D operations. Cs-137 activity was detected at six discrete locations on the floor and was decontaminated during D&D operations.

Laboratory rooms M008 and M005 are located on the mezzanine floor level, west end of the UVAR facility. The rooms contain painted block and poured concrete walls, a drop panel ceiling and poured concrete floor covered in tile. The areas contain laboratory soap stone counter tops, sinks, and fume hoods with HEPA filtered ventilation to the building exterior. Fume hoods and associated ventilation up to the HEPA filter housings were replaced in both rooms prior to the start of this D&D project. Counter tops were decontaminated and elevated floor tile, cabinet sections, a sink and sink drain, HEPA filters, and parts of the ventilation systems were removed during D&D operations.

The Mezzanine Crawl Space is located on the mezzanine level of the UVAR facility. Access to the area is located in the main stairwell. The room contains concrete block walls, precast concrete panel ceiling with steel support beams and a dirt floor. The dirt floor survey is described in Section 4.8.

The reactor pool structure is approximately four meters wide, ten meters long and nine meters in depth. The pool is separated into two halves by a concrete buttress that housed the reactor pool gate. The pool is oriented slightly off from true North to South. For the purpose of this survey, the pool structure interior surface was referenced in two sections consisting of the north section and its three walls, and the south section and its three walls. Half of each buttress wall was included in its adjoining wall and floor section. There are two beam ports located in the south section, on the west wall, approximately 2 meters off the pool floor. The tangential beam port area (referred to as the hot beam port or #1 beam port) was activated by the reactor's neutron beam and has been removed along with the activated concrete surrounding the port.

The second beam port area (referred to as the cold beam port or #2 beam port) was not used as extensively and only the first 18-inches was removed from the poolside with a portion of surrounding concrete. The pool surface paint was removed by means of a hydrolazer. The knee wall surrounding the pool was cut off flush to the reactor room floor. Five, full-width by 12 J , inches-high, sections were removed from the aluminum gate guide, to provide access that 4-25

CH2MHILL allowed the determination of the condition of the underlying concrete surface. Post-remediation monitoring identified small areas of residual elevated beta activity, requiring further decontamination.

The Source room, G022 is located on the ground floor level, next to the biological shield, on the east side of the UVAR facility. The room contains painted block, ceramic and poured concrete walls, painted precast concrete panel ceiling and poured concrete floor. This room was originally a restroom and contains sanitary sewer piping that was used recently as a discharge point for treated and filtered radioactive liquids. The original rest room was converted to store high-level radioactive materials. All associated materials were removed during D&D operations. Small areas of elevated activity were identified by post-remediation surveys and further decontamination was performed.

The GTS Duratek initial characterization and continuing characterization by the CH2M HILL team showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. Major structural contamination was generally limited to surfaces exposed to or in contact with reactor coolant, reactor neutron fields, and materials containing high levels of activity (e.g., the Hot Cell). Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclide was Co-60 with smaller activities of Cs-137. Remaining structural components did not contain detectable levels of activation products. Ni-63 and Tc-99 contaminants were present on facility surfaces from research projects in labs M008 and M005, respectively.

The Decommissioning Plan established NRC default screening values as the criteria for residual radioactive material contamination on WVAR facility surfaces. Structure surfaces did not have sufficient activity levels to enable a meaningful determination of the facility contaminant mixture - particularly with respect to hard-to-detect radionuclides. Final Status Survey Plan-Addendum 004 (Appendix A), establishes an beta DCGLadjusted go of 6320 dpm/100 cm2 as the basis for evaluating the final radiological status of the structure surfaces. An exception to this guideline is labs M005 and M008, where the contaminants are Ni-63 and Tc-99, respectively; DCGLs for these radionuclides are 1.8 E+6 dpm/100 cm2 for Tc-99 and 1.3 E+6 dpm/100 cm2 for Ni-63. A DCGL of 1.3 E+06 dpm/100 cm 2 will be applicable for both of these facilities.

Guidelines for removable structure contamination are 10% of the value for total surface activity.

This assures a conservative approach for satisfying the NRC dose-based criteria for future facility use.

4.5.2 Survey Activities One meter reference grids were established on Class 1 and Class 2 surfaces and 5 meter grids in Class 3 floor and lower wall surfaces. Upper walls surface (ceiling and overhead) locations were referenced to the grid established for the floor beneath.

A listing of building interior surfaces and their MARSSIM classifications by contamination potential is contained in Table 4-9. Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities were the bases for these classifications. Table 4-9 also lists the survey units for building interior surfaces.

4-26

CH2MHILL Table 4-9 Survey Units for UVAR Building Interior Surfaces Surface Survey Room or Area Surface Class Area Unit

( . 2) No.

131 Reactor Room, West Floor 1 92 1 131 Reactor Room, East Floor 1 92 2 131 Reactor Room Lower Walls 1 103 3 Reactor Pool, North Floor and Walls 1 117 4 Reactor Pool, South Floor and Walls 1 117 5 M005/005A Floor and Lower Walls 1 65 6 M008 Floor and Lower Walls 1 89 7 M019 Floor and Lower Walls 1 107 8 M020 Floor and Lower Walls 1 104 9 M021/MO21A Floor, Walls, and Ceiling 1 158 10 Bio Shield Surfaces Wall 1 54 11 G005 Floor, Walls, and Ceiling 1 99 12 G007/GO07A Floor, Pit and Lower Walls 1 167 13 G018 Floor,-Walls, and Ceiling 1 92 14 G020, West Floor and Lower Walls 1 55 15 G020, Center Floor and Lower Walls 1 67 16 G020, East Floor and Lower Walls 1 120 17 G022 Floor, Walls, and Ceiling 1 48 18 G024 Floor, Walls, and Ceiling 1 105 19 G025/G026/G027 Floor, Walls, and Ceiling 1 146 20 131 Reactor Room Upper Walls and Ceiling 2 691 34 127/128/130 Floor, Walls, and Ceiling 2 176 35 107/124/124A/124B Floor and Lower Walls 2 311 36 M005/005A Upper Walls and Ceiling 2 50 37 M008 Upper Walls and Ceiling 2 56 38 M019 Upper Walls and Ceiling 2 72 39 M020 Upper Walls and Ceiling 2 76 40 M006/M014/M015/M030/

M031 Floor and Lower Walls 2 259 41 MCS (crawl space) Floor, Walls, and Ceiling 2 153 42 G004/GO05A Floor and Lower Walls 2 154 43 G006 Floor and Lower Walls 2 '64 44 G007B/G008/G0O8A/G016

/G017/G019 Floor and Lower Walls 2 362 45 & 45A Stairwell 1 Floor and Lower Walls 2 119 46 Stairwell 2 Floor and Lower Walls 2 184 47 G007/GO07A Upper Walls and Ceiling 3 104 54 G020 Upper Walls and Ceiling 3 437 55 107/124/124A/124B Upper Walls and Ceiling 3 220 56 M006/M014/M015/M030/ Upper Walls and Ceiling 3 192 57 M031 4-27

CH2MHILL Surface Survey Room or Area Surface Class Area Unit

_ ) No.

G004/GO05A Upper Walls and Ceiling 3 107 58 G006 Upper Walls and Ceiling 3 31 59 G007B/G008/GOO8A Upper Walls and Ceiling 3 280 60

/G016/G017/G019 G002 All 3 71 63 Elevator All 3 21 64 Mezzanine Offices All 3 1190 65 First Floor Offices All 3 1934 66 & 66A Due to the variability in background levels resulting from construction materials and radioactive sources stored within the facility, it was not practical to establish meaningful reference areas. Instead, unshielded and shielded measurements were performed at each surface activity data point, and the Sign test was used for evaluating direct measurements, relative to the established criteria. The Null Hypothesis is that the activity levels in the survey unit exceed the criteria. Rejection of the Null Hypothesis is required to demonstrate that the release criteria are satisfied. Decision errors are 0.05 (Type 1 and Type 2).

The number of systematic data points required for the Sign test evaluation was determined to be 14 (refer to Section 4.6 of FSSP Addendum 004). For Class 1 and 2 survey units the data point pattern was triangular with a spacing determined on a case- by- case basis, depending on the survey unit surface area. Data points in Class 3 survey units were selected by the survey supervisor, based on judgment as to the contamination potential.

Gamma surface scans were performed using a 2"X2' NaI detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5 to 10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Gamma scanning coverage was minimum 100 % for Class 1, 25% for Class 2 and 10% for Class 3 surfaces.

Beta scans of surfaces were performed using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 1 cm of the surface while advancing the detector at a rate of approximately on detector width per second. Scan speed was adjusted, as necessary to assure detection sensitivities were less than 50% of the release criteria. Audible response was monitored for indication of an elevated count rate. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Beta scanning coverage was 100% for Class 1 surfaces and a minimum of 25% for Class 2 and 10% for Class 3 surfaces.

Surface beta activity measurements were performed at the systematic and judgmental locations.

One-minute static measurements were conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler. Smears for removable activity were performed at locations of direct activity measurements.

4-28

CH2MHILL 4.5.3 Survey Results Field data forms, containing detailed results of surveys, are included in Appendix B.

Identification numbers for surveys of interior structure surfaces are listed in Table 410.

Table 4-10 Survey Results Forms for UVAR Building Interior Surfaces Survey Room or Area Surface Survey Form Unit Number 1 131 Reactor room, West Floor UVA-FS-20 2 131 Reactor Room, East Floor UVA-FS-19 3 131 Reactor Room Lower Walls UVA-FS-17 4 Reactor Pool North Floor and Walls UVA-FS-14

.5 Reactor Pool South Floor and Walls UVA-FS-15 6 M005 / M005A Floor and Lower Walls. UVA-FS-22 7 M008 Floor and Lower Walls UVA-FS-16 8 M019 Floor and Lower Walls UVA-FS-24 9 M020 Floor and Lower Walls UVA-FS-27 10 M021/MO21A Floor, Walls and Ceiling UVA-FS-21 11 Bio Shield Surfaces Wall UVA-FS-66 12 G005 Floor, Walls and Ceiling UVA-FS-32 13 G007 /G0O7A Floor, Pit and Lower Walls UVA-FS-37 14 G018 Floor, Walls and Ceiling UVA-FS-13 15 G020 West Floor and Lower Walls UVA-FS-47 16 G020 Center Floor and Lower Walls UVA-FS-48 17 G020 East Floor and Lower Walls UVA-FS-65 18 G022 Floor, Walls and Ceiling UVA-FS-10 19 G024 Floor, Walls and Ceiling UVA-FS-23 20 G026 Floor, Walls and Ceiling UVA-FS-01 34 131 Reactor Room Upper Walls and Ceiling UVA-FS-55 35 127/128/130 Floor, Walls and Ceiling UVA-FS-54 36 107/124/124A/124B Floor and Lower Walls UVA-FS-51 37 M005 /M0O5A Upper Walls and Ceiling UVA-FS-25 38 M008 Upper Walls and Ceiling UVA-FS-18 39 M019 Upper Walls and Ceiling UVA-FS-26 40 M020 Upper Walls and Ceiling UVA-FS-28 41 M006/M014/M015/M030/M031 Floor, Walls and Ceiling UVA-FS-29 42 MCS (crawl space) Floor and Lower Walls UVA-FS-11 43 G004 / G005A Floor and Lower Walls WVA-FS-34 44 G006 Floor and Lower Walls WVA-FS-35 G007B/G008/GO08A/G016/G017 45 / G019 Floor and Lower Walls WVA-FS-41 G007B/G008/GOO8A/G016/G017 45A / G019 Floor and Lower Walls WVA-FS-43 46 Stainvell 1 Floor and Lower Walls UVA-FS-63 47 Stairwell 2 Floor and Lower Walls UVA-FS-62 54 G007 / G007A Upper Walls and Ceiling UVA-FS-40 4-29

CH2MHILL SuvySurvey Form SUni Room or Area Surface Surver Nme Unit 55 G020 Upper Walls and Ceiling UVA-FS-67 56 107/124/124A/124B Upper Walls and Ceiling UVA-FS-41 57 M006/M014/M015/M030/M031 Upper Walls and Ceiling VA-FS-30 58 G004/G005A Upper Walls and Ceiling UVA-FS-50 59 G006 Upper Walls and Ceiling UVA-FS-38 G007B/GO08/G0O8A/G016/G017 60 /G019 Upper Walls and Ceiling UVA-FS-46 63 G002 All UVA-FS-60 64 Elevator All UVA-FS-61 65 Mezzanine Offices All UVA-FS-33 66 First Floor Offices All UVA-FS-42 66A First Floor Offices All UVA-FS-53 Results of gamma scans are summarized in Table 4-11. Elevated levels were noted in survey units 13,20,42,43,59,58 and 60. Those in survey unit 13,43,54,58 and 60 are associated with concrete walls on one or more sides of these areas; the concrete appears to be a different mix than other concrete in the building and has a uniform gamma level of about 20,000 cpm; about twice ambient background levels. The elevated level (up to 32,000 cpm) in survey unit 20 is due to a sealed Co-60 source, stored in the Hot Cell area, and the generally elevated level (up to 31,000 cpm) in survey unit 42 (the mezzanine crawl space) is associated with the exposed soil floor in this area. No specific locations of elevated gamma levels, indicating potential residual surface contamination, were noted.

Floor monitor scan results are summarized in Table 4-12. Generally elevated levels were noted in survey unit 17 (G020 East) as a result of a sealed source stored in the Hot Cell, which is immediately adjacent to this survey unit. No specific locations of elevated levels, indicating potential surface contamination were noted by the floor monitor scans.

Beta scan results with the hand-held 43-68 detectors are summarized in Table 4-13. Elevated scan levels were identified in survey units 4,11,13,17,38,42,43,44,54,58,60 and 63.

Investigations of these scans identified stored sealed sources, air conditioner filters, supply ventilation grill work and construction material (concrete block and soil) with naturally occurring radioactive material content as the source of the elevated beta responses.

The highest beta result of 1916 cpm was in Room M008. This room had potential Ni-63 contamination with a DCGL of 1.8E6 dpm/100 cm 2 . None of the maximum levels indicated by these scans were above the DCGLadjustdg of 6,320 dpm/100 cm 2 .

4-30

CH2MHILL Table 4-11 Results of Gamma Scans of UVAR Building Interior Surfaces Unit Gamma Scans (cpm)

Number Room or Area Surface Minimum Maximum 1 Reactor Room West Floor 14000 15800 2 Reactor Pool East Floor 12000 14600 3 Reactor Room Floor 10800 16400 4 Reactor Pool North Floor and Walls 8500 12500 5 Reactor Pool South Floor and Walls 9500 15000 6 M005 & M005A Floor and Lower Walls 8100 13200 7 M008 Floor and Lower Walls 8500 15400 8 M019 Floor and Lower Walls 9200 15900 9 M020 Floor and Lower Walls 8000 14200 10 M021 & M021A Floor, Walls and Ceiling 10200 13000 11 Bioshield Surfaces Wall 5900 12600 12 G005 Floor, Walls and Ceiling 10000 13400 13 G007 & G007A Floor, Pit and Lower Walls 10200 21900 14 G018 Floor, Walls and Ceiling 10500 16100 16 G020, Center Floor and Lower Walls 9500 14300 17 G020, East Floor and Lower Walls 14500 17500 18 G022 Floor, Walls and Ceiling 14500 17500 19 G024 Floor, Walls and Ceiling 1100 15500 20 G026 Floor, Walls and Ceiling 8400 27000 34 131 Reactor Room Upper Walls and Ceiling 9400 15500 35 127/128/130 Floor, Walls and Ceiling 11300 16700 36 107/124/124A/124B Floor and Lower Walls 8000 15600 37 M005 / M005A Upper Walls and Ceiling 7400 14900 38 M008 Upper Walls and Ceiling 10100 13900 39 M019 Upper Walls and Ceiling 10100 14600 40 M020 Upper Walls and Ceiling 9100 16000 41 M006/M014/M015/M030/M031 Floor and Lower Walls 9000 14900 42 MCS (crawl Space) Floor, Walls and Ceiling 18000 31000 43 G004 /G005A Floor and Lower Walls 7900 20400 44 G006 Floor and Lower Walls 9000 14100 45&45A G007B/G008/GO08A/G016 G017/G019 Floor and Lower Walls 11900 16200 46 Stairwell 1 Floor and Lower Walls 11000 16900 47 Stairwell 2 Floor and Lower Walls 11200 15800 54 G007 / G007A Upper Walls and Ceiling 10200 22200 55 G020 East Upper Walls and Ceiling 10200 15000 56 107/124/124A/124B Upper Walls and Ceiling 8000 15600 57 M006/M014/M015/M030/M031 Upper Walls and Ceiling 9500 15100 59 G006 Upper Walls and Ceiling 9900 13100 60 GOO7B/G008/GO08A/G016/G017/GO19 Upper Walls and Ceiling 10000 20400 63 G002 Walls All 9700 11800 64 Elevator All 10100 16200 65 Mezzanine Offices All 9200 14700 66&66A First Floor Office All 5800 13000 431

CH2MHILL Table 4-12 Results of Floor Monitor Beta Scans of UVAR Building Interior Surfaces Unit Scan results (cpm)

Number Room or Area Surface Minimum Maximum 1 131 Reactor Room West Floor 1250 1700 2 131 Reactor Room East Floor 1100 1505 4 Reactor Pool North Floor 930 1230 8 M019 Floor 1215 1550 9 M020 Floor 1150 1570 10 M021/MO21A Floor 1100 1520 13 G007/GO07A Floor 419 1095 14 G018 Floor 950 1225 15 G020 West Floor 990 1310 16 G020 Center Floor 940 1540 17 G020 East Floor 960 2820 19 G024 Floor 1150 1500 20 G025/G026/G027 Floor 680 1100 35 127/128/ 130 Floor 748 1260 36 107/124/124A/124B Floor 645 1058 41 M006/M014/M015/ M030/M031 Floor 775 1567 43 G004 / G005A Floor 800 1200 45 G007B/G008/G0O8A/G0016/G017/G019 Floor 845 1380 Table 4-13 Results of 43-68 Detector Beta Scans of UVAR Building Interior Surfaces Survey Scan Results (cpm)

Unit Room or Area Minimum Maximum 3 131 Reactor Room 250 490 4 Reactor Pool North 260 1000 5 Reactor Pool South 270 600 6 M005/M005A 232 549 7 M008 286 590 8 M019 250 630 9 M020 200 581 10 M021/M021A 250 650 11 Bio Shield Surfaces 180 800 12 G005 95 614 13 G007/GO07A 248 1221 14 G018 200 570 15 G020, West 223 367 16 G020, Center 248 619 17 G020, East 140 900 18 G022 210 680 19 G024 280 620 20 G025/G026/G027 200 525 34 131 Reactor Room 200 550 35 127/128/130 282 609 36 107/124/124A/124B 220 567 37 M005/M005A 322 601 4-32

CH2MHILL Survey Scan Results (cpm)

Unit Minimum Maximum 38 M008 325 1916 39 M019 250 620 40 M020 250 613 41 M006/M014/M0l5/M030/M031 239 538 42- MCS (crawl space) 470 1070 43 G004/GO05A 300 860 44 G006 315 860 45&45A G007B/G008/G008A/G016/G017/G019 246 605 46 Stairwell 1 190 450 47 Stairwell 2 180 530 54 G007/GO07A 255 838 55 G020 180 500 56 107/124/124A/124B 220 608 57 M006/M014/M015/M030/M031 247 629 58 G004/GO05A 296 1022 59 G006 320 598 60 G007B/G008/G008A/G016/G017/G019 284 1016 63 G002 200 800 64 Elevator 190 450 65 Mezzanine Offices 95 290 66&66A First Floor Offices 117 470 Table 4-14 contains a summary of the beta activity measurements in each of the building interior survey units. Activity in Room M008 (survey units 7 and 38) ranged up to 34,982 dpm/100 cm 2 . Ni-63 is the contaminant in that facility and the maximum level measured is less than the DCGL established for Rooms M005 and M008 of 1.3 E+6 dpm/100 cm 2 . Survey unit 20 (Hot Cell) also contained systematic measurements above the adjusted gross DCGL of 6,320 dpm/100 cm 2 . The maximum level was 8,804 dpm/100 cm 2 . Operating history and characterization sampling have identified the contaminant in the Hot Cell as Cs-137 with a DCGL of 28,000 dpm/100 cm2 . All Hot Cell surface measurements are within that guideline. All other surface activity beta measurements were within the adjusted gross DCGL of 6,320 dpm/100 cm2 .

With few exceptions, removable beta contamination was less than the detection sensitivity of 28 dpm/100 cm 2 . Those exceptions were survey units 17 (maximum of 29 dpm/100 cm 2 ), 34 (maximum of 36 dpm/100 cm2 ), 40 (maximum of 29 dpm/100 cm2 ), 54 (maximum of 29 dpm/100 cm2 ), 55 (maximum of 139 dpm/100 cm 2 ) and 63 (maximum of 39 dpm/100 cm2 ). All levels were below the removable criteria of 10% of the activity DCGL's.

Table 4-14 Summary of Beta Activity Measurements for WVAR Building Interior Surfaces Beta Activity (d;-cm2)

Survey Room or Area # of Unit Meas. Minimum Maximum Mean Std Dev.

1 131 Reactor room, West 21 510 1522 987 298 2 131 Reactor Room, East 21 15 1413 638 349 3 131 Reactor Room 21 -167 772 22 232 4133

CH2MHILL Beta Activity (d _ cm2)

Survey Room or Area # of Unit . Meas. Minimum Maximum Mean Std Dev.

4 Reactor Pool North 15 -146 1187 774 320 5 Reactor Pool South 15 -116 1704 854 438 6 M005/M005A 24 -549 2198 651 830 7 M008 24 -244 25495 4524 8402 8 M019 19 -73 1529 700 505 9 M020 16 -116 1442 565 509 10 M021/M021A 16 313 1296 661 217 11 Bio Shield Surfaces 21 -277 1471 531 467 12 G005 16 -204 1689 497 585 13 G007/GO07A 26 -400 2643 705 789 14 G018 21 -146 1056 436 351 15 G020,West 16 -29 1165 477 355

-16 G020, Center 18 138 2876 1004 627 17 G020, East 18 -102 1405 698 365 18 G022 17 -109 2796 1060 905 19 G024 19 182 1879 782 417 20 G025/G026/G027 23 1 -245 8804 1051 1751 34 131 Reactor Room 144 -218 1667 362 321 35 127/128/130 28 -175 1558 796 527 36 107/124/124A/124B 44 -648 881 100 242 37 M005/M005A 18 12 3455 1576 494 38 M008 18 -342 34982 3673 7629 39 M019 17 -116 1930 746 136 40 M020 17 -44 1500 706 513 41 M006/M014/M015/M030/M031 29 -459 1172 220 454 42 MCS (crawl space) 16 735 3262 1654 689 43 G004/GO05A 16 -204 1689 497 585 44 G006 20 175 1646 926 506 45 G007B/G008/G008A/G016/G017/G019 26 -58 2548 657 496 45A G007B/G008/G008A/G016/G017/G019 20 -248 1602 642 503 46 Stairwell 1 20 -111 1231 0 362 47 Stairwell 2 19 -82 1875 536 243 54 G007/G007A 15 277 2629 1471 720 55 G020 17 -334 1179 588 387 56 107/124/124A/124B 44 -648 882 100 247 57 M006/M014/MO15/M030/M031 19 -189 1835 677 587 58 G004/G005A 17 313 3386 1250 903 59 G006 15 160 1580 818 492 60 G007B/G008/G008A/G016/G017/G019 38 182 3551 886 604 63 G002 15 -74 4020 1438 1090 64 Elevator 25 -237 1713 770 625 65 Mezzanine Offices 25 -128 393 87 131 66 First Floor Offices 34 -274 1105 110 269 66A First Floor Offices 15 -245 438 68 145 4-34

CH2MHILL Because all surface activity measurements were below the applicable guideline levels for the K> contaminants present, the established project criteria is satisfied; statistical testing to demonstrate compliance is not necessary.

The minimum relative shift, based on the actual survey data is 4.48 (survey unit 63); this is greater than the design basis relative shift of 3 and the number of data points obtained for each survey unit is therefore adequate for demonstrating compliance.

4.5.4 Conclusion Surveys demonstrate that residual contamination of license origin on interior building surfaces satisfies established project decommissioning criteria.

4.6 Exterior Soils and Paved Areas 4.6.1 Description The UVAR Facility includes UVAR building, a small pond, and asphalt paved road, parking areas, and equipment/materials storage pads, situated on a land area of approximately 9390 m2 (see Figure 4-12). The site terrain generally slopes from north to south. The east and south portions of the site are wooded; the northern portion of the site surface is dominated by rock outcroppings. A low (-1 m high) fence encompasses the site.

During facility operation, several small spills of contaminated liquids occurred in the vicinity of the waste collection systems. Equipment, materials, and wastes with a potential for low-level contamination were stored on surfaces south of the building during facility operations and in connection with the facility remediation. In addition, several liquid discharge points from the building to the pond terminate on the hillside north of the pond.

Waste tanks have been excavated and the pond has been drained; final surveys of soils and sediments in those areas were performed and are described in Sections 4.2 and 4.4. Potentially contaminated wastes have also been removed from storage pads outside the building.

Initial characterization by GTS Duratek and follow-on monitoring during the decommissioning actions has identified Co-60 and Cs-137 as the dominant contaminants from facility operations.

Significant levels of other site-related radionuclides were not identified by this monitoring; adequate activity levels were not available to enable meaningful determination of a radionuclide mixture for the balance of exterior rocks and paved areas.

Decommissioning project criteria are the NRC default screening guidelines. The default screening guideline levels for soil for Cs-137 and Co-60 are 11 pCi/g and 3.8 pCi/g, respectively. Default screening surface activity guidelines is 28,000 dpm/100 cm 2 for Cs-137 and 7,100 dpm/100 cm 2 for Co-60.

To demonstrate compliance with project criteria, final status soil samples were analyzed for specific gamma emitting contaminants of license origin and contaminant concentrations compared with respective screening default guideline levels; sum-of-ratios must satisfy the Unity Rule. The restrictive beta DCGt4 djutedgr.osof 6,320 dpm/100 cm2 , used for other facility surfaces (refer to FSSP Addendum 002 in Appendix A), was the guideline for comparison with direct measurements on paved surfaces.

4-35

CH2MHILL 4.6.2 Survey Activities A 10-meter grid was established over the entire site and referenced to the federal planar coordinate system. Figure 4-13 indicates the reference grid system. Further grid identification (e.g., northing and easting from a southwest origin point) was assigned to each node to facilitate location of sampling/measurement points.

For survey design purposes the planning area of the total site (excluding the pond and building footprint) is 6860 m2 . The site is thus comprised of two survey units; one is the paved surfaces of approximately 2500 m2 , and the other is the soil surfaces of approximately 4360 m2 .

Based on the facility use history and characterization and remediation control monitoring, the exterior soil and paved surfaces sediments were designated Class 3 for FSS planning and implementation purposes.

Two survey units were identified; they are:

Survey Unit Description Area (m2) 50 Paved Surface 2500 52 Soil Area 4360'-

Gamma walkover surface scans were performed using a 2"x2" NaI detector (Ludlurn Model 4-10) coupled within 5-10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Gamma scanning coverage was a minimum of 50% of the soil and paved surfaces.

Beta scans of paved surfaces were performed using a large area (-580 cm 2 ) gas proportional detector (Ludlum Model 43-37) coupled with a Ludlurn Model 2221 ratemeter/scaler and advancing the detector at a rate of approximately one detector width per second. Audible response was monitored for indication of elevate count rate. Results (count rate) were documented on survey maps. Locations of elevated response were noted for further investigation. Beta scanning coverage was a minimum of 50% of the paved surfaces.

The scans identified an area of elevated activity on the asphalt pad outside the Reactor Containment Room. Further investigation indicated an impacted area about 4m x 4m in size.

Remediation was performed and the pad was designated survey unit 62A, reclassified as Class I, and rescanned at 100% coverage.

The number of data points required for the Sign Test was determined to be 15 for soil areas and 14 for paved surfaces (refer to FSSP Addendum 005 in Appendix A). Systematic soil sampling and direct measurement locations are indicated on Figure 4-13. Surface (0 to 15 cm) soil samples of approximately 500 g were collected at the systematic sampling locations for the soil area (17 samples were obtained). Soil samples were analyzed by an off-site commercial laboratory for gamma emitters and a composite was analyzed to confirm the absence of hard-to-detect (10 CFR Part 61) radionuclides.

4-36

CH2MHILL Figure 4-12 University of Virginia Reactor Facility and Environs I,.-- * -

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• Removed during decommissioning 4-37

CH2MHILL Surface activity measurements were performed at the systematic sampling locations for the paved areas. One-minute static measurements were conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler.

Because the radionuclides identified as potential contaminants are not present in background at concentrations, which are significant fractions of the release guidelines, correction of FSS sample data for background levels are not required. Adjustment of direct measurement results for background was through use of unshielded/shielded measurements at each data point.

4.6.3 Survey Results Detailed survey results are provided in field data forms in Appendix B.

Surveys of exterior soil and paved surfaces are:

Survey Unit Survey Number Description 50 UVA-FS45 Paved surfaces other than those in 62A 52 UVA-FS-36 Soil 62A UVA-FS-80 Asphalt Pad Gamma scans of the soil ranged from 11,500 cpm to 60,000 cpm. Elevated gamma scan readings were associated with rock outcroppings along the north parking lot area, the west roadway fill area and the east side of the facility. These outcroppings contain concentrations of potassium, natural thorium, and natural uranium that are higher than typical site surface soils. Increased gamma levels were also noted at the southeast corner of the building where sealed sources were stored inside the Hot Cell. Gamma scans did not identify soil area which might potentially be contaminated with radionuclides of license origin.

Results of gamma analyses of systematic soil samples are summarized in Table 4-15. Cs-137 was identified in the samples at detectable concentrations; the maximum level measured was 1.09 pCi/g. The sum-of-fractions values for gamma emitters only range from 0.028 to 0.131; all are well below the Unity Rule criterion of 1.0. The average and standard deviation are approximately 0.07 and 0.03, respectively. The retrospective relative shift of 31 is much greater than the design value, thus indicating that adequate data points were obtained for this evaluation.

Analyses of a composite of systematic samples for non-gamma emitting radionuclides of potential facility origin are summarized in Table 4-16. Only three radionuclides were identified at concentrations above the method detection sensitivities; they are Pu-238, Pu-239, and Sr-90. These results confirm that significant concentrations of hard-to-detect radionuclides of facility origin are not present in site soils.

Gamma scans of paved surfaces ranged from 10,000 to 42,000 cpm. As with scans of soil surfaces, gamma levels were generally elevated in the vicinity of rock outcroppings and portions of the building where radioactive sources are being stored (e.g., Hot Cell). Gamma scans did not identify any areas of paved surfaces which might be contaminated as a result of licensed operations.

4-38

110 v. a(

so* srrw -w; wvww

. . .SITE N 100 " 0.N t.BOUNDARYFENCE ICONv. O. t SOIL SAMPLE - I

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• POND 20 v. 0 E Of*". to/.... 20 t °N. 40 t _ O0 OOO 3£N. _0N. £° 0 N. To N ° O 00 0 A tM. ;O E° N. 140° 01.0 N.O N 40£ON ?O ON9£O.O.10 r 4I Figure 4-13 Plot of Site, Indicating Reference Grid System, and Measurement/Sampling Locations.

CH2MHILL The paving near the door to the Reactor Contaminant Room was remediated and beta scans were repeated using 125 cm2 detectors. Levels ranged from 320 to 1,100 cpm. Beta scans of other paved surfaces (using large-area floor monitor detectors) ranged from 960 cpm to 4,000 cpm.

Elevated readings were noted in the vicinity of the Hot Cell facility and are attributed to direct radiation from sealed sources stored in this facility. Otherwise, no specific locations of elevated levels were identified by the beta scans.

Table 4-15 Summary of Gamma Spectrometry Analysis for Soil Samples from Site Open Land Areas Grid Coords (a) Radionuclide Concentration (pCi/g)

Sample N E Co-60 Cs-137 Others Sum of Fractions 1 9 30 < 0.12 0.24 +/- 0.15 None Detected 0.054 2 9 48 < 0.18 0.22 +/- 0.24 None Detected 0.067 3 9 66 < 0.11 0.10 +/- 0.06 None Detected 0.038 4 24 21 < 0.08 < 0.08 None Detected 0.028 5 24 57 < 0.24 0.22 +/- 0.15 None Detected 0.083 6 24 111 < 0.12 0.60 +/- 0.15 None Detected 0.087 7 39 84 < 0.14 0.86 +/- 0.20 None Detected 0.015 8 54 111 < 0.10 0.71 +/- 0.25 None Detected 0.091 9 54 93 < 0.09 < 0.09 None Detected 0.032 10 54 21 < 0.14 < 0.09 None Detected 0.045 11 69 12 < 0.08 0.50 +/- 0.13 None Detected 0.067 12 85 39 < 0.11 0.35 +/- 0.09 None Detected 0.061 13 99 48 < 0.12 1.09 +/- 0.21 None Detected 0.131 14 99 66 < 0.18 < 0.12 None Detected 0.065 15 84 93 < 0.12 0.27 +/- 0.14 None Detected 0.057 16 69 84 < 0.09 0.31 +/- 0.13 None Detected 0.052 17 69 101 < 0.16 0.43 +/- 0.13 None Detected 0.081 (a) refer to Figure 4-13 (b) DCGL for Cs-137 = 11 pCi/g DCGL for Co-60 = 3.8 pCi/g "O

CH2MHILL Table 4-16 Concentrations of Radionuclides in Composite of Systematic Soil Samples Radionuclide Concentration DCGL (pCig) (pCi/g)

Am-241 < 0.09 2.1 Fe-55 < 1.46 10,000 H-3 < 6.66 110 1-129 < 0.34 0.5 Ni-63 < 3.62 2100 Pu-238 0.09 +/- 0.07 2.5 Pu-239 0.04 +/- 0.05 2.3 Pu-241 < 3.60 72 Sr-90 0.72 +/- 0.37 1.7 Tc-99 < 0.21 19

• Exterior paved surfaces beta activity measurements are summarized in Table 4-17. Levels ranged from 459 to 4631 dpm/100 cm2 with the maximum activity on the pad outside the Reactor Containment Room roll-up door. All systematic measurements were below the guideline of 6,320 dpm/100 cm2 and statistical testing is therefore not necessary to demonstrate compliance with the guideline. The averages for survey units 50 and 62A are 1,459 and 2,722 dpm/100 cm 2 , respectively. Standard deviations are 714 and 701 dpm/100 cm 2 ; relative shifts for these levels are approximately 10, which are much greater than the design value, thus indicating adequate data points were obtained for evaluation.

Table 4-17 Summary of Beta Surface Activity Measurements on Exterior Paved Surfaces Survey . No. of Beta Activity (dpmnlOO cm 2 )

Unit Meas. Minimum Maximum Mean Std. Dev.

50 Bulk of paved surfaces 17 459 2614 1459 714 62A Pad outside containment 15 1456 4631 2722 701 4.6.4 Conclusion Exterior soil surfaces did not contain contaminants of license origin in excess of project decommissioning guidelines. One small area of pavement, adjacent to a door to the Reactor Containment Room, was identified as having surface Cs-137 contamination. This area was remediated, reclassified, and resurveyed. Final surveys of paved areas indicated beta surface activity is within the conservative DCGLadjuswd go.of 6,320 dpm/100 cm 2 . These results demonstrate that the exterior paved surfaces and soil areas at the UVAR site satisfy the established project decommissioning criteria.

441

CH2MHILL 4.7 Exterior Structure Surfaces 4.7.1 Description Figure 4-14 is a plot plan of the UVAR building. The UVAR building is of concrete block construction with brick veneer. Floors are concrete slab. There is approximately 1190 m2 of roof area, at two elevations; one covers the Reactor Confinement structure - a surface area of approximately 175 m2 , and the other (approximately 1,015 m2 ) covers the remainder of the structure. During operation there was a cooling tower on the roof to the southeast of the Reactor Room; this structure was removed during decommissioning. Roofs are of tar-and-gravel composition. The roofs are essentially clear of obstructions such as items of HVAC equipment.

There are multiple sewer line vents and rainwater drains on the roofs.

Other exterior building surfaces of concern include discharge grills and stacks servicing small laboratory exhaust ventilation systems; some of these, e.g., those from rooms M005 and M008, were known to have at one time been internally contaminated. Doors at exits from areas handling radioactive and/or potentially contaminated materials were also surfaces of interest.

These exterior locations are identified on Figures 4-15 to 4-17.

The Decommissioning Plan established the criteria for residual radioactive material contamination on UVAR facility surfaces. UVAR facility criteria, also referred to as derived concentration guideline levels (DCGLs), are selected from the table of NRC default screening values. Exterior structure surfaces did not have sufficient activity levels to enable a meaningful determination of the facility contaminant mixture - particularly with respect to hard-to-detect radionuclides. Therefore, the contaminant mixture for facility drain systems was assumed for the exterior surfaces (refer to Final Status Survey Plan (FSSP) Addenda 001 and 002, in Appendix A). The principal radionuclides in this mixture are Co-60 and H-3, resulting in an beta DCGLadju! go of 6,320 dpm/lOOcm 2 as the basis for evaluating the final radiological status of the exterior structure surfaces. The guideline for removable surface contamination is 10% of the total surface activity guideline, i.e., 632 dpm/100cm 2 . Use of these guidelines assures a conservative approach for satisfying the NRC dose-based criteria for future facility use.

4.7.2 Survey Activities Reference grids (1 m) were established on the roof surfaces to identify survey locations. Other exterior structure surfaces were not gridded, due to their limited surface areas of *10 m2 ;

instead, survey locations were referenced to pertinent building features.

Two survey units were established for exterior structure surfaces. They were:

Survey Unit Description 48 Reactor Containment Roof 49 Main Building Roof Impacted structure surfaces of < 10 m 2 were not designated as survey units. Instead, from 1 to 4 measurements were obtained from such areas, based on judgment and surface area, for comparison individually with the DCGLs. Such surfaces include exterior surfaces of vents, "2

CH2MHILL stacks, and exit doors, from areas of former radioactive materials use and facilities that required remedial action during this decommissioning project.

The roofs were designated MARSSIM Class 2 surfaces; other exterior surfaces were designated Class 3. Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities were the bases for these classifications.

Gamma surface scans were performed using a 2"X 2" NaI detector (Ludlum Model 4-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5 to 10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate. Scans included the area out to 1 m beyond doors and vents. Results (count rate) were documented on survey area maps. Gamma scanning coverage was a minimum of 25% for Class 2 and 10% for Class 3 surfaces.

Beta scans of roof surfaces and exterior structure surfaces were performed using a Ludlum Model 43-68 gas proportional detector coupled with Ludlum Model 2221 ratemeter/scaler. The detector was maintained within -1 cm of the surface while advancing the detector at a rate of approximately one detector width per second. Scans included the area out to 1 m beyond doors and vents. Scan speed was adjusted, as necessary, to assure detection sensitivities were less than 50% of the release criteria. Audible response was monitored for indication of elevated count rate. Results (count rate) were documented on survey area maps. Beta scanning coverage for roof and wall surfaces was a minimum of 25% for Class 2 and 10% for Class 3 surfaces.

Due to the variability in background levels, resulting from construction materials and radioactive sources stored within the facility, it was not practical to establish meaningful reference areas. Instead, unshielded and shielded measurements were performed at each surface activity data point location, and the Sign Test was used for evaluating direct measurements, relative to the established criteria. The Null Hypothesis is that activity levels in the survey unit exceed the criteria: Rejection of the Null Hypothesis is required to demonstrate that the release criteria are satisfied. Decision errors are 0.05 (Type 1 and Type 2).

The number of systematic data points required for the Sign test evaluation was determined to be 14 for a relative shift of 3 (refer to Section 4.6 of FSSP Addendum 006). To provide a high degree of coverage, data points on the roof surfaces were obtained at a spacing of 2 to 2.5 m, resulting in a number of data points significantly larger than the required number. Random start points were determined for establishing measurement patterns.

Surface beta activity measurements were performed at the systematic and judgmental locations.

One-minute static measurements were conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler. Both shielded and unshielded measurements were performed at each location.

Smears for removable activity were performed at locations of direct activity measurements.

4-43

N 0 Motors

• .aConfiement Room Roof Main Building Roof Figure 4-14 UVA Reactor Floor Plan View Indicating Roof Areas.

I I-

8 Reactr Room Aaso Door I0 lot 10 loI3 It 12 7 l*11 "l 14sK~>

Figures 4-15 UVA Reactor First Floor Plan View Indicating Survey Locations on Exterior Building Surfaces.

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HoodVVett Hood Ventb5 MOlDA M017: MOIB j__ F-M019 Figure 4-16 UVA Mezzanine Floor Plan View Indicating Survey Locations on Exterior Building Surfaces.

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z. __ _ __ _ ., . . oo Exit~~xi Doroor20C Figure 4-17 WVA Reactor Ground Floor Plan View Indicating Survey Locations on Exterior Building Surfaces.

I-

CH2MHILL 4.7.3 Survey Results Detailed survey results are provided in field data forms in Appendix B. Surveys of exterior structure are:

Description Survey Number Survey Unit Reactor Containment Roof UVA-FS-52 48 Main Building Roof UVA-FS-53 49 Exterior Doors and Vents TVA-FS-59 N/A Gamma scans of the Reactor Containment Roof ranged from 8,700 cpm to 10,500 cpm. Gamma Scans of the Main Building Roof ranged from 9,500 cpm to 15,000 cpm. No locations of elevated activity were noted on the roof surfaces. Gamma scans of exterior doors and exhaust vents ranged from 8,700 cpm to 88,000 cpm. Elevated gamma levels were noted in the vicinity of the Hot Cell area in the southeast corner of the building. These elevated levels were due to Co-60 and PuBe sealed sources stored in the Hot Cell. No other elevated locations were noted.

Beta scans of the Reactor Containment Roof ranged from 250 cpm to 800 cpm, and beta scans of the Main Building Roof area ranged from 250 cpm to 675 cpm. No locations of elevated beta activity were noted on the roof surfaces. Beta scans of the building exterior doors and vents ranged from 160 cpm to 14,000 cpm. As with the gamma scans, beta scans were elevated in the vicinity of the Hot Cell, due to the sources stored inside this area. No other locations of elevated beta activity were noted on building exterior surfaces.

Surface beta activity measurements are summarized in Table 4-18. The maximum measurement was 3494 dpm/lOOcm 2 . All measurements were below the guideline value of 6320 dpm/lOOcm 2 and thus demonstrated the established criteria are satisfied without need of statistical testing.

Reassessment of the relative shift, using actual survey data yielded values greater than 6. The data point requirements, based on a relative shift of 3 were thus adequate.

Table 4-18 Exterior Structure Surfaces Beta Activity Summary Survey Description Number of Beta Activit (dpm/100 cm2)

Unit Measurements Minimum Maximum Mean Std Dev 48 Reactor Containment Roof 57 200 3353 1905 695 49 Main Building Roof 207 -7 3494 1337 588 N/A Exterior Doors and Vents 19 -96 2018 427 678 All smears contained less than the detectable level of removable beta activity (<28dpm/lOOcm 2 ).

4.7.4 Conclusion Survey results demonstrate that exterior building surfaces do not contain radioactive material contamination of license origin in excess of established project guidelines.

448

CH2MHILL 4.8 Special Soils Areas 4.8.1 Description Several soils areas inside the UVAR building have had a potential for radioactive contamination, based on the operating history of the facility. One of these is a small crawl space adjacent to the Reactor Confinement Room. This space, located between the first and Mezzanine levels, is accessed from the stairwell between these two floors. The crawl space is of masonry construction with a dirt (soil) floor, covering an area of approximately 50 m2 . This crawl space was used for storage of equipment, materials, and supplies, including some radioactive sources and potentially contaminated components and miscellaneous materials. Characterization surveys of this crawl space identified slightly elevated direct radiation levels, due to the masonry construction and the presence of elevated radon progeny, which is believed to originate from naturally occurring radionuclides in the soil floor and which accumulate in this unventilated space.

The soil surrounding the reactor pool is another area of potential soil contamination. The reactor pool structure is approximately 10 m x 4 m and extends approximately 8 m below the reactor room floor level. The reactor pool is located inside the circular Reactor Confinement structure, which has a diameter of approximately 16 m. The space between the outer pool walls and the Confinement structure contains soil fill. Since the base of the Confinement structure does not incorporate a floor, the pool therefore is underlain with soil and bedrock. During reactor operations, losses of pool water were a common occurrence. Specific locations of any pool leakage have not been identified; however, such leakage potentially could have resulted in contamination of soils around and beneath the pool. Breaks in piping beneath the Reactor Room floor were identified during facility remediation. Leakage of contaminated liquids from floor, sink, and pool overflow drains could have contaminated surface soils in the vicinity of these breaks. Characterization of surface and subsurface soils beneath the Reactor Room floor identified small, localized areas of contaminated surface soil; these areas were remediated.

Characterization of the fill around the pool and in the soil, bedrock, and groundwater beneath the pool did not identify contamination of these media requiring remediation.

Figures 4-18 to 4-20 indicate the locations of these soil areas inside the UVAR building.

In preparation for implementing the Final Status Surveys, materials and equipment were removed from the crawl space and piping and other potentially contaminated items and components were removed from the fill area beneath the Reactor Room floor and around the reactor pool. Soils in the vicinity of piping leaks and pool water leaks, identified by scoping and characterization efforts as containing elevated activity were excavated. Areas of particular note that required soil excavation were around a section of piping near the Demineralization Room (MO21A) wall on the east side of the Reactor Containment structure (Figure 4-21) and at piping leaks beneath the Reactor Room floor at core locations "M" and 'B" (Figure 4-22).

449

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CH2MHILL Demineralizer Room Wall Core Access Wall Opening I 3 wAeLFS-M U4LPS-92

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The initial characterization by GTS Duratek showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. Major structural contamination was generally limited to surfaces exposed to or in contact with reactor coolant, reactor neutron fields, and materials containing high levels of activity (e.g.,

the Hot Cell). Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclides were Co-60 and Cs-137; smaller activities of some other gamma emitters and hard-to-detect radionuclides were identified in samples from certain facility locations and media.

Continuing characterization by CH2M HILL identified small localized areas of elevated direct gamma radiation on the surface of the soil fill beneath the Reactor Room floor and the south end of the reactor pool. This contamination was at locations of piping breaks and leaks. Remediation of these areas of elevated activity appeared to eliminate contamination of the fill soil. Characterization did not identify any contamination of the Mezzanine crawl space by radionuclides of license origin.

Addendum 007 to the Final Status Survey Plan (available in Appendix A) describes the contaminants and guidelines for these soil areas. The relatively low activity levels in the surface soil beneath the Reactor Room and pool floor did not enable a meaningful determination of the complete mix; particularly of hard-to-detect radionuclides. Therefore, because of the dominance of Co-60 in the surface samples and because the source of the contamination was liquids from the reactor facility, the same contaminant mixture is 4-53

CH2MHILL assumed for the surface of the fill soil as used for the waste tank remediation and reactor facility piping FSSP. A Co-60 DCGLs.rrogate of 3.4 pCi/g is thus used for these soils.

Compliance with decommissioning requirements was demonstrated by comparing the results of final status survey sample analyses with the Co-60 DCGLuogate of 3.4 pCi/g and by furthermore demonstrating that hard-to-detect radionuclides are not present in significant concentrations. Subsurface soils surrounding and beneath the reactor pool were evaluated over 1-meter thick intervals. Because the radionuclides identified as potential contaminants are not present in background samples at concentrations, which are significant fractions of the release guidelines, correction of FSS sample data for background levels was not be required.

4.8.2 Survey Activities A 1-meter interval grid system was established on surfaces to provide a means for referencing measurement and sampling locations.

Based on facility operating history, characterization survey results, and findings during remediation, the crawl space was designated MARSSIM Class 2 contamination potential, and the soil areas around and beneath the reactor pool were designated Class 1 for survey planning purposes.

For final evaluation, interior soils are divided into the following five groupings:

1) Mezzanine crawl space

2) Surface soil at piping excavations beneath the reactor room floor

3) Surface soil at demineralizer excayation

4) Surface/subsurface fill around pool

5) Surface/subsurface fill beneath pool Because of their small surface areas and location (inside the building), and inclusion of subsurface material, these soils were not evaluated as survey units. Although the FSS, differed slightly from traditional MARSSIM approaches, the survey frequency, survey methods, and data evaluation were consistent with the intent of MARSSIM (refer to Addendum 007 in Appendix A for further information).

Gamma scans of accessible surfaces were performed using a 2" X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5 to 10 cm of the surface and moved across the surface while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Gamma scanning coverage was 4-

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CH2MHILL 100% of accessible surface soils surfaces. Gamma logs of boreholes for subsurface sampling were conducted using the NaI detector (Ludlum Model 44-10 or Ludlum Model 44-2) coupled with a Ludlum Model 2221 ratemeter/scaler. Gamma levels (c/1 min) at 1-m depth intervals were obtained throughout the length of the borehole. Audible response was monitored during detector movement for indication of elevated count rate, which might indicate the presence of radioactive contamination. Gamma scan and logging results were documented on survey maps. Locations of elevated response were noted for further investigation.

Sampling/measurements were performed at uniformly spaced intervals throughout the soil areas or volumes of interest. This spacing between data points was determined by the surface area or volume.

For the small Mezzanine crawl space area, samples of surface soil were obtained on the same pattern and at the same intervals (about 3.5 m) as the surface activity measurement data points on the non-soil surfaces of this area. Thus 5 samples were obtained from this soils area. Five samples were also collected from the small excavation adjacent to the Demineralizer Room. Figures 4-23 and 4-21, respectively indicate these sampling locations.

Due to their larger area/volume, the number of samples from the remaining areas of reactor pool fill were based on the MARSSIM guidance for Sign Test evaluation, which yielded a minimum sample requirement of 15.

Seventeen samples were obtained from surface soil at the locations beneath the Reactor Room floor of the "M" and "B" cores, where remediation was performed. Figure 4-22 indicates these sampling locations.

Soil beneath the reactor pool was sampled at 12 locations, spaced to provide coverage across the pool floor area and to address locations where characterization identified contamination and/or remediation was performed. At each location, surface soil beneath the pool floor was sampled. At 4 locations, boreholes were drilled to the undisturbed soil surface or refusal, whichever was encountered first, using a 2" hand-auger; depth of boreholes ranged from about 4 depths in each borehole. This resulted in a total of 24 samples from this soil region. Because of groundwater infiltration, gamma logging of these boreholes could not be performed. Samples representing the upper 1 m soil layer were also collected at three locations at the extreme south end of the pool, where water leakage was suspected to allow for averaging contamination over a 1-m soil thickness, as described in NUREG-1727 (Ref. 11). Sampling locations are indicated on Figures 4-24 and 4-25.

For the remainder of the fill around the sides of the reactor pool, borings were performed at 16 access locations, spaced throughout the reactor room to provide relatively uniform coverage, while also addressing locations that required remediation. Borings and sampling were performed with a 2" hand-auger. Depths of borings ranged from 1 m to 6.7 m. A total of 14 surface and 16 subsurface samples were obtained from the fill around the reactor pool.

Gamma scans and static counts inside the bore holes were performed by using a 1" Nal detector to identify potential locations of residual concentration. Figures 4-26 and 4-27 indicate the locations of surface and subsurface sampling, respectively.

4-56

CH2MHILL Soil samples of approximately 500 g each were collected at sampling locations. Surface samples were obtained from the upper 15 cm soil layer, using trowels or bucket augers.

Subsurface samples were obtained using bucket augers, split spoon samplers, or other methods consistent with the drilling technique and equipment, and homogenized over a depth interval of 1 meter.

All individual samples were analyzed by gamma spectrometry. Composite samples were prepared from Mezzanine Crawl Space surface soil, surface soil beneath the Reactor Room Floor, subsurface soil from beneath the Reactor Room Floor, and all soil beneath the Reactor Pool. These composite samples were analyzed to confirm the absence of significant levels of hard-to-detect radionuclides.

4.8.3 Survey Results Detailed survey results are provided in field data forms in Appendix B. Surveys of interior soil are:

Survey Number Description UVA-FS-039 Soil beneath reactor room floor UVA-FS-079 Soil beneath pool LJVA-FS-057 Demineralization Excavation UVA-FS-056 Sub-floor cores "M" and "B" UVA-FS-042 & 011 Mezzanine Crawl Space 4-57

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CH2MHILL Gross gamma levels, noted by gamma scans, were generally higher for all interior soil surfaces than for exterior soil areas and structure surfaces. Ranges of these levels are summarized in Table 4-19. Such gamma levels are the result of elevated concentrations of naturally occurring K-40 and uranium and thorium decay series radionuclides in site soils and the enhanced geometry of the detector with soils and structural concrete. The gamma scans did not identify specific locations of significantly elevated levels, which would be suggestive of residual contamination. Detection sensitivity for a 2" x 2" Nal is adequate to identify areas of surface-contaminated soil at the Co-60 and Cs-137 default DCGLs (i.e., 3.8 pCi/g and 11 pCi/g, respectively) in the presence of an ambient level of 30,000 to 40,000 cpm.

Table 4-19 Range of Gamma Scan Levels on Surfaces of Interior Soil Gross Gamma Levels (c/m)

Area Minimum Maximum Pool fill soil 14100 30000 Soil beneath pool 13456 40495 Demnineralizer excavation 19637 26525 Soil at Cores "M" and "B" 19749 30397 Mezzanine Crawl Space 27000 31000 Results of gamma analyses of Mezzanine Crawl Space soil samples are summarized in Table 4-20. These samples contained maximum Co-60 and Cs-137 concentrations of 0.34 Ci/g and 0.18 pCi/g, respectively, but no other detectable radionuclides of license origin.

Comparing these results with the Co-60 DCG4srgIate of 3.4 pCi/g indicates that residual activity is well below the guidance level.

Gamma analysis results for the Demineralizer Room wall excavation contained a maximum Co-60 level of 0.39 pCi/g and no other detectable concentrations of gamma emitters (see Table 4-21). All samples thus satisfied the Co-60 DCGLsur1ogat value of 3.4 pCi/g.

Table 4-20 Results of Gamma Spectrometry of Mezzanine Crawl Space Soil Samples Samplea Depth Radionuclide Concentration (pCi/g)

Location (cm) Co-60 Cs-137 Other 13 0-15 <0.13 <0.13 None Detected 14 0-15 0.34 +/- 0.13 <0.19 None Detected 15 0-15 <0.09 <0.08 None Detected 16 0-15 <0.19 <0.22 None Detected 17 0-15 <0.11 0.18 +/- 0.09 None Detected aRefer to Figure 4-23.

4Z3

CH2MHILL Table 4-21 Results of Gamma Spectrometry of Demineralizer Wall Excavation Soil Samples Samplea Depth Radionuclide Concentration (pCig)

Location (cm) Co-60 Cs-137 Other 1 0-15 <0.13 <0.12 None Detected 2 0-15 <0.23 <0.19 None Detected 3 0-15 0.39 +/- 0.12 <0.12 None Detected 4 0-15 <0.11 <0.12 None Detected 5 0-15 <0.20 <0.16 None Detected aRefer to Figure 4-21.

Table 4-22 summarizes gamma spectrometry of surface soil samples from the remediated areas beneath the Reactor Room floor, near "M" and "B" cores. The maximum Co-60 level was 2.0 pCi/g in the sample from location 17. The maximum Cs-137 level was 7.10 pCi/g at' sample location 3. No other gamma emitting radionuclides of license origin were detected.

All samples contained less than the Co-60 DCGLumnogate of 3.4 pCi/g.

Gamma spectrometry results for samples from beneath the reactor pool are summarized in Table 4-23. Samples from the 0-15 cm depths at locations 2 and 9 contained 4.23 pCi/g of Co-60 and 4.99 pCi/g of Co-60, respectively. Both of these concentrations exceed the Co-60 DCGLsurrogate of 3.4 pCi/g. However, when averaged over an interval of 1 m, the resulting Co-60 concentrations are 1.09 pCi/g and 0.32 pCi/g, respectively. Both of these values and all other Co-60 concentrations in samples from below the pool are below the 3.4 pCi/g DCGLtnrogate value. The maximum Cs-137 concentration in these samples was 0.38 pCi/g; Co-57 was identified in 3 samples at a maximum level of 0.15 pCi/g. No other gamma emitting radionuclides of license origin were identified.

Concentrations of gamma emitters in soil samples from the surface of fill beneath the Reactor Room floor are summarized in Table 4-24. The maximum Co-60 concentration in these samples was 0.40 pCi/g; all samples were therefore well below the Co-60 surrogate DCGL of 3.4 pCi/g. No detectable levels of Cs-137 or other radionuclides of license origin were identified.

Table 4-22 Results of Gamma Spectrometry of Soil Samples From "M" and "B" Areas Samplea Depth Radionuclide Concentration (pCi/g)

Location (cm)

Co-60 Cs-137 Other 1 0-15 <0.14 0.25 +/- 0.13 None Detected 2 0-15 <0.14 0.19 +/- 0.10 None Detected 3 0-15 <0.11 7.10 +/- 1.05 None Detected 4 0-15 <0.22 0.30 +/- 0.15 None Detected 5 0-15 <0.13 <0.14 None Detected 6 0-15 <0.21 <0.21 None Detected 7 0-15 <0.10 0.24 +/- 0.10 None Detected 8 0-15 <0.14 <0.12 None Detected 4-64

CH2MHILL 9 0-15 <0.12 <0.12 None Detected 10 0-15 0.19 +/- 0.15 <0.15 None Detected 11 0-15 <0.13 <0.11 None Detected 12 0-15 <0.14 <0.14 None Detected 13 0-15 <0.23 1.94 +/- 0.35 None Detected 14 0-15 <0.11 0.11 +/- 0.05 None Detected 15 0-15 0.29 +/- 0.10 <0.16 None Detected 16 0-15 0.65 +/- 0.11 <0.11 None Detected 17 0-15 2.00 +/- 0.26 <0.24 None Detected aRefer to Figure 4-22.

Table 4-23 Results of Gamma Spectrometry of Soil Samples from Beneath Reactor Pool Samplea Depth Radionuclide Concentration (pCi/g)

Location (cm) Co-60 Cs-137 Other 1 0-15 <0.18 0.24 +/- 0.14 Co-57 (0.10 +/- 0.08) 2 0-15 4.23 +/- 0.37 <0.23 None Detected 2 0-100 1.09 +/- 0.21 <0.21 None Detected 3 0-15 0.18 +/- 0.09 0.38 +/- 0.11 Co-57 (0.14 +/- 0.08) 4 0-15 <0.19 <0.14 None Detected 5 0-15 <0.20 <0.15 None Detected 6 0-15 <0.17 <0.12 None Detected 7 0-15 0.97 +/- 0.17 <0.18 None Detected 7 0-100 0.67 +/- 0.16 <0.18 None Detected 8 0-15 <0.22 <0.18 None Detected 8 45-75 <0.19 <0.23 None Detected 8 135-165 <0.21 <0.22 None Detected 8 225-255 <0.21 <0.26 None Detected 9 0-15 4.99 +/- 0.43 <0.22 None Detected 9 75-105 0.24 +/- 0.13 <0.21 None Detected 9 165-195 <0.23 <0.17 None Detected 9 285-315 <0.25 <0.20 None Detected 9 0-100 0.32 +/- 0.13 <0.18 Co-57 (0.15 +/- 0.10) 10 0-15 <0.23 <0.25 None Detected 10 75-105 <0.22 <0.22 None Detected 10 165-195 <0.17 <0.15 None Detected 10 225-255 <0.21 <0.18 None Detected 11 0-15 <0.19 <0.15 None Detected 11 15-45 <0.25 <0.22 None Detected 11 45-75 <0.21 <0.17 None Detected 11 75-105 <0.22 <0.21 None Detected 12 0-15 0.75 +/- 0.17 <0.21 None Detected a Refer to Figures 4-24 and 4-25.

4-65

CH2MHILL Concentrations of gamma emitters in soil samples from the surface of fill beneath the I

Reactor Room floor are summarized in Table 4-24. The maximum Co-60 concentration in these samples was 0.40 pCi/g; all samples were therefore well below the Co-60 DCGLsurrogate of 3.4 pCi/g. No detectable levels of Cs-137 or other radionuclides of license origin were identified.

Table 4-24 Results of Gamma Spectrometry of Surface Soil Samples From Beneath Reactor Room Floor.

Sample' Depth Radionuclide Concentration (pCi/g)

Location (cm) Co-60 Cs-137 Other D 0-15 <0.11 <0.10 None Detected E 0-15 <0.14 <0.11 None Detected G 0-15 <0.16 <0.14 None Detected H 0-15 <0.08 <0.07 None Detected I 0-15 <0.10 <0.11 None Detected IB 0-15 <0.09 <0.08 None Detected IC 0-15 <0.09 <0.09 None Detected 1 0-15 <0.12 <0.10 None Detected L 0-15 <0.11 <0.14 None Detected N 0-15 <0.20 <0.14 None Detected P 0-15 <0.10 <0.07 None Detected R 0-15 <0.08 <0.07 None Detected T 0-15 0.40 +/- 0.11 <0.13 None Detected V 0-15 <0.09 <0.12 None Detected a Refer to Figure 4-25.

Results of gamnma spectrometry analyses on samples representing 1-m intervals in fill beneath the Reactor Room floor are summarized in Table 4-25. No radionuclides of license origin were detected in these samples; maximum Co-60 was <0.27 pCi/g and maximum Cs-137 was <0.20 pCi/g. These results are well below the Co-60 surrogate DCGL of 3.4 pCi/g.

Table 4-25 Results of Gamma Spectrometry of Soil Samples Representing 1-m Fill Intervals Beneath Reactor Room Floor Samplea Depth Radionuclide Concentration (pCi/g)

Location (cm) Co-60 Cs-137 Other B 460-560 <0.27 <0.18 None Detected D 30-130 <0.17 <0.15 None Detected E 0-100 <0.13 <0.11 None Detected F 190-290 <0.23 <0.18 None Detected G 210-310 <0.16 <0.16 None Detected H 250-350 <0.23 <0.18 None Detected I 0-100 <0.24 <0.19 None Detected lB 570-670 <0.23 <0.18 None Detected IC 130-230 <0.16 <0.16 None Detected 1 290-390 <0.13 <0.12 None Detected L 310-410 <0.17 <0.15 None Detected 4-66

CH2MHILL Samplea Depth Radionuclide Concentration (pCi/g)

Location (cm) Co-60 Cs-137 Other M 30-130 <0.18 <0.14 None Detected N 380-480 <0.18 <0.15 None Detected P 300-400 <0.24 <0.20 None Detected R 110-210 <0.12 <0.12 None Detected T 0-100 <0.13 <0.11 None Detected a Refer to Figure 4-27.

Table 4-26 summarizes the results of analyses for hard-to-detect radionuclides, performed on composite samples from the interior soil areas. Samples of surface and subsurface soil from beneath the Reactor Room floor did not contain detectable concentrations of radionuclides of license origin. The sample of soil from beneath the reactor pool had detectable concentrations of Pu-238 and Pu-241 at 0.08 pCi/g and 2.43 pCi/g, respectively.

The composite of Mezzanine Crawl Space samples contained detectable concentrations of H-3 and Tc-99 at 7.48 pCi/g and 0.22 pCi/g, respectively. These positive concentrations are all well below their respective default screening DCGL values.

Table 4-26 Analyses of Composite Samples from Interior Soil Areas Concentration (pCi/)

Radionuclide Sample A* Sample B* Sample C* Sample D*

Am-241 <0.13 <0.05 <0.13 <0.06 Fe-55 <1.03 <1.47 <1.36 <1.04 H-3 <5.48 <4.99 <7.38 7.48 +/- 2.66 1-129 <0.23 <0.29 <0.24 <0.22 Ni-63 <9.61 <3.37 <4.95 <10.6 Pu-238 <0.13 <0.01 0.08 +/- 0.05 <0.12 Pu-239 <0.12 <0.02 <0.06 <0.10 Pu-241 <7.96 <1.14 2.43 +/- 1.42 <5.42 Sr-90 <1.23 <0.57 <0.87 <0.58 Tc-99 <0.30 <0.20 <0.31 0.22+/- 0.13

• Sample A: Surface soil beneath Reactor Room Floor (14 individual locations)

Sample B: Subsurface soil beneath Reactor Room Floor (16 individual locations)

Sample C: Soil beneath reactor pool (20 individual locations)

Sample D: Soil from Mezzanine Crawl Space ( 5 individual locations) 4.8.4 Conclusion Several individual 0-15 cm samples from areas of piping leaks beneath the reactor pool contained Co-60 concentrations slightly above the DCGLsu .af. However, when averaged over 1-in soil intervals, the concentrations were well within the DCGL value. Other interior soils areas did not contain levels of Co-60 above the DCGLs,.gate value. Composite samples confirmed the absence of significant hard-to-detect radionuclides of license origin. Based on these findings, the interior soils satisfy established project decommissioning criteria.

4-67

CH2MHILL 4.9 Facility Ventilation 4.9.1 Description Several systems provided ventilation for facilities having a potential for airborne radioactivity. The systems/components remaining after decommissioning, which were potentially radiologically impacted, are:

• Exhaust for fume hood in Room M005.

• Exhaust for fume hood in Room M008.

• Blower for fume hoods (2) in Room M019.

• Exhaust for source storage Room G022.

• Hot Cell exhaust.

• Reactor Room recirculation and exhaust.

Because the exhaust ventilation systems in laboratories M005 and M008 had become contaminated with Tc-99 and Ni-63, respectively, during research projects in those facilities, new fume hoods and ductwork between the hoods and the exhaust fans were installed in these rooms a short time before the reactor decommissioning activities began. The blower assembly was removed from Room M-008 during D&D operations; the original squirrel-cage blower for the M005 exhaust system remains, along with the ductwork downstream of both fan units. During facility operation, these systems exhausted through the outside laboratory walls and into vertical ducts on the building exterior; the vertical ducts discharged above the roof level through rain-cap covered stacks. The remaining exhaust ventilation systems in laboratories M005 and M008 are potentially impacted and were included in this survey. Because the new hoods and ductwork were never used for contaminated operations, the potential for contamination of those surfaces is considered negligible.

Fume hoods in Room M019 became contaminated with Tc-99. Hood baffles were removed and cleaned. Ductwork from the rear of the hood was removed up to and including the HEPA filter and housing. A short section of ductwork, which connected the exhausts from this facility to the former exhaust ventilation from the Hot Cell, remains. The Hot Cell exhaust duct from inside the Hot Cell to the blower in Room M020, remains; the HEPA filter box has been removed from the point where the ductwork joins the blower. The combined Hot Cell and M019 fume hood exhausts pass through a duct inside the Reactor Stack and discharge into the plenum of the Reactor Room exhaust fan.

Reactor Room air is exhausted through a duct near the ceiling of the Reactor Room into the suction plenum of the Reactor Room exhaust fan at the top of the Reactor Stack. At this location the duct from the Hot Cell/M019 hood and the Reactor Room are combined and exhausted through the plenum vertically on the roof of the Reactor Room.

48

CH2MHILL There was a small exhaust from the source storage room (Room G022). The blower has been removed, but the ductwork which discharges at the Mezzanine level on the east end of the building remains.

Reactor Room air is conditioned by a recirculating system. This system draws make-up fresh air through the Reactor Room doorways and combines the fresh air with room air.

This air stream is heated as needed and then discharged back into the Reactor Room through 12 vents, located at the base of the Reactor Room wall.

Figures 4-28 to 4-30 indicate the locations of the remaining potentially impacted ventilation system surfaces. Except for portions of the recirculating air vents, which are encased in concrete, there is access to interior surfaces of components of these ventilation systems to conduct surface activity scans and measurements. That access was adequate to demonstrate that radiological conditions satisfy decommissioning criteria.

The GTS Duratek initial characterization and continuing characterization by the CH2M

-HILL team showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. The overall predominate radionuclides were Co-60 and Cs-137. Contamination of Hot Cell surfaces is primarily Cs-137. Ni-63 was present on facility surfaces from research projects in lab M008, and Tc-99 was present on facility surfaces in lab M005, respectively. Sufficient activity levels were not present on facility surfaces to enable a meaningful determination of the contaminant mixture -

particularly for hard-to-detect radionuclides. Therefore an adjusted gross beta DCGL of 6320 dpm/100 cm2 was developed for surfaces, based on the contaminant mix resulting from reactor effluents. With exception of the systems in labs M008 and M005, this adjusted gross DCGL is the basis for evaluating the final radiological status of ventilation system surfaces. For the systems in M008 and M005, the default screening value of 1.3E+6 dpm/100 cm2 for Ni-63 is applicable (this is more restrictive than the value of 1.3E+6 dpm/100 cm 2 for Tc-99 and is used for simplicity in total surface activity.

4.9.2 Survey Activities The following 5 survey units were established for remaining ventilation systems:

Survey Unit Description 24 Reactor Stack 25 Hot Cell Exhaust Ventilation/M019 Blower 26 Reactor Room Recirculation Ventilation 61 Rooms M005 & M008 Exhaust Ventilation 61A Room G022 Exhaust Ventilation All ventilation system surfaces were classified as Class 1 for final status survey design and implementation.

Ventilation surveys were performed in a similar manner as surveys of facility piping. The number of data points required for the Sign Test was determined to be 18 (FSSP Addendum 008 in Appendix A). Direct measurements were obtained at equally spaced intervals to assure a minimum of 18 data points. Although the relative shift would be higher and the number of data points required would be lower for Ni-63 and Tc-99 as the contaminates, for consistency the number of data points (i.e., 18) remained the same for all survey units.

4-69

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Recdreulatlon Vent Itts 1125 IN la?124 IS1 ot Col MM an Figure 4-28 1A First Floor Indicating Potentially Impacted Ventilation Systems.

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000 -e0-. -------- 0 12b" LIO UEi Figure 4-30 UVA Reactor Ground Floor Indicting Potentially Impacted Ventilation Systems.

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CH2MHILL Gamma and beta scans of accessible Reactor Stack surfaces were performed inside the plenum on the roof, inside the air intake in the Reactor containment room, and at inlets to the stack plenum in the boiler room. Beta scans were performed on all other interior system surfaces.

Beta scans and surface activity measurements of interior surfaces of 6 in (or larger) ID ductwork were performed using a Model 43-68 gas proportional detector. Ductwork, which was not accessible with this detector, was surveyed using a Victoreen Model 491-30 GM detector (refer to FSSP Addendum 002 in Appendix A).

Interior duct surfaces were scanned by passing the detector through the duct. The rate of detector movement was approximately 1 detector width/sec for the gas proportional and pancake GM detectors and 2.5 to 3.0 cm/sec for the 491-30 GM detector. Model 2221 scaler/ratemeters used with the detectors were monitored for changes in audible signal and any indication of elevated count rate, suggesting possible presence of radioactive contamination, were noted for further investigation. Scan coverage was 100% of the accessible surfaces.

One minute static counts were performed at the designated systematic locations and at locations of elevated count rate identified by scans. Because of the variability in instrument background levels due to varying levels of naturally occurring radionuclides in building construction materials, appropriate reference areas were not applicable. Instead, unshielded and shielded measurements were performed at the same locations and difference compared to the contamination criteria. Smears and Masslinn swabs of duct surfaces were performed to identify the presence of removable activity.

Compliance with decommissioning requirements was demonstrated by comparing the beta activity measurements with the applicable guideline values, i.e., 1.3E+6 dpm/100 cm 2 for systems in Rooms M005 and M008 and 6320 dpm/100 cm2 for all other ventilation system surfaces.

4.9.3 Survey Results Detailed results of ventilation surveys are presented in Appendix B; the specific survey numbers are Survey Unit Survey ID Number Description 24 UVA-FS-68 Reactor Stack 25 TVA-FS-69 Hot Cell Exhaust Ventilation/M019 Blower 26 UVA-FS-71 Reactor Room Recirculation Ventilation 61 UVA-FS-70 Rooms MOOS & M008 Exhaust Ventilation 61A UVA-FS-70 Room G022 Exhaust Ventilation Gamma scans of the Reactor Stack ranged from 9,200 to 15,900 cpm; these levels are comparable to background gamma levels on masonry and brick materials. No locations of elevated direct radiation, which would suggest residual contamination, were identified.

Results of beta scans are summarized in Table 4-27.

4-73

CH2MHILL Table 4-27 Beta Scans of Facility Ventilation Survey Survey Location Counts Per Minute Instrument Unit Unit Minimum Maximum Set LTVA-FS-68 24 Reactor Stack 300 650 9 (b)

UVA-FS-69 25 Hot Cell Ventilation 170 390 9 (b)

UVA-FS-71 26 Reactor Rm Recirc Ventilation 190 460 9 (b)

UVA-FS-70 61 M005 & M008 Ventilation 200 500 9 (b)

UVA-FS-70 61A Rm G022 Ventilation 10 26 15(a)

(a) 491-30 GM Detector _ _ _

(b) 43-68 Gas Proportional Detector Scans utilizing a Model 43-68 gas proportional detector ranged from 170 cpm to 650 cpm; those utilizing a Model 491-30 detector ranged from 10 cpm to 26 cpm. Higher ambient levels were observed in the stack of masonry/brick construction. No specific locations of elevated activity were identified.

Total activity measurement results are summarized in Table 4-28. The maximum activity level was 2,433 dpm/IOOcm 2 in the Reactor room recirculation ductwork. All systematic measurements were below the adjusted gross DCGL of 6320 dpm/100cm2 , (1.3 E+6 dpm/100 cm2 for survey unit 61) thus statistical testing is not necessary to demonstrate compliance with the guidelines.

Table 4-28 Ventilation Beta Activity Measurement Summary Beta Activity (dp /100cmn Survey Instrument Number of Std Unit Type Measurements Minimum Maximum Mean Dev 24 43-68 18 393 2351 1202 578 25 43-68 20 -326 1276 203 348 26 43-68 30 215 2433 595 542 61 43-68 18 -260 883 230 277 61A 491-30 19 -169 576 52 181 Smears obtained at the locations of direct measurements in ventilation systems contained a maximum level of 46 beta dpm/100 cm2 ; all but a couple of these smears were below the detection sensitivity of 28 dpm/100 cm 2 .

A Masslinn swab of the entire G022 duct had a positive beta activity of 990 dpm. This activity is representative of the entire system interior surface, estimated at between 3 and 4 m2 , and is therefore well within the guideline level of 632 dpm/100 cm 2 . Smears at the inlet and outlet of this duct had <28 dpm/100 cm 2 .

4.9.4 Conclusion Surveys demonstrated that remaining potentially impacted ventilation systems do not contain residual contamination in excess of the applicable guideline levels and that the established project decommissioning criteria are satisfied.

4-74

5. Quality Assurance Final status survey activities were performed by qualified and trained personnel, following documented procedures and using properly calibrated instrumentation. Activities were in accordance with the Master Final Status Survey Plan and the area/media-specific Addenda to that Plan. No deviations from plans or procedures that might adversely impact final status survey data quality or its evaluation were identified.

In addition, all activities were in accordance with the Quality Assurance Project Plan, developed specifically for this decommissioning project.

Instrumentation and other measuring devices were properly calibrated and data quality was assured through daily performance testing.

Measurements were duplicated at a frequency of 5% for quality control purposes. Table 5-1 summarizes comparisons of 30 duplicate measurements, from the initial FSS Survey Data Forms. Results are evaluated by Normalized Absolute Difference (NAD), where:

Measurement' - Measurement2 N I NAD =

(Ul)2 + (a2)2 NAD should be <1.96. All of the QC measurement pairs satisfied this criterion.

Sampling was in accordance with documented plans. Equipment was decontaminated and monitored, where appropriate, to prevent cross contamination. Samples were controlled under chain-of-custody until transferred to a commercial laboratory that utilized industry-recognized analytical methods. With several minor exceptions, prescribed measurement sensitivities were met. The laboratory followed appropriate internal QA/QC procedures to assure data accuracy and defensibility.

5-1

CH2MHILL Table 5-1 Normalized Absolute Difference for Duplicate Measurements Survey Initial Duplicate ID Measurement Measurement NAD NAD UVA-FS- Unshield Shield Unshield Shield <1.96 ?

005 328 262 339 250 0.67 Y 010 544 754 551 356 0.12 Y 011 425 289 466 321 0.11 Y 013 439 391 469 364 1.40 Y 014 434 271 426 281 0.22 Y 015 346 259 354 277 0.28 Y 016 2429 346 2463 375 0.07 Y 017 413 401 410 337 1.54 Y 019 408 373 402 348 0.49 Y 021 442 322 410 310 0.52 Y 022 488 397 433 368 0.58 Y 023 496 337 466 360 1.30 Y 024 530 423 507 367 0.77 Y 025 419 297 432 329 OA9 Y 026 307 294 322 291 0.52 Y 027 301 291 279 326 1.65 Y 028 371 305 388 300 0.54 Y 029 335 285 326 287 0.04 Y 029 325 306 320 294 0.19 Y 030 296 314 301 327 0.23 Y 032 322 306 375 300 1.63 Y 033 122 128 127 127 0.27 Y 034 409 333 406 364 0.26 Y 035 302 258 316 71 0.03 Y 040 508 333 502 401 1.77 Y 041 396 288 387 270 0.25 Y 041 342 313 356 320 0.19 Y 042 182 180 180 180 0.07 Y 043 340 264 341 275 0.14 Y 044 727 348 701 313 0.20 Y 5-2

6. Summary Final status surveys were conducted on building surfaces and soils, potentially impacted by licensed activities of the University of Virginia pool-type 2 Megawatt research reactor. These surveys were designed, implemented, and evaluated, following the guidance of MARSSIM and NRC supporting documents. Project decommissioning criteria were the conservative NRC default-screening values.

Monitoring before, during, and after remediation indicated that contamination was primarily low-level and was limited to a small portion of the facility. Contaminants were primarily Co-60 and Cs-137; a small number of samples also contained some hard-to-detect radionuclides, but levels were low relative to guidelines and occurrence was spotty.

Results of the FSS identified one small area of paving, requiring remediation and resurvey.

Otherwise, the FSS demonstrated that remedial contamination of license origin is well below the default-screening guideline levels and that the facility satisfies project decommissioning objectives and criteria and qualifies for termination of NRC License No. R-66.

6-1

7. Works Cited

1. Characterization Survey Report for the University of Virginia Reactor Facility, GTS Duratek, March 2000.

2. University of Virginia Reactor Decommissioning Plan, GTS Duratek, February 2000.

3. Evaluation of Radiological Characterization Results Relative to Termination of NRC License R-123, University of Virginia, Charlottesville, Virginia, Safety and Ecology Corporation, January 2003.

4. University of Virginia Decommissioning Plan Performance Summary, CH2M HILL, April 2004.

5. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575, June 2001.

6. Consolidated NMSS Decommissioning Guidance, NUREG-1757, September 2002.

7. Residual Radioactive Contamination from Decommissioning, NUREG/CR-5512, Vol. 3, Parameter Analysis, October 1999 (Draft).

8. Evaluation of Surface Contamination -Part 1: Beta Emitters), and Alpha Emitters, ISO-7503-1, First Edition 1988-08-01.

9. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, December 1997.

10. Approval of Final Status Survey Plan Coverage Change for License No. R-66-University of Virginia (TAC No. MA3737), Letter from D. E. Hughes (U.S. Nuclear Regulatory Commission) to P. Benneche (University of Virginia), March 31, 2004.

11. NMSS Decommissioning Standard Review Plan, NUREG-1727, September 2000.

7-1

Appendix A -- Final Status Survey Plan and Addenda

1. Master Final Status Survey Plan, UVA-FS-002, Revision 1, April 2004

2. Addendum 001: Underground Waste Tank Excavation

3. Addendum 002: Reactor Facility Piping

4. Addendum 003: Pond Sediments

5. Addendum 004: Interior Structure Surfaces

6. Addendum 005: Exterior Soil and Paved Surfaces

7. Addendum 006: Exterior Structure Surfaces

8. Addendum 007: Special Soils Areas

9. Addendum 008: Ventilation Systems

Master Final Status Survey Plan UVA-FS-002 Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Master Final Status Survey Plan Prepared for the University of Virginia Reactor Facility Decommissioning Project UVA-FS.002 Revision I April 2004 Client 621( q go, a ooq OEHS Date

--PGWQ, A9aed 3c0, Zoom Technical Director Date

.is CH12MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830

Master Final Status Survey Plan for the University of Virginia Research Reactor Facility UVA-FS-002 Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 NedHealthi Pfysicist /1 xla/Y DaI6 /

ES&H &QA Dire tor Date Proje tEneer Date Project Manager Date CH2MHILL

Contents Contents ................................................................. iv Acronyms ............................................................... vi

1. Introduction ............................................................... 1-1

2. Facility Description ............................................................... 2-1

3. Survey Objective ............................................................... 3-1

4. Organization and Responsibilities ............................................................... 4-1 4.1 Project Manager ............................................................... 4-1 4.2 Project Engineer ............................................................... 4-1 4.3 Environmental Safety and Health Manager ............................................................... 4-1 4.4 Certified Health Physicist/Characterization and Final Survey Supervisor ............... 4-1 4.5 Radiological Control Supervisor ............................................................... 4-2 4.6 Radiation Control Technicians ............................................................... 4-2 4.7 Quality Assurance Specialist ............................................................... 4-2 4.8 UVAR Technical Director ............................................................... 4-2

5. Radiological Contaminants and Criteria ............................................................... 5-1

6. General Survey Approach ............................................................... 6-1

7. Survey Plan and Procedures .............................................. 7-1 7.1 Data Quality Objectives.....................................................................................................7-1 7.2 Classification of Areas by Contamination Potential ............................................. 7-1 7.3 Identification of Survey Units ............................................. 7-2 7.4 Demonstrating Compliance with Guidelines ............................................. 7-2 7.5 Background Reference Areas and Materials................................................................... 7-5 7.6 Instrumentation ............................................. 7-5 7.7 Survey Reference Systems .............................................. 7-6 7.8 Determining Data Requirements ............................................. 7-7 7.9 Determining Data Point Locations ............................................. 7-7 7.10 Integrated Survey Strategy .. ........................................... 7-8 7.10.1 Beta Surface Scans .............................................. 7-8 7.10.2 Gamma Surface Scans......................................................................................7-8 7.10.3 Surface Activity Measurements .............................................. 7-9 7.10.4 Removable Activity Measurements ............................................. 7-9 7.10.5 Soil Sampling .............................................. 7-9

8. Data Evaluation and Interpretation .............................................. 8-1 8.1 Sample Analysis ............................................. 8-1 8.2 Date Conversion .............................................. 8-1 8.3 Data Assessment ............................................. 8-1 8.4 Determining Compliance with Guidelines .............................................. 8-1 8.4.1 WRS Test ............................................. 8-1 8.4.2 Sign Test ............................................. 8-2 8.4.3 Unity Rule Sign Test ............................................. 8-3

9. Final Status Report .............................................. 9-1

10. Works Cited ............................................. 10-1 Masli Fnal Stalis Surey Plan WVAR iv

TABLES Table 5-1 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces ..........................................  ; 5-2 Table 5-2 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil....................................................................................... ;..........................................................53 Table 7-1 MARSSIM - Recommended Survey Unit Areas .7-2 Table 7-2 UVAR Survey Areas and Classification .7-3 Table 7-3 Instrumentation for UVAR Final Status Survey .7-6 FIGURES Figure 2-1 Map of Charlottesville Area Surrounding the UVAR Site .2-3 Figure 2-2 Western Grounds of the University of Virginia.2-4 Figure 2-3 University of Virginia Reactor Facility .2-5 Figure 2-4 WVA Reactor Figue 24 UVA eactr First Fist Floor Foor Plan lan View.2-6 iew..................................................................................26 Figure 2-5 UVA Reactor Mezzanine Floor Plan View .2-7 Figure 2-6 UVA Reactor Ground Floor Plan View............................................................................2-8 APPENDICES Appendix A Approach for Development of FSS Guidelines . A-1 Appendix.B List of Procedures Applicable to UVAR Final Status Survey .B-1 MasteW Fin Status Soe PlaW L)VAR V

Acronyms ALARA As Low As Reasonably Achievable CAVALIER Cooperatively Assembled Virginia Low Intensity Educational Reactor bq becquerel cm centimeter cm2 square centimeters DCGL Derived Concentration Guideline Level dpm disintegrations per minute DQO Data Quality Objective ES&H Environmental Safety and Health ESHM Environmental Safety and Health Manager FSS Final Status Survey g gram km kilometer LBGR Lower Bound of the Gray Region m meter m2 square meters MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDA Minimum detectable activity MDC Minimum detectable concentration MIF mineral irradiation facility mrem millirem MW Megawatt NRC Nuclear Regulatory Commission Norn number of data points NIST National Institute of Standards and Technology OEHS UVa Office of Environmental Health and Safety pCi picocurie QA Quality Assurance QC Quality Control RCS Radiological Control Supervisor RCT Radiological Control Technician RWP Radiation Work Permit S+ Sign test statistic SEC Safety and Ecology Corporation UCL Upper Confidence Level UVAR University of Virginia Reactor WR Wilcoxon Rank Sum test statistic WRS Wilcoxon Rank Sum Master Fnal Status Survey Plan WAR an

1. Introduction The University of Virginia began operating a light-water cooled, moderated, and shielded pool-type nuclear research reactor at its Department of Nuclear Engineering in June 1960.

Reactor uses included radiation research, activation analysis, isotope production, neutron radiography, radiation damage studies, and training of Nuclear Engineering students. The reactor was initially commissioned to operate at a maximum power of 1 Megawatt (MW) thermal; it was upgraded to a power level of 2 MW in January 1971. Aluminum-dad, high-enriched uranium fuel was initially used; the reactor was converted to low-enriched uranium fuel in early 1994. The reactor operated under Nuclear Regulatory Commission (NRC) License No. R-66.

In June 1998, the reactor was permanently shut down and the fuel was removed and shipped off site between the shutdown date and mid-1999. Beginning in July 1999, GTS Duratek performed a radiological characterization of the reactor, the facility housing the reactor, and the surrounding fenced and gated land area, collectively referred to as the University of Virginia Reactor (UVAR) facility. Results of that characterization are presented in a March 2000 Characterization Survey Report (Ref. 1.). The University of Virginia submitted a Decommissioning Plan for the UVAR facility to the NRC in February 2000 (Ref. 2.).

Beginning in March 2002, the University of Virginia contracted with CH2M HILL to conduct the decommissioning of the UVAR. Other contractors teamed with CH2M HILL to accomplish this effort are WMG, Inc. Safety and Ecology Corporation (SEC), Bartlett Nuclear, Inc., and Parallax, Inc. (see Section 4). This team conducted additional characterizations, as required; surveyed and released or disposed of materials, depending on radiological conditions; and performed decontamination of components, where appropriate. Following the removal or decontamination of surfaces and materials, a Final Status Survey (FSS) of the facility will be performed to demonstrate that the radiological conditions satisfy NRC-approved criteria for use without radiological restrictions and termination of License No. R-66. This document describes the methodologies for conducting and evaluating that FSS of the UVAR facility.

The UVAR facility also housed the smaller, 100-Watt Cooperatively Assembled Virginia Low Intensity Educational Reactor (CAVALIER) (License No. R-123), located on the ground floor in Room G007. The approval authorizing the Decommissioning Plan for the UVAR facility also required that the CAVALIER facility would first be decommissioned to satisfy Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors," and would then be integrated into the UVAR site. A thorough characterization survey of the CAVALIER facility has been conducted. Results were evaluated relative to the NRC-authorized radiological criteria, Regulatory Guide 1.86, and the Decommissioning Plan for termination of the CAVALIER License. This evaluation indicates that the CAVALIER facility satisfies requirements for termination of the license and a report has been prepared for submission to the NRC. Once the CAVALIER License is terminated, that facility becomes integrated into the UVAR site for FSS under this Plan.

Master Fina Stau Sumy PW LVAR 1-1

2. Facility Description The 2030 m2 (interior floor space) UVAR facility is located on Old Reservoir Road on the western grounds of the University of Virginia in Charlottesville, Virginia (Figure 2-1). The UVA Research Reactor and the decommissioned CAVALIER facility, as well as former offices for faculty and students of the Department of Nuclear Engineering, and the reactor staff, are housed in the facility. The UVAR facility is sited approximately 0.6 kilometers (km) west of the city limits of Charlottesville in Albemarle County, Virginia. To the north, east, and south of the facility (no closer than 0.5 km) there are city residential districts. The only access to the facility is by way of Old Reservoir Road. The land and facilities are the property of the University of Virginia, which is responsible for facility oversight and support. Figures 2-2 and 2-3 are plans showing the UVAR facility and environs.

Figures 2-4 through 2-6 show the three levels of the UVAR facility. The Reactor Confinement Room (Rm 131), which housed the former UVA Research Reactor, is located on the upper floor (first floor). This room contained the 9.8-m-long by 3.7-m-wide by 8.2-m-deep reactor pool, associated operating equipment and systems, the operating controls, and some research/experimental equipment. This room is circular and has an elevated (-10-m) ceiling. In addition, the Instrument Shop (Rm 128), Shipping Area (Rm 127), and multiple offices and other support facilities for staff and students are located on this building level.

On the Mezzanine level were located the Demineralizer (Rm M021), Mechanical Room (Rm M020), HP Laboratory (Rm.M019), several partially contaminated laboratories (Rms M005 [Tc-99 contamination] and M008 [Ni-63 contamination]), and multiple offices and other support facilities for staff and students. A crawl space (MCS) is accessed from the stairwell on the Mezzanine level.

The ground floor contained the Heat Exchanger (Rm G024), Rabbit Room (Rm G005),

Beamport/Experimental area (Rm G020), Hot Cell (Rms G025, G026, and G027), Counting Room (Rm G004), Woodworking and Machine Shop (Rm G008), Source Storage (Rms G022, G018, and G007A), the former CAVALIER facility (Rm G007), and miscellaneous support facilities and areas.

There was a cooling tower located on the roof of the mezzanine level, adjacent to the Reactor Confinement room; this facility provided cooling for the reactor secondary system water.

The UVAR facility building is situated on a 9390-m 2 fenced parcel of land. This land area included two sets of underground tanks for collection of potentially contaminated facility liquid wastes, a pond used for collection and holdup of facility discharges with no or low potential for containing radioactive contamination, a water tank for fuel transfer at ground level at the front of the building, underground storm and sanitary sewer drainage systems, and miscellaneous larger materials and equipment with little or no potential for being radiologically impacted.

The UVAR facility building is of concrete block construction with brick veneer. Floors are concrete slab. Internal walls are block and drywall.

Master Fin Status Suvey Plan tVAR 2-1

CH2MHILL All areas or items in the area that were contaminated above the release limits were either decontaminated to below the release limits or physically removed and processed as radioactive waste. Waste minimization on a cost-to-ultimate-disposal basis was implemented. A few pieces of slightly radioactive equipment (mineral irradiation facility (MIF), transfer casks, etc.) desired by other facilities, were physically transferred to other licensed research reactors for their use.

MaseinFai Stats Surey Pa UVAR 2-2

CH2MHILL Figure 2-1 Map of Charlottesville Area Surrounding the UVAR Site Master FRn4 Status Surmey Plain UVAR 243

CH2MHILL Figure 2-2 Western Grounds of the University of Virginia Location 12: Reactor Facility Location 1: Aerospace Research Laboratory Location 2: Alderman Observatory Residence Area Location 6: Hereford Residential College Location 7: High Energy Physics Laboratory Location 9: McCormick Observatory Location 10: National Radio Astronomy Observatory Location 14: Shelbourne Hall Location 15: Slaughter Recreation Facility Location 16: Special Materials Handling Facility WAR Master Fin Staths Sumy PRan 24

CH2MHILL

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• I. /UND ROUOPm Figure 2-3 University of Virginia Reactor Facility Master Fmib Stats Suey PlS UVAR 245

CA, 1128 lu ~131 106 1105 j 104 103 102 1 ISt 112 lISA115 108 103 24 123 109

__J 114 1 215 I16BL 1204A 12 Figure 2-4 tJVA Reactor First Floor Plan View

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001 M020l Figure -5 VAReatrM ezzanin Flo Pla Viewm

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I' 0008 U v G020 GOOMS 02C102 r-6_V R Go un d Pa .Floo Figure 2-6 WVA Reactor Ground Floor Plan View FJ I-

3. Survey Objective The objective of the FSS is to demonstrate that the radiological conditions of the UVAR facility building and grounds satisfy the approved radiological guidelines for unrestricted release and termination of NRC License No. R-66.

Master Fia Status Survey Pla WVAR 31

4. Organization and Responsibilities A multi-organization team, led by CH2M HILL, is assisting the University of Virginia in this decommissioning project. Other team members include SEC, Bartlett Nuclear Inc., WMG, Inc., and Parallax, Inc. This Section describes the organizational structure for FSS survey activities.

4.1 Project Manager The Project Manager, Patrick Ervin, manages the day-to-day planning, organizing, scheduling, directing, coordinating, and controlling of project resources and budget and is the primary point of contact for the University of Virginia regarding project-related matters.

The Project Manager monitors project status and performance to ensure implementation of the required technical, environmental, safety, health, radiation protection, quality assurance, and safeguards elements of the project.

4.2 Project Engineer The Project Engineer, Patrick Ervin, assists the Certified Health Physicist and Radiological Control Supervisor (RCS) in development of survey work packages, plans, and procedures; and coordinates with the RCS to schedule and implement survey work activities. The Project Engineer reviews and approves FSS Plans.

4.3 Environmental Safety and Health Manager The Environmental Safety and Health Manager (ESHM), Mike Anderson, reports to the Project Manager and is responsible for oversight of all site radiological controls and radiation protection activities, as well as implementation of all industrial safety and industrial hygiene, and environmental monitoring requirements. The ESHM evaluates potential health and safety concerns, prepares hazard assessments for the activities, and assists in training personnel in the safe performance of these activities. The ESHM reviews and approves FSS Plans.

4.4 Certified Health Physicist/Characterization and Final Survey Supervisor The Certified Health Physicist/Characterization and Final Survey Supervisor, Jim Berger, assists the Project Engineer and RCS in development of survey work packages, plans, and procedures; provides oversight of the RCS to assure project requirements are satisfied and that radiological surveys are implemented in accordance with applicable work packages, plans, and procedures; and provides technical expertise in selection of survey methodologies and evaluation and interpretation of survey findings. He is responsible for Master Foa Sabz Se Plan UVAR 4-1

CH2MHILL technical adequacy of survey results and approves the FSS report. He also provides assistance in resolution of NRC issues related to the FSS.

4.5 Radiological Control Supervisor The RCS, Frank Myers, develops survey work packages, plans, and procedures; develops Radiation Work Permit (RWP) and As Low as Reasonably Achievable (ALARA) plans for surveys; determines that prerequisites for survey activities are satisfied; determines the level of survey coverage for various applications; selects appropriate survey instrumentation; oversees the performance of the surveys; evaluates survey results; and documents conclusions of survey evaluation.

4.6 Radiation Control Technicians Radiation Control Technicians (RCTs) conduct survey activities in accordance with hazard assessment, RWP, and ALARA requirements and approved plans and operating procedures. RCIs analyze smears, convert data (as required), and document survey findings.

4.7 Quality Assurance Specialist The Quality Assurance Specialist (Parallax, Inc.) provides periodic performance audits, inspections, and surveillances to assure the requirements of the Quality Assurance Project Plan and quality procedures are satisfied and that work plans and procedures being followed.

4.8 UVAR Technical Director The UVAR Technical Director (Paul Benneche) is the University of Virginia's representative for technical oversight of this decommissioning project. He will review and approve FSS Plans and reports for technical adequacy in satisfying project and regulatory requirements.

Master FMnStatus Suvey PKaLWAR 42

5. Radiological Contaminants and Criteria The GTS Duratek initial characterization survey and the continuing characterization by the CH2M HILL team showed that radiological contamination was generally low level and was limited to a small portion of the structure and grounds. Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclide was Co-60; smaller activities of fission and activation products, namely Cs-137, C-14, Fe-55, and Eu-152 were identified in some media. Ni-63 and Tc-99 contaminants were present on facility surfaces from research projects in Labs M008 and M005, respectively. Low levels of uranium and thorium decay series nuclides were identified in the pond sediments and some backfill material; however, these are of natural origin, rather than from licensed reactor operations.

The Decommissioning Plan established the criteria for residual radioactive material contamination on UVAR facility surfaces and in facility soil. UVAR facility criteria, also referred to as derived concentration guideline levels (DCGLs) are selected from the tables of NRC default screening values (refer to NUREG-1757, Ref. 3, and NUREG/CR-5512 (Volume 3), Ref. 4). The screening values for total surface contamination by potential residual contaminants from licensed activities are listed in Table 5-1; guideline levels for removable activity are 10 percent of the values in that table. Screening values for contaminants in soil are listed in Table 5-2. These screening criteria are based on assuring that estimated doses to facility occupants and the public during future facility use do not exceed annual doses exceeding 25 mrem; default screening criteria are based on conservative exposure scenario and pathway parameters and are generally regarded as providing a high level of confidence that the annual dose limits will not be exceeded.

Aside from the contamination in Labs M005 and M008, multiple radionuclides constituted the contamination on most UVAR surfaces and in facility soils and sediments. Review of historic data and analyses of characterization samples indicated different radionuclide mixtures associated with the following facility media and surfaces:

• Reactor pool surfaces

• Reactor Confinement Room surfaces

• Reactor coolant processing system surfaces

• Soil around the reactor pool

• Pond sediment

• Soil and components associated with liquid waste tanks Additional radionuclide mixes were identified during this project. When such situations were encountered, representative samples were collected and analyzed for specific contents (gamma spectrometry and hard-to-detect analyses) and results were used to develop application-specific DCGLs.

Master FnaI Status Survey PKiLWAR S-1

CH2MHILL Table 5-1 Acceptable License Termination Screening Values of Common Radionuclides for Structure Surfaces Radionuclide Symbol Screening Value Source (dprn/100 cm2 )

Hydrogen-3 (Tritium) 3H 1.2E+08 NUREG-1757 Carbon-14 14C 3.7E+06 NUREG-1757 Sodium-22 22Na 9.5E+06 NUREG-1757

. Sulfur-35 35S 1.3E+07 NUREG-1757 Chlorine-36 36C1 5.OE+05 NUREG-1757 54 Manganese-54 Mn 3.2E+04 NUREG-1757 Iron-55 55 Fe 4.5E+06 NUREG-1757 Cobalt-60 60Co 7.1E+03 NUREG-1757 Nickel-63 63Ni 1.8E+06 NUREG-1757 Strontium-90 90Sr 8.7E+03 NUREG-1757 Technetium-99 99Tc 1.3E+06 NUREG-1757 Iodine-129 129I 3.5E+04 NUREG-1757 1 37 Cesium-137 Cs 2.8E+04 NUREG-1757 152 Europium-152 Eu 1.3E+04 NUREG/CR-5512, Vol. 3 Plutonium-238 Z8PU 3.1E+01 NUREG/CR-5512, Vol. 3 Plutonium-239 239Pu 2.8E+01 NUREG/CR-5512, Vol. 3 Plutonium-241 241Pu 1.4E+03 NUREG/CR-5512, Vol. 3 Americium-241 241Am 2.7E+01 NUREG/CR-5512, Vol. 3 Notes:

a Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume for screening purposes that 100 percent of the surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. Users may calculate site-specific levels using available data on the fraction of removable contamination and DandD Version 2.

b Units are disintegrations per minute (dpm) per 100 square centimeters (dpm/100 cm2 ). One dpm is equivalent to 0.0167 becquerel (Bq). Therefore, to convert to units of Bq/m2 , multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSv/y (25 mrem/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies (see Part 20; Appendix B, Note 4).

Master Fal Status Survey PLanWVAR 52

CH2MHILL Table 5-2 Acceptable License Termination Screening Values of Common Radionuclides for Surface Soil Radionuclide Symbol Screening Values Source (pCi/g)

Hydrogen-3 3H 1.1E+02 NUREG-1757 Carbon-14 14C 1.2E+01 NUREG-1757 54 Manganese-54 Mn 1.5E+01 NUREG-1757 Iron-55 55Fe 1.OE+04 NUREG-1757 Cobalt-60 6 0Co 3.8E+oo NUREG-1757 Nickel-63 63Ni 2.1E+03 NUREG-1757 Strontium-90 90 Sr 1.7E+00 NUREG-1757 Technetium-99 99Tc 1.9E+01 NUREG-1757 Iodine-129 129I 5.OE-01 NUREG-1757 Cesium-137 137CS 1.1E+01 NUREG-1757 Europium-152 152 Eu 8.7E+00 NUREG-1757 Plutonium-238 M38Pu 2.5E+OO NUREG-1757 Plutonium-239 239Pu 2.3E+OO NUREG-1757 Plutonium-241 241Pu 7.2E+01 NUREG-1757 Americium-241 24IAm 2.1E+O0 NUREG-1757 Notes:

a These values represent surficial surface soil concentrations of individual radionuclides that would be deemed in compliance with the 25 mrem/y (0.25 mSv/y) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies; see Part 20, Appendix B, Note 4.

b Screening values are in units of (pCi/g) equivalent to 25 mrem/y (0.25 mSv/y). To convert from pCi/g to units of becquerel per kilogram (Bq/kg), divide each value by 0.027. These values were derived using DandD screening methodology (NUREG/CR-5512, Volume 3). They were derived based on selection of the 90"' percentile of the output dose distributionfor each speciflc radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at "Standard Man"'or at the mean of the distribution for an average human.

Mater Fnal Slatus Suey Pln UVAR -3

CH2MHILL Specific guidelines for different surfaces and media depend on the nature of the surface or medium and the radionuclide mix. For soil and sediment contamination, concentrations of specific significant contaminants in FSS samples will be determined to demonstrate satisfying the unity rule; gross beta measurements will be used to demonstrate compliance with surface activity guidelines, with the gross beta DCGL based on measurements of surrogate contaminants with known relationships to the total contamination mix. Appendix A describes the approach for determining and implementing area-specific guidelines for FSS.

The criteria described in this section are net (above background) concentrations and activity levels of radionuclides; appropriate adjustments for instrument background levels and naturally-occurring radionuclide concentrations in various local media will be made to FSS data before comparing data to the respective criteria.

Use of default screening values as decommissioning guidelines does not allow for areas of elevated activity. Therefore, there are no area factors for small areas of contamination, and all surface activity levels and radionuclide concentrations in soil must satisfy those guideline levels. In addition, because of use of the conservative default screening values, further evaluations and actions to demonstrate that the final conditions satisfy ALARA are not required.

Plan WAR Master Fmal Staus SuWvey 5-4

6. General Survey Approach This FSS Plan was prepared in accordance with the guidelines and recommendations presented in the "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSIM),

NUREG-1575, (Ref 5). Guidance provided in NUREG-1757, "Consolidated NMSS DeconmnissioningGuidance" (Ref. 3), has also been followed in the design, implementation, and evaluation of this FSS. The process emphasizes the use of Data Quality Objectives (DQOs) and Data Quality Assessment, along with a quality assurance/quality control program. The graded approach concept has been followed to assure that survey efforts are maximized in those 'areas having the greatest potential for residual contamination or the highest potential for adverse impacts of residual contamination.

Final Status Surveys will be performed by trained radiation control technicians (RCTs), who are following standard, written procedures and using properly calibrated instruments, sensitive to the potential contaminants. Appendix B lists the procedures applicable to this FSS.

The medium, dimensions, contamination potential (i.e., classification), and contaminant mix may differ for each area undergoing FSS; all factors influencing a specific survey design are typically not available until immediately before an area is turned over for FSS. Therefore, designs for specific surveys, including determination of specific guidelines, sampling/measurement methods, survey unit identification and classification, and data evaluation techniques, will be developed at the time of survey in accordance with the guidance presented in this Master Plan. Each design will be documented as an Addendum to the Master FSS Plan.

Mater Fnal Stabs Survey Plan UVAR &I

7. Survey Plan and Procedures 7.1 Data Quality Objectives The objective of the FSS is to demonstrate that the radiological conditions of the facility satisfy the decommissioning criteria (see Section 5) established in the NRC-approved Decommissioning Plan. The DQOs permit demonstration at the 95 percent confidence level that these criteria are met. Decision errors are 5 percent for both Type I and Type II errors.

Such a Type I (alpha) decision error provides a confidence level of 95 percent that the statistical tests do not incorrectly determine that a surveyed area satisfies criteria when, in fact, it does not. The Type II (beta) decision error provides a confidence level of 95 percent that the statistical tests do not incorrectly determine that a surveyed area does not satisfy criteria when, in fact, it does. Target measurement sensitivities < 25 percent of DCGLw's enable quantification of contaminants at or below the guideline values at the 95 percent confidence level.

Data quality indicators for precision, accuracy, representativeness, comparability, and completeness are as follows:

• Precision is determined by comparison of replicate values from field measurements and sample analyses; the objective is a relative percent difference of 20 percent or less at 50 percent of the guideline value.

• Accuracy is the degree of agreement with the true or known value; the objective for this parameter is +/- 20 percent at 50 percent of the guideline value.

• Representativeness and comparability do not have numeric values; they are assured through selection and proper implementation of sampling and measurement techniques.

• Completeness refers to the portion of the data that meets acceptance criteria and is thus acceptable for statistical testing; the objective for this survey is 90 percent.

7.2 Classification of Areas by Contamination Potential For the purposes of guiding the degree and nature of FSS coverage, MARSSIM first classifies areas as impacted, i.e., areas that may have residual radioactivity from licensed activities, or non-impacted, i.e., areas that are considered unlikely to have residual radioactivity from licensed activities. Non-impacted areas do not require further evaluation. For impacted areas MARSSIM identifies three classifications of areas, according to contamination potential.

• Class 1 Areas: Impacted areas that, prior to remediation, are expected to have concentrations of residual radioactivity that exceed the guideline value.

• Class 2 Areas: Impacted areas that, prior to remediation, are not expected to have concentrations of residual radioactivity that exceed the guideline value.

Master Fial Stabs Sure Plan VAR 7-1

CH2MHILL

• Class 3 Areas: Impacted areas that have a low probability, typically on the order of containing residual activity. Typically levels will not exceed 25-35 percent of the guideline value.

Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities are the bases for classification.

7.3 Identification of Survey Units Impacted areas are divided into survey units for implementing the FSS. A survey unit is a portion of a facility with common contaminants and contamination potential. A survey unit consists of contiguous surfaces or areas. Table 7-1 lists the survey unit areas suggested by MARSSIM; for application at the UVAR facility, areas of survey units will follow these suggested maximum sizes.

Table 7-2 contains a list of facility areas by classification and a projected number of survey units within those areas. Classifications and survey unit boundaries may change based on results as the FSS progresses; if classifications or boundaries change, the survey for affected areas, as described in the applicable FSS Addendum, will be redesigned and the survey and data evaluation repeated.

Table 7-1 MARSSIM - Recommended Survey Unit Areas Class Recommended Survey Unit Area ass Structures Land 1 Uptol00M 2 up to 2000 m2 2 100 to 1000 m2 2000 to 10,000 m2 3 no limit no limit Impacted structure surfaces of *10 m2 and impacted land surfaces of s100 m2 will not be designated as survey units. Instead, a minimum of four measurements (or samples) will be obtained from such areas, based on judgment, and compared individually with the DCGLW.

7.4 Demonstrating Compliance with Guidelines MARSSIM recommends the use of non-parametric statistical tests for demonstrating that radiological conditions satisfy the established project guideline levels. One of the recommended tests is the Wilcoxon Rank Sum (WRS) test; this test may be used when a specific radionuclide of concern is present in background at a concentration greater than 10 percent of the guideline level and when the measurement is not radionuclide specific, e.g., for direct measurements of total surface activity. The other recommended test is the Sign test; this test is used when the radionuclide of concern is not present in background at a significant fraction (i.e., <10 percent) of the guideline level. The Sign test is also used when evaluating data based on the Unity Rule (Sum of Ratios) and may be used for surface activity data representing multiple surface media. Both of these tests are applicable to UVAR facility FSS. The selection of a specific test method will be designated at the time of area FSS design and documented in individual FSS Addenda. MARSSIM Section 8 and Master FmaI Status Survey Ptan UVAR 7-2

( ( CH2M..-L C.

Table 7-2 UVAR Survey Areas and Classifications Room or Area Surface Class Approximate No. of Remarks Surface Area Survey Units (m2) 131 Reactor Room Floor 1 130 2 131 Reactor Room Lower Walls 1 100 1 Reactor Pool Floor and Walls 1 150 2 M005/005A Floor and Lower Walls 1 45 1 M008 Floor and Lower Walls 1 60 1 M019 Floor and Lower Walls 1 80 1 M020 Floor and Lower Walls 1 85 1 M021/021A Floor, Walls, and Ceiling 1 100 1 Bio Shield Surfaces Wall 1 100 1 G005 Floor, Walls, and Ceiling 1 85 1 G007/GO07A Floor, Pit and Lower Walls 1 100 1 G018 Floor, Walls, and Ceiling 1 110 1 G020 Floor and Lower Walls 1 300 3 G022 Floor, Walls, and Ceiling 1 60 1 G024 Floor, Walls, and Ceiling 1 100 1 G025/G026/G027 Floor, Walls, and Ceiling 1 70 1 Pond Sediments 1 1600 1 Waste tank area Soil 1 350 1 Outside Drains Basins and piping 1 N/A 1 Sanitary, storm, and surface Reactor Stack 1 N/A 1 Ductwork, stacks, blowers Other Vent systems 1 N/A 2 Ductwork, stacks, blowers Reactor Piping Interior 1 N/A 2 Coolant piping, drain Piping/Drain systems Systems Maste Final Status Survey Plan UVAR 7-3

( ( CH2MV-.~L C Table 7-2 UVAR Survey Areas and Classifications (continued)

Room or Area Surface Class Approximate No. of Remarks Surface Area Survey (m2) Units 131 Reactor Room Upper Walls and 2 420 1 Ceiling 127/128/130 Floor, Walls, and Ceiling 2 180 1 107/124/124A/124B Floor and Lower Walls 2 250 1 M005/005A Upper Walls and 2 30 1 Ceiling M008 Upper Walls and 2 40 1 Ceiling M019 Upper Walls and 2 60 1 Ceiling M020 . Upper Walls and 2 65 1 Ceiling M006/M014/M015/M030/M031 Floor and Lower Walls 2 250 1 Includes catwalk over G020 MCS (crawl space) Floor, Walls, and Ceiling 2 100 1 Possible soil samples also G004/GO05A Floor and Lower Walls 2 100 1 G006 Floor and Lower Walls 2 70 1 G007B/G008/G0O8A/G016/G017/G019 Floor and Lower Walls 2 150 1 Stairwell 1 Floor and Lower Walls 2 300 1 Stairwell 2 Floor and Lower Walls 2 300 1 Reactor Room Roof all 2 140 1 Building Roof all 2 700 1 Outside Paved Surfaces all 2 2000 2 Rear Loading Area/Remainder Soil Area NE of Reactor soil 2 800 1 Remainder of Structure Floors, walls, and 3 4500 3 Ceiling Remainder of Property Soil 3 4000 1 Master Final Status Survey Plan VAR 74

CH2MHILL.

NUREG-1505 (Ref. 6) contain details on data assessment/interpretation and selection and application of these statistical tests. Also refer to Section 8 of this Plan.

The null hypothesis (H.) for each survey unit is that residual activity exceeds the guideline levels. Therefore, the rejection of the null hypothesis by the statistical test demonstrates that the residual activity does not exceed guidelines and the survey unit satisfies requirements for unrestricted release.

7.5 Background Reference Areas and Materials In addition to the instrumentation background response, many construction materials and environmental media (e.g., soil/sediment) contain naturally-occurring levels of radioactive materials which contribute to a survey measurement. Background contributions must therefore be determined if: (1) the residual contamination includes a radionuclide that occurs in background; or, (2) measurements are not radionuclide-specific. Multiple reference areas and materials are anticipated to be required for the UVAR FSS. For applications involving the WRS test, reference areas must be of the same material as the survey unit being evaluated, but without a history of potential contamination by licensed operations; the number of reference data points must be the same (+/- 20 percent) as the number of data points required from the survey unit. A set of reference measurements must be obtained for each instrument being used for survey unit evaluation. For applications involving the Sign test, sufficient background determinations should be made for each media or surface material and with each instrument to provide an average background level that is accurate to within +/- 20 percent; this usually requires 8 to 10 measurements, which are then evaluated using the procedure described in Draft NUREG/CR-5849 (Ref. 7), and additional data points are obtained, as necessary. Reference area and background requirements will be identified in individual FSS Addenda.

7.6 Instrumentation Table 7-3 lists the instrumentation to be used for survey activities described in this Plan, along with nominal operating parameters and estimated detection sensitivities. Because radionuclides present as contaminants emit (with few exceptions) beta particles with maximum energy greater than 0.300 MeV, detector efficiencies for measuring surface activity are generally determined using Tc-99 (maximum beta energy of approximately 0.292 MeV). For situations where contaminants emit beta particles of lower energy, e.g.,

facilities con'taminated with Ni-63, detector efficiencies are specifically determined for those contaminants. Effects of surface conditions on measurements are integrated into the overall instrument response through use of a "source efficiency" factor, in accordance with the guidance in ISO-7503-1 (Ref. 8) and NUREG-1507 (Ref. 9). Default source efficiency factors, of 0.5 for beta-emitters > 0.4 MeV E. and 0.25 for beta-emitters between 0.150 MeV and 0.400 MeV E. (per SO-7503-1) are generally applicable to UVAR contaminants and surface conditions. However, if contaminants and/or conditions are not consistent with use of these default values, specific source efficiency factors will be determined and documented in the FSS Addendum.

Master Fmi Status Sovey Plan VAR 7-5

CH2MHILL Detection sensitivities are estimated using the guidance in NUREG-1575 (MARSSIM) and NUREG-1507. Instrumentation and survey techniques are chosen with the objective of achieving detection sensitivities of

• 25 percent of the criteria for structure surfaces, for both scanning and direct measurement. This assures identification of areas of elevated activity, that is, having a size and activity level that could adversely impact the average for the survey units.

All instruments have current calibrations using National Institute of Standards and Technology (NISl)-traceable standards. Operational and background checks will be performed at the beginning of each day of FSS activity and whenever there is reason to question instrument performance. Defective instruments will be removed from service and data obtained with that instrument since its last acceptable performance will not be accepted.

Table 7-3 Instrumentation for UVAR Final Status Survey Detector Type Make Meter Application Sensitivity (dpmI100 cm 2 , exc ept as noted)

Scanning Static Count (1 minute) 43-68 Gas Ludlum 2221 Beta scan and 1200 500 Proportional measurement 43-68 Gas Ludlum. 2221 Ni-63 Beta scan 5000 2000 Proportional and measurement 43-37 Floor Ludlum 2221 Beta scan 800 N/A Monitor 43-68 Gas Ludlum 2221 Alpha 200 70 Proportional measurement Tennelec Gas Tennelec N/A Alpha smear N/A 5 LB5100 proportional measurement Tennelec Gas Tennelec N/A Beta smear N/A 10 LB5100 proportional measurement 44-10 NaI Ludlum 2221 Gamma scan 3.3 pCi/g Co-60 N/A 6.4 pCi/g Cs-137 491-30 GM Victoreen 2221 Beta Scan & 4600 2300

_ _ _ _ Measurement I_ I 7.7 Survey Reference Systems A grid system is established on surfaces to provide a means for referencing measurement and sampling locations. On Class 1 and 2 structure surfaces, a 1-m interval grid will be established; a 5-m interval grid will be established on Class 3 structure surfaces; and a 10-m interval grid will be established on land area surfaces. Grid systems typically originate at the southwest corner of the survey unit, but specific survey unit characteristics may necessitate alternate grid origins. Grids are assigned alphanumeric indicators to enable survey location identification. Structure grids are referenced to building features; open land grids are referenced to the state or federal planar grid system. Maps and plot plans of survey areas will include the grid system identifications. Systems and surfaces of less than Master Fmal Salus SurMey Plan AR 746

CH2MHILL 20 m2 will not be gridded, but survey locations will be referenced to prominent facility features.

7.8 Determining Data Requirements Data needs for statistical tests are determined as follows:

1. Calculate the relative shift (A/a)

A/c = DCGL- LBGR The DCGL is the gross or nuclide specific guideline (per Appendix A)

The LBGR (the Lower Bound of the Gray Region) is initially selected as 1/2 of the DCGL as recommended by MARSSIM.

a should be determined empirically from actual survey data, however, for planning purposes and lacking empirical survey unit data, the value of a will be set at 25 of the DCGL or the MDA of the measurement method, whichever is greater.

The resulting relative shift is 2, which is within the range of 1 to 3 recommended by MARSSIM.

2. Determine decision errors The DQOs for this project establish decision errors of 0.05 for both Type I and Type II errors.

3. Determine the number of data points required The number of data points required for statistical testing is obtained from MARSSIM Tables 5.3 (WRS test) and 5.5 (Sign test). For a relative shift of 2 and decision errors of 0.05, the number of data points for the WRS test is 13 and the number for the Sign test is 15. These numbers of data points include an additional 20 percent to allow for potential sample loss and quality control (QC).

The number of data points is determined in this manner for each survey unit undergoing FSS. The determination is documented in the FSS Addendum applicable to that survey unit.

Actual survey unit FSS data will be used to recalculate the relative shift and confirm that adequate data were collected for evaluation.

7.9 Determining Data Point Locations MARSSIM recommends a triangular measurement or sampling pattern to increase the probability of identifying small areas of residual activity. This type of triangular pattern is used for this FSS, except where dimensions and/or other factors related to a specific survey unit require use of an alternate pattern. The spacing (L) between data points on a triangular pattern is determined by:

L = [(Survey Unit Area)/ (0.866 x number of data points)11/ 2 Master Fnal Status Survey Plan LNAR 7-7

CH2MHILL To simplify the designation of data points while assuring that a sufficient number of data points are obtained for statistical purposes, the value of L is rounded to the nearest whole meter. If the systematic pattern does not provide sufficient data points to satisfy the number determined in Section 7.8, additional data points will be identified, using a random-number technique.

7.10 Integrated Survey Strategy Data collected for FSS of structure surfaces will consist of scans to identify locations of residual contamination; direct measurements of beta surface activity; and measurements of removable beta surface activity. FSS of open land (soil) areas will consist of scans to identify locations of residual contamination and samples of soil, analyzed for potential contaminants. Additional measurements and samples will be obtained, as necessary, to supplement the information from these typical survey activities. Survey techniques are described in more detail in this Section.

7.10.1 Beta Surface Scans Beta scanning of structure surfaces will be performed to identify locations of residual surface activity. Gas-flow proportional detectors will be used for beta scans. Floor monitors with 580 cm2 detectors will be used for floor and other larger accessible horizontal surfaces; hand-held 125 cm2 detectors will be used for surfaces not assessable by the floor monitor.

Scanning will be performed with the detector within 0.5 cm of the surface (if surface conditions prevent this distance, the detection sensitivity for an alternate distance will be determined and the scanning technique adjusted accordingly). Scanning speed will be no greater than 1 detector width per second. Audible signals will be monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage will be 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces (Ref. 10). Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

7.10.2 Gamma Surface Scans Gamma scanning will be performed on structure and land surfaces to identify locations of residual surface activity. NaI gamma scintillation detectors (2-in x 2-in) will be used for these scans. Scanning will be performed by moving the detector in a serpentine pattern, while advancing at a rate of approximately 0.5 m per second. The distance between the detector and the surface will be maintained within 5 cm. Audible signals will be monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage will be 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces (Ref. 10). Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

MasrFra Status Survey Mmn WAR 7-8

CH2MHILL 7.10.3 Surface Activity Measurements Direct measurement of beta surface activity will be performed in designated locations (see Section 7.9) using a 125-cm2 gas flow detector. Measurements will be conducted by integrating the count over a 1-minute period. Where adverse surface conditions may result in underestimating activity by direct measurements, surface samples will be obtained for laboratory analyses. Need for such sampling will be identified in FSS Addenda for specific survey units.

7.10.4 Removable Activity Measurements A smear for removable activity will be performed at each direct surface activity measurement location. A 100-cm 2 surface area will be wiped with a 2-in diameter cloth or paper filter, using moderate pressure.

7.10.5 Soil Sampling Sample of surface (upper 15 cm) soil will be obtained from selected locations (Section 7.9) using a hand trowel or bucket auger. Subsurface soil will be collected in continuous cores, homogenized over 1-m depth intervals (NUREG-1727, Appendix E [Ref. 11]).

Approximately 500 to 1000 g of soil will be collected at each sampling location.

Master Fnal Stats Suniey Plan WVAR 7-9

8. Data Evaluation and Interpretation 8.1 Sample Analysis Smears for removable activity will be analyzed by the on-site laboratory for gross alpha and gross beta activity. Analyses of samples of soil and other volumetric media may include gamma spectrometry and/or wet chemistry analyses, depending on radionuclides anticipated. Individual FSS Addenda will describe analyses to be performed.

8.2 Data Conversion Measurement data will be converted to units of dpm/100 cm2 or pCi/g for comparison with guidelines and/or for statistical testing. Where appropriate for Sign tests, data will be adjusted for material and instrument background contributions; data for WRS tests will not be corrected for background, but, instead, will be compared with the data from a reference area.

8.3 Data Assessment Data will be reviewed to assure that the type, quantity, and quality are consistent with the FSS Plan and design assumptions. Data standard deviations will be compared with the assumptions made in establishing the number of data points. Individual and average data values will be compared with guideline values and proper survey area classifications will be confirmed. Individual measurement data in excess of the guideline level for Class 2 areas and in excess of 25 percent of the guideline for Class 3 areas will prompt investigation.

Patterns, anomalies, and deviations from design assumption and Plan requirements will be identified. Need for investigation, reclassification, remediation, and/or resurvey will be determined; a resolution will be initiated and the data conversion and assessment process repeated for new data sets.

8.4 Determining Compliance with Guidelines 8.4.1 WRS Test For a structure surface survey unit to be evaluated using the WRS test, individual survey unit net total activity measurements and the average of the total net activity measurements will be calculated using the average reference area level; also, the difference between the highest survey unit and lowest reference area measurements will be calculated.

If the difference between the highest survey unit and lowest reference area measurements is less than the guideline level, the survey unit satisfies the criterion and no further evaluation will be necessary.

Master Final Status S Prve ltan WAR 8-1

CH2MHILL If the average net surface activity value is greater than the guideline, the survey unity does not satisfy the criterion, and further investigation, remediation, and/or resurvey is required.

If the average net surface activity value is less than the guideline value, but the difference between any survey unit and reference area activity measurement is greater than the guideline, data evaluation by the WRS test proceeds, as follows:

• List each of the survey unit measurements and reference area measurements; do not correct these data for background.

• Add the guideline value to each reference area measurement (for surface activity, add the calculated instrument response equivalent of the guide line to the reference area measurements); these are known as adjusted reference area measurements.

• Rank all (survey unit and reference area) measurements in order of increasing size from 1 to N, where N is the total number of pooled measurements.

• If several measurements have the same value, assign them the average ranking of the group of tied measurements.

• If there are "less-than" values, they are all assigned the average of the ranks from 1 to t, where t is the number of "less-than" values.

• Sum the ranks of the adjusted reference area measurements; this value is the test statistic, WR.

• Compare the value of WR to the critical value in MARSSIM Table 1.4 for the appropriate sample size and decision level.

If WR is greater than the critical value, the null hypothesis is rejected, and the survey unit meets the established criteria. If WR is smaller than the critical value, the null hypothesis is accepted, and the survey unit does not meet the established criteria; investigation, remediation, reclassification, and/or resurvey should be performed as appropriate.

8.4.2 Sign Test For an open land or structure surface survey unit to be evaluated using the Sign test, individual activity values and the average activity value will be calculated.

If all values for a survey unit are less than the guideline level, that survey unit satisfies the criterion and no further evaluation is necessary.

If the average activity value is greater than the guideline, the survey unit does not satisfy the criterion, and further investigation, remediation, and/or resurvey is required.

If the average activity value is less than the guideline level, but some individual values are greater than less than the guideline, data evaluation by the Sign test proceeds, as follows:

• List each of the survey unit measurements

• Subtract each measurement from the guideline level

• Discard all differences which are "0"; determine a revised sample size

• Count the number of positive differences; this value is the test statistic, S+

• Compare the value of S+ to the critical value in MARSSIM Table I.3 for the appropriate sample size and decision level.

Masser Fina Stabs Swo PMI WAR &2

CH2MHILL If S+ is greater than the critical value, the null hypothesis is rejected, and the survey unit meets the established criteria. If S+ is smaller than the critical value, the null hypothesis is accepted, and the survey unit does not meet the established criteria; investigation, remediation, reclassification, and/or resurvey should be performed, as appropriate.

8.4.3 Unity Rule Sign Test For an open land or structure surface survey unit to be evaluated using the Unity Rule Sign test, individual activity values and the ratios of the activity values to their respective guideline values will be calculated. For each data location add the ratios together to determine the Sum of Ratios.

If all Sum of Ratios values for the survey unit are less than 1, that survey unit satisfies the criterion and no further evaluation is necessary.

If the average Sum of Ratios value is greater than the guideline, the survey unit does not satisfy the criterion, and further investigation, remediation, and/or resurvey is required.

If the average Sum of Ratios value is less than 1, but some individual values are greater than 1, data evaluation by the Sign test proceeds, as follows:

• List each of the survey unit Sum of Ratios value.

• Subtract each value from 1.

• Discard all differences which are "O" to determine a revised sample size.

• Count the number of positive differences; this value is the test statistic, S+.

• Compare the value of S+ to the critical value in MARSSIM Table 1.3 for the appropriate sample size and decision level.

If S+ is greater than the critical value, the null hypothesis is rejected, and the survey unit meets the established criteria. If S+ is smaller than the critical value, the null hypothesis is accepted, and the survey unit does not meet the established criteria; investigation, remediation, reclassification, and/or resurvey should be performed, as appropriate.

Master Final Status Survey Wtan WAR B3

9. Final Status Report A report describing the survey procedures and findings will be prepared and provided to the University of Virginia for submission in support of license termination. Report format and content will be consistent with the recommendations presented in Chapter 9 of Draft NUREG/CR-5849.

Mater Fml Statu Suw Mm UVAR 91

<, 10. Works Cited

1. Characterization Survey Report for the University of Virginia Reactor Facility, GTS Duratek, March 2000.

2. University of Virginia Reactor Decommissioning Plan, GTS Duratek, February 2000.

3. Consolidated NMSS Decommissioning Guidance, NUREG-1757, September 2002.

4. Residual Radioactive Contamination from Decommissioning, NUREG/CR-5512, Vol. 3, Parameter Analysis, October 1999 (Draft).

5. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575, June 2001.

6. A Non-parametric Statistical Methodology for the Design and Analysis for Final Decommissioning Surveys, NUREG-1505.

7. Manual for Conduction Radiological Surveys in Support of License Termination, NUREG/CR-5849, June 1992.

8. Evaluation of Surface Contamination -Part 1: Beta Emitters) and Alpha Emitters, ISO-7503-1, First Edition 1988-08-01.

9. Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG-1507, December 1997.

10. Approval of Final Status Survey Plan Coverage Change for License No. R-66-University of Virginia (TAC No. MA3737), Letter from D. E. Hughes (U.S. Nuclear Regulatory Commission) to P. Benneche (University of Virginia), March 31, 2004.

11. NMSS Decommissioning Standard Review Plan, NUREG-1727, September 2000.

Master Fnal Stabas Survey Pn UVAR 10fi1

APPENDIX A Approach for Development of FSS Guidelines .

Master Final Status Swvey Plan UVAR

Appendix A Approach for Development of FSS Guidelines Introduction Application of dose-based UVAR facility FSS guidelines for situations where multiple radionuclide contaminants are present requires consideration of the contribution of each radionuclide, relative to its specific guideline value. The particular radionuclides present at significant levels, the nature of the contamination (surface activity or volumetric), and the methods used to evaluate the radiological conditions (e.g., direct measurement of gross activity, radionuclide-specific analyses, or use of surrogate measurements) necessitate different approaches for developing appropriate guidelines for implementing the FSS. This document describes the methods for determining guidelines.

A. Determine the Mix of Radionuclide Contaminants

1. Tabulate the sample results (activity or concentration) for each of the analytes from nuclide-specific analyses. Consider non-detects (MDAs or MDCs) as actual levels.

(Note: Samples with low-activity levels may result in MDA orMDC values which are a significantfractionof the total activity and will incorrectlyoverestimate the contributions from non-detectableradionuclidesin the total contaminant mix. Therefore, samples containinghigher levels of the representativeradionuclidemix ofinterest should be selected for this determination.)

2. Adjust these results by eliminating radionuclides not associated with the licensed operation and by subtracting average natural background levels.

3. Calculate the total activity or concentration of adjusted levels in the sample and the individual fractional contribution of each radionuclide of interest.

4. Repeat steps 1-3 for all samples from the area of interest.

5. Calculate the average and standard deviation of the fractional contribution of each radionuclide of concern.

6. Calculate the 95 percent upper confidence level (UCL) fractional contribution of each radionuclide of concern that is potentially present, using the method described in Section 8.5.5 of NUREG/CR-5849.

7. Calculate the total of the radionuclide UCL fractions, and normalize the individual UCL values, based on a total of 1 (i.e., unity). The resulting values represent the fractional activity contributions (f, through fn) for radionuclides 1 through n in the survey area of interest.

Master Fnal Status Surey Mm UVAR A-l

Note: Some FSS areas have few, if any, locations with activity of most hard-to-detect radionuclides above analytical detection levels. Therefore, there may be limited data available for determining the average and variability of relative radionuclide ratios. In such situations, radionuclide mixes for other survey areas with the potential for similar contamination will be used, if available. If multiple data sets are not available, radionuclide mixes will be based on a single sample, and analyses of FSS samples will be used to confirm (or modify) the radionuclide mix, used for survey planning and design and to evaluate the final status, relative to criteria.

B. Establish the Gross Beta Surface Activity Guideline of a Mixture

1. Using the fractional activity contributions of radionuclides, determined from Section A, calculate the gross activity guideline value (DCGLgroIs) by:

DCGLvoss= F fl + f2 + ... + fn DCGLI DCGL2 DCGL, Where f, through f. are the activity fractions of radionuclides 1 through n, with DCGLs, DCGLI through DCGLTI respectively and F represents the total fraction of detectable radionuclides in the mixture.

An alternative to deriving a DCGLgJOsS based on the fractional activity contributions is to identify the most conservative DCGL for the identified radionuclides present and use the DCGL value for that radionuclide in the above calculation. Use of this approach will be indicated in Addenda for survey areas with potential surface activity, where applicable.

2. When one or more of the radionuclides present will not be detected by the gross measurement, the gross measurement may serve as a surrogate for the undetected radionuclides by adjusting the DCGLgoSS to account for the activity fractions of the undetected radionuclides by:

DCGLadjgross = 1 1 + R2 + ... + Rn DCGLgrS DCGL 2 DCGLn Where R2 through Rn represent the ratio of the activity fractions f2 through fn, of the non-detectable radionuclides, 2 through n, respectively, to the total fraction of detectable radionuclides in the mixture, i.e., fm/F.

Mater Fmal Status Srey Ptn UVAR A-2

C. Establish a Soil Guideline

1. For multiple contaminants in soil, the Unity Rule is applicable. This means that the sum of the ratios of concentrations present to their respective DCGL,'s from the NRC Table of soil default screening values must be *1.

Cl + C2 +...+ Cn <1 DDCLI DCGL 2 DCGL.

Where C. =concentration of each individual radionuclide (1, 2, ...n)

DCGL= guideline value for each individual radionuclide (1,2,..., n)

In other words, there is not a single soil guideline for the radionuclide mix, but, rather, a group of guidelines applicable to each radionuclide and a Unity Rule applicable to the sum of ratios.

2. Using the fractional activity contributions of radionuclide in a mixture, determined from Section A, levels of certain contaminants (e.g., hard-to-detect radionuclides) can be inferred, based on analyses of contaminants that are easier to measure. The measured radionuclide is referred to as the surrogate. The DCGL for the surrogate radionuclide is adjusted for the contributions of inferred contaminants, following the approach described in MARSSIM Appendix I, Section 1.11.2. If C1 and DCGLI are the concentration and guideline value, respectively, for the surrogate radionuclide, and C2 through Cn, DCGL2 through DCGL,, and R2 through Rn are the concentrations, guideline values, and fractional contributions (ratios of Cy through C/ ), respectively, the adjusted DCGL for the surrogate radionuclide is calculated by:

DCGLogate = 1/[1/DCGL1 + R 2 /DCGL2 + ... + Rn/DCGI]

The ratio of the concentration of the surrogate radionuclide to its DCGLsnate thus accounts for all radionuclides for which contributions are inferred by the surrogate measurement.

Master Fua States Svey Plan WAR A3

APPENDIX B List of Procedures Applicable to UVAR Final Status Surveys Master Fmal Status Survey Pian UVAR

List of Procedures Applicable to UVAR Final Status Surveys BJC-EH-4536 Portable Instrument Response and Operability Tests SEC-IO-701 Radiological Measurement Instrumentation Quality Content Checks (forms only, used with procedure BJC-EH-4536)

SEC-IO-702 Calculating Detection Sensitivity SEC-IO-703 Reference Grid System SEC-IO-704 Surface Scanning SEC-IO-705 Background Measurement and Baseline Sampling SEC-IO-706 Direct Surface Activity Measurement SEC-IO-707 Removable Surface Activity Measurement SEC-IO-723 Operating Instructions for the Ludlum Model 2221 SEC-IO-726 Operating Instructions for the Ludlum 239-1F Floor Monitor SEC-IO-727 Operation of the Tennelec LB5 100W Automatic Counting System SEC-IS-401 Instrumentation and Calibration SEC-IS-402 Training of Instrumentation and Calibration Personnel SEC-IS-403 Calibration of Ludlum Model 2221 SEC-IS407 Calibration of Beta/Gamma Pancake Probes SEC-IS415 Calibration of High Energy Gamma Scintillation Probes SEC-IS417 Calibration of Gas Flow Proportional Probes SEC-IS-426 Calibration of the Tennelec Low Background Counting System SEC-EM-301 Waste Sampling SEC-EM-302 Surface Water Sampling SEC-EM-303 Groundwater Sampling SEC-EM-304 Sediment Sampling SEC-EM-305 Soil Sampling SEC-EM-308 Laboratory Requests and Chain of Custody SEC-EM-311 Sampling Information and Data Management SEC-EM-312 Documentation and Logbooks SEC-QA-001 Corporate Quality Management Program SEC-QA-901 Audit Program SEC-QA-908 Storage and Maintenance of Records SEC-QA-909 Field Change Notice and Field Change Request Mast Fmal Status Swvey Mmai WAR B1

Final Status Survey Plan Addendum 001: Underground Waste Tank Excavation Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2M H LL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 001: Underground Waste Tank Excavation Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

6OXi OEHS Date A&PP3v 2CoLf Technical Director Date CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830

• 1 Fma Satus Sufty RaI Addendum 001: U pdwWasb Tax ExmarAm

Final Status Survey Plan Addendum 001: Underground Waste Tank Excavation Revision 1 Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 reified Health Phy -st Kate Project Manager Date 0

CH2MH ILL Master Final Status SwLey Plan Addendurn 001: Undergrwund Waste Tank Excavation

Contents Contents . . . . . . . . . . . . . . . . . . . iu

1. Introduction l. 1-

2. Area Description .2-1 3.Contaminants of Concern and Guidelines..........................................................................3-1

4. Survey Approach .4-1 4.1 Survey Reference .4-1 4.2 Survey Classification .4-1 4.3 Survey Unit Identification .4-1 4.4 Demonstration of Compliance with Release Guidelines .4-1 4.5 Number of Required Data Points .4-1 4.6 Sampling Pattern .4-1 4.7 Survey Methods .4-2 4.8 Sample Analyses .4-4 4.9 Investigation .4-4

5. Data Evaluation .5-1 FIGURES 2-1 University of Virginia Reactor Facility Indicating Location of the Underground Waste Tanks .2-2 2-2 Waste Tank Excavation, Indicating the Grid System for Survey Reference . 2-3 4-1 Waste Tank Survey Unit, Indicating FSS Soil Sampling Locations .4-3 TABLES 3-1 Results of Analyses of Waste Tank Sludge Sample to Establish Potential Contaminant Mixture .3-2 Master Final Status Survey Plan Addendum 001: Underground Waste Tank Excavat1on

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 001) applies to the area of the underground waste tank excavation.

It should be noted that the field survey activities for the waste tank excavation described in this addendum were implemented during the week of January 20, 2003, to enable timely backfill of the excavation and to coordinate with the NRC for confirmatory scans and sampling.

Master Fmal Status Survey Plan Addendum 001: UrKlergrourd WasteTank Excavation 1-1

23 Area Description Two sets of underground metal tanks, located southeast of the UVAR facility adjacent to the pond, were used for collection/holdup of liquid wastes, which were potentially contaminated with low concentrations of radioactive materials (Figure 2-1). Two of these tanks serviced the hot cell facility and two were used for collection of demineralizer regeneration liquids from the 2-MW WVA Reactor. Both of these tank sets were initially equipped for environmental discharge to the pond, provided the liquid met appropriate release criteria following dilution with adjoining pond water. However, the demineralizer regeneration liquid tanks were later replumbed so that they discharged directly into the pond spillway. Both sets of tanks and associated piping, valves, pumps, etc., have been removed along with their concrete foundations. Small quantities of contaminated soil in the vicinity of the tanks and discharge lines were removed. The resulting excavation is approximately 175 m2 in area and ranges up to approximately 3 m in depth; including the unexcavated soil edges. The area to be addressed by this survey is approximately 350 m2 (Figure 2-2).

Master Final Status SUrvey Plan Addendum 001: Undergriund WasteTank Excavation 2-1

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Meters Figure 2-2 Waste Tank Excavation, Indicating the Grid System for Survey.

Master Final Status Survey Plan 2-3 Addendum 001: Underground Waste Tank Excavation

3. Contaminants of Concern and Guidelines Contaminated soil was identified at the base of the demineralizer regeneration waste tank blockhouse. During excavation of the tanks, samples of the soil were collected at depths down to approximately 3 m. The samples were analyzed by gamma spectrometry to determine the radiological nature of the contaminants. These analyses identified only Co-60 and Cs-137 at detectable concentrations. The two samples with detectable levels of these radionuclides contained: 1) 3.00 pCi/g of Co-60 and 4.26 pCi/g of Cs-137 and 2) 1.84 pCi/g of Co-60 and 1.74 pCi/g of Cs-137. During excavation, approximately 0.5 m of additional soil was removed from the surface where these samples were obtained. Hand augered samples from three other locations did not contain detectable concentrations of facility related gamma-emitting radionuclides. Approximately 40 additional samples of excavated (non-impacted) soil were collected and analyzed by gamma spectrometry. No gamma-emitting radioactive contaminants were identified at detectable levels in these samples.

Therefore, the maximum reported Co-60 and Cs-137 concentrations in the excavated soils were the MDA's of <0.83 pCi/g and <0.87 pCi/g, respectively. A sample of waste tank sludge was collected and analyzed for gamma emitters and hard-to-detect radionuclides.

Results, presented in Table 3-1, identified 6930 pCi/g of Co-60, 8142 pCi/g of H-3, 1110 pCi/g of Fe-55, and smaller positive concentrations of several additional nuclides. Five of the fourteen potential radionuclides of concern were not present above their method detection sensitivities. Based on these ratios, only Co-60, H-3, Sr-90, and Mn-54 would have a relative dose contribution above 1 %; the total dose contribution from all other contaminants would be less than 1.6%. Additional samples, containing sufficient activity levels to enable a meaningful determination of the radionuclide mixture for this survey unit, were not available. Based on these results and the history of reactor operations, the predominant dose contributing radioactive contaminant in soils from the waste tank excavation is expected to be Co-60.

NRC default screening criteria will be utilized as release criteria for the soil in the waste tank excavation area. The default screening criteria concentration for Co-60 is 3.8 pCi/g. The Co-60 DCGL Iogate is calculated to be 3.4 pCi/g, based on the mixture determined from the single sample available. For the final status survey, soil samples from the excavation area will be analyzed for gamma emitting contaminants and the Co-60 result compared to the DCGLsurroSate. (Also refer to Section 4.8.)

Master Fnal Status Survey PMa Addendm 001: Undenground Waste Tank Excavation 3-1

CH2MHILL Table 3-1 Results of Analyses of Waste Tank Sludge Sample to Established Potential Contaminant Mixture Radionuclide Sample Activity Fraction of Soil DCGL Fraction of (pCi/g) Activity (pCi/g) Mixture Dose Am-241 <0.3 <1.7E-5 2.1 0.000 C-14 108 6.1E-3 12 0.005 Co-60 6930 3.9E-1 3.8 0.914 Cs-137 <27.7 <1.6E-3 11 0.001 Fe-55 1110 6.3E-2 11000 0.000 H-3 8142 4.6E-1 110 0.037 I-129 8.56 4.9E-4 0.5 0.009 Mn-54 352 2.OE-2 15 0.012 Ni-63 577 3.3E-2 2100 0.000 Pu-238 <0.3 <1.7E-5 2.5 0.000 Pu-239 <0.3 <1.7E-5 2.3 0.000 Pu-241 <200 <1.1E-2 72 0.001 Sr-90 71 4.OE-3 1.7 0.021 Tc-99 26.7 1.5E-3 19 0.001 MasterFa Stats SurM Kin 32

4. Survey Approach 4.1 Survey Reference System A 5-meter grid was established over the excavation area and extended to unexcavated soil surrounding the excavation. This grid is an extension of the reference grid established for survey of the sediments in the adjacent pond, thus enabling the sampling locations to be related to the federal and/or state planar coordinate system. Figure 2-2 illustrates the reference grid system's relations ship to the excavation.

4.2 Survey Classification Based on the facility use history and identification of contaminants of license origin in the-soils in this area, the survey area is designated Class I for FSS planning and implementation purposes.

4.3 Survey Unit Identification The area of the waste tank excavation and surrounding soil is approximately 350 m2 ; this is within the MARSSIM-recommended area of 1000 m2 for Class 1 open land survey units.

Therefore, the area is a single survey unit.

4.4 Demonstration of Compliance with Release Guidelines Compliance with decommissioning requirements will be demonstrated by comparing the results of FSS sample analyses with the surrogate Co-60 DCGL of 3.4 pCi/g. Because the radionuclides identified as potential contaminants are not present in background at concentrations, which are significant fractions of the release guidelines, correction of FSS sample data for background levels will not be required. Statistical testing of results will utilize the Sign Test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors will be 0.05 (Type 1 and Type 2).

4.5 Number of Required Data Points (Refer to Master FSS Plan, Section 7.8).

DCGL = 3.4, LBGR = 0.5 DCGL A = DCGL - LBGR = 1.7 a = 1.0 (based on values for Co-60 in excavated backfill samples)

A/oY = 1.7 From MARSSIM Table 5.5, N = 17; i.e., 17 samples required from survey unit for Sign test.

Actual FSS data will be used to recalculate the relative shift and confirm an adequate number of data points for compliance evaluation.

Master Final Status Suvey Plan Addendum 001: Underground Waste Tank Excavaton 4-1

CH2MHILL 4.6 Sampling Pattern A triangular pattern, based on 17 samples and 350 m2 areas, was used to determine sampling locations. The distance (L) between samples is:

L = [350/(0.866 x 17)] 0.5 = 4.9 m (rounded to 5.0 m for ease of field implementation)

A random start point for the pattern is based on survey unit dimensions of 10 m N/S and 25 m E/W and random numbers from the MARSSIM random number table of 0.793416 and 0.448970. The resulting start point was 7.9 m N and 11.2 W.

Sampling locations are:

3.6N,1.3W 3.6 N, 3.7 W 3.6 N, 8.7 W 3.6 N, 13.7 W 3.6 N, 18.7 W 3.6 N, 23.7W 7.9 N, 1.2 W 7.9 N, 6.2 W 7.9 N, 11.2 W 7.9 N, 16.2 W 7.9 N, 21.2W 12.2 N, -1.3 W 12.2 N, 3.7W 12.2N, 8.7 W 12.2 N, 13.7 W 12.2 N, 18.7 W 12.2N, 23.7 W Because only 14 of these locations fell onto soil surfaces, an additional line of locations at

-0.7 N was added.

-0.7 N, 1.2 W

-0.7 N, 6.2 W

-0.7 N, 11.2 W

-0.7 N, 16.2 W

-0.7 N, 21.2 W The resulting total number of sampling locations is 19; Figure 4-1 indicates the sampling locations.

4.7 Survey Methods Gamma walkover surface scans were performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5 to 10 cm of the soil surface and moved from side to side in a serpentine Master Fmal Status Survey Plan Addendun 001: Underground Waste Tank Excavaton 4-2

CH2MHILL pattern while noting any indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Scanning coverage was 100% of the soil surface.

Surface (0 tol5 cm) soil samples of at least 500 g were collected at the 19 discrete sampling locations identified above. If a sample could not be obtained from a pre-identified location one was obtained from the nearest soil location available. The survey/sampling record noted this deviation. The licensee and the NRC Inspector witnessed the soil sampling and selected samples for confirmatory purposes. Requested samples were homogenized and split. This process was used because the excavation required accelerated backfilling to maintain slope stability. Samples were assigned unique identification numbers and a chain of custody record and analytical request were prepared.

4.8 Sample Analyses Samples were screened by on-site gamma spectrometry and then sent to an off-site commercial laboratory for individual gamma spectral analysis. Co-60 will be used as a surrogate with a DCGL urngate of 3.4 pCi/ to infer the compliance of each sample with criteria. A composite, consisting of an equal amount (about 10 grams) from each survey unit sampling location, was also prepared for off-site analysis by gamma spectrometry and for hard-to-detect (10 CFR Part 61) radionuclides. Results of these analyses of the composite will be used to confirm the absence of hard-to-detect contaminants not identified in the tank sludge sample.

4.9 Investigation If surface gamma scans, sample analysis, or statistical data evaluation identify residual contamination exceeding release criteria, the source of the residual contamination will be determined and characterized. Remedial action will be conducted as required, and the FSS activities repeated utilizing a newly determined sampling pattern and random start point.

Mastr FialStabjsSurvey Pain Addendun 001: Underground Waste Tank Excavation 4-3

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4 Meters -* 0.6m4-Figure 4-1 Waste Tank Survey Unit, Indicating FSS Soil Sampling Location Master Fina Stabus Survey Plan 44 Addendum 001: Underground Waste Tank Excavallon

5. Data Evaluation Sample Co-60 content will be compared with the DCGLsnrrogate of 3.4 pC/g to determine if activity results meet release criteria for radionuclides that are of facility origin. The Sign Test will be performed and the results compared with the critical value for the appropriate number of samples and decision errors. A composite sample will be analyzed for hard-to-detects to confirm the absence of other potential contaminants.

Master Final Status Survey Plan Addendun 001: Uderground WasLe Tank Excavation i

Final Status Survey Plan Addendum 002: Reactor Facility Piping Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 002: Reactor Facility Piping Revision i Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

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OEHS Date

-Po- Q Technical Director Date CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 Frd Stius Suvey PRam AdendnmO02 FcO Fwiby PFwq

Final Status Survey Plan Addendum 002: Reactor Facility Piping Revision 1 Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 rtified Health Physici ate I

Project Manager Date CH2MHILL Final Status Survey Plan Addendurn 002: Reator FacibTy Piping

J Contents Contents . .............................................................. iii

1. Introduction ............................................................. 1-1

2. Description .............................................................. 2-1

3. Contaminants of Concern and Guidelines ............................................................. 3-1

4. Survey Approach .............................................................. 4-1 4.1 Survey Reference System .............................................................. 4-1 4.2 Survey Classification ............................................................. 4-1 4.3 Survey Unit Identification ............................................................. 4-1 4.4 Demonstrating Compliance with Release Guidelines ............................................. 4-1 4.5 Number of Required Data Points .............................................................. 4-2 4.6 Sampling Pattern ............................................................. 4-2 4.7 Survey Methods ............................................................. 4-2

.4.8 Sample Analyses .............................................................. 4-3 4.9 Investigation ............................................................. 4-3

5. Data Evaluation .............................................................. 5-1 FIGURES 2-1 UVAR Reactor Room Floor Showing Remaining Drain Piping . ................................. 2-3 2-2 Ground Floor Indicating Remaining Drain Piping ......................................................... 2-5 2-3 Drains Servicing the Former CAVALIER Facility .......................................................... 2-6 2-4 Sanitary and Storm Drains and Drain Outfalls for the UVAR Facility ........................ 2-7 TABLES 3-1 Results of Analyses of Waste Tank Sludge Sample to Establish Piping Potential Contaminant Mixture ............................................................... 3-1 Attachment A Response and Field Use of the Victoreen Model 491-30 GM Detector for FSS of Reactor Facility Piping ................................................................ A-1 Feial Status Survey Plan iv Addendum 002 Reactr Faity iong

1.Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 002) applies to the potentially impacted buried and embedded piping, associated with the reactor facility.

Fird StuSuey Plan Addendumn

02. Reactor FacHity Piping 1-1

2. Description The bulk of known potentially contaminated piping was removed from the UVAR Facility during remediation activities, but sections of radiologically impacted piping previously associated with the reactor coolant system and various drains from the reactor facility remain. This remaining piping is embedded in concrete or buried beneath concrete or asphalt paving and soil. The piping is generally of small diameter (2 in to 4 in ID); however there are several short sections of larger diameter. Remaining piping includes portions or all of the following:

• Heat exchanger lines: Stainless Steel (SS), 6 in ID x 22 ft and 6 in ID x 32 ft.

• Reactor pool drains: SS, 2 in ID x 32 ft and 2 in ID x 36 ft.

• Reactor Room floor drains: Cast Iron (CI), 2 in ID x -160 ft (multiple sections).

• Ground floor drain to Pond standpipe: CI, 2 in ID x 40 ft and 4 in ID x 140 ft.

• Reactor Demineralizer drain to outside underground collection tanks: CI, 2 in ID x 75 ft.

• Hot Cell drain to outside underground collection tanks: Duriron with PVC repair, 2 in ID x 55 ft.

• Ground floor Bulk Access Facility drains to Pond hillside: CI, 2 in ID x 40 ft and terra cotta, 4 in ID x 80 ft.

• Sanitary sewer from liquid release point to sewer manway: 4 in CI by 40 ft.

• Drain lines from CAVALIER facility to Pond hillside.

In addition, the facility and property are serviced by building and pool footing drains, storm drains and sanitary sewer drains, located beneath the paved area, south of the building.

There is no history to suggest these systems may have become contaminated as a result of licensed facility operations.

Figures 2-1 through 2-4 illustrate the locations of the reactor facility piping.

Visual (boroscope) inspection of the internal surfaces of reactor room drain piping revealed breaks or blockages in the floor drain piping beneath the Reactor Room floor. Indications of potential breaks in the floor drain piping beneath the ground floor, and in the piping to the Hot Cell collection tanks under the roadway (tanks were removed during remediation). This inspection also identified accumulations of scale and loose debris, concentrated on the bottom surfaces of the piping. Visual inspection of storm and sanitary system piping was not conducted.

Fial Status Survey PLan Addendun 002Q Rewt Fadrty Piig 2-1

CH2MHILL Broken or damaged areas of piping were accessed, and contaminated pieces of pipe and soil were identified and removed. The locations in the ground floor drain and hot cell piping that indicated potential damage, were excavated, and determined to be intact. Hydrolazing of reactor piping internal surfaces was performed to remove scale and loose debris. Piping access points have been created to enable final status survey and NRC confirmatory activities.

Final Status Survey FPan 2-2 Addendurn 002 Reactor FacTey Firg

( ( CH2M;,LL LEGEND X 4 INCH SAMPLE CORE (SC)

D CLEAN OUT (CO) o FLOOR OR SINK DRAIN (FO OR SO) o ACCESS CORES (LETTERED)

SCALE BAR EGUALS 3 METERS (PLUGGED 5-12 03)

FD;

-N DOWN TO POND STANOPIPE PIPING REMOVED FROM REACTOR ROOM TO HEAT EXCHANGER ROOM SUMP CO X 1B REMAINING PIPING SHOWN SOLID DOTTED LINE SHOWS REMOVED PIPING BREAK- INDICATES SEPARATED LINE LOCATION Figure 2-1 UVAR Reactor Room Floor Showing Remaining Drain Piping.

FiniStts Suvey Plan Addendum 002. Reactor Faclity Piping 243

( ( C142M.et-L Figure 2-2 Ground Floor Indicating Remaining Drain Piping.

Final Status Survey Plan Addendum O02 Reator Facility Piong 24

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-100 ft to Pond Outfall Figure 2-3 Drains Servicing the Former CAVALIER Facility.

Final Stats Survey Plan 2-5 Addendum 002 ReawWet Fadlity Piping

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I ,.

L-_ - _

i Figure 2-4 Sanitary and Storm Drains and Drain Outfalls for the UVAR Facility.

Final Status Survey Plan 246 Addendum 002 Reador Fadity Pping

3. Contaminants of Concern and Guidelines Piping did not contain sufficient activity levels to enable meaningful determinations of the contaminant mixture, particularly for the hard-to-detect radionuclides. It is logical to assume that the mixture in the reactor piping is the same as that reaching the waste tanks (see Addendum 001). Soil removed during excavation of the underground waste tanks, soil from the vicinity of piping breaks, debris collected from piping, and pieces of removed piping were analyzed by gamma spectroscopy. These analyses identified Co-60 as. the primary potential contaminant in most of the remaining piping. Cs-137 is the major potential contaminant associated with the Hot Cell drain; however, because Co-60 is potentially present in the drain, the more conservative approach of assuming Co-60 as the dominant contaminant on a dose basis in all drains was chosen. Table 3-1 presents the radionuclide analysis of waste tank sludge, used as the basis for the radionuclide mnixture on interior piping surfaces. This activity mnixture is dominated by Co-60 (39%) and H-3 (46%).

On a dose basis, Co-60 contributes 83.7% of the dose and Pu-241 contributes 12.2% of the dose; the remainder of the radionuclides potentially contributes less than 5% of the total dose from this mnixture.

Table 3-1 Results of Analyses of Waste Tank Sludge Sample to Establish Piping Potential Contaminant Mixture Radionuclide Sample Activity Fraction of ufDCGIU Fraction of (pCi/g) Activity (dpn/l100cm2) Mixture Dose Am-241 <0.3 <1.7E-5 2.7E+1 0.006 C-14 108 6.1E-3 3.7E+6 0.000 Co-60 6930 3.9E-1 7.1E+3 0.837 Cs-137 <27.7 <1.6E-3 2.8E+4 0.000 Fe-55 1110 6.3E-2 4.5E+6 0.000 H-3 8142 4.6E-1 1.2E+8 0.000 I-129 8.56 4.9E-4 3.5E+4 0.000 Mn-54 352 2.0E-2 3.2E+4 0.012 Ni-63 577 3.3E-2 1.8E+6 0.000 Pu-238 <0.3 <1.7E-5 3.1E+1 0.008 Pu-239 <0.3 <1.7E-5 2.8E+1 0.008' Pu-241 <200 <1.1E-2 1.4E+3 ;122 Sr-90 71 4.OE-3 8.7E+3 0.006 Tc-99 26.7 1.5E-3 1.3E+6 0.000 The gross activity DCGL for all radionuclides at the activity fractions in Table 3-1 is 15,200 dpm/100cm2. Based on only beta emissions from Co-60, Mn-54 and Sr-90 being measurable (i.e. 41.4% of the radionuclides present will be detectable), the approach described in Appendix A of the Master Final Status Survey Plan yields a DCGL~OSS of 7390 dpm/lOOcm 2 and an adjusted gross of 6320 dpm/lOOcm 2 . This latter value (6320 dpm/lOOcm 2 ) will be used as the applicable total gross ,Bcriteria for all facility piping. Removable activity criteria ForaStatus Survey Flan Addendum 002 Reactor Faty Ping 3-1

are 10% of this value. This criteria represents a conservative approach for Hot Cell piping, in which the contaminant is more likely to be Cs-137 with a less restrictive guideline value.

Final Stylus Survey Plan 3-2 Addendum 002 Reactr Facility iping

4. Survey Approach 4.1 Survey Reference System Locations of survey measurements and sampling will be referenced to piping access points and identified on facility drawings.

4.2 Survey Classification Based on the facility use history and identification of contaminants of license origin in the remaining impacted piping, the reactor facility piping surfaces are designated Class 1 for FSS planning and implementation purposes. Storm drains, building and pool footing drains, the CAVALIER Facility drains, and the non-release path portion (west line) of the sanitary sewer system are designated Class 2.

4.3 Survey Unit Identification Piping has been grouped into the following survey units:

• Reactor Room floor drain system.

• Heat Exchanger piping.

• Reactor pool drains.

• Hot Cell drain.

• Drain to liquid waste storage tanks.

• Reactor drains to pond.

• Sanitary sewer release path.

• CAVALIER Facility drains.

• Storm drains, building and pool footing drains and the non release path portion (west line) of the sanitary sewers.

4.4 Demonstrating Compliance with Release Guidelines Compliance with decommissioning requirements will be demonstrated by comparing the results of FSS measurements with the adjusted gross DCGL surface activity criteria for the anticipated contaminant mixture. Instrument background may be a significant fraction

(>10%) of the count rate representing the total activity criteria and highly variable, due to a variety of piping materials and facility locations. Background correction of the direct measurements will be performed by the shielded/unshielded measurement approach at Final Staus Survey Plan Addendum 002: Reacor Faclity iping 4-1

CH2MHILL each measurement location. Statistical testing of results will utilize the Sign Test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria.

Decision errors will be 0.05 (Type 1 and Type 2).

4.5 Number of Required Data Points DCGL = 6320, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL - LBGR = 3160 cr = 2300 (based on the MDA for the least sensitive measurement technique; refer to Attachment)

A/r= 1A From MARSSIM Table 5.5, N = 20; i.e., 20 measurements required for each survey unit for Sign Test evaluation.

Actual survey unit FSS data will be used to recalculate the relative shift and confirm adequate data were collected for evaluation.

4.6 Sampling Pattern Direct measurements will be obtained at equally spaced intervals along the piping to assure a minimum of 20 data points. For example, for the 55 ft (16.7 m) length of Hot Cell piping measurements will be obtained at 0.8 m intervals along the entire length of this piping.

4.7 Survey Methods Scans and surface activity measurements of interior surfaces of 6 in (or larger) ID piping will be performed using Model 43-68 gas proportional or Model 44-9 pancake GM detectors, depending on accessibility with such detectors. Piping which is not accessible with those detectors will be surveyed using a Victoreen Model 491-30 GM detector. This latter detector has a 30 mg/cm 2 wall thickness and in an unshielded configuration has an effective field of view of slightly more than 100 cm 2 in a 2 in ID pipe. The overall diameter of the 491-30 detector assembly is approximately 2.9 cm, enabling access to most piping surfaces. Detector response to Co-60 in piping was determined by cross calibration, using a section of contaminated piping containing a measured activity level.

Following removal of contaminated piping and soil and hydrolazing to remove scale and loose contamination from internal surfaces, the interior surfaces will be scanned by passing the detector through the pipe. The rate of detector movement will be approximately 1 detector width/sec for the gas proportional and pancake GM detectors and 2.5 to 3.0 cm/sec for the 491-30 GM detector. Model 2221 scaler/ratemeters used with the detectors will be monitored for changes in audible signal and any indication of elevated count rate, suggesting possible presence of radioactive contamination, will be noted for further investigation. Scan coverage will be 100% of the length of Class 1 piping and 25% of the length of Class 2 piping.

Fira Status Swvey FRan 4-2 Addendum 002: Reatr Faity FoxphV

CH2MHILL Unshielded and shielded one-mninute static counts will be performed at the designated systematic locations (see Sections 4.5 and 4.6) and at locations of elevated count rate identified by scans.

A swab will be passed through each pipe section to collect removable activity.

Where contaminated piping has been removed, gamma scans of the excavation soil surface will be performed to confirm the effectiveness of remediation. Samples of soil and gravel will be obtained from such excavations.

4.8 Sample Analyses Swabs for removable activity will be scanned with a Model 43-68 detector to identify areas of elevated beta activity. This same detector will be used to perform a one-minute count to measure activity levels at each elevated beta activity location (if any) identified by the scans, or at a random location on the swab (if no areas of elevated activity are noted).

Samples of soil and gravel will be analyzed by a commercial laboratory by gamma spectrometry. Analyses for hard-to-detect radionuclides will also be performed on selected samples to confirm the radionuclide mixture assumed for the piping.

4.9 Investigation If surface scans, direct measurements, swabs, samples, or statistical data evaluation identify residual contamination on Class 1 surfaces exceeding release criteria, remediation will be performed, followed by resurvey, as appropriate. Identification of contamination exceeding release criteria on Class 2 surfaces will require investigation, remediation, reclassification to Class 1, and resurvey in accordance with the higher rigor for Class 1 surfaces.

Following FSS activities, piping access points will be covered to prevent recontamination and to allow for future NRC confirmatory actions.

Fa Slatus Sawu Plan 4-3 Addendum 002 Rieacor Faciity Piping

5. Data Evaluation Total and removable net activity levels will be calculated. Data will be assessed for conformance with the FSS Plan and design DQOs. Additional data will be obtained, if required, and the assessment repeated. The Sign Test will be performed and results compared with critical values for the appropriate number of data points and decision errors.

Samples of soil and gravel will be compared directly with established Co-60 DCGLIurogate of 3.4 pCi/g.

FMi Status Swey Plan Addendum 002 Reacto Faclity Piping 5-1

ATTACHMENT A To Addendum 002 Response and Field Use of the Victoreen Model 491-30 GM Detector for FSS of Reactor Facility Piping Flrd Stafts Survey FPan AdlendumO02 Reator Falty F g Al1

Response and Field Use of the Victoreen Model 491-30 GM Detector for FSS of Reactor Facility Piping: Attachment A to Addendum 002 for the UVAR Facility Master FSS Plan Introduction Small-diameter piping and piping with internal obstructions is not adequately accessible by the 125 cm 2 gas proportional detector or the 15.5 cm2 pancake GM detector, commonly used for final status survey (FSS) of other surfaces. Therefore a smaller detector is needed to perform scans and direct measurements of the internal surfaces of most of the impacted reactor piping remaining at the UVAR facility. A Victoreen Model.

491-30 GM detector with a 30 mg/cm2 window thickness, a tube length of 6.3 cm, and an effective field of view of approximately 100 cm2 in the unshielded detector configuration was selected for such applications. The entire detector assembly, including the integral sliding shield, is approximately 2.9 cm in diameter. Piping with a diameter of 5.1 cm (2 in) or greater will therefore be accessible by this detector.

Detector response to Co-60 contamination in pipe scale on the interior of a 2 in ID pipe was determined as follows:

• A 20 to 25 cm section of contaminated cast iron pipe was obtained from beneath the Reactor Room floor.

• A gamma spectrum of this section was obtained and Co-60 was identified as the dominant (>80%) contaminant.

• The piping was cut lengthwise into 4 strips. These strips were arranged side-by-side to provide a "flat" contaminated surface source.

• The activity level on this surface was measured, using a 125 cm2 Model 43-68 gas proportional detector; this gas proportional detector had a total efficiency factor of 0.10 for Tc-99, which has a beta energy spectrum, which is very similar to the energy spectrum of Co-60. The average source count was 358 c/mi and the average detector background count was 208 c/m, resulting in a net count rate of 150 c/m. The resulting activity level determined for the piping surface was 1200 dpm/100 cm 2 .

• The piping pieces were then reassembled into their original configuration, i.e., a 2 in diameter pipe with internal surface contamination; the pipe was positioned with the scale concentrated on the bottom of the pipe.

• The Victoreen detector (Serial No. 339) was covered with a thin plastic sleeve to prevent contamination and inserted into the pipe; the detector was in close contact with the contaminated scale on the bottom of the pipe, at the same location where the gas proportional detector measurement had been performed.

Final Status Survey APm Addendum 002 Reactor Faty Piping A-2

• Ten-minute counts were performed with the Victoreen 491-30 detector shielded (background) and unshielded (source response). The resulting count rates were:

background, 36.1 c/m, and source, 52.3 c/m.

• The net response of the Victoreen detector was calculated to be (52.3 c/m -

36.1 c/m)/1200 dpm/100 cm2 or (0.0135 c/m)/dpm/100 cm 2 . This response can also be expressed as 74.1 dpm/100 cm 2 of Co-60 activity per 1 net c/mr on the Victoreen 491-30 detector.

MDA's are estimated using the following relationships:

MDA Static count 3 + 4.65[RB *

t. Es Where RB = Background count rate (C/nin) t = Counting time (min)

Es = Instrument response (c/m/dpm/100 cm2 )

MDAscan 6]O 60 60[RB liee Where RB = Background count rate (c/mmn) i = interval in contact with source (sec) p = Surveyor efficiency (0.5) d' = index of sensitivity (1.38)

ES = Instrument response (c/m/dpm/100 cm 2 )

Based on the detector response and a background of 36.1 c/m, the estimated MDA for a 1-minute static measurement with the Victoreen 491-30 detector is:

MDA1.miinutecount = [3 + 4.65(36.lc/m)0-5]/[0.0135 c/m/dpm/100 cm2 ]

= 2292 dpm/100 cm 2 The MDA for a scan at the detector movement rate of 3 cm/sec (scan interval of 6.3 cm/3 cm/sec = 2.1 sec) is:

MDAsca = (1.38)[(36.1)/ (2.1/60)]0.5(60/2.1)/ (0.5)05(0.0135)

= 4643 dpm/100 cm 2 Both the static count and scan MDA's are below the adjusted gross DCGL of 6320 dpm/100 cm2 .

Fina Status Survey Plan Addendun 0002Reactor Facilty Rpiing A-3

For piping contaminated predominantly with Cs-137, use of the response factor developed for Co-60 will overestimate the contamination level, because of the higher energy of the Cs-137 beta particles and the resulting increases in detector efficiency and source efficiency.

Fma Status Survey Plan Addendun 002 Reator Fzay Polng A4

Final Status Survey Plan Addendum 003: Pond Sediments Revision 1 Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MH ILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 003: Pond Sediments Revision I Prepared for University of Virginia Reactor Facility Decominissioning Project April 2004 Client Approvals:

OEHS

'4 &Wi30jg2ooy D te AiLLQ a(QAfyy 4s A?&30, 2oo-t Technical Director Date

• CH2MHIll 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 Mriesndw Soca PWn Addo~ 003:Po Se ws

Final Status Survey Plan Addendum 003: Pond Sediments Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 ertified Health Pwsicist Da~e

- /zo /! 7/

Project Manager Date CH2MHILL FwNal Status Survey Plan Addendum 003: Pond Sediments

Contents Contents....

Cotns................................................................................................

1. Introduction ....................................................... 1-1

2. Description ....................................................... 2-1

3. Contaminants of Concern and Guidelines ....................................................... 3-1

4. Survey Approach ....................................................... 4-1 4.1 Survey Reference System ....................................................... 4-1 4.2 Survey Classification ....................................................... 4-1 4.3 Survey Unit Identification ....................................................... 4-1 4.4 Demonstrating Compliance with Release Guidelines ....................................... 4-1 4.5 Number of Required Data Points ....................................................... 4-1 4.6 Sampling Pattern ....................................................... 4-2 4.7 Survey Methods ....................................................... 4-2 4.8 Sample Analyses ....................................................... 4-3 4.9 Investigation ....................................................... 4-3

5. Data Evaluation ....................................................... 5-1 FIGURES 2-1 University of Virginia Reactor Facility and Environs .................................................... 2-2 2-2 Plot Plan of Pond, Indicating Reference Area and Sampling Locations ...................... 2-3 3-1 Results of Analyses of Pond Sediment Samples to Establish Potential Contaminant Mixture ....................................................... 3-2 3-2 Fractional Contribution by Radionuclides Detected in Pond Sediments .................... 3-2 Fira Stus Survey Pan Addendum 001 Pond Sediments

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum..

This addendum (Addendum 003) applies to the sediments in the small on-site pond to the south of the UVAR Facility.

It should be noted that a characterization survey of the pond sediments was conducted during late September 2002 to coordinate drainage of the pond with the time of the year when reduced precipitation is typically expected. The drainage and characterization were also timed to facilitate excavation of the adjacent underground waste tanks (refer to FSS Addendum 002). Design and implementation of the characterization survey of the pond sediments were based on providing adequate data to also enable evaluation as a final status survey, in case the results indicated the decommissioning criteria were satisfied, no remediation was required, and the sediment medium was not subject to potential future contamination during the decommissioning activities performed after the survey.

Rd Status Survey Plan Addendum 003: Pond Sediments 1-1

2. Description Storm runoff from the adjacent land areas and overflow from the storm drain on the UVAR site are collected in a small pond, located to the south of the UVAR Building (see Figure 2-1).

Some laboratory drains, floor drains, and other sources of non-sanitary wastewater with low potential for radiological or other hazardous constituents also routinely discharged to this pond. Two underground waste tanks serviced the Hot Cell, and two tanks were used for collection of demineralizer regeneration liquids from the reactor. Both of these sets of tanks were originally plumbed to allow the contents to be discharged to the pond, provided the liquid met appropriate release criteria following dilution with the pond water; the demineralizer regeneration tanks were later replumbed so they could be discharged directly into the pond spillway.

The pond covers a surface area of about 1450 m2 , and ranges in depth from approximately 2 to 4 m. The pond bottom is covered with sediments, ranging from a few cm to several m thick. Figure 2-2 is a map of the pond, indicating pertinent features.

During the late summer of 2002, the pond was drained. This allowed the sediments to dry and exposed the sediment and various piping surfaces to facilitate radiological monitoring and sampling.

Fil SLalus Survey Plan Addendum 003: Pond Sediments 21

CH2MHILL Figure 2-1 University of Virginia Reactor Facility and Environs i/~-S MIGSPENT IUL

• ~

• i/UNDERROUND*

• / /

• Removed during decommissioning Fbii Stats SurMePbn Addendurn 003: Pond Sediments 2-2

CH2MHILL Figure 2-2 Plot Plan of Pond, Indicating Reference Area and Sampling Locations FainStatus Survey Plan Addendurn DU Pond Sediments 2-3

3. Contaminants of Concern and Guidelines During facility operation, there were several intentional and unintentional discharges of low-level contaminated liquids to the pond. Two of these occurred in laboratories M005 and M008, and involved contamination by Tc-99 and Ni-63, respectively. Reactor Pool water discharges to the pond were made in the 1960's. A break in the piping from the demineralizer regeneration tanks resulted in release of low-level contaminated liquids, containing primarily Cs-137 and Co-60, onto the bank of the pond. Because of this history, there is a potential for the sediments to be contaminated with facility-derived radionuclides.

Pond sediments analyzed during the 1999 GTS Duratek characterization identified positive levels of Cs-137, Co-60, Eu-152, and Pu-241 in some samples; however, many analyses did not identify activity levels above the analytical method detection limits.

Samples of pond sediments obtained during the CH2M HILL characterization did not contain sufficient activity levels (particularly of hard-to-detect radionuclides) to enable a meaningful determination of the contaminant mixture for this survey unit. Table 3-1 summarizes results of analyses of samples selected on the basis of surface gamma scans, gamma borehole logs, and locations relative to drain outfalls. Most analyses indicated radionuclide concentrations below the method detection sensitivity; the only clearly positive results were for Cs-137 (maximum 3.46 pCi/g), Ni-63 (maximum 22.9 pCi/g), and Pu-241 (15.8 pCi/g).

No individual radionuclide concentration in the results in Table 3-1 is above its specific default screening DCGL and none of the samples yielded a sum of fractions value greater than 1 (one). These results further confirm the overall low contamination levels in the pond sediments. There does not appear to be a correlation among the contaminants of interest for the three radionuclides which were identified in the samples. Table 3-2 presents the average activity, activity fraction, and dose contribution fraction for these three radionuclides. Based on these ratios, Cs-137 will be used as a surrogate for all potential contaminants. Pond sediments will be evaluated using the calculated Cs4137 DCGLsurrogate of 5.9 pCi/g.

FMal Status Surey Plan Addendwn 003: Pond Sediments 3-1

CH2MHILL Table 3-1 Results of Analyses of Pond Sediment Samples to Establish Potential Contaminant Mixture Radionuclides Concentration (pCi/g)(a) 131 B1* B2 B2* B3 B3* B4 B5 B6 Am-241 <0.01 - <0.02 - 0.02 - <0.12 <0.13 <0.03 Co-60 <0.15 <0.22 <0.31 <0.17 <0.15 <0.19 <0.15 <0.61 <0.14 Cs-137 <0.15 0.86 2.22 0.64 2.08 <0.17 <0.12 3.46 <0.12 Fe-55 <20.3 - <18.0 - <15.5 - <20.3 <17.9 <13.4 H-3 - <5.41 - <2.77 - <5.88 - - -

I-129 <0.21 - <0.51 - <0.22 - <0.27 <0.45 <0.22 Ni-63 <0.92 - 3.06 - 1.07 - <0.82 22.9 <0.84 Pu-238 <0.02 - 0.02 - <0.02 - <0.12 <0.27 <0.02 Pu-239 <0.02 - 0.02 - <0.02 - <0.25 <0.13 <0.02 Pu-241 <0.72 - <0.71 - 0.95 - 13.1 15.8 1.07 Sr-90 <0.61 - <0.72 - <0.65 - <0.60 <0.86 0.88 Tc-99 <0.75 - <0.76 - <0.76 - <0.76 <0.80 <0.76 (a) Analyses represent maximum activity core interval except for those indicated with (*)

which are averages over 1-meter core interval.

Table 3-2 Fractional Contribution by Radionuclides Detected in Pond Sediments 5Radionuclides Average Activity DCGL (pC/g) Dose Fraction Activity (pCi/g) Fraction Cs-137 1.09 0.0905 11 0.540 Ni-63 4.59 0.4154 2100 0.013 Pu-241 5.37 0.4860 72 0.447 FMS Sbhts Survey Plan Addendurn 0O Pond Sediment 32

4. Survey Approach 4.1 Survey Reference System A 10-m grid was established over the pond area, extending onto the banks surrounding the pond. This grid will be referenced to the reference grid established for survey of the entire site land area, thus enabling the survey locations to be related to the federal and/or state planar coordinate system. Figure 2-2 indicates the reference grid system.

4.2 Survey Classification Based on the facility use history and identification of contaminants of license origin in the pond sediments, the sediments are designated Class 1 for FSS planning and implementation purposes.

4.3 Survey Unit Identification The sediments comprise one survey unit. For survey design purposes the planning area of the survey unit is 1450 m2 .

4.4 Demonstrating Compliance with Release Guidelines Compliance with decommissioning requirements will be demonstrated by comparing the results of characterization/final status survey sample analyses with the Cs-137 surrogate DCGL of 5.9 pCi/g. Because the radionuclides identified as potential contaminants are not present in background at concentrations, which are significant fractions of the release guidelines, correction of FSS sample data for background levels will not be required.

Statistical testing of results will utilize the Sign Test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors will be 0.05 (Type 1 and Type 2).

4.5 Number of Required Data Points DCGL - 5.9, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL - LBGR = 2.95 af = 1 (based on the levels of Cs-137 detected in non-impacted site soil).

A/a = 2.95 From MARSSIM Table 5.5, N = 15; i.e., 15 data points required in the survey unit for statistical evaluation using the Sign test. Actual FSS data will be used to recalculate the relative shift and confirm that adequate data points were obtained for compliance evaluation.

Find Status Survey Plan Addendum 03: Pond Sediments 4-1

CH2MHILL 4.6 Sampling Pattern Sampling was performed at systematically spaced intervals on a triangular pattern throughout the pond. The spacing between data points was determined to be:

L = [1450/(0.866 x 15)] 0.5 = 10.56 m (A spacing of 10.5 m between samples and a spacing of 9.0 m between N-S lines of sampling points was used for ease of field implementation).

A random start point for the pattern was based on survey unit dimensions and random numbers from the MARSSIM random number table of 0.7337 and 0.4872. The resulting start point was 7.3 m W and 4.9 m S.

Systematic sampling locations were:

4.9 S, 7.3 W 4.1 N, 2.0.W 4.9 S, 17.8 W 4.1 N, 12.5 W 4.9 S, 28.3 W 4.1 N, 23.0 W 4.9 S, 38.8 W 4.1 N, 33.5 W 4.9 S,49.3 W 4.1 N, 45.0 W 13.9 S,12.5 W 4.1 N, 55.5 W 13.9 S, 23.0 W 4.1 N, 66.0 W 13.9S,33.5 W 13.1 N 49.3 W In addition to the 16 systematic sample locations, samples were obtained at 18 judgmental

("biased") locations in the vicinity of inlet and outlet piping.

4.7 Survey Methods Gamma walkover surface scans were performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector was maintained within 5-10 cm of the sediment surface and moved from side to side in a serpentine pattern while noting any indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) were documented on survey area maps. Locations of elevated response were noted for further investigation. Scanning coverage was 100% of the sediment surface.

Surface (0-15 cm) sediment samples of approximately 500 g were collected at the 16 systematic and 18 judgmental sampling locations identified above. In soft sediments, sediment columns were obtained by driving PVC pipe to refusal, capping and removing the pipe and extruding the core into a half pipe. In more resistant sediments, boreholes were augered through the sediments to the underlying soil using a 2-in diameter bucket auger.

Some locations required a combination of both methods. The resulting samples were obtained from depths of 15 to 45 cm, 45 to 75 cm, and 75 to 105 cm, where thickness of FMal Status Sumvey Plan Addendum 003: Pond Sediments 4-2

CH2MHILL sediment allowed. If a sample could not be obtained from a pre-identified location, one was obtained from the nearest sediment location available; the survey/sampling record noted this situation. A total of 92 samples were obtained. Duplicate samples were collected at 4 locations. Samples were assigned unique identification numbers and a chain of custody record and analytical request were prepared.

Boreholes were gamma logged at 30 cm intervals from the surface to the bottom of the borehole; where necessary to maintain a borehole open, thin-walled PVC piping was inserted into the borehole as the auger was advanced.

4.8 Sample Analyses Sample cores were scanned for gamma and beta activity. All samples were analyzed in the on-site laboratory by gamma spectrometry. Based on the results of surface scans, borehole logging, sample core scans, and on-site analyses, samples from 6 locations were sent to an off-site commercial laboratory for gamma spectrometry and analysis for hard-to-detect (10 CFR Part 61) radionuclides. Results of these analyses were used to develop a surrogate DCGL for Cs-137. All FSS samples will be analyzed for gamma emitters and results compared with the Cs-137 surrogate DCGL. If analyses identify the presence of other gamma emitters, the assumptions of radionuclide mix and surrogate DCGL will be reevaluated.

4.9 Investigation If sample analyses do not identify radionuclide concentrations in sediments exceeding release criteria, the characterization data will be regarded as FSS data and assessment and statistical evaluation will be performed (refer to Section 5). If sample analyses indicate or statistical data evaluation identifies residual radioactivity concentrations exceeding release criteria, contaminated sediments will be remediated and FSS activities repeated, utilizing a newly determined sampling pattern and random start point.

Fin Status Survey Plan Addendum 001 Pond Sedimerts 4-3

5. Data Evaluation Cs-137 concentrations will be compared with the surrogate DCGL of 5.9 pCi/g. The Sign Test will be performed and the results compared with the critical value for the appropriate number of samples and decision errors.

F'ma Status Suvey Plan AddendUM 003: Pond Sdments 5-1

Final Status Survey Plan Addendum 004: Interior Structure Surfaces Revision 1 Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 004: Interior Structure Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

OEHS Date A ID 2ooit Technical Director Date CH2MHILL 151 Lafayette Drive, Suite 110

. Oak Ridge, TN 37830 ddSo S0vke ra Mde n004: IMecS h~tr Sot ae

Final Status Survey Plan Addendum 004: Interior Structure Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Alth PhyZt~ Dd te i/3/04/

Project Manager Date CH2MHILL Fma Status Survey Plan Addendum 004: Interior Structure Surfaces

Contents' Contents ................................. iv

1. Introduction ................................. 1-1

2. Description ................................ 2-1

3. Contaminants of Concern and Guidelines ............................... 3-1

4. Survey Approach ............................... 4-1 4.1 Survey Reference System ............................... 4-1 4.2 Survey Classification ............................... 4-1 4.3 Survey Unit Identification ............................... 4-1 4.4 Demonstrating Compliance with Release Guidelines ....................................... 4-1 4.5 Background Reference Areas and Materials ......................................................... 4-5 4.6 Number of Required Data Points ............................................................ 4-5 4.7 Sampling Pattern .............................................................. -5 4.8 Survey Methods ............................................................ 4-6 4.9 Sample Analysis ............................................................ 4-6 4.10 Investigation ............................................................. 4-6

5. Data Evaluation ............................................................ 5-1 FIGURES 2-1 University of Virginia Reactor Facility and Environs .................................................... 2-2 2-2 UVA Reactor First Floor Plan View ............................................................ 2-3 2-3 UVA Reactor Mezzanine Floor Plan View.......................................................................24 2-4 UVA Reactor Ground Floor Plan View ............................................................ 2-5 TABLES 4-1 UVAR Building Interior Surface Survey Areas and Classifications .......... 4-2 4-1 UVAR Building Interior Surface Survey Areas and Classifications (continued) ....... 4-3 FiM Stalus SUweyPlan IV Addendum 004: Interior Stucture Surfaces

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 004) applies to the interior surfaces of the UVAR Facility building.

FnStal us Suvey PMn Addendum 004: Intenor Stuxu:e Surtaes 1-1

2. Description The UVAR Facility is located on Old Reservoir Road, approximately 0.6 kilometers (km) west of the Charlottesville, VA city limits. The Facility includes the UVAR building, a small pond, and asphalt-paved roads, parking areas, and equipment/materials storage pads, situated on a land area of approximately 9030 m2 (see Figure 2-1). The three-story building housed the UVA Research Reactor and the CAVALIER facility, as well as offices for the reactor staff and faculty and students of the Department of Nuclear Engineerin&

miscellaneous laboratories, and other support facilities for the reactors and Department of Nuclear Engineering.

Figures 2-2 through 2-4 show the three levels of the UVAR facility. The upper level has approximately 620 m2 of floor area. The Reactor Confinement Room (Rm 131), which housed the former UVA Research Reactor, is located on the upper floor (first floor). This room contained the 9.8 m long by 3.7 m wide by 8.2 m deep reactor pool, associated operating equipment and systems, the operating controls, and some research/experimental equipment. This room is circular and has an elevated (-10 m) ceiling. In addition, the Instrument Shop (Rm 128), Shipping Area (Rrn 127), and multiple offices and other support facilities for staff and students are located on this building level.

On the approximately 670 m2 Mezzanine level were located the Demineralizer (Rm M021),

Mechanical Room (Rm M020), HP Laboratory (Rm. M019), several partially contaminated laboratories (Rms M005 [Tc-99 contamination] and M008 [Ni-63 contamination]), and multiple offices and other support facilities for staff and students. A crawl space (MCS) is accessed from the stairwell on the Mezzanine level.

The 740 m2 ground floor contained the Heat Exchanger (Rm.G024), Rabbit Room (Rm.G005),

Beamport/Experimental area (Rm.G020), Hot Cell (Rms G025, G026, and G027), Counting Room (Rm.G004), Woodworking and Machine Shop (Rm G008), Source Storage (Rmss G022, G018, and G007A), the former CAVALIER facility (Rmn G007), and miscellaneous support facilities and areas.

The UVAR building is of concrete block construction with brick veneer. Floors are concrete slab. Internal walls are block and drywall. Most offices, hallways, and small laboratories have a dropped ceiling of acoustical tile, and tile floors.

In preparation for implementing the Final Status Survey, impacted reactor and support systems and components were removed and disposed of as radioactive waste or surveyed and released for use without radiological restrictions. Contaminated facility surfaces and materials were removed or decontaminated.

Fma Statuis Survey Plan Addendum 004: Intenor Structure Surfaces 2-1

CH2MHILL Figure 2-1 University of Virginia Reactor Facility and Environs CA/

BULDN TRAN/zRArM

• UNDR/ROUND e/v- d

• Removed during decommissioning Fir Staus Survey Plan Addendum 004: Inte>r" Suture Surfaces 2-2

C ( (

1225 129 13112

'1 27 tits 113 116B 14 uS6C 120 109 114 1 15 Figure 2-2 UVA Reactor First Floor Plan View n

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( ~(C M005 0212MOA M004...................

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moI-Figure 2-3 WVA Mezzanine Floor Plan View MI- I

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0006] lG02400. 018 01 000784 a SCO 0007 SfillI 020 008 QOOZ'41 000

.In Figure 2-4 UVA Reactor Ground Floor Plan View

3. Contaminants of Concern and Guidelines The GTS Duratek initial characterization and continuing characterization by the CI2M HILL team showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. Major structural contamination was generally limited to surfaces exposed to or in contact with reactor coolant, reactor neutron fields, and materials containing high levels of activity (e.g., the Hot Cell). Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclide was Co-60 with smaller activities of Cs-137. Remaining structural components did not contain detectable levels of activation products. Ni-63 and Tc-99 contaminants were present on facility surfaces from research projects in labs M008 and M005, respectively.

The Decommissioning Plan established the criteria for residual radioactive material contamination on UVAR facility surfaces. UVAR facility criteria, also referred to as derived concentration guideline levels (DCGLs), are selected from the table of NRC default screening values. Structure surfaces did not have sufficient activity levels to enable a meaningful determination of the facility contaminant mixture - particularly with respect to hard-to-detect radionuclides. Because the nature of contaminants on surfaces should be similar to that in effluents from the reactor facility, the mixture determined for waste tank sludge and facility piping (refer to Addenda 001 and 002) will be assumed. Unless there is evidence to the contrary, the adjusted gross beta DCGL of 6320 dpm/100 cm 2 (see Addendum 002) will be the basis for evaluating the final radiological status of the structure surfaces. Guidelines for removable structure contamination are 10% of the value for total surface activity. This assures a conservative approach for satisfying the NRC dose-based criteria for future facility use. Appendix A of the Master FSS Plan describes the method for establishing guidelines for other radionuclides and combinations of radionuclides. If other guidelines are to be used, a justification will be prepared and included with the FSS documentation.

Fri Status Survey Plan Addendum 004: Interior Sbtucture Surfaes 3-1

4. Survey Approach 4.1 Survey Reference System A grid system will be established on surfaces to provide a means for referencing measurement and sampling locations. On Class 1 and 2 structure surfaces, a 1-m interval grid will be established; a 5-meter interval grid will be established on Class 3 structure surfaces. Upper surface (ceiling and overhead) locations may be referenced to the grid established for the floor beneath. Grid systems will originate at the southwest corner of the survey unit, except where specific survey unit characteristics necessitate alternate grid origins. Grids are assigned alphanumeric indicators to enable survey location identification and are referenced to building features. Maps and plot plans of survey areas will include the grid system identifications. Systems and surfaces of less than 20 m2 will not be gridded, but survey locations will be referenced to prominent facility features.

.4.2 Survey Classification A listing of building interior surfaces and their MARSSIM classifications by contamination potential is contained in Table 4-1. Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities are the bases for these classifications. Classification changes that indicate a lower potential for contamination (and reduced FSS rigor) will require justification and concurrence by the NRC.

4.3 Survey Unit Identification Table 4-1 lists the survey unit, based on MARSSIM classification and the recommended survey unit area limitations, suggested by MARSSIM. Contiguous structure surfaces will be grouped into survey units to satisfy the Class and area criteria. Classifications and survey unit boundaries may change, based on results as the FSS progresses; if classifications or boundaries change, the survey will be redesigned and the survey and data evaluation repeated.

Impacted structure surfaces of

• 10 m2 and impacted land surfaces of

• 100 m2 will not be designated as survey units. Instead, a minimum of 4 measurements (or samples) will be obtained from such areas, based on judgment, and compared individually with the DCGLW.

4.4 Demonstrating Compliance with Release Guidelines The Wilcoxon Rank Sum (WRS) test will be used for evaluating direct measurements of total surface activity, relative to the established criteria, where all survey unit measurements are on the same type of surface medium. Where multiple media are involved, the Sign test will be used. The selection of the test method will be survey unit-specific. The Null Hypothesis will be that activity levels in the survey unit exceed the release criteria. Rejection of the Null FidnStatus Sumt Pla Addendum 004: Interior Stnjcture Surfaces 4A1

CH2MHILL Hypothesis will be required to demonstrate that the release criteria are satisfied. Decision errors will be 0.05 (Type 1 and Type 2).

4.5 Background Reference Areas and Materials For applications of the WRS test, reference areas of the same material as the survey unit being evaluated, but without a history of potential contamination by licensed operations, will be identified. The number of reference data points will be the same (+/- 20%) as the number of data points required from the survey unit. A set of reference measurements will be obtained for each instrument being used for survey unit evaluation. For applications involving the Sign test, sufficient background determinations will be made for each media or surface material and with each instrument to provide an average background level that is accurate to within

+/- 20%; this usually requires 8 to 10 measurements, which are then evaluated using the procedure described in draft NUREG/CR-5849 and additional data points obtained, as necessary. Reference area and background requirements will be survey-unit-specific.

4.6 Number of Required Data Points The following calculation is based on an adjusted gross DCGL of 6320.

DCGL = 6320, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL - LBGR = 3160 a = 500 (based on the MDA for the 43-68 gas flow proportional detector)

A/a = 6.32 (adjusted to 3 per MARSSIM guidelines)

From MARSSIM Table 5.3, N/2 = 10; i.e., 10 data points each required for the survey unit and the reference area (WRS test). If the Sign test is to be used, 14 data points will be required for the survey unit. Actual survey data will be used to recalculate the relative shift and confirm an adequate number of survey locations to evaluate compliance.

4.7 Sampling Pattern Sampling/measurements will be performed at systematically spaced intervals on triangular patterns throughout the soil and paved areas. The spacing between data points will be determined by the area of the survey unit and will therefore be survey unit-specific.

Random start points for the systematic sampling pattern will be determined for each survey unit, based on the overall survey unit dimensions and random numbers from the MARSSIM random number table.

In addition to the systematic locations, samples and measurements will be obtained at "biased" locations, identified by scanning or professional judgment of the FSS field supervisor as having the greatest potential for contamination. Examples of such locations include work area where radioactive materials were handled, high traffic areas, locations which required remediation to reduce or remove residual activity, and locations of elevated direct radiation, identified during the FSS.

Frd Status SurreyfPn AMendum 004: Interbr Scnture Surfaces 2-5

(j ( CH2MC.-LL Table 4-1 UVAR Building Interior Surface Survey Areas and Classifications Room or Area Surface Class Approximate No. of Remarks Surface Area Survey Units 131 Reactor Room Floor 1 130 2 131 Reactor Room Lower Walls 1 100 1 Reactor Pool Floor and Walls 1 150 2 M005/005A Floor and Lower Walls 1 45 1 M008 Floor and Lower Walls 1 60 1 M019 Floor and Lower Walls 1 80 1 M020 Floor and Lower Walls 1 85 1 M021/021A Floor, Walls, and Ceiling 1 100 1 Bio Shield Surfaces Wall 1 100 1 G005 Floor, Walls, and Ceiling 1 85 1 G007/GO07A Floor, Pit and Lower Walls 1 100 1 G018 Floor, Walls, and Ceiling 1 110 1 G020 Floor and Lower Walls 1 300 3 G022 Floor, Walls, and Ceiling 1 60 1 G024 Floor, Walls, and Ceiling 1 100 1 _

G025/G026/G027 Floor, Walls, and Ceiling 1 70 1 Final Status Survey Plan Addendurn 004: Interdr Structure Surfaces 4.3

Q ( CH2MC..L Table 4-1 UVAR Building Interior Surface Survey Areas and Classifications (continued)

Room or Area Surface Class Approximate No. of Remarks Surface Area Survey (m2) Units 131 Reactor Room Upper Walls and 2 420 1 Ceiling 127/128/130 Floor, Walls, and Ceiling 2 180 1 107/124/124A/124B Floor and Lower Walls 2 250 1 M005/005A Upper Walls and 2 30 1 Ceiling.

M008 Upper Walls and 2 40 1 Ceiling -

M019 Upper Walls and 2 60 1 Ceiling M020 Upper Walls and 2 65 1 Ceiling .

M006/M014/M015/M030/M031 Floor and Lower Walls 2 250 1 Includes catwalk over G020 MCS (crawl space) Floor, Walls, and Ceiling 2 100 1 Soil samples per Addendum 006 G004/GO05A Floor and Lower Walls 2 100 1 G006 Floor and Lower Walls 2 70 1 G007B/G008/G008A/G016/G017/G019 Floor and Lower Walls 2 150 1 Stairwell 1 Floor and Lower Walls 2 300 1 Stairwell 2 Floor and Lower Walls 2 300 1 Remainder of structure Floors, walls, and 3 4500 3 Ceiling FPnal Status Survey Plan Mdendurn 004: Interor Srcture Suraesm 44

CH2MHILL .

4.8 Survey Methods Gamma walkover surface scans will be performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlurn Model 2221 ratemeter/scaler. The detector will be maintained within 5-10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Gamma scanning coverage will be 100% of Class 1 floor surfaces and a minimum of 25% for Class 2 and 10% for Class 3 floor surfaces. Where conditions allow, beta scans of structure floor surfaces will be performed using a large area (-580 cm 2 ) gas proportional detector (Ludlum Model 43-37) coupled with a Ludlum Model 2221 ratemeter/scaler and Model 239-1 floor monitor; smaller Ludlum Model 43-68 gas proportional or Ludlum Model 44-9 pancake GM detectors coupled with Ludlum Model 2221 ratemeter/scalers will be used, as required by accessibility limitations. Ludlum Model 43-68 gas proportional or Ludlum Model 44-9 pancake GM detectors coupled with Ludlum Model 2221 ratemeter/scalers will be used to scan other structure surfaces. The detector will be maintained within -1 cm of the surface while advancing the detector at a rate of approximately one detector width per second. Scan speed will be adjusted, as necessary, to assure detection sensitivities are less than 50% of the release criteria. Audible response will be monitored for indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Beta scanning coverage for wall and ceiling surfaces will be 100% of Class 1 surfaces and a minimum of 25% for Class 2 and 10% for Class 3 surfaces.

Surface activity measurements will be performed at the systematic and judgmental locations (see Sections 4.6. and 4.7); 1-minute static measurements will be conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler.

Smears for removable activity will be performed at locations of direct activity measurements.

4.9 Sample Analyses Smears will be analyzed for gross alpha and gross beta activity in the on-site counting facility.

4.10 Investigation If measurements, sample analyses, or statistical data evaluation identifies residual radioactivity exceeding 50% of the applicable DCGL, the source of the contamination will be investigated. Remediation will be performed, as necessary. The survey unit will be reclassified in accordance with the Master FSS Plan and FSS activities repeated, utilizing a newly determined sampling pattern and random start point.

Fial StaUs Survey Plan Addendum 004: Interr Sbucdure Surfaces 45

5. Data Evaluation Measurements will be compared with adjusted gross DCGL of 6320 dpm/100 cm 2 using the WRS test or Sign test, depending on the particular survey unit design. Results will be compared with the critical value for the appropriate number of samples and decision errors.

Judgmental measurements and measurements from non-survey unit surfaces will be individually compared with the adjusted gross DCGL.

Fmal Stat Survey Plai Addencdum 004: Interior Sbuctue Surfaces 5-1

Final Status Survey Plan Addendum 005: Exterior Soil and Paved Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2M HILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 005: Exterior Soil and Paved Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

6?t Date 3X9, real1 Date FoAQ' Aq\SQ30 20oLD Date Technical Director CHI2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 FwnStaM &swey Plan Adndn005-Exberbr So aid P&ed 9wfam

Final Status Survey Plan Addendum 005: Exterior Soil and Paved Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 rtified Health Physit Date Project Manager Date CH2MHILL Final Status Survey Plan Addendum 005: Extenor SoAand Paved Surfaes

Contents Contents ............ iii

1. Introduction ........... 1-1

2. Description .2-1

3. Contaminants of Concern and Guidelines 1. 3-

4. Survey Approach .4-1 4.1 Survey Reference System .................... 4-1 4.2 Survey Classification .4-1 4.3 Survey Unit Identification .4-1 4.4 Demonstrating Compliance with Release Guidelines .4-1 4.5 Number of Required Data Points .4-2 4.6 Sampling Pattern .4-2 4.7 Survey Methods .4-3 4.8 Sample Analyses .4-4 4.9 Investigation............................................................................................................4-4

5. Data Evaluation .5-1 FIGURES 2-1 University of Virginia Reactor Facility and Environs .2-2 2-2 Plot Plan of Site, Indicating Reference Grid System, and Measurement/

Sampling Locations .2-3 Final Stafts Survey Plan IV Addendum 005: Eteror Sol ad Paed Strazes

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 005) applies to the open land and paved exterior surfaces surrounding the UVAR Facility building.

Fra Stat Survey Plan Addendum 005: Exterior Sci and Paved Surfaces 1-1

2. Description The UVAR Facility is located on Old Reservoir Road, approximately 0.6 kilometers (km) west of the Charlottesville, VA city limits. The Facility includes the 880 m2 UVAR building footprint, a small (-1600 m2 ) pond, and asphalt-paved roads, parking areas, and equipment/materials storage pads, situated on a land area of approximately a 9390 M2 (see Figure 2-1). The site terrain generally slopes from north to south. The east and south portions of the site are wooded; the northern portion of the site surface is dominated by rock outcroppings. A low (-1 m high) fence encompasses the site.

During Facility operations, there were two underground liquid-waste collection tanks and two underground Hot-Cell Drain Tanks, located southeast of the building, near the edge of the pond. These tanks have been removed and the excavation has been surveyed in accordance with FSS Addendum 001. The pond received site runoff and some facility liquid releases during operations. This pond has been drained and sediments were surveyed in accordance with FSS Addendum 003. The remainder of the exterior property is addressed by this Addendum.

Frd SaMus Survey Plan Addendun 005: Exterior Sol and Paved Surfaces 2.1

CH2MHILL

/

• k ,

SPETnUEL TRNSFER TANK

.I i1 Figure 2-1 University of Virginia Reactor Facility and Environs Final Status Survey Plan Addendum 005: Extefi Soi and Paved Surfaes 2-2

( ( CH2MEL Figure 2-2 Plot of Site, Indicating Reference Grid System, and Measurement/ Sampling Locations.

Final Status Survey Plan Addendum 005: Exteror Soil and Paved Surfaces 2.3

3. Contaminants of Concern and Guidelines During facility operation, several small spills of contaminated liquids occurred in the vicinity of the waste collection systems. Equipment, materials, and wastes with a potential for low-level contamination were stored on surfaces south of the building during facility operations and in connection with the facility remediation. In addition, several liquid discharge points from the building to the pond terminate on the hillside north of the pond.

Initial characterization by GTS Duratek and follow-on monitoring during the decommissioning actions included scanning and sampling of potentially affected surfaces.

Cs-137 was identified in surface soil, but at concentrations typical of background soil.

Surveys of waste tank excavations and pond sediments have identified Co-60 and Cs-137 as the dominant contaminants from facility operations. Significant levels of other site-related radionuclides were not identified by this monitoring; adequate activity levels were not available to enable meaningful determination of a radionuclide mixture.

Release levels for site are the NRC default screening criteria; the default screening criteria concentrations for Cs-137 and Co-60 are 11 pCi/g and 3.8 pCi/g, respectively. Default screening criteria surface activity is 28,000 dpm/100 cm2 for Cs-137 and 7100 dpm/100 cm 2 for Co-60. For the final status survey, soils will be analyzed for specific gamma emitting radionuclide contaminants, and results must satisfy the Unity Rule for the sum of ratios of radionuclide concentrations present to respective screening default guideline levels. In addition, a composite sample will be analyzed for hard-to-detect radionuclides to confirm the absence of significant levels of these contaminants. The restrictive adjusted gross DCGL of 6320 dpm/cm2 will be used to demonstrate compliance for contamination on paved surfaces (refer to Addendum 002).

FRd Status Survey Plan Addendum 005: Exteri SWl ard Paved Surfaces 3-1

4. Survey Approach 4.1 Survey Reference System A 10-meter grid has been established over the entire site. This grid has been referenced to the federal planar coordinate system. Figure 2-2 indicates the reference grid system. Further grid identification (e.g., northing and easting from a southwest origin point) will be assigned to each node to facilitate location of sampling/measurement points.

4.2 Survey Classification Based on the facility use history, characterization, and remediation control monitoring, the exterior soil and paved surfaces sediments are designated Class 3 for FSS planning and implementation purposes.

4.3 Survey Unit Identification For survey design purposes the planning area of the total site (excluding the pond and building footprint) is 6860 m2 . The site is comprised of two survey units; one is the paved surfaces of approximately 2500 M2 , and the other is the soil surfaces of approximately 4360 iM2 .

4.4 Demonstrating Compliance with Release Guidelines Compliance with decommissioning requirements will be demonstrated by comparing the results of final status survey measurements and sample analyses with default screening criteria. For soil samples, the sum of ratios of identified radionuclides will be used; the sum of ratios must satisfy the Unity Rule. The radionuclides identified as potential contaminants are not present in background at concentrations, which are significant fractions of the release guidelines. Therefore, correction of FSS sample data for background levels will not be required. Statistical testing of results will utilize the Sign Test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors will be 0.05 (Type 1 and Type 2).

For activity measurements on paved surfaces measurements will be compared with the adjusted gross DCGL criteria of 6320 dpm/100 cm 2 . Because the direct measurement procedure includes inherent detector background contributions, correction of FSS measurement data for background levels will be required, through use of an appropriate reference area of similar paving material in a non-impacted location or by using the shielded/unshielded measurement technique. Statistical testing of results will utilize the WRS Test or Sign test to reject or accept the null hypothesis that the residual contamination exceeds the release criteria. Decision errors will be 0.05 (Type 1 and Type 2).

Fua SItus Srvey Pan Addendum 005 Elderor Scl and Paved Surfaces 4-1

CH2MHILL 4.5 Number of Required Data Points Soil (Samples)

The following calculation is based on use of Unity Rule (i.e., DCGL = 1).

DCGL = 1, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL- LBGR = 0.5 cf = 0.25 (based on the sum of ratios of maximum levels of radionuclides detected in characterization samples to respective DCGL's)

A/cr = 2 From MARSSIM Table 5.5, N = 15; i.e., 15 data points required for the survey unit for Sign Test.

Paved Surfaces (Activity Measurements)

The following calculation is based on an adjusted gross DCGL of 6320 dpm/100 cm 2 .

DCGL = 6320, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL - LBGR = 3160 a = 500 (based on the MDA for the 43-68 gas flow proportional detector)

A/a = 6.32 (adjusted to 3 per MARSSIM guidance)

From MARSSIM Table 5.3, N/2 = 10; i.e., 10 data points each required for the survey unit and the reference area for WRS Test (14 data points will be required if the Sign test is used).

Actual FSS data will be used to recalculate the relative shift and confirm adequate data points for evaluation 4.6 Sampling Pattern Sampling/measurements will be performed at systematically spaced intervals on triangular patterns throughout the soil and paved areas. The spacing between data points for soil is:

L = [4360/(0.866 x 15)] 0.5 = 18.3 m (A spacing of 18 m between samples and 15 m between E/W sampling lines will be used for ease of field implementation).

The spacing between data points for paved surfaces is:

L = [2500/(0.866 x 10] 0.5 = 16.99 m (A spacing of 17 m between samples and 14 mn between E/W sampling lines will be used for ease of field implementation). If the Sign test is used, the spacing will be adjusted to assure the larger number of samples.

Random start points for the systematic sampling patterns have been selected, using overall survey unit dimensions of 100 mn x 140 m and random numbers from the MARSSIM random number table. The resulting start point for soil area survey is 39 N, 84 E; the start point for paved surface survey is 46 N, 27 E (refer to Figure 2-2).

Based on these parameters the following systematic sampling/measurement data points have been identified (refer to Figure 2-2):

Firmi Stalus Survey Plan Addendum 005: Exterior Soil and Paved Surfa=es 42

CH2MHILL Soil Surfaces Paved Surfaces 9N,30E 32N,18.5E 9 N,48 E 32 N, 35.5 E 9 N, 66 E 32 N, 52.5 E 24 N, 21 E 32 N, 137.5 E 24N,57E 46N,10E 24N,111 E 46 N, 61 E 39 N, 84 E 46 N, 78 E 54 N, 21 E 46 N, 129 E 54 N, 93 E 60N,18.5E 54 N, 111 E 60 N, 86.5 E 69 N, 12 E 60 N,120.5 E 69 N, 84 E 74 N, 27 E 69N,102E 74 N, 61 E 84 N, 39 E 74 N, 95 E 84 N, 93 E 88 N, 52.5 E 99 N, 48 E 88 N, 69.5 E 99 N, 66 E In addition to the systematic sample/measurement locations, samples and measurements will be obtained at "biased" locations, identified by scanning or professional judgment of the FSS field supervisor as having the greatest potential for contamination. Examples of such locations include egress points from the reactor room on the upper level, egress points from the beamport/experimental facilities on the ground level, collection areas for natural drainage pathways, and drainage outfalls to the pond hillside.

4.7 Survey Methods Gamma walkover surface scans will be performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector will be maintained within 5-10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Gamma scanning coverage will be a minimum of 50% of the soil and paved surfaces. Beta scans of paved surfaces will be performed using a large area (-580 cm 2 ) gas proportional detector Fnal Staus Survey Plan Addendum 005: Exterior Sal and Paved Surfaces 4-3

CH2MHILL (Ludlum Model 43-37) coupled with a Ludlum Model 2221 ratemeter/scaler and Model 239-1 floor monitor. The detector will be maintained within -1 cm of the surface while advancing the detector at a rate of approximately one detector width per second. Audible response will be monitored for indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Beta scanning coverage will be a minimum of 50% of the paved surfaces.

Surface (Oto 15 cm) soil samples of approximately 500 g will be collected at the systematic and 4 to 6 judgmental sampling locations (see Sections 4.5. and 4.6). Surface activity measurements will be performed at the systematic and 4-6 judgmental locations (see Sections 4.5. and 4.6); 1-minute static measurements will be conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler.

Samples will be assigned unique identification numbers and a chain of custody record and analytical request will be prepared.

4.8 Sample Analyses Soil samples will be analyzed by an off-site commercial laboratory for gamma emitters (gamma spectrometry) spectrometry. A composite will be analyzed for hard-to-detect (10 CFR Part 61) radionuclides.

4.9 Investigation If measurements, sample analyses, or statistical data evaluation identifies residual radioactivity concentrations exceeding 50% of the release criteria, the source of the contamination will be investigated. If results are confirmed, remediation will be performed, as necessary, the impacted areas reclassified, and FSS activities repeated, utilizing a newly determined sampling pattern and random start point.

Fra Status Survey Plan Addendum 005: Exterior Soa and Paved Surfaces 4-4

5. Data Evaluation For soil samples, the sum of ratios will be calculated for gamma emitting radionuclides, which are of facility origin. The Sign Test will be performed for systematic samples and the results compared with the critical value for the appropriate number of samples and decision errors. Judgmental sample results will be individually compared with DCGL's using the Unity Rule.

Systematic paved surface measurements will be compared with the adjusted gross DCGL of 6320 dpm/100 cm 2 and results will be compared with the critical value of the WRS (or Sign) test for the appropriate number of samples and decision errors. Judgmental measurement results will be individually compared with the adjusted gross DCGL.

Final Status Survey Plan Addendum 005: Exleror Sol and Paved Surfaces 51

Final Status Survey Plan Addendum 006: Exterior Structure Surfaces Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2BlHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

• Final Status Survey Plan Addendum 006: Exterior Structure Surfaces Revision I Prepared for.

University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

4HA§20 051 OEHS Date a 0 2Ote Technical Director Date is CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830

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Final Status Survey Plan Addendum 006: Exterior Structure Surfaces Revision I Prepared for University of Virginia Research Reactor Facility April 2004 H -!9o/ eF rtified Health Physi W'te/e Project Manager Date CH2MHILL Fmal Status Survey Pan Addendun 006: Exterior Shcture Sufaces

Contents Contents . ...........................................................

ii

1. Introduction ........................................................... 1-1

2. Description ............................................................ 2-1

3. Contaminants of Concern and Guidelines ........................................................... 3-1

4. Survey Approach ........................................................... 4-1 4.1 Survey Reference System ........................................................... 4-1 4.2 Survey Classification ........................................................... 4-1 4.3 Survey Unit Identification ........................................................... 4-1 4.4 Demonstrating Compliance with Release Guidelines ............................................. 4-1 4.5 Number of Required Data Points ........................................................... 4-1 4.6 Sampling Pattern ........................................................... 4-2 4.7 Survey Methods ........................................................... 4-2 4.8 Sample Analyses ........................................................... 4-3 4.9 Investigation ........................................................... 4-3

5. Data Evaluation ........................................................... 5-1 FIGURES 2-1 University of Virginia Reactor Facility and Environs .................................................... 2-2 2-2 UVA Reactor Floor Plan View Indicating Roof Areas ................................................... 2-3 2-3 UVA Reactor First Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ........................................................... 2-4 2-4 UVA Mezzanine Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ........................................................... 2-5 2-5 UVA Reactor Ground Floor Plan View Indicating Survey Locations on Exterior Building Surfaces ........................................................... 2-6 Fmal Stats Survey Pbn IV Addendum 06: Eeor Stuchre Surfds

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each

'impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 006) applies to the exterior surfaces of the UVAR Facility building. Other related Addenda include 004 (Interior Structure Surfaces) and 005 (Exterior Soil and Paved Areas).

FrmdStatus Survey Plan Addendum 006: Exterir Structure Surfaces 1-I

2. Description The UVAR Facility is located on Old Reservoir Road, approximately 0.6 kilometers (km) west of the Charlottesville, VA city limits. The Facility includes the UVAR building, a small pond, and paved roads, parking areas, and equipment/materials storage pads, situated on a land area of approximately 9390 m2 (see Figure 2-1). The three-story building housed the UVA Research Reactor and the CAVALIER facility, as well as offices for the reactor staff and faculty and students of the Department of Nuclear Engineering, miscellaneous laboratories, and other support facilities for the reactors and Department of Nuclear Engineering.

Figure 2-2 is a plot plan of the UVAR building. The UVAR building is of concrete block construction with brick veneer. Floors are concrete slab. There is approximately 1190 m2 of roof area, at two elevations; one covers the Reactor Confinement structure - a surface area of approximately 175 m2 , and the other (approximately 1015 m2) covers the remainder of the structure. During operation there was a cooling tower on the roof to the southeast of the Reactor Room; this structure was removed during decommissioning. Roofs are of tar-and-gravel composition. The roofs are essentially clear of obstructions such as items of HVAC equipment. There are multiple sewer line vents and rainwater drains on the roofs.

Other exterior building surfaces of concern include discharge grills and stacks servicing small laboratory exhaust ventilation systems; some of these, e.g., those from rooms M005 and M008, were known to have at one time been contaminated. Doors at exits from areas handling radioactive and/or potentially contaminated materials are also surfaces of interest.

These exterior locations are identified on figures 2-3 to 2-5.

In preparation for implementing the Final Status Survey, impacted reactor and support systems and components were removed and disposed of as radioactive waste or surveyed and released for use without radiological restrictions. Additional characterization surveys were performed to identify potentially contaminated surfaces; any such surfaces were removed or decontaminated.

FAendSta:us Sven Plbn Addendum D06: Exterior Struture Surfaces 2-1

CH2MHILL Figure 2-1 University of Virginia Reactor Facility and Environs 0~

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N 0 Motors Confinement Room Roof Main Building Roof Figure 2-2 UVA Reactor Floor Plan View Indicating Roof Areas.

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00000h 00007 ExexttDDoor Figure 2-5 UVA Reactor Ground Floor Plan View Indicating Survey Locations on Exterior Building Surfaces.

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3. Contaminants of Concern and Guidelines The GTS Duratek initial characterization and continuing characterization by the CH2M Hill team showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. Major structural contamination was generally limited to surfaces exposed to or in contact with reactor coolant, reactor neutron fields, and materials containing high levels of activity (e.g. the Hot Cell). Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclide was Co-60 with smaller activities of Cs-137. Remaining structural components did not contain detectable levels of activation products. Ni-63 and Tc-99 contaminants were present on facility surfaces from research projects in labs M008 and M005, respectively. Characterization did not identify any contamination of exterior building surfaces.

The Decommissioning Plan established the criteria for residual radioactive material contamination on UVAR facility surfaces. UVAR facility criteria, also referred to as derived concentration guideline levels (DCGLs), are selected from the table of NRC default screening values. Structure surfaces (interior and exterior) did not have sufficient activity levels to enable a meaningful determination of the facility contaminant mixture -

particularly with respect to hard-to-detect radionuclides. Therefore, unless there is evidence to the contrary, the adjusted gross beta DCGL of 6320 dpm/lOOcm 2 will be the basis for evaluating the final radiological status of the exterior structure surfaces. Guidelines for removal structure contamination are 10% of the value for total surface activity. This assures a conservative approach for satisfying the NRC dose-based criteria for future facility use.

Appendix A of the Master FSS Plan describes the method for establishing guidelines for other radionuclides and combinations of radionuclides. If other guidelines are to be used, a justification will be prepared and included with the FSS documentation.

Final Staus Survey Plan Addendum 006 Exterior Structure Surfaces 3-1

4. Survey Approach 4.1 Survey Reference System A grid system will be established on roof surfaces to provide a means for referencing measurement and sampling locations. On Class 1 and 2 structure surfaces, a 1-meter interval grid will be established; a 5-meter interval grid will be established on Class 3 structure.

surfaces. Vertical and overhead surfaces may be referenced to the grid established for the area beneath. Grid systems will originate at the southwest corner of the survey unit, except where specific survey unit characteristics necessitate alternate grid origins. Grids are assigned alphanumeric indicators to enable survey location identification and are referenced to building features. Maps and plot plans of survey areas will include the grid system identifications. Systems and surfaces of less than 20 m2 will not be gridded, but survey locations will be referenced to prominent facility features.

4.2 Survey Area Classification The roofs (main building and Reactor Confinement Room) are designated MARSSIM Class 2 surfaces; other exterior surfaces are designated Class 3. Facility history (including the Historic Site Assessment) and radiological monitoring conducted during characterization and remedial activities are the bases for these classifications.

4.3 Survey Unit Identification Based on the MARSSIM classification by contamination potential and the survey unit area limitations, suggested by MARSSIM, the following survey units have been identified:

• Reactor Confinement Room roof

• . Main building roof Impacted structure surfaces of *10 m2 and impacted land surfaces of

• 100 m2 will not be designated as survey units. Instead, a minimum of 4 measurements will be obtained from such areas, based on judgment, for comparison individually with the DCGLs. Such surfaces include exterior surfaces of vents, stacks, and doors, exiting from areas of former radioactive materials use and facilities requiring remedial action during this decommissioning project.

Classifications and survey unit boundaries may change, based on results as the FSS progresses; if classifications or boundaries change, the survey will be redesigned and the survey and data evaluation repeated.

Final Status Svey Plan WVAR Addendum 006: Ext"r Stucture Surfaces 4-1

4.4 Demonstrating Compliance with Release Guidelines The Wilcoxon Rank Sum (WRS) test will be used for evaluating direct measurements of total surface activity, relative to the established criteria, where all survey unite measurements are on the same type of surface medium. Where multiple media are involved, the Sign test will be used. The selection of the test method will be survey unit-specific. The Null Hypothesis will be that activity levels in the survey unit exceed the release criteria. Rejection of the Null Hypothesis will be required to demonstrate that the release criteria are satisfied. Decision errors will be 0.05 (Type 1 and Type 2).

4.5 Background Reference Areas and Materials For applications of the WRS test, a reference area of the same material as the survey unit being evaluated, but without a history of potential contamination by licensed operations, will be identified. The number of reference data points will be the same (+/- 20%) as the number of data points required from the survey unit. A set of reference measurements will be obtained for each instrument being used for survey area evaluation. For applications involving the Sign test, the shielded/unshielded measurement approach will be used.

Sufficient background determinations will be made for each media or surface material and with each instrument to provide an average background level that is accurate to within +/-

20%; this usually requires 8 to 10 measurements, which are then evaluated using the.

procedure described in draft NUREG/CR-5849 and additional data points obtained if necessary. Reference area and background requirements rill be survey-unit specific.

4.6 Number of Required Data Points The following calculation is based on adjusted gross DCGL of 6320.

DCGL = 6320, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL- LBGR = 3160 a = 500 (based on the MDA for the 43-68 gas flow proportional detector)

A/ea = 6.32 (adjusted to 3 per MARISSM guidelines)

From MARSSIM Table 5.3, N/2 = 10; i.e., 10 data points each required for the survey unit and the reference area (WRS test). If the Sign test is to be used, 14 data points will be required for the survey unit. Actual survey data will be used to recalculate the relative shift and confirm an adequate number of survey locations to evaluate compliance.

4.7 Sampling Pattern Measurements will be performed at systematically spaced intervals on triangular patterns throughout the roof surfaces. The spacing between data points for the main building roof is:

L = [1015/ (0.866 x 10)] 0-5 = 10.8 m (A spacing of 10 m between samples will be used for ease of field implementation).

The spacing between data points for the Reactor Confinement Room roof is:

Fidnd Sutrve Pun 0a0us Addendum 006: Extedir Stntue Surfaces 4-2

L = [175/(0.866 x 10] 05 = 4.5 m (A spacing of 4 m between samples will be used for ease of field implementation).

Intervals will be adjusted to assure adequate data points if the Sign test is to be used.

Random start points for the systematic sampling patterns on the roofs have been selected, using survey unit dimensions and random numbers (N-0.501578, E-0.204221 and N-0.644294, E-0.821341) from the MARSSIM random number table.

In addition to the systematic locations, measurements will be obtained at "biased" roof locations, identified by scanning or professional judgment of the FSS field supervisor as having the greatest potential for contamination. On other building exterior surfaces, measurements will be obtained at "biased" locations, identified by scanning or professional judgment of the FSS field supervisor as having the greatest potential for contamination.

4.8 Survey Methods Gamma walkover surface scans will be performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector will be maintained within 5 to 10 cm of the surface and moved from side to side in a serpentine pattern while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Gamma scanning coverage will be 100% of Class 1 surfaces and a minimum of 25% for Class 2 and 10% for Class 3 floor surfaces. Where conditions allow, beta scans of roof surfaces will be performed using a large area (-580 cm2 ) gas proportional detector (Ludlum Model 43-37) coupled with a Ludlum Model 2221 ratemeter/scaler and Model 239-1 floor monitor; smaller Ludlum Model 43-68 gas proportional or Ludlum Model 44-9 pancake GM detectors coupled with Ludlum Model 2221 ratemeter/scalers will be used, as required by accessibility limitations. Ludlum Model 43-68 gas proportional or Ludlum Model 44-9 pancake GM detectors coupled with Ludlum Model 2221 ratemeter/scalers will be used to scan other exterior structure surfaces. The detector will be maintained within -1 cm of the surface while advancing the detector at a rate of approximately one detector width per second. Scan speed will be adjusted, as necessary, to assure detection sensitivities are less than 50% of the release criteria. Audible response will be monitored for indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Beta scanning coverage for wall surfaces will be 100% of Class 1 surfaces and a minimum of 25% for Class 2 and 10% for Class 3 surfaces.

Surface activity measurements will be performed at the systematic and judgmental locations (see Sections 4.6. and 4.7); 1-minute static measurements will be conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler.

Based on ambient background variability and surface media, unshielded and shielded measurements may be performed at each survey unit data point.

Smears for removable activity will be performed at locations of direct activity measurements.

Final Status Survey Plan Addendum 006: Exlerior Structure Surfaces 4-3

4.9 Sample Analyses Smears will be analyzed for gross alpha and gross beta activity in the on-site counting facility.

4.10 Investigation If measurements, sample analyses, or statistical data evaluation identifies residual radioactivity exceeding 50% of the release criteria, the source of the contamination will be investigated. Remediation will be performed, as necessary. The survey unit will be reclassified in accordance with the Master FSS Plan and FSS activities repeated, utilizing a newly determined sampling pattern and random start point.

Frd Status Survey PMm Addendum 006: Exteru Shnxtue Suraes 44

5. Data Evaluation Measurements will be compared with release criteria using the WRS or Sign test. Results will be compared with the critical value for the appropriate number of samples and decision errors. Judgmental measurements and measurements from non-survey unit surfaces will be individually compared with DCGL.

Final Status Survey Plan UVAR Addendurn 006: Eterior Struture Swfaces 5-1

Final Status Survey Plan Addendum 007: Special Soils Areas Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette Drive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

Final Status Survey Plan Addendum 007: Special Soils Areas Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project April 2004 Client Approvals:

IZ1al-6-t OEHS Date TedmicaQ DiretrQ e A?'Lte ZOLf Technical Director Date CH2MHILL 151 Lafayette Drive, Suite 110.

Oak Ridge, TN 37830 Fri StaWus SWmey Pm A~kiuW 007: Sp SofthMm

Final Status Survey Plan Addendum 007: Special Soils Areas Revision 1 Prepared for University of Virginia Reactor Facility Decommissioning Project 4..v/Fu-12OO4 Da{/, I tified Health Physi Project Manager Date CH2m HILL Status Survey Pgn dnal Addendum 007: SpeciW Soils Aeas

Contents Contents .... ............................................................. iii 1...I ntroduction .................................................................................................................................. 1-1 2.Description .................................................................................................................................... 2-1

3. Contaminants of Concern and Guidelines ................................................................. 3-1

4. Survey Approach ................................................................. 4-1 4.1 Survey Reference System .................... 4-1 4.2 Survey Classification ..................... 4-1 4.3 Survey Area Identification .................... 4-1 4.4 Demonstrating Compliance with Release Guidelines .......................................................... 4-1 4.5 Number of Data Points and Sampling Patter ................................................................ 4-1 4.6 Survey Methods ................................................................. 4-3 4.7 Sample Analyses ................................................................ 4-3 4.8 Investigation ................................................................ 4-3 5.Data Evaluation ................................................................. 5-1 TABLES Table 3-1 Radionuclide Concentration on the Surface of Fill Soil around the Reactor Pool 3-1 Table 3-2 Radionuclide Concentration in Subsurface Fill Soil around the Reactor Pool ....... 3-2 FIGURES 2-1 University of Virginia Reactor Facility and Environs.. ............ 2-3 2-2 UVA Reactor First Floor Plan View Indicating the Location of Soil Fill Around Reactor Pool ................................................................ 2-4 2-3 UVA Mezzanine Floor Plan View Indicating the Location of the Mezzanine Crawl Space ................................................................ 2-5 2-4 UVA Reactor Ground Floor Plan View Indicating the Location of the Soils Beneath the Reactor Pool ................................................................ 2-6 Master Fina Status Swe im Addend um 007: Speci Sofs Areas

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 007) applies to the soils in the Mezzanine crawl space, in the space between the reactor pool and the walls of the reactor room, and beneath the reactor pool.

Non-soil structure surfaces in the crawl space will be surveyed in the same manner as other building interior surfaces. Addendum 004: InteriorStructure Surfaces describes the process for FSS of those surfaces.

Fina Status Survey Plan Addendum 007: Spedal Soas Areas 1-1

2. Description The UVAR Facility is located on Old Reservoir Road, approximately 0.6 kilometers (km) west of the Charlottesville, VA city limits. The Facility includes the UVAR building, a small pond, and roads, parking areas, and equipment/materials storage pads, situated on a land area of approximately 9390 m2 (see Figure 2-1). The three-story UVAR building housed the UVA Research Reactor and the CAVALIER facility, as well as offices for the reactor staff and faculty and students of the Department of Nuclear Engineering, miscellaneous laboratories, and other support facilities for the reactors and Department of Nuclear Engineering.

The UVAR building is of concrete block construction with brick veneer. Floors are concrete slab; the floors of facilities with heavy equipment use and/or higher potential for radioactive contamination, including the Reactor Room, ground level experimental (beam port) area, maintenance facilities, and workshops are painted or bare concrete. Internal walls are block and drywall. Most offices, hallways, and small laboratories have a dropped ceiling of acoustical tile, and tiled floors. There are also several soils areas inside the building, which have a potential for radioactive contamination, based on the operating history of the facility. One of these is a small crawl space adjacent to the Reactor Confinement Room. This space, located between the first and Mezzanine levels, is accessed from the stairwell between these two floors. The crawl space is of masonry construction with a dirt (soil) floor covering an area of approximately 50 m2 . This crawl space was used for storage of equipment, materials, and supplies, including some radioactive sources and potentially contaminated components and miscellaneous materials. Characterization surveys of this crawl space identified slightly elevated direct radiation levels, due to the masonry construction and the presence of elevated radon progeny, which is believed to originate from naturally occurring radionuclides in the soil floor and accumulate in this unventilated space.

The soil surrounding the reactor pool is another area of potential soil contamination. The reactor pool is approximately 10 m x 3.6 m and extends approximately 7.5 m below the reactor room floor level. The reactor pool is located inside the circular Reactor Confinement structure, which has a diameter of approximately 16 m. The space between the outer pool walls and the Confinement structure contains soil fill. Since the base of the Confinement structure does not incorporate a floor, the pool therefore is underlain with soil and bedrock.

During reactor operations, small losses of pool water were a common occurrence. Specific locations of any pool leakage have not been identified; however, such leakage potentially could have resulted in contamination of soils around and beneath the pool. Breaks in piping.

beneath the Reactor Room floor were identified during facility remediation. Leakage of contaminated liquids from floor, sink, and pool overflow drains could have contaminated surface soils in the vicinity of these breaks. Characterization of surface and subsurface soils beneath the Reactor Room floor identified small, localized areas of contaminated surface soil; these areas were remediated. Characterization of the fill around the pool and in the soil, bedrock, and groundwater beneath the pool did not identify contamination of these media requiring remediation.

Fird Staus Survey Plan Addendurn 007: Spedc Sols Areas 2-1

CH2MHILL Figures 2-2 to 2-4 indicate the locations of these soils areas inside the UVAR building.

In preparation for implementing the Final Status Surveys, materials and equipment were removed from the crawl space and piping and other potentially contaminated items and components were removed from the fill area beneath the Reactor Room floor and around the reactor pool.

FMaStatus Survey Pla Addendumn 007: Spedal Soils Areas 2-2

-I CH2MHILL Figure 2-1 University of Virginia Reactor Facility and Environs ROOM~p

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127 102 111 ziia II l1S 1US 124 Figure 2-2 UVA Reactor First Floor Plan View Indicating the Location of Soil Fill Around Reactor Pool.

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IF moo Mogsai Figure 2-3 WVA Mezzanine Floor Plan View Indicating the Location of the Mezzanine Crawl Space.

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EIED Figure 2-4 WVA Reactor Ground Floor Plan View Indicating the Location of the Soils Beneath the Reactor Pool. H

3.Contaminants of Concern and Guidelines The initial characterization by GTS Duratek showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. Major structural contamination was generally limited to surfaces exposed to or in contact with reactor coolant, reactor neutron fields, and materials containing high levels of activity (e.g.,

the Hot Cell). Depending on the mechanism of contamination and the medium, radionuclides and their relative ratios varied. The overall predominant radionuclides were Co-60 and Cs-137; smaller activities of some other gamma emitters and hard-to-detect radionuclides were identified in samples from certain facility locations and media.

The continuing characterization by CH2M HILL identified small localized areas of elevated direct gamma radiation on the surface of the soil fill beneath the Reactor Room floor and the south end of the reactor pool. This contamination was at locations of piping breaks and leaks. Remediation of these areas of elevated activity appeared to eliminate contamination of the fill soil. Characterization did not identify any contamination of the Mezzanine crawl space by radionuclides of license origin. Tables 3-1 and 3-2 summarize results of sampling of pool fill soil.

Table 3-1 Radionuclide Concentration on the Surface of Fill Soil around the Reactor Pool Concentration (pCi)

Radionuclide Sample Sample Sample Sample Sample Sample Sample 021 038 056 052 047 104 139 Co-60 85.5 45.5 19.2 18.8 9.74 4.99 4.23 Cs-137 1.2 1.0 0.3 1.11 <0.26 <0.22 <0.23 Co-57 0.3 0.6 0.2 0.71 0.16 <0.12 <0.13 Eu-152 8.4 3.5 3.0 2.33 <1.22 <1.18 <1.06 The relatively low activity levels in the surface soil beneath the Reactor Room and pool floor did not enable a meaningful determination of the complete mix; particularly of hard-to-detect radionuclides. Therefore, because of the dominance of Co-60 in the surface samples and because the source of the contamination was liquids from the reactor facility, the same contaminant mixture will be assumed for the surface of the fill soil as used for the waste tank remediation (refer to Addendum 001) and reactor facility piping (refer to Addendum 002). A Co-60 DCGL unogate of 3.4 pCi/g will thus be used for the soils on the fill surface. A composite of surface FSS samples will be analyzed to confirm the absence of significant concentrations of hard-to-detect radionuclides.

Analyses of subsurface soil fill samples (Table 3-2) identified positive levels of Co-60, Am-241, Pu-241, and H-3. However, only one of six samples contained a positive level (16.3 pCi/g of Co-60) of any gamma emitters and only one of the three samples analyzed for hard-to-detects contained a positive level of H-3 (95.5 pCi/g). There does not appear to be a correlation among the relative radionuclide levels in these samples and a meaningful contaminant mixture was not identified for subsurface fill soil. Because of the marked difference between samples PSL 2-5 and 071, an analytical error or cross-contamination is Ferid Wm7:

Suet Mmb Addeum007: SpeMa Soft Arms 31

CH2MHILL suspected. Since the source of contamination in surface fill would be the same as that for the surface soil, the Co-60 DCGLsurrogate of 3.4 pCi/g will be used for these soils also, and a composite of subsurface FSS samples will be analyzed to confirm the absence of significant concentrations of hard-to-detect radionuclides.

Table 3-2 Radionuclide Concentration in Subsurface Fill Soil around the Reactor Pool Concentration (pCig_

Radionuclides Sample Sample Sample Sample Sample Sample 058 064 070 PSL 2-5 PSL 3-4 071 Co-60 <0.16 <0.18 <0.15 16.3 <0.15 <0.12 Cs-137 <0.15 <0.14 <0.14 <0.18 <0.18 <0.10 Eu-152 <0.95 <0.82 <0.86 <0.84 <1.06 <0.61 Co-57 <0.10 <0.09 <0.09 <0.09 <0.12 <0.08 Am-241 NP NP NP 0.71 1.22 <0.06 Fe-55 NP NP NP <1.29 <1.54 <0.92 H-3 NP NP NP <3.26 <3.29 95.5 I-129 NP NP NP <0.47 <0.57 <0.20 Ni-63 NP NP NP <1.0 <1.0 <11.2 Pu-238 NP NP NP <0.03 <0.03 <0.17 Pu-239 NP NP NP <0.02 <0.02 <0.14 Pu-241 NP NP NP 1.23 1.73 <7.03 Sr-90 NP NP NP <0.15 <0.14 <0.62 Tc-99 NP NP NP <0.15 <0.15 <0.22 NP = Analysis Not Performed Frd Stabz Srv Plan Addendum 007: Specal Softs Area 3-2

4.Survey Approach 4.1 Survey Reference System A 1-meter interval grid system will be established on surfaces to provide a means for referencing measurement and sampling locations. Grid systems will originate at the southwest corner of the survey unit, except where specific survey unit characteristics necessitate alternate grid origins. Grids are assigned alphanumeric indicators to enable survey location identification and are referenced to building features. Maps and plot plans of survey areas will include the grid system identifications.

4.2 Survey Classification Based on facility operating history, characterization survey results, and findings during remediation, the crawl space is designated MARSSIM Class 2 contamination potential, and the soil areas around and beneath the reactor pool is designated Class 1 for survey planning purposes.

4.3 Survey Area Identification For final evaluation, interior soils are divided into the following five groupings:

1) Mezzanine crawl space

2) Surface soil at piping excavations beneath the reactor room floor

3) Surface soil at demineralizer excavation

4) Surface/subsurface fill around pool

5) Surface/subsurface fill beneath pool Because of their small surface areas and location (inside the building), and inclusion of subsurface material, the FSS will not follow traditional MARSSIM approaches; however, the survey frequency and data analyses and evaluation will be consistent with the intent of MARSSIM.

4.4 Demonstrating Compliance with Release Guidelines Compliance with decommissioning requirements will be demonstrated by comparing the results of final status survey sample analyses with the Co60 DCGLsurrogate of 3.4 pCi/g and by furthermore demonstrating that hard-to-detect radionuclides are not present in significant concentrations. Subsurface soils surrounding and beneath the reactor pool will be evaluated over 1-meter thick intervals. Because the radionuclides identified as potential contaminants are not present in background samples at concentrations, which are Fial Stalus Survey Plan Addendum 007: Special Soils Aeas 4-1

CH2MHILL significant fractions of the release guidelines, correction of FSS sample data for background levels will not be required.

4.5 Number of Data Points and Sampling Pattern Sampling/measurements will be performed at uniformly spaced intervals throughout the soil areas or volumes of interest. The spacing between data points will be determined by the surface area or volume.

For the small Mezzanine crawl space area, samples of surface soil will be obtained on the same pattern and at the same intervals (about 3.5 m) as the surface activity measurement data points on the non-soil surfaces of this area. Thus 5 samples are expected from this soils area. Five samples will also be collected from the small excavation adjacent to the Demineralizer Room.

Due to their larger area/volume, the number of samples from the remaining areas of reactor pool fill will be based on the MARSSIM guidance for Sign Test evaluation as follows:

The following is based on a Co-60 DCGLsulrroSate of 3.4 pC/g:

DCGL = 3.4, LBGR = 0.5 DCGL (Refer to Master FSS Plan, Section 7.8)

A = DCGL - LBGR = 1.7 a = 0.85 (based on 25% of the DCGLsurrogate)

A/a=2 From MARSSIM Table 5.5, N = 15; i.e., 15 data points required for statistical evaluation of a survey unit using the Sign Test.

The sampling pattern for the soils around and beneath the reactor pool will be determined on the basis of soil surface area and volume. The surface area of soil beneath the Reactor Room floor is approximately 140 m2 , and the pool fill volume is assumed to occupy the entire space between the pool and outer room walls to the depth of the pool; the resulting volume is approximately 1000 m3 . Fifteen surface samples from the top of the fill under the reactor room floor will be obtained. The specific sampling locations will be determined on the basis of soil conditions and accessibility and will be distributed to provide reasonably uniform coverage. In addition, 15 subsurface samples (1 at each surface sampling location, at a randomly selected depth of 1 m to 7 m) will be obtained.

Sample borings will be performed through the pool bottom at a minimum of 4 existing locations, into the soils beneath the reactor pool. Four samples will be obtained from each of the access locations at depths extending to bedrock "refusal" for a total of 16 samples.

Sample depths at each location will be at the surface immediately beneath the pool; at the maximum depth obtained; and at two equally spaced depths between the pool and the maximum level.

In addition to the systematic locations, samples will be obtained at "biased" locations (if any), identified by scanning or professional judgment of the FSS field supervisor as having the greatest potential for contamination.

Fma Su0 SM AW Addedum 007: Specal Soft Arma 42

CH2MHILL 4.6 Survey Methods Gamma scans of accessible surfaces will be performed using a 2"X 2" Nal detector (Ludlum Model 44-10) coupled with a Ludlum Model 2221 ratemeter/scaler. The detector will be maintained within 5 to 10 cm of the surface and moved across the surface while noting any indication of audible elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Gamma scanning coverage will be 100% of accessible surface soil surfaces. Gamma logs of boreholes for subsurface sampling will be conducted using the Nal detector (Ludlum Model 44-10 or Ludlum Model 44-2) coupled with a Ludlum Model 2221 ratemeter/scaler. Gamma levels (c/I min) at 1-m depth intervals will be obtained throughout the length of the borehole. Audible response will be monitored during detector movement for indication of elevated count rate, which might indicate the presence of radioactive contamination. Gamma scan and logging results will be documented on survey maps. Locations of elevated response will be noted for further investigation.

Soil samples of approximately 500 g each will be collected at systematic and judgmental sampling locations. Surface samples will be obtained from the upper 15 cm soil layer, using trowels or bucket augers. Subsurface samples will be obtained using bucket augers, split spoon samplers, or other methods consistent with the drilling technique and equipment, and homogenized over a depth interval of 1 meter.

Samples will be assigned unique identification numbers and a chain of custody record and analytical request will be prepared.

4.7 Sample Analyses Samples will be sent to an off-site commercial laboratory for gamma spectrometry analysis.

Gamma analyses will be surrogates for calculating soil activity concentrations, assuming the previously analyzed mix of radionuclides remains proportional to the concentrations of principal gamma emitters. Composite samples, consisting of equal amounts from individual samples representing the survey areas, will be prepared and analyzed for hard-to-detect (10 CFR Part 61) radionuclides. Results of these analyses will be used to develop fractional contributions of non-gamma emitting radionuclides in sediments. These contributions will be used to adjust the results of individual sample gamma analyses for the presence of other contaminants.

4.8 Investigation If measurements, sample analyses, or statistical data evaluation identifies residual radioactivity exceeding the release criteria, the source of the contamination will be investigated. Remediation will be performed, as necessary, and FSS activities repeated, utilizing a newly determined sampling pattern.

Fmal Status Survey Plan Addendun 007: Special Sols Areas 4-3

5. Data Evaluation Sample gamma spectrometry results will be compared individually with the Co-60 DCGLsursogate of 3.4 pCi/g. Results of hard-to-detect analyses of composite samples will be evaluated for the presence of any contaminant that might exceed default screening DCGLs.

Final Status Survey Plan WAR Addendum 007: Special Soils Areas 5-1

Final Status Survey Plan Addendum 008: Ventilation Systems Revision I Prepared for University of Virginia Reactor Facility Decommissioning Project Prepared by CH2MHILL 151 Lafayette brive, Suite 110 Oak Ridge, TN 37830 With assistance from Safety and Ecology Corporation 2800 Solway Road Knoxville, TN 37931 April 2004

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Contents Contents . ............................................................... i

1. Introduction ............................................................... 1-1

2. Description ............................................................... 2-1

3. Contaminants of Concern and Guidelines .............................................................. 3-1

4. Survey Approach ............................................................... 4-1 4.1 Survey Reference System .............................................................. 4-1 4.2 Survey Classification ............................................................... 4-1 4.3 Survey Unit Identification ............................................................... 4-1 4.4 Demonstrating Compliance with Release Guidelines ....................................... 4-1 4.5 Background Reference Areas and Materials ......................................... ............. 4-1 4.6 Number of Required Data Points .............................................................. 4-2 4.7 Sampling Pattern ............................................................... 4-2 4.8 Survey Methods .............................................................. 4-2 4.9 Sample Analyses .............................................................. 4-2 4.10 Investigation ............................................................... 4-2

5. Data Evaluation ............................................................... 5-1 FIGURES 2-1 University of Virginia Reactor Facility and Environs .......................................... .......... 2-3 2-2 UVA Reactor First Floor Indicating Polentially Impacted Ventilation Systems ........ 2-4 2-3 UVA Mezzaine Floor Indicting Potentially Impacted Ventilation Systems ................ 2-5 2-4 UVA Reactor Ground Floor Indicating Potentially Impacted Ventilation Systems ..2-6 Final Satus Survey Plan Addendum 008: Ventlafion Systems

1. Introduction The Master Final Status Survey Plan (UVA-FS-002) identifies the Data Quality Objectives for Final Status Survey (FSS) activities, together with the underlying technical assumptions, approaches, and methodologies for designing, implementing, and evaluating a FSS on each impacted area of the University of Virginia Research Reactor (UVAR) Facility. A separate survey area-specific addendum is prepared for each area or group of areas with common media, contaminants, and other characteristics, prior to beginning FSS. The FSS for the specific area or group of areas will then be performed in accordance with that addendum.

This addendum (Addendum 008) applies to the interior surfaces of the potentially impacted ventilation systems in the UVAR Facility building.

Fmal Status Survey Plan Addendum 008: Ventlation Systets 1-1

2. Description The UVAR Facility is located on Old Reservoir Road, approximately 0.6 kilometers (km) west of the Charlottesville, VA city limits. The Facility includes the UVAR building, a small pond, and asphalt-paved roads, parking areas, and equipment/materials storage pads, situated on a land area of approximately 9390 m2 (see Figure 2-1). The three-story building housed the UVA Research Reactor and the CAVALIER facility, as well as offices for the reactor staff and faculty and students of the Department of Nuclear Engineering miscellaneous laboratories, and other support facilities for the reactors and Department of Nuclear Engineering.

Several systems provided ventilation for facilities having a potential for airborne radioactivity. The remaining systems/components, which are potentially radiologically impacted, are:

• Exhaust for fume hood in Room M005.

• Exhaust for fume hood in Room M008.

• Exhaust for fume hoods (2) in Room M019.

• Exhaust for source storage Room G022.

• Hot Cell exhaust.

• Reactor Room recirculation and exhaust.

Because the exhaust ventilation systems in laboratories M005 and M008 had become contaminated with Tc-99 and Ni-63, respectively, during research projects in those facilities, new fume hoods and ductwork between the hoods and the exhaust fans were installed in these rooms a short time before the reactor decommissioning activities began. The blower assembly was removed from Room M-008 during D&D operations; the original squirrel-cage blower for the M005 exhaust system remains, along with the ductwork downstream of both fan units. During facility operation, these systems exhausted through the outside laboratory walls and into vertical ducts on the building exterior; the vertical ducts discharged above the roof level through rain-cap covered stacks. The remaining exhaust ventilation systems in laboratories M005 and M008 are potentially still impacted and will be included in this survey. Because the new hoods and ductwork were never used for contaminated operations, the potential for contamination of those surfaces is considered negligible.

Fume hoods in Room M019 became contaminated with Tc-99. Hood baffles were removed and cleaned. Ductwork from the rear of the hood was removed up to and including the HEPA filter and housing. A short section of ductwork, which connected the exhausts from this facility to the former exhaust ventilation from the Hot Cell, remains. The Hot Cell exhaust duct from inside the Hot Cell to the blower in Room M020, remains; the HEPA filter box has been removed from the point where the ductwork joins the blower. The combined FmaSubs Survey Plan Addendum 008: Venblaton Systems 2-1

CH2MHILL Hot Cell and M019 fume hood exhausts pass through a duct inside the Reactor Stack and discharge into the suction pleinum of the Reactor Room exhaust fan.

Reactor Room air is exhausted through a duct near the ceiling of the Reactor Room into the suction plenum of the Reactor Room exhaust fan at the top of the Reactor Stack. At this location the duct from the Hot Cell/M019 hood and the Reactor Room are combined and exhausted through the plenum vertically on the roof of the Reactor Room.

There was a small exhaust from the source storage room (Room G022). The blower has been removed, but the ductwork which discharges at the Mezzanine level on the east end of the building remains.

Reactor Room air is replenished by a recirculating system. This system draws fresh air in through reactor room door and combines it with room air. This stream is heated as needed and then discharged back into the Reactor Room through 12 vents, located at the base of the Reactor Room wall.

Figures 2-2 to 2-4 indicate the locations of the remaining potentially impacted ventilation system surfaces. Except for portions of the recirculating air vents, which are encased in concrete, there is access to interior surfaces of components of these ventilation systems to conduct surface activity scans and measurements. It is anticipated that access will be adequate to demonstrate that radiological conditions satisfy decommissioning criteria.

Final Status Survey Plan Addendum 008: Ventilaton Systems 2-2

CH2MHILL Figure 2-1 University of Virginia Reactor Facility and Environs 0 =0 -. .

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3. Contaminants of Concern and Guidelines The GTS Duratek initial characterization and continuing characterization by the CH2M HILL team showed that radiological contamination was generally low level and was limited to a small portion of the structure interior. The overall predominant radionuclides were Co-60 and Cs-137. Ni-63 and Tc-99 contaminants were present on facility surfaces in labs M008 and M005, respectively. Sufficient activity levels were not present on facility surfaces to enable a meaningful determination of the contaminant mixture - particularly for hard-to-detect radionuclides. Therefore an adjusted gross beta DCGL of 6320 dpm/100cm 2 was developed for surfaces, based on the contaminant mix resulting from reactor effluents (refer to Addendum 002). With exception of the systems in labs M008 and M005, this adjusted gross DCGL will be the basis for evaluating the final radiological status of ventilation system surfaces. For the systems in M008 and M005, the default screening value of 1.3E+6 dpm/lOOcm 2 for Ni-63 will be applicable (this is more restrictive than the value of 1.8E+6 dpm/1OOcm 2 for Tc-99 and is used for simplicity in application and conservatism).

Guidelines for removable activity are 10% of the levels for total surface activity. If there is specific evidence during the FSS that contamination of a surface is comprised of different radionuclides, the appropriateness of these guideline levels will be reevaluated. Appendix A of the Master FSS Plan describes the method for establishing guidelines for other radionuclides and combinations of radionuclides. If guidelines other than those indicated here are to be used, a justification will be prepared and included with the FSS documentation.

rd Statu Survey PMm Mdendun ON: Venblabon System - 31

4. Survey Approach 4.1 Survey Reference System Because of the nature and limited area of the ventilation system components, reference grid systems will not be established on surfaces to provide a means for referencing measurement and sampling locations. Instead, survey locations will be referenced to prominent facility features.

4.2 Survey Classification Based on facility history (including the Historic Site Assessment), radiological monitoring conducted during characterization, and remedial activities, all potentially impacted exhaust ventilation systems are Class 1.

4.3 Survey Unit Identification Impacted structure surfaces of *10 m2 are not designated as survey units. Instead, a minimum of 1 measurement per m2 or a total of 4 measurements, whichever is greater, will be obtained from such areas and compared individually with the applicable DCGL.

4.4 Demonstrating Compliance with Release Guidelines -

Total and removable surface activity measurements from ventilation system surfaces will be individually compared directly with the applicable DCGL. Because of the small surfaces and possible small number of data points, statistical tests will not be performed.

4.5 Background Reference Areas and Materials If ambient direct radiation levels necessitate a shielded/unshielded measurement approach, a material or instrument background value will not be required for correction of surface activity data. Otherwise, an instrument and/or material background will be determined on a similar material as the surface being surveyed, but without a history of potential contamination by licensed operations. If required, sufficient background determinations will be made for each instrument to provide an average background level that is accurate to within +/- 20%; this usually requires 8 to 10 measurements, which are then evaluated using the procedure described in draft NUREG/CR-5849 and additional data points obtained, as necessary.

Final Slatts &ne Ra Addendu= OM0: Ventsadon Sysbems 4-1

CH2MHILL 4.6 Number of Required Data Points Impacted structure surfaces of < 10 m2 are not designated as survey units. Instead, a minimum of 1 measurement per m2 or a total of 4 measurements, whichever is greater, will be obtained from such areas and compared individually with the applicable DCGL.

4.7 Sampling Pattern To the extent that the surfaces are accessible, measurements will be performed at locations uniformly spaced throughout the ventilation system surfaces. Measurements will be biased to locations which, based on scanning results and professional judgment of the FSS field supervisor, have the greatest potential for contamination. Examples of such locations include duct inlet and discharge points, bends, seams, and locations of discoloration and accumulations of dirt and debris.

4.8 Survey Methods Where conditions allow, beta scans of interior system surfaces will be performed using Ludlum Model 43-68 gas proportional or Ludlum Model 44-9 pancake GM detectors coupled with Ludlum Model 2291 ratemeter/scalers. The detector will be maintained within -1 cm of the surface while advancing the detector at a rate of approximately one detector width per second. Scan speed will be adjusted, as necessary, to assure detection sensitivities are less than 50% of the release criteria. Audible response will be monitored for indication of elevated count rate, which might indicate the presence of radioactive contamination. Results (count rate) will be documented on survey area maps. Locations of elevated response will be noted for further investigation. Beta scanning coverage will be 100% of accessible surfaces.

Surface activity measurements will be performed at uniformly distributed and judgmental locations; 1-minute static measurements will be conducted using a Ludlum Model 43-68 gas proportional detector coupled with a Ludlum Model 2221 ratemeter/scaler.

Smears for removable activity will be performed at locations of direct activity measurements.

4.9 Sample Analyses Smears will be analyzed for gross alpha and gross beta activity in the on-site counting facility.

4.10 Investigation If measurements, sample analyses, or statistical data evaluation identifies residual radioactivity exceeding 50% of the release criteria, the source of the contamination will be investigated. Remediation will be performed, as necessary, and FSS activities repeated, utilizing a newly determined sampling pattern and random start point.

Fdal Status Sure PFn Addendurn 008: Ventlaton Systems 42

5. Data Evaluation Measurements will be individually compared directly with the adjusted gross DCGL of 6320 dpm/100 cm2 . Only if all measurements are less than the criteria, will the surface satisfy the requirements for release.

Fird Statu Swvey Pa UVAR Adendum O Venbiatio Systems 5-1