ML12030A229

Plum Brook Reactor Facility Final Status Survey Report Attachment 12, Reactor Building (Building 111), Revision 0, Part 1 of 3

ML12030A229
Person / Time
Site:	Plum Brook
Issue date:	01/25/2012
From:	US National Aeronautics & Space Admin (NASA)
To:	NRC/FSME
References
Download: ML12030A229 (77)

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Plum Brook Reactor Facility Final Status Survey Report Attachment 12 Revision 0 Reactor Building (Building 1111)

FINAL STATUS SURVEY REPORT ROUTING AND APPROVAL SHEET Document

Title:

Final Status Survey Report, Attachment 12 Reactor Building (Building 11I11)

Revision Number: 0 ROUTING SIGNATURE DATE Prepared By B. Mann 7 1/24/12 Prepared By N/A REVIEW & CONCURRENCE Independent Technical Reviewer R. Case 2 1/24/12 Other Reviewer, QA Manager J. Thomas  : L,.<

• , 1/24/12 Other Reviewer . N/A FSS/Characterization Manager W. Stoner 1/24/12 NASA Project Radiation Safety Officer W. Stoner 1/24/12 ii

NASA PBRF DECOMMISSIONING PROJECT CHANGE/CANCELLATION RECORD DOCUMENT TITLE: Final Status DOCUMENT NO: N/A REVISION NO: 0 Survey Report, Attachment 12, Reactor Building (Building 1111)

Revision 0: Initial issue of Report IAD-01/31 Rev 1 iii

Plum Brook Reactor Facility FSSR Attachment 12 Rev. 0 LIST OF EFFECTIVE PAGES DOCUMENT NO: N/A REVISION NO: 0 Page No. Revision Level Page No. Revision Level Page No. Revision Level Cover Page 0 Routing & Approval 0 Sheet Change/Cancellation 0 Record LOEP 0 TOC 0 List of Tables & List 0 of Figures List of Acronyms & 0 Symbols, 3 pages Text, 48 pages 0 Appendix A 0 20 pages Appendix B 0 150 pages Appendix C 0 4 pages Appendix D 0 19 pages

+ 4 4 a s

+ 4 4 a s

+ 4 4 a 4

+ 0 4 0 4 a a a a a AD-01/5 IForm Rev 2 iv

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 TABLE OF CONTENTS 1.0 Introd uction ........................................................................................................ . . . .. 1 2.0 Reactor Building Description .................................................................................. 2 2.1 Building Layout and Construction ........................................................................... 2 2.2 Building Systems and Services ................................................................................ 7 2.3 Building Modifications ............................................................................................ 8 2.4 Final Configuration and Scope ................................................................................ 9 3.0 History of Operations ................................................................................................. 9 3.1 Chronology ................................ !......................................................................... . . 10 3.2 Startup and Operations ............................................................................................ 10 3.3 Radioactive Materials in the Reactor Building ...................................................... 11 3.4 Post-Shutdown Materials Disposition and Characterization .................................. 13 3.5 Decommissioning ................................................................................................... 15 4.0 Survey Design and Implementation for the Reactor Building ............................. 16 4.1 FSS Plan Requirements ......................................................................................... 17 4.2 Area Classification and Survey Unit Breakdown .................................................... 18 4.3 Number of Measurements and Samples ............................................................... 23 4.4 Instrumentation and Measurement Sensitivity ...................................................... 26 5.0 Reactor Building Survey Results ........................................................................... 30 5.1 Scan Surveys ......................................................................................................... . . 31 5.2 Systematic Measurements and Tests ...................................................................... 34 5.3 Investigations, Additional Measurements and Tests .............................................. 38 5.4 Soil Survey Unit Results ........................................................................................ 41 5.5 QC Measurements ................................................................................................... 42 5.6 ALARA Evaluation ................................................................................................ 43 5.7 Comparison with EPA Trigger Levels ................................................................. 45 5.8 Conclusions .......................................................................................................... . . 46 6 .0 Referen ces ..................................................................................................................... 47 7 .0 A p p en d ices .................................................................................................................... 48 Appendix A - Exhibits Appendix B - Survey Unit Maps and Tables Showing Measurement Locations and Results Appendix C - QC Measurements Appendix D - Evaluation of Revised DCGLs v

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 LIST OF TABLES Table 1, Reactor Building Activities Covered in Unusual Incident Reports (1960-1973) ..... 12 Table 2, Summary of 1998 Surface Beta Contamination Measurements in the RB ................. 14 Table 3, Reactor Building Radionuclide Activity Fractions and Gross Activity DCGLs ........ 17 Table 4, Class-Based Survey Scan Coverage and Action Level Requirements ....................... 18 Table 5, RB Survey U nits for FSS ............................................................................................ 19 Table 6, Reactor Building Survey Unit Breakdown by Major Elevation ................................ 23 Table 7, Reactor Building Survey Design Summary ................................................................ 24 Table 8, Sensitivity Analysis for Reactor Building FSS Design .............................................. 26 Table 9, Typical Detection Sensitivities of Field Instruments ................................................. 28 Table 10, Typical Detection Sensitivities of Field Instruments used for Soil Scans ................ 30 Table 11, Scan Survey R esults ................................................................................................. 31 Table 12, Total Surface Beta Activity Measurement Summary and Test Results ................... 35 Table 13, Summary of Scan Investigations and Static Measurements ..................................... 38 Table 14, Elevated Measurement Comparisons and Tests ..................................................... 40 Table 15, Removable Surface Activity Measurements above MDA ........................................ 40 Table 16, Reactor Building Soil Survey Unit Sample Results ................................................. 42 Table 17, Replicate QC Measurement RPD Evaluation ........................................................... 43 Table 18, Screening Level Values for Reactor Building and Radionuclide Activity Fractions ... 44 Table 19, NRC Soil Screening Level ALARA Comparison ................................................... 45 Table 20, Comparison of Soil Sample Results with EPA Trigger Levels ............................... 45 LIST OF FIGURES Figure 1, PBRF NW Area Showing Reactor Building and Other Support Buildings ................ 3 Figure 2, E-W Cross Section View of Reactor Building and CV ............................................... 4 vi

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 LIST OF ACRONYMS & SYMBOLS A detector open area (cm1) a alpha; denotes alpha radiation, also type I error probability in hypothesis testing AEC Atomic Energy Commission ALARA As Low As Reasonably Achievable AF Area Factor ATS Assembly Test and Storage (Building 1121)

P3 beta; denotes beta radiation, also type II error probability in hypothesis testing bi background counts in observation interval BR Background count rate BPL Byproduct License cm centimeters cm2 square Centimeters cpm counts per Minute CPT Cold Pipe Tunnel CRB Cold Retention Basins (Building 1154)

A delta, DCGLw - LBGR d9 Scan surveyor sensitivity index DCGL Derived Concentration Guideline Level DCGLEMC DCGL for small areas of elevated activity, used with the Elevated Measurement Comparison test (EMC)

DCGLw DCGL for average concentrations over a survey unit, used with statistical tests.

(the "W" suffix denotes "Wilcoxon)"

dpm disintegrations per minute Ei Detector, or instrument efficiency Es Surface efficiency Et Total efficiency EMC Elevated Measurement Comparison EP embedded piping EPA US Environmental Protection Agency FH Fan House (Building 1132)

FSS Final Status Survey FSSP Final Status Survey Plan FSSR Final Status Survey Report ft. foot (or feet) 7 gamma, denotes gamma radiation gcpm gross counts per minute g gram h hour HL Hot Laboratory (Building 1112)

HPT Hot Pipe Tunnel HRA Hot Retention Area (Building 1155)

HSOO Health Safety Operations Office HTD Hard To Detect vii

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 LIST OF ACRONYMS & SYMBOLS. Continued i observation counting interval during scan surveys in. inch KVA electrical system power rating in units of 1000 volt-amperes LMI Ludlum Measurements, Inc.

LBGR Lower Bound of the Gray Region m2 square meters MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC Minimum Detectable Concentration MDCscan Minimum Detectable Concentration for scanning surveys MDCs~tati,: Minimum Detectable Concentration for static surface activity measurements MDCR Minimum Detectable Count Rate MOU Memorandum of Understanding MUR Mockup Reactor mrem millirem MW Megawatt NASA National Aeronautics and Space Administration N Number of FSS measurements or samples established in a survey design N/A Not Applicable ncpm net counts per minute NRC US Nuclear Regulatory Commission PBRF Plum Brook Reactor Facility PNL Pacific Northwest Laboratory Standard normal distribution function p surveyor efficiency for scan surveys percent pCi/g picocuries per gram PCW Primary Cooling Water PPP process piping pump QC Quality Control RAMS Remote Area Monitoring System RB Reactor Building (Building 1111)

RESRAD RESidual RADioactive - a pathway analysis computer code developed by Argonne National Laboratory for assessment of radiation doses. It is used to derive cleanup guideline values for soils contaminated with radioactive materials RESRAD-BUILD A companion code to RESRAD for evaluating indoor building contamination and developing site-specific DCGLs ROLB Reactor Office and Laboratory Building (Building 1141)

RPD relative percent difference s seconds a generic symbol for standard deviation of a population viii

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 LIST OF ACRONYMS & SYMBOLS. Continued SNL Sandia National Laboratory SR Survey Request SMTA surface measurement test area tb background count time ts sample count time TBD Technical Basis Document P Mean activity concentration jiCi/ml microcuries per milliliter gtR/h microRoentgen per hour UCM Unusual Condition Measurement UL Upper limit of the confidence interval about the mean VSP Visual Sample Plan WHB Waste Handling Building (Building 1133)

Zia. (100-a) percentile of normal distribution Zl.* (100-P3) percentile of normal distribution 00 Mathematical symbol for infinity ix

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 1.0 Introduction This report presents the results of the final status radiological survey of the Plum Brook Reactor Facility (PBRF) Reactor Building (RB, Building 1111). It is Attachment 12 of the PBRF Final Status Survey Report (FSSR). 1 This attachment describes the RB, its operational history and final condition for the final status survey (FSS). It describes the methods used in the FSS and presents the results. 2 As stated in the PBRF Final Status Survey Plan (FSSP) [NASA 2007], the goal of the decommissioning project is to release the facility for unrestricted use in compliance with the requirements of US NRC 10CFR20 Subpart E. The principal requirement is that the dose to future site occupants will be less than 25 mrem/y. Subpart E also requires that residual contamination be reduced to levels as low as reasonably achievable (ALARA). Derived Concentration Guideline Levels (DCGLs) for residual surface contamination have been established for the Reactor Building.

Considering the radionuclide mixtures established for areas within the Reactor Building, gross beta DCGLs range from 14,382 to 40,500 dpm/100-cm2 .

The survey measurement results and supporting information presented herein demonstrate that residual contamination levels in each survey unit of the RB are well below the DCGL. Additionally, it is shown that residual contamination has been reduced to levels that are consistent with the ALARA requirement. Therefore, the Reactor Building meets the criteria for unrestricted release.

Section 2.0 of the report provides a description of the RB. This includes the building layout, its relation to other PBRF buildings and facilities, design and materials of construction, building contents and use, systems and services, building modifications, final configuration and scope of the FSS for this building.

A brief history of operations is presented in Section 3.0. A chronology of significant milestones is followed by history of operations with radioactive materials. Post shutdown and decommissioning activities are summarized. Results of radiological characterization surveys in support of decommissioning are presented.

Section 4.0 presents the FSS design for the Reactor Building. This section includes applicable FSS Plan requirements, breakdown into survey units and assignment of MARSSIM classifications. The survey design approach, instrumentation and measurement sensitivities are described.

Survey results are presented in Section 5.0. This section includes a summary of the FSS measurements performed in the Reactor Building survey units, comparison to the DCGLw, tests performed and an evaluation of residual contamination levels relative to the ALARA criterion.

I The PBRF Final Status Survey Report comprises the report main body and several attachments. The attachments present survey results for individual buildings, outdoor structures and open land areas. The entire final report will provide the basis for requesting termination of NRC Licenses TR-3 and R-93 in accordance with 10CFR50.82 (b) (6).

2 It is noted that the Reactor Building and Containment Vessel (CV) are integrated structures, sharing a common footprint and many systems and services. However, due to their combined size and complexity, the two structures were administratively separated for conduct and reporting of the FSS. The FSS of the CV is reported in Attachment 11 of the FSS Report.

3 The DCGLs listed in this paragraph are the gross activity DCGLs for the radionuclide mixtures assigned to the specified areas with adjustments to account for the dose contribution from deselected radionuclides. Also it is noted that the values listed above are modified from the values originally used, to correct a calculation error in the Technical Basis Document TBD-07-001 [PBRF 2007].

1

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Supporting information is contained in Appendices. Appendix A contains photos and drawings to supplement the text. Survey design maps, tables of coordinates and total surface beta measurement results for each survey unit are provided in Appendix B. The QC measurements are presented in Appendix C. Appendix D presents an evaluation of the impacts of DCGLs that were revised to correct errors noted in the original Technical Basis Document, PBRF-TBD-07-001 [PBRF 2007].

2.0 Reactor Building Description The Reactor Building is a large four story structure which housed the Plum Brook 60 MW Test Reactor, the 100 kW Mockup Reactor (MUR) and associated experimental and test facilities. It is 162 ft. (E-W) by 149 ft. (N-S) with two levels below grade (-15 ft. and -25 ft.) and two levels above grade (0 ft. and a mezzanine floor at 12 ft.) The building layout, history of construction, systems and services and configuration for FSS are described below.

2.1 Building Layout and Construction Figure 1, a map of the PBRF main facility area, shows the Reactor Building, and other principal support buildings. Figure 2, shows the vertical arrangement of the RB and CV in an east-west cross section view. A photo of the Reactor Building exterior (taken in early 2011) is provided in Exhibit 1 of Appendix A.

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Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Figure 1, PBRF NW Area Showing Reactor Building and Other Support Buildings l HL CV ~ ~ E5ECOTANEN C

51 i HL MOT LAB PPM PRiMARY PUMP HOUSE HWMB "DEMOLISMED SEB SERVICET BUILDING5 EQUIPMENT BUILDING FM FAN MOUSE RO5 WASTE REACTOR HANDLING OFFICE BUILDING LABORATORY BUILDING HRA HOT RLTENTION AREA "THESE BUILDINGS WERE DEMOUSHED TO -3 ft.

ELEVATION AND BACKFILLED.

3

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Figure 2, E-W Cross Section View of Reactor Building and CV CONTAINMENT VESSEL.

OREACTO SUNDM* REACTOR BUILDING 0" / -AN Kz * -15'-0"

-25"-0"O-4

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 As seen in Figure 1, the Reactor Building is the center of the four-building complex which comprises the heart of the PBRF. Following fabrication and installation of the CV shell, the RB was constructed surrounding the CV. Photos showing the RB under construction are provided in Exhibits 2 and 3 of Appendix A.

The RB was directly connected with passages for personnel, materials and equipment to: the Hot Laboratory (HL) on the south, the Primary Pump House (PPH) on the east and the Reactor Office and Laboratory Building (ROLB) on the west. It was also connected via an underground tunnel to the Assembly Test and Storage (ATS) Building located about 150 ft.

north of the RB. The RB was also connected to other nearby support buildings and structures:

the Service Equipment Building (SEB), Fan House (FH), Waste Handling Building (WHB),

Hot Retention Area (HRA) and Cold Retention Basins (CRBs) via underground pipe chases, or tunnels and underground piping and utility connections. Most, if-not-all, of the residual contamination at the PBRF originated in the Reactor Building (and CV).

The Reactor Building was designed as a multi-level facility to support operations of the PBRF 60 MW Test Reactor and the MUR. It included the reactor and experiment control rooms, process systems piping and sumps, guaranteed power conversion equipment, offices for shift and technical services personnel, decontamination and change rooms, equipment and electronics storage areas, a room for facility drawings and vendor catalog files, a counting room, an area for remotely controlling hot laboratory equipment, and restrooms. Plan views of the major elevations which illustrate the RB layout are shown in Exhibits 4 through 7 of Appendix A.

Canals F, G and H were located in the southeast portion of the RB. These were 25 foot deep concrete lined structures that normally were water filled. Each canal was connected by 10 by 24 ft. openings with water-tight removable lift doors. Canal F was connected to Canal E, located inside the CV; it was utilized to transfer radioactive material into and out of the CV.

Canal F also connected to Canal J in the Hot Lab, providing an underwater transfer path for irradiated materials and test equipment from the 60 MW reactor and the MUR. Canal H housed the MUR core and control rod drive mechanisms and a fuel storage rack. The MUR control room was located about six feet above the 0 ft. elevation directly north of Canal H.

The photo in Exhibit 8 of Appendix A shows the MUR circa 2000. Canal G served as the spent fuel storage pool for both reactors. It contained cadmium lined fuel storage racks, a cut-off saw for removing fuel element end pieces and a hot plug storage rack. The photo in Exhibit 9 of Appendix A shows the fuel storage racks in Canal G circa 2000.

The Reactor Building contained two below grade areas. The -25 ft. elevation, also called the sub-basement, contained areas for support equipment, the process piping pump (PPP) room and sump and trenches to the PPP room that bottomed at -30 ft. Access to the -25 foot level was by a stairwell on the -15 foot level on the west side of the CV. There was also a stairway from the -25 foot level to the top of the PPP room. About two thirds of the sub-basement elevation, the area comprising the northwest, north and northeast sectors, was unexcavated.

The ft. elevation housed piping and equipment for process and utility-service systems.

These included primary cooling water, de-ionized and process water, secondary cooling water supply and return, utility and equipment air, steam and radioactive waste water systems. Most of the supply piping for these services came from the SEB through the Cold Pipe Tunnel 5

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 (CPT). The -15 ft. elevation also included the emergency power system with batteries and motor-generator sets, electrical control cabinets, cable trays, air conditioning equipment and a cage for storing radioactive sources. The Canal H water cleanup system and Westinghouse experimental equipment were also located on the -15 ft. elevation.

The -15 ft. elevation was accessed from the RB 0 ft. elevation via stairways on the east and west side exterior to the CV. See Exhibit 10 of Appendix A for a view of the 0 ft. elevation showing the west stairwell to the -15 ft. elevation. The -15 ft. elevation was connected to the ROLB via a double doorway, to the ATS Building via a tunnel on the northwest side and to the SEB via the CPT on the east side. There were two hatchways in the floor for moving hardware from the ground level to lower levels - one on the northeast side for the -15 foot level and another hatchway to the -25 foot level on the south side west of Canal F.

Entrance to the CV was provided through three metal airlocks and a large truck door.

Personnel airlocks were located on the west and east side of the CV and the truck door was on the north side. The experiment airlock, a large cylindrical airlock, was located in the southwest part of the CV. A change room and decontamination room were located adjacent to the east airlock. Controls for Hot Laboratory equipment and electrical cabinets were located along the south wall of the 0 ft. elevation and on an elevated platform. Three leaded glass windows located in that wall section provided viewing into the Hot Dry Storage area, into Canal J and the Hot Handling Area, all in the HL. This area of the RB also contained control consoles for remotely operating cranes in the Hot Dry Storage and Hot Handling rooms and Canal J.

Access to the RB 0 ft. elevation was provided through personnel doors on the west side (an exterior door in the southwest comer and a double door into the ROLB). On the RB north side two personnel doors exited to the outside. On the east side, a personnel door exited to the outside and a personnel door connected to the PPH. Double personnel doors on the south side connected to the Hot Laboratory. Two rollup truck doors provided access for large items: one was located in the southwest comer with a 20-ton overhead crane (bridge and trolley) and one on the northeast side of the building with a 5-ton overhead monorail crane.

A mezzanine area ran along the west and north side of the RB. The west side contained two office areas (housing facility drawings and equipment manuals) and an electronic equipment and parts storage area. The Reactor Control Room was located in the NW comer. A storage room, rest room, janitor's closet and four offices were located on the north side.

Three control rooms were located in the Reactor Building. The 60 MW Test Reactor Control Room was located in the northwest comer on the mezzanine level. It housed all the controls for operating the 60 MW Test Reactor and the associated support systems. These included the reactor control rod drive actuators, nuclear instrumentation, primary cooling water instrumentation and essential pump and valve controls. The Experiment Control Room, located on the first floor directly beneath the 60 MW Test Reactor Control Room, housed numerous experiment consoles and a data logging computer facility. The MUR Control Room located north of Canal H housed the MUR nuclear and process instrumentation, mechanical rod drives and experiment instrumentation.

The exterior of the RB was primarily fluted metal siding with steel sash windows (Solex brand). The original RB roof was a one and three-quarter inch metal deck covered with one 6

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 and one half inch fiberglass insulation plus four ply built-up roofing materials (roof paper, gravel and tar). There were two penthouses on the Reactor Building roof, one on the east side and one on the west side, both provided normal ventilating air intake for the Reactor Building areas.

2.2 Building Systems and Services Basic Reactor Building services such as heating, ventilating and air conditioning (HVAC),

fire protection, domestic water, deionized water and service air were provided through common PBRF systems, mostly originating in the SEB. Of special note are radioactive process waste handling and monitoring systems associated with the Reactor Building. These are the ventilating air system, electrical power, the hot drain system and the Remote Area Monitoring System (RAMS).

The RB was heated by steam supplied from boilers in the SEB. The 0 ft. and mezzanine open areas were heated by seven steam unit heaters and two steam heaters which heated fresh air from two intake units on the building roof. Reactor Building air exhausted through a pipe chase connection on the south wall of the building to the Hot Pipe Tunnel (HPT) and on to the Fan House where it was monitored, filtered and released (or diverted to holdup tanks) via the PBRF 100 ft. exhaust stack. The Control Rooms and offices were heated and cooled by unit air conditioners Electrical power was supplied to the PBRF via two 34.5 kV, 3 phase 60 cycle transmission lines from nearby local utility substations. These two lines were independent and each was of sufficient capacity to carry the entire electrical load of the PBRF. Originally, four 100 KVA continuously operating diesel generators located in the SEB supplied emergency power in the event of loss of power from the outside lines. Subsequently, the four emergency diesel generators were replaced by two larger capacity units. In addition, a "guaranteed power system" provided backup power upon loss of both commercial power and emergency diesel power. The guaranteed power system consisted of two battery powered motor-generator sets that provided power for vital reactor instrumentation and control functions such as reactor core instrumentation and control rod drive motors.

Potentially contaminated waste water from the RB was collected in sumps and sent to the PBRF Hot Drain System for processing. Floor and equipment drains from the -15 foot level drained to the cold sump at the -25 ft. level under the west stairway. Floor and equipment drains at the -25 ft. level drained to the hot sump in the PPP Room. Three other sumps were located in the RB: a hot sump on the east side at the -15 ft. elevation (below the decontamination room) which pumped directly to the HPT-Hot Lab hot drain line, a sanitary sump on north side -15 ft. level and a cold sump south of the east stairway -15 ft. elevation.

There also were two deep well sumps - one on the -15 and one on the -25 ft. elevation.

A significant amount of embedded piping was located below grade level in the Reactor Building. The primary cooling water supply and return was embedded in concrete sheathing between the Reactor Building and the primary pump house. The quadrant and canal system piping was embedded below the -25 foot floor level between the CV and the PPP room.

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Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 The RAMS provided essential information on radiological conditions in the CV and Reactor Building and throughout the PBRF. The system provided continuous radiation monitoring information to the Health Safety Operations Office (HSOO) in the Reactor Office and Laboratory Building (ROLB) and to a panel in the Reactor Control Room. The RAMS had the capacity to receive information from up to 100 field monitoring units located throughout the PBRF. Field units included radiation detection instruments for monitoring airborne activity (particulate and radioiodine), activity in process water and direct radiation levels. The base units (containing receiver channels, recorders and alarm annunciators) were located in the HSOO. Detectors for airborne particulate activity and direct radiation were located in the following areas of the RB: 4

• - 25 ft. sub-basement - particulate:

o PPP Room - particulate and direct radiation.

0 - 15 ft. CPT entrance - particulate.

• 0 ft. MUR area (Canal H) - water activity and direct radiation (beta-gamma):

o over Canal G - direct radiation (slow neutron) and o Canal G water activity (beta-gamma).

• 12 ft. (mezzanine) - particulate and gaseous.

2.3 Building Modifications Following initial criticality of the 60 MW Test Reactor in 1961, modifications to the Reactor Building and facilities within occurred throughout the period 1961 to 1973. The modifications were made to improve the efficiency and safety of operations and to support the experimental programs. Modifications included:

• A decontamination room with an emergency shower was added to the area outside the east airlock on the 0 ft. elevation (1969).

• Partitions were installed on the mezzanine level to convert a shop area to offices to house shift support staff (Rooms 1 through 4).

" Shelving and storage racks were added to Rooms 202 - 204 in the Mezzanine for storage of drawings, electrical and electronics parts.

" Modifications to the Experiment Control Room were made to centralize the operation of experiments and to upgrade the computerized data logging system. This included modification and expansion of Rooms 102 - 103 to serve as the Experiment Control Room annex.

• A fenced-locked controlled area was added to the - 15 ft. elevation, northwest side for storing radioactive sources and miscellaneous small shielded and unshielded radioactive material items.

• A recycle cleanup system for Canal H was added to the -15 ft. elevation south of the east stairwell.

4 Descriptions of the RAMS are provided in the PBRF Operations Procedures Manual, system Procedure OSY-7617 located in PBRF Records Management Files Box #23, Remote Area Monitoring System. Also, PBRF Training Manual, Section 8.4, Remote Area Monitoring System.

8

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Spent fuel storage capacity was expanded by the installation of additional cadmium-lined spent fuel storage racks in Canal G.

A major modification involved installation of the cathodic protection system in 1961. Twenty anode stations were installed around the perimeter of the CV approximately 18 degrees apart.

Anode stations #1, #2, and #3 are located on the 0' level between the CV and Canal G &

Canal H. Anode stations #4-15 are located on the -15' level and encircle the CV from the vicinity of the east stairwell to the west stairwell. The last five anode stations (#16-20) are located on the -25' level and circle the CV from the vicinity of the cold sump to the Canal F wall.

The anode stations are constructed of 6 in. pipe casing extending down to bedrock at depths between the -35 ft. 6 in. and -50 ft. 6 in. elevations. They contain from one to three mixed metal oxide anodes 2 inches in diameter and 5 ft. long. The stations were filled to the top with a dry calcined petroleum coke breeze. The stations were topped with 6 x 6 x 4 in. junction boxes embedded in the concrete floor for termination of the anode wiring. The stations 5 are all connected with either 3/4or 1 inch conduit which is embedded in the concrete floor.

2.4 Final Configuration and Scope All systems and equipment including associated mechanical and electrical equipment have been removed from the Reactor Building. The scope of FSS reported in this attachment includes building interior surfaces and exterior surfaces. The FSS also covers remaining surface attachments, temporary safety covers and small embedded fixtures, for example unistruts, pipe and conduit stubs. It does not include piping embedded in Reactor Building concrete or piping buried beneath or adjacent to the building. These results are reported in separate attachments to the FSS Report.6 Exhibits 11 through 19 of Appendix A show the condition of the principal Reactor Building areas at the time of the FSS (2011). See Section 3.5 for a discussion of equipment and materials removed during RB decommissioning.

3.0 History of Operations A chronology of major milestones is given below. This is followed by a discussion of Reactor Building operations, post-shutdown and decommissioning activities. Emphasis is on operations with radioactive materials that could affect the final building condition and final status survey. 7

' The FSS of RB cathodic protection wells is covered in the Final Report Attachment 17, Buried and Miscellaneous Piping.

6 The FSS of embedded piping is reported in Attachment 9 of the FSS Report. The FSS of buried and miscellaneous piping is reported in Attachment 17 of the FSS Report.

7 Information sources for the history and pre-decommissioning period include, construction photos, construction drawings, PBRF operating cycle reports, Radiochemistry periodic reports, PBRF Annual Reports, Unusual Occurrence Files, memoranda and other historical files maintained by PBRF Document Control.

9

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 3.1 Chronology Major PBRF and Reactor Building milestones are listed below:

1956 - September, groundbreaking for PBRF.

1959 - Main Reactor Building structure completed.

1961 - June, 60 MW Test Reactor critical.

1961- 1962 Preoperational testing.

1963 - Full power 60 MW Test Reactor operations begin.

8 1973 - January 5th, Reactor shutdown (after 152 operating cycles).

1973 - June 30, PBRF facilities placed in "standby condition.

1985 - Initial radiological characterization, Teledyne Isotopes Inc.

2002 - Decommissioning Plan approved.

2003 - 2005 - Equipment removal and initial building decontamination.

2006 - 2011 - Remediation of contaminated areas and preparation for FSS.

2011 - FSS measurements completed.

3.2 Startup and Operations Construction photos show that the exterior of the RB was complete in late 1959. Interior work, such as completion of the process systems, installation of the Mockup Reactor and its control room, the 60 MW Test Reactor and experiment control rooms and installation of the air lock doors continued into 1963 or later.

The Reactor Building was the hub of PBRF activity from 1959 through 1973. The staff located in the building included technical services managers, shift operations supervisors, mechanics, reactor operators, experiment operators, maintenance and support staff, electronic technicians, and MUR operations staff. Operations in the RB included a large number of personnel during the day shift (30-50) Monday through Friday and a limited number (10 to

15) on the back shifts seven days per week.

Routine operations activities included:

• operation of the 60 MW Test Reactor and MUR,

" unloading spent fuel into Canal G, 8 The length of an operating cycle was determined by fuel bum-up in the 60 MW Test Reactor. Loss of reactivity, usually driven by xenon poisoning, dictated when the reactor was shut down and refueled. The typical cycle duration was two weeks; three days for refueling and 11 days of operating time. Some shutdown periods extended longer than three days, for example for experiment installation, reactor modifications and maintenance.

10

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0

• cropping spent fuel elements in preparation for shipment and loading spent fuel into casks for shipment off-site,

" transferring irradiated materials to and from the HL through Canal F using underwater carts,

• filling canals with de-ionized water; draining the canals when needed,

" performing remote operations for the HL through the RB south wall,

• operating and maintaining auxiliary equipment such as the quadrant and canal chillers and the guaranteed power system,

• moving un-irradiated experiments and reactor hardware into the CV,

• experiment control and data collection and

" decontamination of small items.

3.3 Radioactive Materials in the Reactor Building The US Atomic Energy Commission (AEC) authorized operations and use of radioactive materials at the PBRF under several licenses. 9 License No. TR-3 (Docket 50-30) authorized the 60 MW Test Reactor. The 100 KW Mock-up Reactor was licensed under License No. R-

93. A broad byproduct license (BPL) No. 34-06706-03, authorized possession and use of radioactive materials (byproduct material) produced by the Plum Brook 60MW and Mockup Reactors and other radioactive materials. Radioactive materials in the Reactor Building were those originating from PBRF operations, tests and experiments [PBRF 2009].

Handling of radioactive materials in the Reactor Building usually involved the use of shielding casks and underwater manipulations. Long handling tools were used to move spent nuclear fuel elements and to load and unload fuel elements and experiments in the MUR.

Certain operations required that personnel wear protective clothing. These included cleaning hot sumps, decontamination work and maintenance on the MUR. Some tasks required safe work permits due to the potential for contamination and radiation exposure, e.g., loading spent fuel for shipment. Frequent shielded transfers of primary cooling water samples were made from the PPH to the ROLB through the RB. Shielded transfers of irradiated rabbits or specimens such as flux wires were made to the ROLB laboratories through the RB from the MUR, CV or HL. Such activities did occasionally result in unintended spills of radioactively contaminated materials in RB areas.

Airborne radioactivity often resulted from off gassing into the PPP Room when draining primary cooling water from the reactor vessel during shutdowns. Leaking valves or contaminated water dripping off underwater handling tools led to occasional surface contamination incidents.

Table 1 lists incidents reported in PBRF Operational Cycle reports that involved Reactor Building operations. The list is not all-inclusive, but rather shows typical events that occurred during operations.

9 Authority for the PBRF reactor and radioactive materials licenses was assumed by the US Nuclear Regulatory Commission in 1975.

11

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 1, Reactor Building Activities Covered in Unusual Incident Reports (1960-1973)

Report Date Description O Cycle No.

The RB floor from the CV entrance to the ROLB hallway was contaminated to 6 9/28/63 7100 dpm/ft. 2. Personnel with contaminated feet apparently were the cause.

The area was decontaminated to less than 50 dpm/ft (P3-y).

Quadrant C water was accidentally released to the PPP Room sump causing it 7 10/21/63 to overflow and contaminate the PPP Room floor and a portion of the RB -25 foot level. The impacted areas were decontaminated on 10/22/63.

A leaking pump contaminated about 500 square feet of floor area adjacent to 12 2/6/64 Canal F. Surface contamination levels of up to 1000 dpm/100-cm 2 (03-y) were measured. The area was decontaminated.

Water was detected leaking from a cathodic protection well at ft. level 16/17 4/18/64 near Canal F. A sample was collected and analyzed. It indicated less than lxl0"7 jtCi/cc (t3-y).

Three incidents occurred where floors were contaminated (one in the CV 19 5/19- change room, one in the RB change room shower and one in the RB 2 6/11/64 decontamination room). Contamination levels of up to 4000 cpM/100-cm (f3-y) were measured The impacted areas were decontaminated.

The floor area and walkway near Canal F were contaminated up to 3500 21 7/14/64 dpm/1 00-cm 2 beta-gamma, from water spilled during an experiment transfer.

The impacted area was decontaminated.

While transferring specimens from the CV, contaminated water spilled on the floor outside the Experiment Control Room. Surface contamination of up to 34 5/12/65 2280 dpm/100-cm 2 (03-y) was measured on the floor. The area was decontaminated to less than 200 dpm/100-cm2 (P--Y).

A rupture disc in the primary cooling emergency fill line leaked primary 45 4/23/66 cooling water into the RB sump at the -15 foot level. The major isotopes involved were Rb-88 and Cs-138. The sump pumped out directly to the storm sewer ditch. A water sample measured 5.9x10-4 tCi/ml (03-y).

49 7/12/66 Heavy rains caused major flooding of RB basement. (Similar heavy rains occurred in July 1969 which also flooded the basement.)

Flushing the reactor tank caused high airborne activity in the HRA and 67 10/15/67 PPP room. The RAMS alarms sounded. Fission products Kr-85, Rb-88, Xe-135 and Cs-138 were the principal isotopes.

Water leaked from the Primary Cooling Water (PCW) makeup line at west end 109 6/9/70 of the Cold Pipe Tunnel. The measured activity level was 1.08x10-1 jCi/ml (03-y). The area was decontaminated. A similar incident occurred the next day.

127 5/21/71 A water leak in the RB/PPH wall was measured at 3x 10' jCi/mI (P3-y). The impacted area was decontaminated.

Table 1 Note:

1. Units of measure and concentration limits reported in event descriptions may not be consistent with those in current PBRF procedures and current regulations.

12

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 3.4 Post-Shutdown Materials Disposition and Characterization In the period between termination of reactor operations in January 1973 and June 30, 1973, the Reactor Building was placed in standby status as was the entire PBRF. End condition statements were prepared which governed the status of each system for the protected safe shutdown mode.

Notification was received on January 5, 1973 that NASA was terminating all nuclear related research operations at PBRF due to budget constraints. The Test Reactor, Mock-up Reactor, and all associated operations were to be shutdown and placed in a standby condition and the reactor staff released by June 30, 1973. Following notification, the 60 MW Test Reactor was immediately shutdown. A Master Plan was developed to address activities associated with terminating the operating licenses for PBRF and placing the facility in a standby status. Plum Brook Reactor Facility End Condition Statements for Protected Safe Storage Mode detailed the facility final condition status goals for mid-1973, including the Reactor Building.

During the period between 1973 and the start of decommissioning, activities at PBRF were controlled in accordance with the modified AEC and NRC licenses: TR-3, R-93 and BPL No.

34-06706-03. These licenses authorized possession only of the remaining radioactive materials on site, i.e., no facility operations were permitted. During this period, selected equipment, materials, and waste (both low-level radioactive and non-radioactive) were removed to other locations or discarded as the projected long-term considerations for the facility changed from possible restart to standby to decommissioning. In 1982, the NRC terminated BPL 34-06706-03 based on NASA's request. Licenses TR-3 and R-93 were amended, transferring any remaining licensed radioactive materials to those licenses. For a brief history of the activities during this period see the NASA PBRF Decommissioning Plan, Section 1.2.1 Decommissioning Historical Overview [NASA 2008].

The radiological status of the Reactor Building has been investigated during the period between shutdown in January 1973 and start of decommissioning in 2002. The Reactor Building was included in an evaluation performed by Teledyne Isotopes, Inc. during 1984-86.

The results were reported in a 1987 Report [TELE 1987]. The Teledyne Isotopes report indicated that the majority of the radionuclide inventory in the Reactor Building (exclusive of the 60 MW test reactor in the CV) was from residual contamination in sumps and drains, piping and equipment. Reactor Building structures were contaminated in areas where radioactive material was handled and areas in contact with Quad and Canal water and where process piping and equipment leaked.

In the Teledyne study, the Reactor Building floors and walls outside of the CV were investigated by collection of measurements and samples at 225 locations. Measured levels of fixed beta-gamma surface contamination ranged from < 100 to 5,000 cpm. 10 Removable beta-gamma surface activity levels ranged from non-detectable to 352 dpm/100-cm 2. Fixed alpha surface contamination levels ranged from non-detectable to 22 cpm. Removable surface 10Fixed contamination alpha and beta-gamma activity measurements were reported in cpm. The Teledyne Report did not specify whether the measurements were net or gross cpm. Detection efficiencies and sensitive areas of detectors used were not reported, so conversion to conventional units used today (dpm/1 00-cm 2) was not attempted.

13

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 alpha contamination levels ranged from non-detectable to 5 dpm/1 00-cm 2 . Gamma radiation levels measured at one meter from surfaces ranged from 5 to 230 ftR/h.

Canals F, G and H were divided into 15 ft. grids for investigation of surface contamination and direct radiation. Measurements were taken at 62 locations. Note that Canal J (in the Hot Lab) measurements were reported with Canal F results and are not separable. Measured levels of fixed beta-gamma surface contamination ranged from 50 to 10,000 cpm. Removable beta-gamma surface activity levels ranged from 2 to 4689 dpm/100-cm 2 . Fixed alpha surface contamination levels ranged from non-detectable to 429 cpm. Removable surface alpha contamination levels ranged from non-detectable to 5 dpm/100-cm 2. Gamma radiation levels measured at one meter from surfaces ranged from 4 to 300 jtR/h.

A characterization survey of the PBRF was performed by GTS-Duratek in 1998. This survey included measurements of total surface beta activity and removable surface alpha and beta activity at several hundred measurement points in the RB. The - 25 ft., -15 ft., 0 ft. and the mezzanine floors were marked off into grids (approximately 175 grids). No surface activity measurements were taken in the canals and only a few measurements were taken on the walls of the major elevations (lower walls - below 2 meters). Results for 107 total surface beta activity measurements are summarized in Table 2.

Table 2, Summary of 1998 Surface Beta Contamination Measurements in the RB Major Elevation Range (dpm/100-cm 2) Average (dpm/100-cm 2)

-25 ft. non-detect to 7,060 740

-15 ft. 180 to 43,000 1860 0 ft. non-detect to 520 195 12 ft. Non-detect to 440 220 The total surface beta activity measurement results reported by GTS-Duratek were quite low in general with only one area showing significant contamination. This was a localized area on the floor of the east side of the -15 ft. elevation. Smears were collected at 136 locations in the gridded areas and counted for alpha and beta activity; all were below MDA (alpha MDA =

8.8 dpm and beta MDA = 14 dpm). Samples of sediment/sludge were collected from two RB sumps and analyzed by gamma spectroscopy. The PPP Room Hot Sump sample results were 379 pCi/g Co-60 and 183 pCi/g Cs-137. The decontamination room sump sample results were 35 pCi/g Co-60 and 10 pCi/g Cs-137. Two concrete core samples were collected from floors and analyzed by gamma spectroscopy. These were four-in, cores drilled to a depth of approximately 3 in. and counted as a whole. One was collected from the area on the -15 ft.

elevation where the high total surface beta activity was measured. The results for this sample were 0.1 pCi/g Co-60 and 0.2 pCi/g Cs-137. The second core was collected from the northeast floor area of Canal F and the results were 156 pCi/g Co-60 and 2.7 pCi/g Cs-137 [GTS 1998].

14

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 3.5 Decommissioning The Decommissioning Plan identified the MUR and all contaminated equipment, piping and drains in the Reactor Building as items to be removed in the decommissioning process

[NASA 2008]. The extent of surface remediation that would be required was not completely defined at the time the Decommissioning Plan was submitted (2000). Most of the accessible floor areas and lower walls in the building experimental areas exhibited low levels of surface contamination. However, the floor trenches and sumps in the -15 and -25 ft. elevations had not been characterized to any significant degree. Canals F, G and H had received limited coverage (in the 1984-86 Teledyne survey). These areas were believed to be contaminated at levels that would require remediation.

During the period of active decommissioning, extensive radiological surveys were performed to define the extent of Reactor Building structures requiring remediation and to guide remediation efforts. Summaries of survey results are provided to illustrate contamination levels in the Reactor Building prior to remediation of contaminated areas. A report by the decommissioning contractor included results of characterization surveys performed on Reactor Building structural surfaces during 2003 and 2004 [PBRF 2004]. This report and the supporting characterization packages provided the basis for the MARSSIM classification of principal PBRF areas in the FSS Plan, Table 2-1, Survey Area Classification. [NASA 2007].

The report addressed structural surface contamination levels in Reactor Building areas that had not previously been well characterized. For example, total surface beta activity levels (in dpm/100-cm 2) of up to 6.1E+06 were measured in Canal G; 5.49E+04 in Canal H; 3.8E+04 in the -15 ft. elevation work area and 2.5E+04 in the -25 ft. elevation PPP Room.

Additional characterization surveys were performed in 2004 and 2005 to determine radionuclide mixtures, and estimate depth of penetration in contaminated PBRF building concrete floors, walls and floor-wall joints (by obtaining core bore samples). It was concluded from these measurements that reactor produced radionuclides had penetrated to depths of up to one-half inch in Reactor Building floors at the - 15 and -25 ft. elevations and into Canal G and H walls (Canal F was not sampled). A core sample taken in a PPP Room trench showed slightly deeper penetration (- 3/4 in.) [PBRF 2005, PBRF 2005a].

Canals F, G and H were normally filled with water during Reactor operations. The PPP Room and the -25 ft. elevation sub-basement floors and trenches were also exposed to contaminated water for extended periods. From detailed characterization surveys performed in 2005 and 2006, it was determined that most of the contamination was in areas where the painted/mastic surface coating had failed, and at wall/floor joints, unistruts, seams, bolt/anchor holes, around penetrations and in the "bathtub rings" [PBRF 2005b, PBRF 2006].

Decommissioning of the RB proceeded in phases. In the initial phase, from 2002 through approximately 2006, the principal focus was on removing free-standing and fixed equipment, including exposed valves, piping and supporting equipment. The RB was cleared of removable items such as piping, pumps, valve extensions and handles, electrical boxes, cable trays, ventilation systems, electronic instrumentation (including control room panels and 15

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 equipment), experiment hardware, etc. This included waste disposal of radioactively contaminated items. Recyclable items were segregated where possible.

Following removal of fixed equipment, asbestos abatement was performed. Asbestos abatement work in the Reactor Building was initiated in late 2007 and continued through April 2008. Asbestos containing materials included mastic-sealant on water-filled area surfaces, floor tiles and thermal system insulation and gasket material that remained after fixed equipment removal. Techniques for removal of asbestos-containing material included sponge-jet blasting, wall and floor shaving and scabbling using "box scabbling" equipment and hand tools." It is noted that asbestos abatement in effect was the first stage of decontamination of RB surfaces because much of the asbestos-containing material was contaminated or close contact with contaminated piping and equipment remnants.

In preparation for the final phase of decommissioning, extensive walk-downs and inspections of the Reactor Building (and other structures) were performed in 2007 to identify items to be removed and surfaces to be prepared for the FSS. These resulted in "work lists" for the decommissioning contractor.

A variety of techniques were used to prepare RB surfaces for FSS. Wall surface coatings of water-filled canals were removed by sponge jet blasting and the underlying concrete was shaved or scabbled. Canals F, G and H, -15 ft. and -25 ft. floors, sumps and trenches were shaved and scabbled. Pipe stubs, equipment mounting brackets and bolts, door frames and other protuberances were removed by flame cutting. Areas where elevated activity was measured and where contamination had penetrated to depths were remediated by over-coring and use of impact-hammering tools. Exhibits 8 through 19 of Appendix A show views of the main RB areas before and after preparation for FSS.

Prior to the FSS of the structure, embedded piping and other wall and floor penetrations in the RB were remediated and surveyed to meet the appropriate release criteria. After completion of FSS measurements, the remediated embedded and buried piping for floor drains, quadrant and canal drains, recirculation lines, and primary cooling lines left in place were grouted to meet FSS Plan requirements.

4.0 Survey Design and Implementation for the Reactor Building This section describes the method for determination of the number of fixed measurements and samples for the Reactor Building FSS. Applicable requirements of the FSS Plan are summarized.

These include the DCGLw 12, the gross activity DCGL, scan survey coverage and action-investigation levels, classification of areas and breakdown into survey units. The radiological instrumentation and their detection sensitivities are discussed.

11 PBRF Work Execution Package, PBRF-WEP-06-004, Asbestos Abatement in Main PBRF Buildings, approved 9/14/07, closed 7/10/08.

12 The convention used in the MARSSIM is to identify the DCGL used as the benchmark for evaluating survey unit measurement results, as the DCGLw. The "W" subscript denotes "Wilcoxon", regardless of the particular test used (Wilcoxon Rank Sum Test, or Sign Test).

16

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 4.1 FSS Plan Requirements The DCGLs for individual radionuclides were established for PBRF structures considering exposure to future site occupants using two exposure scenarios. Single radionuclide DCGLs were calculated using RESRAD-BUILD Version 3.22 for a building reuse scenario. Single radionuclide volumetric DCGLs were calculated for subsurface structures using RESRAD Version 6.21 for a resident farmer scenario.' 3 The volumetric DCGLs (in pCi/g) were converted to "effective surface" DCGLs (in dpmr/100-cm 2) using surface-to-volume ratios for the assumed volume of contaminated subsurface concrete. The DCGL calculations are described in the FSSP, Attachment B. To obtain the DCGLs for PBRF structures, the smaller of the two DCGLs calculated for each of the radionuclides of concern was selected.

A gross activity DCGL is used for structural surfaces in the PBRF, where multiple radionuclides are potentially present in residual contamination. The gross activity DCGL accounts for the presence of multiple radionuclides, including beta-gamma and alpha emitters.

The gross activity DCGL can also account for so-called hard-to-detect (HTD) radionuclides.

The latter are not detected, or detected with very low efficiency, by the beta detectors selected for the FSS of structures.

The gross activity DCGL for the Reactor Building is calculated using equations in Section 3.6 of the FSSP for gross beta, gross alpha and surrogate DCGLs, based on the radionuclide mixture in residual contamination. Activity fractions and the gross activity DCGLs for the Reactor Building are shown in Table 3.

Table 3, Reactor Building Radionuclide Activity Fractions and Gross Activity DCGLs Radionuclides DCGL Location H-3 Co-60 Sr-90 1-129 Cs-137 Eu-154 U-234 U-235 (dpm/100-cm2) (2)

Activity Fractions Assigned to Reactor Building (%) 1 Canals F, G, H 1.16 11.69 0 0 86.99 0 0.16 0 31,711 Mezzanine & 0 54.05 5.30 11.45 0.09 28.72 0.39 0 0 29,060 ft. el.

-15 ft. el. 0 0 0 0 48.40 0 51.60 0 35,296

- 25 ft. Sub-basement, PPP 7.29 53.89 22.22 0 12.05 1.71 2.42 0.42 14,600 Rm. & pipe trench RB Exterior, Default (3) 27.07 9.65 7.88 1.42 46.71 0.12 6.98 0.17 27,166 Table 3 Notes:

1. Activity profiles and gross activity DCGLs for structures are reported in the Technical Basis Document PBRF-TBD-07-001 [PBRF 2007]. As discussed in Appendix D, the gross activity 13Potential exposure to future occupants from subsurface structures could occur from contaminated concrete rubble placed as fill and from contaminated intact structures such as the below-grade portion of the Reactor Bioshield.

17

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 DCGLs and radionuclide activity fractions have been revised to correct errors discovered in TBD-07-001. However, the information in Table 3 was used in the design and evaluation of the FSS measurements and is reported here as such. The re-evaluation reported in Appendix D shows that the FSS Plan requirements remain satisfied, considering all the effects of revised DCGLs, radionuclide mixtures and activity fractions.

2. These are gross activity DCGL values calculated for the radionuclide mixtures shown in the table.

In Reactor Building survey designs, the DCGL values may be adjusted to account for dose contributions from "insignificant radionuclides" and embedded piping (see Table 7 for details).

3. The default radionuclide mixture for PBRF structures and the associated DCGL in TBD-07-001 were applied to the RB roof and exterior walls.

Survey designs incorporate requirements for scan coverage and investigation levels derived from the MARSSIM classification of survey units. The values applicable to the Reactor Building are shown in Table 4.

Table 4, Class-Based Survey Scan Coverage and Action Level Requirements Static Measurement Classification Scan Survey Scan Investigation or Sample Result Coverage Levels Investigation Levels Class 1 100% > DCGLEMC > DCGLEMC Class 2 10 to 100% >if DCGLw MDCscan or is >> MDCscan DCGLw > DCGLw Class Clas_3_ 3 Minimum of 10%

Mnimu of_____ 10%_ >ifDCGLw MDCcan or is >> MDCscan DCGLw >- 50% of the DCGLw 4.2 Area Classification and Survey Unit Breakdown The Reactor Building was divided into 34 areas for initial classification and final status survey planning as shown in Table 2-1 of the FSS Plan. Areas identified as Class 1 included the Canals, the PPP Room and sump, adjacent areas on the -25 ft. elevation, and the -15 ft.

elevation equipment areas. The Reactor Control Room and offices on the Mezzanine (12 ft.

elevation) were identified as Class 2. Offices and the Experiment Control room on the 0 ft.

elevation (Rooms 102 through 105) were identified as Class 3. The Reactor Building roof was classified as Class 2 and the building exterior walls as Class 3. As part of FSS implementation, individual survey units were established and their final MARSSIM classification assigned. The Reactor Building was divided into 130 survey units for the FSS (129 structure survey units and one soil survey unit). The breakdown by MARSSIM classification is: 116 Class 1, 13 Class 2 and one Class 3. Table 5 shows the survey unit breakdown for the FSS. For each survey unit, the table identifies the final MARSSIM classification, the surface area (in M 2 ), the Survey Design and Survey Request Nos., the description and for comparison, the classification assigned in the FSSP.

18

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 6 summarizes the survey unit breakdown by major elevation. 14 The convention used for survey unit identification is: RB-1 = 0 ft. elevation, RB-2 = 12 ft. elevation (mezzanine), RB-3 = -15 ft. elevation, RB-4 = -25 ft. elevation and RB-5 = Roof and Exterior Walls.

Table 5, RB Survey Units for FSS D nAreaSR No. Class in Unit Class (M) Design () Description FSSP (2)

UNO.2 RB-I-1 1 71.8 66 327/356 Floor- Section #1 1 RB-1-2 1 75.0 66 327/356 Floor - Section #2 1 RB-1-3 1 72.6 66 327/356 Floor - Section #3 1 RB-1-4 1 72.0 66 327/356 Floor - Section #4 1 RB-1-5 1 57.3 66 327/356 Floor - Section #5 1 RB-1-6 1 71.6 66 327/356 Floor - Section #6 1 RB-1-7 1 60.0 66 327/356 Floor - Section #7 1 RB-1-8 1 69.9 66 327/356 Floor - Section #8 1 RB-1-9 1 72.4 66 327/356 Floor - Section #9 1 RB-I-10 1 71.9 66 327/356 Floor- Section#10 1 RB-1-11 1 68.3 66 327/56 Floor - Section #11 1 RB-I-12 1 53.6 66 327/356 Floor- Section #12 1 RB-I-13 1 66.1 66 327/356 Floor- Section#13 1 RB-1-14 1 66.6 66 327/356 Floor - Section #14 1 RB-1-i5 1 94.7 66 327/ 57 CV Dome Section #1 Exterior

& ColumnsWall - Lower Wall 2 RB-1-16 95.8 CV Dome Exterior Wall - Lower Wall 2 1 66 327/357 Section #2 RB-1-17 2 620 65 328/355 CV Dome Exterior Wall - Upper Wall 2 RB-1-18 2 481 65 328/355 RB East Wall - Upper Wall 2 RB-1-19 2 768 65 328/355 RB South Wall - Upper Wall 2 RB-1-20 2 492 65 328/355 RB West Wall - Upper Wall 2 RB-1-21 2 449 65 328/355 RB North Wall - Upper Wall 2 RB-1-22 1 48.5 66 327/357 RB Lower South Wall - West End 2 RB-1-23 1 67.2 66 327/357 RB Lower South Wall - East End 2 RB-1-24 1 85.7 66 327/357 RB Lower East Wall - South End 2 RB-1-25 1 85.4 66 327/357 RB Lower East Wall, North End, North 2 Wall, East End RB-1-26 1 99.4 66 327/3572 RB Lower North Office Wall & Lower Stairway RB-1-27 1 99.7 66 327/357 RB Lower West Wall 2 RB-1-28 2 615 65 328/355 RB North Ceiling 2 RB-1-29 2 737 65 328/355 RB Southeast Ceiling 2 RB-1-30 2 627 65 328/355 RB Southwest Ceiling 2 RB-1-31 1 40.7 66 327/356 RB 0' Offices - Floor Section #1 2 14 The calculations performed in preparation of this report are documented in a memorandum to the PBRF Decommissioning Project File [PBRF 20121.

19

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 5, RB Survey Units for FSS Survey Area SurveyClass in Unit Class (mi) Desuvn SN) Description FSSP (2)

No.

RB-1-32 1 61.2 66 327/356 RB 0' Offices - Floor Section #2 2 RB-1-33 1 51.8 66 327/356 RB 0' Offices - Floor Section #3 2 RB-1-34 1 62.6 66 327/356 RB 0' Offices - Floor Section #4 2 RB-1-35 1 49.4 66 327/356 RB 0' Offices - Floor Section #5 2 RB-1-36 1 69.1 66 327/356 RB 0' Offices - Floor Section #6 2 RB-1-37 1 88.0 66 327/357 RB 0' Wal2 Offices - Rooms 7 - 10 Lower Walls 2 RB-1-38 1 25.3 66 327/357 RB 0' Offices - Room 6 Lower Walls 2 RB-1-39 1 97.0 66 327/357 RB 0' Offices Lower Walls - Room 5 Control Room 2 RB-1-40 1 78.7 66 327/357 RB 0' Offices - Room 102/103 Lower 3 Walls RB-1-41 1 49.1 66 327/357 RB 0' Offices - Room 104 Lower Walls 3 RB-1-42 1 54.5 66 327/357 RB 0' Offices - Room 105 Lower Walls 3 RB-1-43 2 872 55 328/355 RB 0' Offices - Upper Walls & Ceilings 3 RB-2-1 2 994 59 310 Mezzanine Rooms 202, 203, 204, 205, 2 RB-2-2 2 998 59 311 Mezzanine Rooms 15, 16, 28, Pipe 2 Chase, 206, 208, 209 RB-3-1 1 10.5 70A 303 RB -15' Cold Sump I RB-3-2 1 6.5 70B 332 RB -15' SAN Sump Soil 1 RB-3-3 1 73.8 61A 312 RB -15' - Floor Section #1 1 RB-3-4 1 74.7 61A 312 RB -15' - Floor Section #2 1 RB-3-5 1 74.9 61A 312 RB -15' - Floor Section #3 1 RB-3-6 1 74 61A 312 RB -15' - Floor Section #4 1 RB-3-7 1 73.1 61A 312 RB -15' - Floor Section #5 1 RB-3-8 1 73.5 61A 312 RB -15' - Floor Section #6 1 RB-3-9 1 74.5 61A 312 RB -15' - Floor Section #7 1 RB-3-10 1 74.5 61A 312 RB -15' - Floor Section #8 1 RB-3-11 1 74 61A 312 RB -15' - Floor Section #9 1 RB-3-12 1 74.4 61A 312 RB -15' - Floor Section #10 1 RB-3-13 1 46.4 61A 335 RB -15' - East Stairway I RB-3-14 1 46.4 61A 312 RB -15' - West Stairway 1 RB-3-15 1 92.2 61B 313 RB -15' - Wall Section #1 1 RB-3-16 1 97.9 61B 313 RB -15' - Wall Section #2 1 RB-3-17 1 91.3 61B 313 RB -15' - Wall Section #3 1 RB-3-18 1 90.8 61B 313 RB -15' -Wall Section #4 1 RB-3-19 1 89.7 61B 313 RB -15' - Wall Section #5 1 RB-3-20 1 97.7 61B 313 RB -15' - Wall Section #6 1 RB-3-21 1 98.4 61B 313 RB -15' - Wall Section #7 1 RB-3-22 1 98.9 61B 313 RB -15' - Wall Section #8 1 20

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 5, RB Survey Units for FSS Survey Area Survey SR No. Class in Unit Class (M) Design No) Description FSSP (2)

Uni Cas ( No.

RB-3-23 1 47.2 61A 312 RB -15' - ATS Tunnel Entrance I RB-3-24 1 96 61B 314 RB -15' - Ceiling Section #1 I RB-3-25 1 97.4 61B 314 RB -15' - Ceiling Section #2 1 RB-3-26 1 99.8 61B 314 RB -15' - Ceiling Section #3 1 RB-3-27 1 96.5 61B 314 RB -15' - Ceiling Section #4 1 RB-3-28 1 94.7 61B 314 RB -15' - Ceiling Section #5 1 RB-3-29 1 91.3 61B 314 RB -15' - Ceiling Section #6 1 RB-3-30 1 82.3 61B 314 RB -15' - Ceiling Section #7 1 RB-3-31 1 91.7 61B 314 RB -15' - Ceiling Section #8 1 RB-3-32 1 82.8 61B 314 RB -15' - Ceiling Section #9 1 RB-3-33 1 79.2 61B 314 RB -15' - Ceiling Section #10 1 RB-3-34 1 94.9 61B 314 RB -15' - Ceiling Section #11 1 RB-3-35 1 91.1 61B 314 RB -15' - Ceiling Section #12 1 RB-3-36 1 88.8 61B 314 RB -15' - Ceiling Section #13 1 RB-3-37 1 74.6 61B 314 RB -15' - Ceiling Section #14 1 RB-3-38 1 53.6 61B 315 RB -15' - East Columns 1 RB-3-39 1 51.9 61B 315 RB -15' - West Columns 1 1 48.5 43 316/337 Canal F - Floor including F-J and F-H doorway 1 RB-4-2 1 96.9 43 316/337 Canal F - West & North Walls 1 RB-4-3 1 97.8 43 316/337 Canal F - East Wall & F-J doorway sides 1 RB-4-4 1 43.4 43 316/337 Canal F - Catwalk steel & undersides 1 RB-4-5 1 55.5 43 316/337 Canal G - Floor & G-H doorway 1 RB-4-6 1 84.2 43 316/337 Canal G - North Wall 1 RB-4-7 1 91.1 43 316/337 Canal G - E & W Walls, & S Wall lower 1 RB-4-8 1 72.2 43 316/337 Canal G - S Wall above 4m, G & H 1 walkway undersides RB-4-9 1 42.3 43 316/337 Canal H - Floor 1 RB-4-10 1 96.2 43 316/337 Canal H - W & N Walls, G-H doorway 1 RB-4-11 1 92.5 43 316/337 Canal H - S & E Walls 1 RB-4-12 1 48.5 60A 306 RB -25' - Trench bottom & Pit sides 1 RB-4-13 1 84.5 60A 306 RB -25' - Trench sides (including RV wall below -25')

RB-4-14 1 74.7 60A 307 RB -25' - Floor Section #1 - East Floor 1 RB-4-15 1 74.7 60A 307 RB -25' - Floor Section #2 - SE & South 1 RB-4-16 1 73.6 60A 307 RB -25' - Floor Section #3 - Center 1 RB-4-17 1 73.9 60A 307 RB -25' - Floor Section #4 - North Floor 1 RB-4-18 1 70.6 60A 306 RB -25' - Pump Room Main Trench Floor & Pipe Chase RB-4-19 1 60.5 60A 306 RB -25' - Pump Room Main Trench 1 _Walls 21

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 5, RB Survey Units for FSS Survey Area Survey SR No. Class in Unit Class (M) Design (1) Description FSSP (2)

No.

RB-4-20 1 71.1 60A 307 RB -25' - Pump Room Floor & Pump 1 RB-4-21 1 62.9 60B 308 RB -25' - Wall Section #1 1 RB-4-22 1 85.5 60B 308 RB -25' - Wall Section #2 1 RB-4-23 1 79.2 60B 308 RB -25' - Wall Section #3 1 RB-4-24 1 92.8 60B 308 RB -25' - Wall Section #4 1 RB-4-25 1 94.2 60B 308 RB -25' - Wall Section #5 1 RB-4-26 1 67.7 60B 308 RB -25' - Wall Section #6 1 RB-4-27 1 65.3 60B 308 RB -25' - Wall Section #7 1 RB-4-28 1 91.4 60B 308 RB -25' - Wall Section #8 1 RB-4-29 1 70.3 60B 308 RB -25' - Columns 1 RB-4-30 1 66.1 60B 309 RB -25' - Pump Room - Mezzanine 1 RB-4-31 1 73.9 60B 308 RB -25' - Pump Room - South & West Walls RB-4-32 1 57.6 60B 308 RB -25' - Pump Room - North, Northeast, & East Walls RB-4-33 1 85.4 60B 309 RB -25' - Pump Room - Ceiling 2 RB-4-34 1 71.6 60B 309 RB -25' - Ceiling Section #1 1 RB-4-35 1 97.8 60B 309 RB -25' - Ceiling Section #2 1 RB-4-36 1 91.9 60B 309 RB -25' - Ceiling Section #3 1 RB-4-37 1 94.3 60B 309 RB -25' - Ceiling Section #4 1 RB-4-38 1 88.8 60B 309 RB -25' - Ceiling Section #5 1 RB-4-39 1 85.5 60B 309 RB -25' - Ceiling Section #6 1 RB-4-40 1 61.5 60B 309 RB -25' - Ceiling Section #7 1 RB-4-41 1 59.1 60B 309 RB -25' - Ceiling Section #8 1 RB-4-42 1 10.5 70C 304 RB -25' - Cold Sump I RB-4-43 1 18.9 70C 304 RB -25' - PPP Room Sump I RB-5-1 3 1384.4 69A 352 RB Exterior Walls 3 RB-5-2 2 729.9 69B 360 RB Roof- North Section 2 RB-5-3 2 712.3 69B 360 RB Roof- South Section 2 Table 5 Notes:

1. Where two SRs are shown for a survey unit, the first one listed covers the post-remediation survey.

The post-remediation surveys included scan surveys and associated investigative measurements.

Scan survey results of the survey unit final condition are included and reported as FSS results.

2. The FSS Plan classification was based on area history and available characterization data.

22

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 6, Reactor Building Survey Unit Breakdown by Major Elevation Major No. of Surface  % of Survey  % of Surface Elevation Survey Units Area (m2) Units Area 0 ft. el 43 7944.8 33.1 42.0 Mezzanine 2 1992.0 1.5 10.5

-15 ft. el. 39 3021.9 30.0 16.0

-25 ft. el 43 3124.9 33.1 16.5 Roof & Ext 3 2826.6 Walls 2.3 14.9 Total 130 18910.2 100 100 4.3 Number of Measurements and Samples The number of measurements and samples for each survey unit was determined using the MARSSIM statistical hypothesis testing framework as outlined in the FSS Plan. The Sign Test is selected because background count rates of instruments used in the FSS are equivalent to a small fraction of the applicable DCGLw.' 5 Decision error probabilities for the Sign Test are set at a = 0.05 (Type I error) and 13= 0.10 (Type II error) in accordance with the FSSP.

The Visual Sample Plan (VSP) software was used to determine the number of FSS measurements in the Reactor Building. 16 When the Sign Test is selected, the VSP software uses MARSSTM Equation 5-2 to calculate the number of measurements. Equation 5-2 is shown below:

N = 1.2 (ZI-a + ZI-,6 L (Equation 1) 4[(,)( A ) -0.5 a I where:

1.2 = adjustment factor to add 20% to the calculated number of samples, per a MARSSIM requirement to provide a margin for measurement sufficiency, N = Number of measurements or samples, a= the type I error probability, 03= the type II error probability, Z1 - = (100-a) percentile of normal distribution (1.6449 for a = 0.05),

ZI.p = (100-3) percentile of normal distribution (1.2816 for 3 = 0.1),

15 Background count rates for the LMI 44-116 detector, the instrument of choice for FSS surface beta activity measurements on structures, are in the range of 300 cpm or less for most materials. This is equivalent to about 2500 dpm/100-cm 2 ; less than 10% of PBRF structure DCGLs (this assumes a detection efficiency of - 12%).

16 The FSS Plan (Section 5.2.4) states that a qualified software product, such as Visual Sample Plan© [PNL 2010], may be used in the survey design process.

23

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 (D(A/a) = value of cumulative standard normal distribution over the interval -

00, A/a, A = the "relative shift", defined as the DCGL - the Lower Bound of the Gray Region (LBGR), and a = the standard deviation of residual contamination in the area to be surveyed (or a similar area). This may include the variation in measured "ambient" background plus the material background (for total surface beta measurements).

The MARSSIM module of VSP requires user inputs for the following parameters: a, 13, LBGR, the DCGLw and cr. The numbers of measurements were calculated for the 130 Reactor Building survey units using the parameters established in 14 survey designs. Table 7 summarizes the Reactor Building survey designs and lists the values of the key VSP input parameters.

Table 7, Reactor Building Survey Design Summary Design No. (1) Survey Units Class DCGLw (2) ( LBGR (2)2 2 A (2) 2 a (2) A/a N 43 (3) Canals F, G, H 1 27,271141 24,166 3105 1035 3.0 11 RB-4-1 thru RB-4-11 59 Area 11 Mezzanine 1 26,154 25,077 1077 359 3.0 15 RB-2-1 & RB-2-2 60A -25 ft. Fl. & Trenches 1 12,556 16) 6,278 6278 2260 2.8 11 RB-4-12 thru RB-4-20 60B -25 ft. Walls & Ceiling RB-4-21 thru 1 12,556 161 6,278 6278 2260 2.8 11 RB-4-41 61A -15 ft. Floor, RB-3-3 thru RB-3-14 & 1 30,355 (7) 26,758 3597 1199 3.0 11 RB 3-23

-15 ft. Walls & Ceiling 61B RB-3-15 thru RB-3-22 & RB3-24 thru 1 30,355 (7) 27,772 2583 861 3.0 11 ERB-3-39 61C -15 ft.E. Stairway resurvey, RB-3-13 1 30,355 (7) 26,758 3597 1199 3.0 11 0 ft. upper walls & ceil.

65 RB-I-17 thru RB-1-21, RB-I-28 thru 2 26,154 (5 25,077 1,077 359 3.0 11 RB-1-30 & RB-1-37 thru RB-1-42 0 ft. Fl. & L.W.

66 RB-1-1 thru RB-1-16, 1 26,154 13,077 13,077 5,230 2.5 11 RB- 1-22 thru RB- 1-27, RB- 1-31 thru RB-1-42 69A Exterior Walls, RB-5-1 3 24,449 ( 12,225 12,225 4890 2.5 11 69B Roof, RB-5-2, RB-5-3 2 24,449 (8) 12,225 12,225 4890 2.5 11 70A Cold Sump -15 ft. RB-3-1 1 31,766 (9) 15,883 15,883 6353 2.5 11 70B San Sump -15 ft. Soil, RB-3-2 1 4.96 (10) 2.48 2.48 0.99 2.5 11 70C -25 ft. Cold Sump & PPP Rm. Sump 1 12,556 (6) 6,278 6278 2511 2.5 11 RB-4-42, RB-4-43 I I II Table 7 Notes:

1. The data reported in this table is obtained from the Survey Design reports listed where the DCGLs published in TBD-07-001 were used to calculate the number of measurements. As shown in Appendix D, the revised DCGLs do not result in any changes in the numbers of measurements shown in this table.

24

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0

2. Units are dpmrIl00-cm2 , except for Survey Design 70B which are pCi/g.

3. This design also covers Canal E in the CV.

4. The gross activity DCGL value in TBD-07-001 (31,711 dpm/100-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides and one mrem/y from embedded piping (EP).

5. The gross activity DCGL value in TBD-07-001 (29,060 dpm/100-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides.

6. The gross activity DCGL value in TBD-07-001 (14,600 dpm/100-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides and one mrem/y from EP.

7. The gross activity DCGL value in TBD-07-001 (35,296 dpm/100-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides and one mrem/y from EP.

8. The gross activity default DCGL value in TBD-07-001 (27,166 dpmrlO00-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides.

9. The gross activity DCGL value in TBD-07-001 (35,296 dpm/100-cm 2) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides.

10. The DCGL of 5.77 pCi/g developed in Att. 6 of the design (Cs-137 Surrogate DCGL for Cs-137 & U-234 mixture) is adjusted to account for 2.5 mrem/y contribution from deselected radionuclides and one mrem/y from EP.

Selection of design input parameters followed guidance in the FSS Plan. The Plan states that "the LBGR is initially set at 0.5 times the DCGLw, but may be adjusted to obtain a value for the relative shift (A/a) between 1 and 3." It is seen in Table 7 that relative shift values in the range of 2.5 to 3.0 were used in the final calculations for determining N.

The VSP software automatically performs an analysis to examine the sensitivity of N, the number of samples, to critical input parameter values. The following is an example obtained from the VSP reports for survey units in Design 43. The sensitivity of N was explored by varying the following parameters: standard deviation, lower bound of gray region (as % of DCGL), beta, probability of mistakenly concluding that the survey unit mean concentration, p, is greater than the DCGL and alpha, probability of mistakenly concluding that the survey unit mean concentration, p, is less than the DCGL. Table 8 summarizes this analysis. The region of interest is for a = 0.05 (required to be fixed), f3 = 0.10 (may be adjusted) and the LBGR at 70% to 90% of the DCGL. In this region (for j3 = 10), doubling a causes an increase in N only when the LBGR is fairly close to the DCGL (90% of the DCGL). For lower values of the LBGR (70 to 80%) N is insensitive to changes in a. The sensitivity of N to an incorrect conclusion that the survey unit will pass (regulator's risk) is quite low; increasing a from 0.05 to 0.10 and 0.15, while holding 3 = 0.10 and a constant at 1035 dpmr/100-cm 2, shows that the number of measurements is 11 or fewer in all cases. These results show that N = 11 represents a conservative design.

25

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 8, Sensitivity Analysis for Reactor Building FSS Design Number of Samples DCGLw = 27,271 () t=0.05 2" a= 0.10 a = 0.15

_= _ 2070 _ o=1035 = 2070 0= 1035 a= 2070 o= 1035 LBGR = 90% (1)(4) 130.05 21 15 16 11 14 10 13=0.10 16 11 12 9 11 8 13=0.15 14 10 11 8 9 6 LBGR = 80% 13=0.05 15 14 11 11 10 10 13=0.10 11 11 9 9 8 8 13=0.15 10 10 8 8 6 6 LBGR = 70% 13=0.05 14 14 11 11 10 10 13=0.10 11 11 9 9 8 8 13=0.15 10 10 8 8 6 6 Table 8 Notes:

1. Units of DCGL, o and LBGR are dpmr/100-cm 2 .

2. a = alpha, probability of mistakenly concluding that p < DCGL.

3. a = Standard Deviation.

4. LBGR = Lower Bound of Gray Region (as % of DCGL).

5. 13= beta, probability of mistakenly concluding that p > DCGL.

Visual Sample Plan was also used to determine the grid size, the random starting location coordinates (for Class 1 and 2 survey units) and to display the measurement locations on survey unit maps drawn to scale. Refer to Appendix B for location coordinate tables and scale VSP maps showing measurement locations for each Reactor Building survey unit.

The survey designs also specify scan survey coverage and action levels based on the MARSSIM classification listed in Table 4. If the scan sensitivity of the detectors used in Class 1 survey units is below the DCGLw, the number of measurements in each survey unit is determined solely by the Sign Test. If the scan sensitivity is not below the DCGLw, the number of measurements is increased as determined by the Elevated Measurement Comparison (EMC). As discussed in the next section, the scan sensitivities of instruments used in the FSS of the Reactor Building are below the DCGLw, and no increase in the number of measurements above the value calculated using the Sign Test was required.

4.4 Instrumentation and Measurement Sensitivity Instruments used in the FSS were selected in the Survey Designs. Their detection sensitivities must be sufficient to meet the required action levels for the MARSSIM class of each survey unit. Minimum detection sensitivities for direct surface activity static and scan measurements are calculated using equations from Section 6.4 of the FSS Plan [NASA 2007]. The static MDC equation is:

26

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 3 + 3.29FBRt,(1l+ts)

MDCstatic = A Tb (Equation 2) 100 where:

MDCstatic = Minimum Detectable Concentration (dpm/ 100-cm 2 ),

BR = Background Count Rate (cpm),

tb = Background Count Time (min),

t, = Sample Count Time (min),

A = Detector Open Area (cm 2 ) and Eto, = Total Detection Efficiency (counts per disintegration). The total efficiency equals the product of Detector Efficiency, Ei and Surface Efficiency, E,.

The scan MDC equation is:

d' 4b- 60 MDC... =i (Equation 3) 100 where:

MDCscan = Minimum Detectable Concentration (dpmr/100-cm 2),

d' = Index of sensitivity related to the detection decision error rate of the surveyor, from Table 6.5 of MARSSIM [USNRC 2000],

i = observation counting interval, detector width (cm) / scan speed (s),

bi= background counts per observation interval, E= Detector Efficiency (counts per disintegration),

Es = Surface Efficiency, typically 50% for beta per ISO 7503-1, Table 2 [ISO 1988],

p = Surveyor efficiency (typically 50%) and A Detector Open Area (cm 2).

A summary of the a priori detection sensitivities of instruments used in the FSS of the Reactor Building is provided in Table 9. Note that the detector sensitivities and other parameter values shown in Table 9 are those published in the survey designs based on the DCGLs in TBD 001. In Appendix D, it is shown that in cases where the DCGLs were reduced in correcting the errors in TBD-07-00 1, scan sensitivities were still sufficient to meet the investigation level requirements in Table 4.

27

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 9, Typical Detection Sensitivities of Field Instruments Detector MDCscan Net cpm MDCstatic Detector Model Efficiency (dm/ 100-cm 2 ) (2) Equivalent to (dpM/100-cm 2)(2)

(c/d) (1) DCGLw LMI 44-116 0.140 2,587) 1 4,066 718(4)

LMI 44-9 0.143 10,025(") 298 3,263(6)

Table 9 Notes:

1. The detector efficiencies listed are total efficiency, i. e., Et = Ei + E,.

2. A priori scan sensitivities are calculated using Equation 3 and static sensitivities are calculated using Equation 2.

3. The scan MDC for the LMI 44-116 is reported in Design No. 70A for background count rate

= 200 cpm; scan speed =15.2 cm/s and E, = 0.5. An efficiency correction factor = 0.8349 is applied to compensate for concrete roughness (the detector-to-surface distance is 0.5 in.).

4. The static MDC for the LMI 44-116 detector is reported in Design No. 70A for background count rate = 200 cpm, E, = 0.5 and the detector-to-surface distance = 0.5 in. (one minute count times are assumed for both the background and sample counts).

5. The scan MDC for the LMI 44-9 is obtained from Survey Design No. 61A. The background count rate is 125 cpm with a scan speed of 4.4 cm/s and the detector in contact with the surface.

6. The static MDC for the LMI 44-9 is obtained from Survey Design No. 61A. The background count rate is 125 cpm and the detector in contact with the surface (one minute count times are assumed for both the background and sample counts).

The scan investigation level for Class 1 survey units listed in Table 4 is the DCGLEMC, as specified in the FSS Plan Section 8.1. However, the scan investigation level is typically set at an instrument count rate corresponding to the DCGLw established in the survey design for each structure survey unit. For example, as seen in Table 9 above, the 44-116 detector count rate that corresponds to the DCGLw is 4066 net cpm. In this design (Design 70A) the DCGLw is 31,766 dpm/l00-cm 2 . The scan investigation level was rounded downward to 3500 gross counts per minute (gcpm) in survey instructions for the survey units covered by this design.

This practice was established early in the FSS of structures and has been continued. It is also noted the FSS Plan states that technicians are to respond to indications of increased count rates even though scan count rates may not be above the investigation level specified in survey instructions. 17 Modifications to survey methods may be required when unusual measurement conditions are encountered. For example, surface irregularities may cause an increase in the effective surface-to-detector distance and hence a decrease in detector efficiency. Measurement detection sensitivities may thus be reduced. It may be necessary to reduce scan speeds to ensure that required scan sensitivities are achieved. The bases for adjustments due to non-17 From FSS Plan Section 7.1.1: "Technicians will respond to indications of elevated areas while surveying. Upon detecting an increase in visual or audible response, the technician will reduce the scan speed or pause and attempt to isolate the elevated area. If the elevated activity is verified to exceed the established investigation level, the area is bounded (e.g., marked and measured to obtain an estimated affected surface area). Representative static measurements are obtained as determined by the FSS/Characterization Engineer. The collected data is documented on a Radiological Survey Form" [NASA 2007].

28

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 standard conditions are provided in a PBRF Technical Basis Document. 18 The FSS Procedure, CS-06, Survey Unit Inspection to Support Survey Design, prescribes standard methods to identify and account for non-standard conditions. In survey units with irregular surfaces, representative areas are established, marked and assigned ID codes (typically about 2 x 3 ft.).

Depths of irregularities are measured to provide a basis for establishing instrument efficiency adjustment factors for scan survey and fixed measurements. These areas are called Surface Measurement Test Areas (SMTAs).

Isolated-small irregularities on the order of a detector probe area or less, are identified in survey unit inspections as requiring consideration of reduced detector efficiency. These are called Unusual Condition Measurement (UCM) locations. They are also marked and assigned ID codes. Examples of locations in Reactor Building survey units where special measurement conditions apply are shown in Exhibits 20 through 27 of Appendix A.

Scan sensitivities for detectors used for gamma scan surveys of soil are determined using the method referenced in the PBRF FSS Plan and described in NUREG-1507 [USNRC 1998].

Scan sensitivities for the Ludlum Model 44-10 Nal detectors used in FSS of soils at PBRF were developed in the PBRF technical basis document TBD-09-002 [PBRF 2009a]. The method is summarized and the key equations presented. The scan MDC is calculated using the following equations adapted from NUREG-1507 for walkover gamma scanning with NaI detectors [USNRC 1998]:

d' 4P (60J i) (Equation 4)

MDCSCO - MDCRsuva Cony

• MSo (Equation 5) where:

MDCsuRv = the minimum detectable count rate in cpm that can be reliably detected by the "surveyor",

d'= index of sensitivity, unitless (MARSSIM default value of 1.38 is assigned),

bi= background counts observed in the interval i, i = observation interval (s),

p = surveyor efficiency, unitless (MARSSIM default value of 0.5 for walkover scans is assigned),

18 The PBRF-TBD-07-004 [PBRF 2007a] presents efficiency correction factors developed for the LMI 44-116 detector.

The correction factors are presented as a function of detector-to-surface distance. Application of the factors requires empirical measurements of the effective detector-to-surface distance for areas with non-standard surface conditions as part of the survey unit inspection process.

29

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 MDCscan = the scan MDC, here in units of pCi/g, Conv = instrument response conversion factor, units of cpm per gR/h and MS, = instrument response in units of gR/h per pCi/g (determined empirically or with a shielding algorithm).

The instrument response factor for Cs-137 is 0.139 pR/h per pCi/g, calculated using the MicroShield code. The most conservative instrument response conversion factor measured for detectors in the PBRF LMI 44-10 inventory is 232.39 cpm per jiR/h for Cs-137.

Using these values, detection sensitivities of the instruments used in the FSS of the RB soil survey unit (RB-3-2) are provided in Table 10. Minimum detectable scan count rates and MDCscan values for 44-10 detectors operated in the Cs- 137 window vs. background count rates are shown in Table 10.

Table 10, Typical Detection Sensitivities of Field Instruments used for Soil Scans LMI 44-10 with Cs-137 Window (1)

Background (cpm) (2) Minimum Detectable MDCscan Count Rate (ncpm) (3) (pCi/g) (4) 100 79 2.4 150 96 3.0 200 111 3.5 250 124 3.9 300 136 4.2 Table 10 Notes:

1. Ludlum Model 44-10 Nal detector with Model 2350-1 data logging scaler-rate meter setup to count in Cs-137 energy window. Data from Survey Design No. 70B. Scan speed

= 0.5 m/s, detector to soil surface = 10 cm.

2. Specified as average background count rate.

3. This is the MDCsuv, the minimum detectable net count rate (ncpm) that can be reliably detected by the surveyor. The scan investigation-action level was set at 135 ncpm. This is equivalent to 85% of the surrogate DCGLw for this design (4.2 pCi/g).

4. The MDCscan sensitivity values shown in the table are for Cs- 137.

5.0 Reactor Building Survey Results Results of the Reactor Building FSS are presented in this section. This includes scan survey frequencies (% of areas covered) for each survey unit and identifies survey units where scan investigations were performed. Investigations performed and the results are summarized. Fixed measurement results for each survey unit and the results of comparison tests of survey unit maximum and average values with the DCGLw are reported. As discussed below, no statistical tests were 30

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 required. It is shown that levels of residual contamination have been reduced to levels that are ALARA. This section closes with a summary which concludes that applicable criteria for release of the Reactor Building for unrestricted use are satisfied and all FSS Plan requirements are met.

5.1 Scan Surveys Scan survey results were reviewed to confirm that the scan coverage requirement (as % of survey unit area) was satisfied for all survey units. The results of QC replicate scan surveys were also reviewed to confirm that the minimum coverage requirement of 5% was satisfied.

Results of the Reactor Building scan surveys are compiled in Table 11. The table shows that scan coverage requirements were satisfied for all survey units. The table also shows that investigations were performed in 11 survey units. The results of these investigations are discussed in Section 5.3.

Table 11, Scan Survey Results Scan Survey Investigation QC Replicate Scan Survey Coverage (%)(1) Performed Coverage (%) (1)(2)(3)

RB-I-1 1 100 No 7.7%

RB-1-2 1 100 No 7.7%

RB-1-3 1 100 No 7.7%

RB-1-4 1 100 No 7.7%

RB-1-5 1 100 No 7.7%

RB-1-6 1 100 No 7.7%

RB-1-7 1 100 No 7.7%

RB-1-8 1 100 No 7.7%

RB-1-9 1 100 No 7.7%

RB-I-10 1 100 No 7.7%

RB-I-11 1 100 No 7.7%

RB-I-12 1 100 No 7.7%

RB-1-13 1 100 No 7.7%

RB-I-14 1 100 No 7.7%

RB-I-15 1 100 No 7.7%

RB-I-16 1 100 No 6.6%

RB-1-17 2 36 No 6.6%

RB-I-18 2 49 No 6.6%

RB-1-19 2 32 No 6.6%

RB-1-20 2 45 No 6.6%

RB-1-21 2 45 No 6.6%

RB-1-22 1 100 No 7.7%

RB-1-23 1 100 No 7.7%

RB-1-24 1 100 No 7.7%

RB-1 -25 1 100 No 7.7%

RB-1-26 1 100 No 7.7%

RB-1-27 1 100 No 7.7%

RB-1-28 2 35 No 6.6%

RB-1-29 2 36 No 6.6%

RB-1-30 2 41 No 6.6%

RB-1-31 1 100 No 6.6%

31

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 11, Scan Survey Results Scan Survey Investigation QC Replicate Scan Survey Unit Class Coverage (%) () Performed Coverage (%) (1 (2)(3)

RB-1-32 1 100 No 7.7%

RB-1-33 1 100 No 7.7%

RB-1-34 1 100 No 7.7%

RB-I1-35 1 100 No 7.7%

RB-1-36 1 100 No 7.7%

RB-1-37 1 100 No 7.7%

RB-1-38 1 100 No 7.7%

RB-1-39 1 100 No 7.7%

RB-1-40 1 100 No 7.7%

RB-1-41 1 100 No 7.7%

RB-1-42 1 100 No 7.7%

RB-1-43 2 37 No 6.6%

RB-2-1 2 50 No 5.3%

RB-2-2 2 52 No 6.1%

RB-3-1 1 100 No 8.5%

RB-3-2 1 100 No 29.2%

RB-3-3 1 100 Yes 7.3%

RB-3-4 1 100 Yes 7.3%

RB-3-5 1 100 No 7.3%

RB-3-6 1 100 No 7.3%

RB-3-7 1 100 No 7.3%

RB-3-8 1 100 No 7.3%

RB-3-9 1 100 No 7.3%

RB-3-10 1 100 No 7.3%

RB-3-11 1 100 No 7.3%

RB-3-12 1 100 No 7.3%

RB-3-13 1 100 Yes 7.3%

RB-3-14 1 100 No 7.3%

RB-3-15 1 100 No 5.9%

RB-3-16 1 100 No 5.9%

RB-3-17 1 100 No 5.9%

RB-3-18 1 100 No 5.9%

RB-3-19 1 100 No 5.9%

RB-3-20 1 100 No 5.9%

RB-3-21 1 100 No 5.9%

RB-3-22 1 100 No 5.9%

RB-3-23 1 100 No 7.3%

RB-3-24 1 100 No 5.2%

RB-3-25 1 100 No 5.2%

RB-3-26 1 100 No 5.2%

RB-3-27 1 100 No 5.2%

RB-3-28 1 100 No 5.2%

RB-3-29 1 100 No 5.2%

RB-3-30 1 ___ 100 No 5.2%

RB-3-31 1 ____100 No 5.2%

32

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 11, Scan Survey Results Scan Survey Investigation QC Replicate Scan Survey Unit Class Coverage (%) (1) Performed Coverage (%) 1)(2)(3)

RB-3-32 1 100 No 5.2%

RB-3-33 1 100 No 5.2%

RB-3-34 1 100 No 5.2%

RB-3-35 1 100 No 5.2%

RB-3-36 1 100 No 5.2%

RB-3-37 1 100 No 5.2%

RB-3-38 1 100 No 9.5%

RB-3-39 1 100 No 9.5%

RB-4-1 1 100 No 8.1%

RB-4-2 1 100 No 8.1%

RB-4-3 1 100 Yes 8.1%

RB-4-4 1 100 Yes 8.1%

RB-4-5 1 100 No 8.1%

RB-4-6 1 100 No 8.1%

RB-4-7 1 100 No 8.1%

RB-4-8 1 100 No 8.1%

RB-4-9 1 100 No 8.1%

RB-4-10 1 100 No 8.1%

RB-4-11 1 100 No 8.1%

RB-4-12 1 100 Yes 5.8%

RB-4-13 1 100 Yes 5.8%

RB-4-14 1 100 No 9.5%

RB-4-15 1 100 No 9.5%

RB-4-16 1 100 No 9.5%

RB-4-17 1 100 No 9.5%

RB-4-18 1 100 Yes 5.8%

RB-4-19 1 100 Yes 5.8%

RB-4-20 1 100 Yes 9.5%

RB-4-21 1 100 No 5.5%

RB-4-22 1 100 No 5.5%

RB-4-23 1 100 No 5.5%

RB-4-24 1 100 No 5.5%

RB-4-25 1 100 No 5.5%

RB-4-26 1 100 No 5.5%

RB-4-27 1 100 No 5.5%

RB-4-28 1 100 No 5.5%

RB-4-29 1 100 No 5.5%

RB-4-30 1 100 No 5.2%

RB-4-31 1 100 No 5.5%

RB-4-32 1 100 No 5.5%

RB-4-33 1 100 No 5.5%

RB-4-34 1 100 No 5.5%

RB-4-35 1 100 No 5.5%

33

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 11, Scan Survey Results Scan Survey Investigation QC Replicate Scan Survey Unit Class Coverage (%)

• Performed Coverage (%) (1)(2)(3)

RB-4-36 1 100 No 5.5%

RB-4-37 1 100 No 5.5%

RB-4-38 1 100 No 5.5%

RB-4-39 1 100 No 5.5%

RB-4-40 1 100 No 5.5%

RB-4-41 1 100 No 5.5%

RB-4-42 1 100 Yes 21.5%

RB-4-43 1 100 No 21.5%

RB-5-1 3 10 No 6.8%

RB-5-2 2 63 No 5.0%

RB-5-3 2 39 No 5.0%

Table 11 Notes:

1. Scan % coverage values for the primary scan surveys are rounded to the nearest whole per cent. Scan % coverage for QC scan surveys are rounded to the nearest /10It %.

Those results where the calculated second decimal is 5, e. g., 5.55, are rounded downward to 5.5.

2. The % QC scan coverage is given as the % of the area scanned in the initial survey.

3. Replicate QC scan results are reported for multiple survey units in typical Survey Requests. For these, the QC scan percentages are reported as % of the scanned area of the survey units combined. So the same % coverage value is assigned to all of the survey units reported in a Survey Request.

4. One hundred percent of the accessible surface area was scanned. A fraction of the surface area of the survey unit is inaccessible for scanning. In most such survey units, it is less than a few percent of the total surface area.

5.2 Systematic Measurements and Tests Reactor Building FSS total surface beta measurement results are summarized in Table 12.

Individual measurements in each survey unit are reported in Appendix B. The table presents the number of measurements, maximum, average and standard deviation for each survey unit.

Table 12 compares the maximum activity measured in each survey unit to the DCGLw. It is demonstrated that all measured activity values are less than the DCGLw, thus all survey units meet the 25 mrem/y release criterion. The mean activity of each survey unit is also compared to the DCGLw, and as expected, are all less than the DCGLw. 19 The average of 1441 systematic total surfake beta measurements reported in the Reactor Building release records 2

is: 451 + 480 dpm/100-cm (one standard deviation) [PBRF 2011]. 20

'9 As shown in Appendix D, the RB systematic measurements are compared to the revised-reduced DCGLs and it is shown that there is no change in the conclusion that the 25 mrem/y dose criterion is satisfied.

20 It is noted that in converting total surface activity measurements in cpm to dpm/100-cm 2, the detector background response from surface materials is not subtracted. As a result, the total surface activity measurement results are biased high.

34

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 12, Total Surface Beta Activity Measurement Summary and Test Results SurvyN. of Maxmum Test Result: Aeae Standard Test Result:

UntIMauemns (I) Maximum < () Deviation Average <

Uni_ I,_ Masremnt DCGLw (1) (2) DCGLw RB-1-i 11 811 Yes 443 301 Yes RB-1-2 1 11 1,007 Yes 595 212 Yes RB-1-3 11 750 Yes 523 128 Yes RB-1-4 11 1,088 Yes 550 316 Yes RB-i-5 11 1,070 Yes 622 265 Yes RB-1-6 11 757 Yes 517 226 Yes RB-1-7 11 1,092 Yes 560 333 -Yes RB-1-8 11 1,039 Yes 666 219 Yes RB-1-9 12 (:3) 961 Yes 580 189 Yes RB-i-10 11 1,465 Yes 654 388 Yes RB-1-1i1 11 915 Yes 564 264 Yes RB-1-12 11 838 Yes 519 191 Yes RB-1-13 11 1,528 Yes 641 326 Yes RB-1-14 123 711 Yes 539 127 Yes RB-1-15 12 197 Yes 34 91 Yes RB-1-16 11 92 Yes -72 140 Yes RB-1-17 i11__ 197 Yes 3 150 Yes RB-1-18 11 1,013 Yes 224 333 Yes RB-1-19 11__i 845 Yes 188 306 Yes RB-1-20 11__i 678 Yes 289 258 Yes RB-1-21 11__i 717 Yes 317 278 Yes RB-1-22 11 913 Yes 475 286 Yes RB-1-23 11 664 Yes 230 243 Yes RB-1-24 11 758 Yes 408 317 Yes RB-1-25 11__i 738 Yes 202 241 Yes RB-1-26 11__i 1,078 Yes 387 320 Yes RB-1-27 11 799 Yes 412 257 Yes RB-1-28 11 550 Yes 171 204 Yes RB-1-29 11 430 Yes 35 177 Yes RB-1-30 11 168 Yes 38 104 Yes RB-1-31 12 ('3) 1,092 Yes 636 225 Yes RB-1-32 11 1,216 Yes 640 266 Yes RB-1-33 11 1,042 Yes 566 191 Yes RB-1-34 12 1,216 Yes 685 315 Yes RB-1-35 11 830 Yes 481 240 Yes RB-1-36 11 987 Yes 624 203 Yes RB-1-37 14 ()2,289 Yes 385 643 Yes RB-1-38 12 ()2,693 Yes 1,751 865 Yes RB-1-39 12 ()441 Yes 149 173 Yes RB-1-40 11 638 Yes 141 246 Yes RB-1-41 11 921 Yes 472 417 Yes RB-1-42 11 826 Yes 264 352 Yes RB-1-43 11 550 Yes 110 245 Yes RB-2-1 16T 898 Yes 377 314Ye 35

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 12, Total Surface Beta Activity Measurement Summary and Test Results Survey No. of Maximum Test Result: Standard Test Result:

(1) Maximum < (1) Deviation Average <

Unit ID Measurements DCGLw (1)(2) DCGLw RB-2-2 15 (3) 715 Yes 229 288 Yes RB-3-1 11 361 Yes 118 188 Yes RB-3-3 11 3,189 Yes 1,210 869 Yes RB-3-4 11 923 Yes 735 162 Yes RB-3-5 11 993 Yes 674 173 Yes RB-3-6 11 930 Yes 631 174 Yes RB-3-7 11 888 Yes 503 164 Yes RB-3-8 11 776 Yes 612 148 Yes RB-3-9 11 2,000 Yes 788 433 Yes RB-3-10 11 867 Yes 636 163 Yes RB-3-11 11 797 Yes 611 143 Yes RB-3-12 11 916 Yes 652 153 Yes RB-3-13 11 3,550 Yes 483 1042 Yes RB-3-14 11 340 Yes 82 164 Yes RB-3-15 11 847 Yes 553 160 Yes RB-3-16 11 803 Yes 372 171 Yes RB-3-17 11 627 Yes 293 182 Yes RB-3-18 11 613 Yes 355 236 Yes RB-3-19 12 () 768 Yes 390 229 Yes RB-3-20 11 662 Yes 117 232 Yes RB-3-21 11 380 Yes 129 136 Yes RB-3-22 11 190 Yes -15 142 Yes RB-3-23 11 797 Yes 532 183 Yes RB-3-24 11 373 Yes 115 184 Yes RB-3-25 11 830 Yes 221 274 Yes RB-3-26 11 654 Yes 205 281 Yes RB-3-27 11 529 Yes 299 180 Yes RB-3-28 11 634 Yes 150 272 Yes RB-3-29 12 (3) 765 Yes 220 256 Yes RB-3-30 11 667 Yes 155 327 Yes RB-3-31 11 627 Yes 165 243 Yes RB-3-32 11 438 Yes 59 220 Yes RB-3-33 11 542 Yes 224 220 Yes RB-3-34 11 503 Yes 267 183 Yes RB-3-35 11 497 Yes 215 200 Yes RB-3-36 11 614 Yes 208 205 Yes RB-3-37 11 438 Yes -34 442 Yes RB-3-38 11 328 Yes 40 152 Yes RB-3-39 11 350 Yes 19 180 Yes RB-4-1 11 1,087 Yes 649 242 Yes RB-4-2 11 797 Yes 541 203 Yes RB-4-3 11 902 Yes 578 212 Yes RB-4-4 11 1,410 Yes 407 546 Yes RB-4-5 11 993 Yes 742 154 Yes 36

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 12, Total Surface Beta Activity Measurement Summary and Test Results Survey No. of Maximum Test Result: Average Standard Test Result:

Measurements (1) Maximum < (1) Deviation Average <

Unit ID Unit DCGLw (1)(2) DCGLw RB-4-6 11 765 Yes 488 100 Yes RB-4-7 11 948 Yes 471 261 Yes RB-4-8 11 1,077 Yes 609 288 Yes RB-4-9 11 1,327 Yes 739 235 Yes RB-4-10 11 1,090 Yes 429 372 Yes RB-4-11 11 762 Yes 568 147 Yes RB-4-12 11 2,100 Yes 818 456 Yes RB-4-13 12 (3) 1,150 Yes 556 293 Yes RB-4-14 11 1,190 Yes 847 208 Yes RB-4-15 11 1,029 Yes 776 237 Yes RB-4-16 11 1,204 Yes 749 189 Yes RB-4-17 11 1,022 Yes 497 418 Yes RB-4-18 11 3,920 Yes 1,109 959 Yes RB-4-19 11 5,458 Yes 2,572 1348 Yes RB-4-20 11 1,420 Yes 961 336 Yes RB-4-21 11 622 Yes 473 115 Yes RB-4-22 11 720 Yes 477 212 Yes RB-4-23 11 888 Yes 466 269 Yes RB-4-24 12 (3) 1,082 Yes 628 245 Yes RB-4-25 11 888 Yes 557 219 Yes RB-4-26 11 820 Yes 440 166 Yes RB-4-27 11 207 Yes -15 157 Yes RB-4-28 11 167 Yes -58 149 Yes RB-4-29 11 287 Yes -39 188 Yes RB-4-30 11 1,042 Yes 625 167 Yes RB-4-31 11 1,094 Yes 705 229 Yes RB-4-32 11 719 Yes 383 221 Yes RB-4-33 11 1,144 Yes 708 228 Yes RB-4-34 11 552 Yes 212 267 Yes RB-4-35 11 706 Yes 343 246 Yes RB-4-36 11 622 Yes 163 251 Yes RB-4-37 11 615 Yes 176 277 Yes RB-4-38 11 615 Yes 197 224 Yes RB-4-39 11 671 Yes 275 274 Yes RB-4-40 11 853 Yes 399 321 Yes RB-4-41 11 748 Yes 122 353 Yes RB-4-42 10 (4) 541 Yes 325 170 Yes RB-4-43 11 2,538 Yes 1,748 432 Yes RB-5-1 11 565 (- Yes 110 232 Yes RB-5-2 11 2,345 Yes 1,231 784 Yes RB-5-3 11 1,115 Yes 618 267 Yes Table 12 Notes:

1. The units for: maximum, average and standard deviation are dpm/1 00-cm 2.

2. Standard deviations of the measurements in each survey unit are reported for comparison to the values used in the survey design. In the Reactor Building structural survey units, values of a 37

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 obtained from the FSS measurements are much less than values used in the survey designs (see Table 7). On average, the actual standard deviations are less than 25% of the design value and in only two survey units did the value exceed the design value. This confirms that the survey designs for the Reactor Building were indeed conservative.

3. In the FSS design calculation for survey units developed using VSP, "extra" fixed measurement locations are sometimes added when "fitting" the calculated grid size onto the survey unit layout.

4. Measurement at location SM-1 in RB-4-42, -25 ft. elevation Cold Sump, not collected due to water in sump bottom.

5. The measurement results from the Class 3 Survey Unit, RB-5-1, were reviewed to ensure that they were all less than 50% of the DCGLw.

5.3 Investigations, Additional Measurements and Tests Investigations were performed in 11 survey units as a result of scan survey observations. The resulting static total surface beta measurements are listed in Table 13. The table shows that 31 measurements were reported as a result of investigations initiated during scan surveys of the 130 RB survey units. Of these measurements, 12 were greater than the survey unit DCGLw.

As discussed below, these required that the elevated measurement comparison (EMC) and elevated measurement test (EMT) be performed.

Table 13, Summary of Scan Investigations and Static Measurements Size of Static Survey Measurement Sz Area Elevated fSai 1 Measurement (1) Comments Unit ID (cm 2) (dpM/100-cm 2)

A 3 m2 area was rescanned using a 44-116 detector after RB-3-3 N/A N/A N/A the scan investigation level was exceeded with the 43-37 detector. The scan investigation level was not exceeded with the 44-116; no further action.

RB-3-4 IM-1 100 19,100 < DCGL (30,355 dpm/100-cm2 ); no further action.

>DCGL (30,355 dpm/100-cm 2 ); EMC/EMT RB-3-13 IM- 1 100 33,700 performed RB-3-13 IM-2 100 45,000 >DCGL (30,355 dpM/100-cm 2); EMC/EMT RB-3-3_M2_10 40performed.

RB-4-3 IM-1 100 15,060 < DCGL (27,271 dpm/100-cm 2); no further action.

> DCGL (27,271 dpm/100-cm 2 ); EMC/EMT RB-4-3 IM-2 100 37,500 performed.

RB-4-3 IM-3 100 14,600 < DCGL (27,271 dpm/100-cm 2); no further action.

RB-4-4 IM-1 1000 22,800 < DCGL (27,271 dpm/100-cm 2); no further action.

RB-4-4 IM-2 700 19,200 < DCGL (27,271 dpm/100-cm 2); no further action.

> DCGL (12,556 dpm/100-cm2 ); EMC/EMT RB-4-12 IM-I 100 20,200 performed.

> DCGL (12,556 dpm/100-cm2); EMC/EMT RB-4-12 IM-2 100 41,100 performed.

R134-12 IM-3 100 16,600 > DCGL (12,556 dpm/1 00-cm 2); EMC/EMT performed.

RB-4-13 IM-1 100 9,250 < DCGL (12,556 dpm/100-cm2 ); no further action.

RB-4-13 IM-2 100 10,100 < DCGL (12,556 dpm/100-cm 2); no further action.

RB-4-18 IM-1 100 7,460 < DCGL (12,556 dpm/100-cm 2); no further action.

> DCGL (12,556 dpm/100-cm2 ); EMC/EMT RB-4-18 IM-2 100 13,600 performed.

38

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 13, Summary of Scan Investigations and Static Measurements Survey Measurement Size of Static Sz Area Elevated fSai 1 Measurement 2(1) Comments Unit ID (cm 2) (dpm/100-cm )

RB-4-18 IM-3 100 10,400 < DCGL (12,556 dpm/I00-cm ); no further action.

RB-4-18 IM-4 100 10,900 < DCGL (12,556 dpmr/1OO-cm 2); no further action.

RB-4-18 IM-5 100 11,700 < DCGL (12,556 dpm/100-cm 2 ); no further action.

RB-4-18 IM-6 100 10,200 < DCGL (12,556 dpm/100-cm 2 ); no further action.

RB-4-18 IM-7 100 4,920 < DCGL (12,556 dpm/100-cm 2 ); no further action.

RB-4-18 IM-8 100 12,000 < DCGL (12,556 dpm/100-cm 2 ); no further action.

RB-4-18 IM-9 100 8,470 < DCGL (12,556 dpm/100-cm 2 ); no further action.

RB-4-18 IM-10 200 7,800 < DCGL (12,556 dpm/100-cm22 ); no further action.

8M-1 100 22,800 > DCGL (12,556 dpm/100-cm ); EMC/EMT RB-4-18 Iperformed.

> DCGL (12,556 dpm/100-cm2); EMC/EMT RB-4-18 IM-12 1550 22,500 performed.

> DCGL (12,556 dpm/100-cm 2); EMC/EMT RB-4-19 IM-1 100 23,523 performed.

RB-4-20 IM-1 100 11,500 < DCGL (12,556 dpm/100-cm 2); no further action.

> DCGL (12,556 dpm/100-cm2 ); EMC/EMT RB-4-20 IM-2 100 14,400 performed.

> DCGL (12,556 dpm/100-cm 2); EMC/EMT RB-4-20 IM-3 100 12,800 performed.

RB-4-20 IM-4 100 5,850 < DCGL (12,556 dpm/100-cm 2); no further action.

RB-4-20 IM-5 100 7,990 < DCGL (12,556 dpm/100-cm 2 ); no further action.

Area in a 6 in. penetration in the sump was measured with 44-10 (NaI). Activity estimated to be equivalent to RB-4-42 N/A N/A See comments 9,000 dpm/100-cm 2. The penetration was remediated and subsequently included in the FSS of PPP Room trenches.

JLaUi*

1. This table includes only investigative measurements that were assigned an IM Number and the measured activity recorded in survey documentation. All are surface activity measurements reported in units of dpm/l00-cm 2.

Elevated measurement comparisons and the EMT were performed in six survey units. These were prompted by investigative measurements which showed elevated activity in excess of the DCGLw in small localized areas. In accordance with the FSS Plan, Section 8.3, the DCGLEMC is calculated as the product of the Area Factor (AF) and the DCGLw. The EMT is defined by the following equation:

+(average concentration in elevated area)- 5

[Equation 6]

DCGLW (AF) (DCGLw)

Where: 6 is the average residual activity concentration in the survey unit.

If more than one elevated area is found in a survey unit, the second term in Equation 6 is calculated for each and summed with the first term to perform the unity rule calculation for 39

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 the EMT. Results of the DCGLEMC and EMT calculations are presented in Table 14. Note that these represent "as-left" conditions in the affected survey units.

Table 14, Elevated Measurement Comparisons and Tests Measured Size of DCGLw Calculated Average EMT Survey Measurement Elevated 2 DCGLEMC Activity in Unity Sunity Meas n Activity 2 (dpml100-cm) (dpm/100-cm 2) Survey Unit Unit ID (dpm/100_cm ) Area (1) Valu (cm2) (1) (2) (dpm/100-cm 2) Value RB-3-13 IM-1/EMA-1 33,700 100 30,355 1.220E+06 4.835E+02 0.080 ()

RB-3-13 IM-2/EMA-2 45,000 100 30,355 1.220E+06 4.835E+02 0.080 ()

RB-4-3 IM-2/EMA-1 37,500 100 27,271 1.096E+06 5.778E+02 0.055 RB-4-12 IM-1/EMA-1 20,200 100 12,556 5.048E+05 8.18 1E+02 0.215 (3)

RB-4-12 IM-2/EMA-2 41,100 100 12,556 5.048E+05 8.181E+02 0.215 ()

RB-4-12 IM-3/EMA-3 16,600 100 12,556 5.048E+05 8.181E+02 0.215 (3)

RB-4-18 IM-2/EMA-1 13,600 100 12,556 5.048E+05 1.109E+03 0.198 (3)

RB-4-18 IM-1 1/EMA-2 22,800 100 12,556 5.048E+05 1.109E+03 0.198 (3)

RB-4-18 IM-12/EMA-2 22,500 1550 12,556 5.048E+05 1.109E+03 0.198 (3)

RB-4-19 IM-1/EMA-1 23,523 100 12,556 5.048E+05 2.572E+03 0.246 RB-4-20 IM-2/EMA-1 14,400 100 12,556 5.048E+05 9.608E+02 0.127 (3)

RB-4-20 IM-3/EMA-2 12,800 100 12,556 5.048E+05 9.608E+02 0.127 (3)

Table 14 Notes:

1. Note that an evaluation of the EMC and EMT calculations in view of the revised DCGLs is reported in Appendix D and no changes occur in the conclusions reported here.

2. Calculated as the product of the AF and the DCGLw. Per Table 3-5 of the FSS Plan, the AF for areas up to 0.25 m2 is 40.2 [NASA 2007].

3. Unity value includes the sum of contributions from each of the individual elevated areas where measurements exceeded the DCGL.

The FSS Plan requires that removable surface activity in each survey unit be less than 10% of the DCGLw. In accordance with the FSS Plan, removable surface activity measurements were taken at each systematic measurement location in structural survey units. Removable surface activity measurements were also taken at locations where static investigative measurements were performed. Removable surface activity is measured by counting 100 cm 2 smear samples 21 for beta and alpha activity. Smear results were below counting instrument MDA values in all but five survey units. Results of removable surface activity measurements in excess of counting instrument MDA values are shown in Table 15. The table includes results of smears taken at systematic and investigative measurement locations. Removable activity measurement results are less than 10% of the applicable survey unit DCGLw for the final conditions in all survey units.

Table 15, Removable Surface Activity Measurements above MDA I Survey I Survey I Measurement I Removable Surface Activity (dpm/100-cm2 )

21 Smears are counted in the PBRF Counting Laboratory on automated sample changer proportional counters. Two such counters are available for this purpose: Tennelec Model LB-5 100 and Tennelec Model 5X-LB.

40

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Unit ID Request Location ID (1) Beta-Gamma (2)(3) Alpha (4)

RB-1-29 328 SM-11 23.3 < MDA RB-3-13 312 IM-1 207.4 < MDA RB-3-13 312 IM-2 30.5 < MDA RB-3-13 312 IM-3 167.9 < MDA RB-4-18 306 IM-6 718.8 <MDA RB-4-18 306 IM-7 33.7 < MDA RB-4-19 306 SM-1 22.4 < MDA RB-4-21 308 SM-4 17.0 < MDA Table 15 Notes:

1.SM denotes systematic measurement; IM denotes investigative measurement.

2.Beta MDA values for these measurements ranged from 15.1 to 22.0 dpm/l100-cm 2.

3. The lowest DCGLw for these survey units was 12,556 dpm/100-cm 2.

4. Alpha MDA values for these measurements ranged from 8.2 to 12.1 dpm/100-cm 2.

5.4 Soil Survey Unit Results A soil survey unit, RB-3-2, was established in the Reactor Building to address soil that was exposed after the sanitary sump located on the -15 ft. elevation was removed. Eleven surface soil sample locations were established in Survey Design 70B. The exposed soil surface was subjected to a 100% gamma scan and the samples were collected and analyzed by gamma spectroscopy at the PBRF on-site counting laboratory in accordance with the instructions in Survey Request SR-332. As reported in the RB-3-2 Release Record, no scan investigations were performed. The sample results are summarized in Table 16.

It is noted that the survey design identified only Cs-137 and U-234 as significant radionuclides for this soil survey unit. This is because it was assumed that any soil contamination would have the same radionuclide profile assigned to the -15 ft. elevation structures where the sanitary sump was located. As shown in Table 3, the radionuclide profile consisted of Cs-137 (48.4%) and U-234 (51.6%). However, Co-60 and Sr-90 results are also included as these are identified in the FSS Plan as the principal radionuclides of concern in PBRF soil. In Table 16, the maximum activity concentrations of Sr-90 and U-234 are inferred from the measured activity of Cs-137, the surrogate radionuclide.

41

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 16, Reactor Building Soil Survey Unit Sample Results No. of Samples Maximum Activity Concentration (pCi/g)

Cs-137 Co-60 I Sr-90 I U-234 11 1.60E-01"I < MDA'2) 1.42E-02 1.71E-01 14)

Table 16 Notes:

1. The single sample result from the 11 samples that was > MDA.

2. The maximum and average Co-60 MDA values for the 11 Survey Unit RB-3-2 samples are 7.57E-02 and 6.2 IE-02 pCi/g, respectively.

3. The Sr-90 concentration is inferred from the maximum measured Cs- 137 concentration and Sr-90: Cs-137 activity ratio of 0.089. This activity ratio is obtained from TBD-09-001 [PBRF 2009b].

4. The U-234 concentration is inferred from the maximum measured Cs-137 concentration and U-234: Cs-137 activity ratio of 1.07. This activity ratio is obtained from the activity fractions in Table 3, used in Design 70B.

5.5 QC Measurements Per FSS Plan requirements, QC replicate measurements were taken for at least 5% of the Reactor Building FSS measurements. This included scan surveys, total surface activity measurements and soil samples. Scan QC survey results are shown in Table 11 wherein the 5% scan QC coverage is confirmed. No QC scan surveys identified areas of elevated activity.

These surveys confirmed the results of the original scan surveys of the areas covered.

Replicate total surface activity measurements were performed at selected measurement locations including systematic and investigative measurements. The 5 % requirement is satisfied in that 90 QC measurements were reported; this represents 6.1 % of the combined total of 1472 systematic and investigative measurements (1441 systematic and 31 investigative). Appendix C contains the individual measurement results for the 90 surface activity original and QC replicate measurement pairs. The FSS Plan (Section 12.7) identifies a target criterion of 20% for the relative percent difference (RPD) between original and replicate measurements [NASA 2007]. Fifty five of the 90 measurement pairs exceeded the 20% criterion. Each measurement pair failing to meet the 20% criterion was individually investigated in accordance with FSS Plan requirements and implementing procedures, resolved and determined to be acceptable. 22 Results of the replicate QC measurement evaluation are summarized in Table 17. It is found that most of the measurement pairs that exceeded the 20% RPD criterion were low activity measurements (below 2,500 dpm/100-cm 2). Experience and the theory of measurement errors 22 When the acceptance criterion is not met, an investigation is performed to determine the cause and corrective actions.

The investigation may include repetition of the replicate QC measurement or other actions determined by the FSS/Characterization Supervisor. If upon repetition, the RPD criterion is still not satisfied, the result may be accepted if the original and QC replicate measurement are in agreement that both are below the DCGLw for the survey unit, the FSS/Characterization Supervisor reviews the investigation and concurs that the measurement is acceptable and the results of the investigation are documented in the Survey Request Summary and Close-out (Procedure CS-01, Survey Methodology to Support PBRF License Termination).

42

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 have shown that low activity measurements such as these are subject to variation which is high relative to the measured activity.

Table 17, Replicate QC Measurement RPD Evaluation No. of No. of Average Average QC Measurement Activity Range Measurement Original Replicate Average Pairs > 20%

(1) (2) Pairs. Activity (1) Activity (1) RPD (%) (3) RPD

< 1000 75 573.0 486.2 51.5 50 1000 to 10000 9 3,541.0 2,931.6 94.8 5

>10,000 6 29,616.7 30,903.3 6.6 0 Table 17 Notes:

1. Units are dpm/100-cm 2.

2. The specified activity range is for the original measurements.

3. Calculated as the average of the individual measurement pair RPDs.

For the soil survey unit (RB-3-2), one QC replicate soil sample was collected (at location SP-

2) and analyzed by gamma spectroscopy at the PBRF on-site counting laboratory. Both original and replicate QC sample results were < MDA for Cs-137 and Co-60. As reported in the RB-3-2 Release Record, the results were evaluated in accordance with Section 12.7.2 of the FSS Plan and the analysis indicated positive agreement between the original and QC sample.

5.6 ALARA Evaluation It is shown that residual contamination in the Reactor Building has been reduced to levels that are ALARA, using a method acceptable to the NRC. The NRC guidance on determining that residual contamination levels are ALARA includes the following:

"In light of the conservatism in the building surface and surface soil generic screening levels developed by the NRC, NRC staff presumes, absent information to the contrary, that licensees who remediate building surfaces or soil to the generic screening levels do not need to provide analyses to demonstrate that these screening levels are ALARA. In addition, if residual radioactivity cannot be detected, it may be presumed that it had been reduced to levels that are ALARA. Therefore the licensee may not 23 need to conduct an explicit analysis to meet the ALARA requirement.",

Screening level values published by the NRC for the mix of radionuclides in structural surface residual contamination potentially present in the Reactor Building are shown in Table 18.

However, since individual radionuclide activity concentrations are not measured in the FSS of structures, a direct comparison of residual contamination levels to individual radionuclide screening level values is not possible. A comparison can be made to an appropriate gross 23 This guidance was initially published in Draft Regulatory Guide DG-4006, but has been reissued in NUREG-1757 Volume 2, Appendix N.

43

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 activity DCGL. A screening level value that is equivalent to the gross activity DCGL was calculated using the equations in Section 3.6 of the FSS Plan. 24 The activity fractions listed in Table 3 (for the -25 ft. elevation PPP Room and pipe trench) were used in the calculation. The activity fractions are also shown in Table 18. On this basis, the screening level equivalent DCGL for the Reactor Building is calculated to be 2355 dpm/100-cm 2 .

The average total surface beta activity measured in the FSS of the Reactor Building surfaces is 451 +/- 480 dpm/100-cm 2 (one standard deviation). The upper limit of the 9 5 th % confidence interval of this mean value is 476 dpnm!100-cm2.25 This value is well below the screening level gross activity DCGL of 2355 dpm/100-cm2 . 22 Soil activity concentrations measured in survey unit RB- 3-2 are compared to NRC surface soil screening level values in Table 18. As shown in the table, the soil activity concentrations are well below their respective screening level values. From these comparisons, it is concluded that the ALARA criterion is satisfied.

Table 18, Screening Level Values for Reactor Building and Radionuclide Activity Fractions Screening Level Value Reactor Building Radionuclide ( 00-CM2) 2 Activity Fraction (dpm/l (%) (1)

H-3 1.2 E+08 ) 7.29 Co-60 7.1E+03 (2) 53.89 Sr-90 8.7E+03 (2) 22.22 1-129 3.5E+047 2 ) 0 Cs-137 2.8E+04 (2) 12.05 Eu- 154 1.2E+041" 1.71 U-234 9.1E+01 (3) 2.42 U-235 9.8E+0lt 3 ) 0.42 Table 18 Notes.

1. Activity fractions used to develop the DCGLw for Reactor Building -25 ft.

elevation PPP Room and trenches per Table 3.

2. Values from NUREG-1757 Vol. 2, Table H.1 [USNRC 2006].

3. Values from NUREG/CR-5512, Vol. 3, Table 5.19 [SNL 1999]. These are 9 0 th percentile values of residual surface activity corresponding to 25 mrem/y to a future building occupant.

24 The equivalent screening level gross activity DCGL is calculated using an EXCEL template [PBRF 2012]. This template incorporates the equations in section 5.3 of the FSS Plan [NASA 2007].

25 The upper limit of the confidence interval, 95th percentile value, is calculated as: UL = mean + 1.96 c/'/n, where n =

1441 measurements.

44

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 Table 19, NRC Soil Screening Level ALARA Comparison NRC LevelScreening (pCi/g) Maximum Measured Concentration (pCi/g)

Co-60 3.8 < MDA`1 Cs- 137 I11 1.60E-01 (2)

Sr-90 1.7 1.42E-02 (3)

U-234 13 1.71E-01 (4)

Table 19 Notes:

1. The maximum and average Co-60 MDA values for the 11 RB-3-2 samples are 7.57E-02 and 6.21E-02 pCi/g, respectively.

2. The result of the single sample with measured concentration > MDA.

3. See Table 16, Note 3.

4. See Table 16, Note 4.

5.7 Comparison with EPA Trigger Levels The PBRF license termination process includes a review of residual contamination levels in soil and groundwater, as applicable, in accordance with the October 2002 Memorandum of Understanding (MOU) between the US NRC and the US Environmental Protection Agency (EPA) [USEPA 2002]. Concentrations of individual radionuclides, identified as "trigger levels for further review and consultation between the agencies", are published in the MOU.

Maximum activity concentrations of radionuclides of concern measured in the Reactor Building FSS are compared to EPA trigger levels. This comparison is shown in Table 20. The table shows that the measured soil activity concentrations are well below EPA trigger levels.

It is noted that groundwater is not within the scope of the Reactor Building FSS.

Table 20, Comparison of Soil Sample Results with EPA Trigger Levels Radionuclide EPA Trigger Maximum Measured Level (pCi/g) Concentration (pCi/g)

Co-60 4 < MDA(')

Cs-137 (2) 6 1.60E-01( 3 )

Sr-90 (2) 23 1.42E-02 (4)

U-234 401 1.71E-01 (5)

Table 20 Notes:

1. The maximum and average Co-60 MDA values for the 11 RB-3-2 samples are 7.57E-02 and 6.2 1E-02 pCi/g, respectively.

2. Specified in the MOU as including daughter activity [USEPA 2002].

3. The result of the single sample with measured concentration > MDA.

4. See Table 16, Note 3.

5. See Table 16, Note 4.

45

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 5.8 Conclusions The results presented above demonstrate that the Reactor Building satisfies all FSS Plan commitments and meets the release criteria in 10CFR20 Subpart E. The principal conclusions are:

" Scan surveys were performed on 100 % of the accessible surfaces of all Reactor Building Class 1 survey units; between 38 and 63% on the Class 2 survey units and 10.2% on the single Class 3 survey unit.

• Investigations were performed following observation of elevated activity during the scan survey in 11 survey units. As a result of these investigations, elevated measurement comparisons and elevated measurement tests were performed in six survey units; all were satisfactory.

• All randomly selected fixed total surface activity measurements are less than the applicable DCGLw.

• All survey unit mean fixed measurement results (total surface beta activity) are below the DCGLw, hence no statistical tests were required.

• Removable surface activity measurements are less than 10% of the DCGLw in all RB survey units.

• Soil sample analysis results from the single RB soil survey unit are much less than the DCGLw values for the principal radionuclides of PBRF origin in soil, Cs-137, Co-60 and Sr-90 and for U-234 also identified as a potential contaminant in this survey unit.

• Residual surface activity concentration measurement results are shown to be less than NRC screening level values - demonstrating that the ALARA criterion is satisfied.

• Only minor changes from what was proposed in the FSS Plan were made - the Reactor Building was divided into 130 survey units, whereas the FSS Plan identified 34 survey areas, not divided into survey units.

" There was one change from initial assumptions (in the FSS Plan) regarding the extent of residual activity in the Reactor Building. One area of potentially contaminated soil was identified underneath the sanitary sump at the 15 ft. elevation. No reclassification of survey units was required as a result of FSS measurements and investigations.

" Errors in several of the DCGLs published in a supporting technical basis document were recently discovered and revised DCGLs were published in a new technical basis document. The potential impacts of revised-reduced DCGLs were evaluated and it is found that all FSS Plan requirements remain satisfied and the conclusion holds that the 25 mrem/y dose criterion is satisfied.

46

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 6.0 References GTS 1998 GTS-Duratek, Inc., NASA/Lewis Plum Brook Facility 1998 Confirmation Survey, Volume 1 - Survey Packages and Survey Results, November 1998.

ISO 1988 International Organization for Standardization, Evaluation of Surface Contamination,Part1. Beta Emitters andAlpha Emitters, ISO-7503-1, 1988.

NASA 2007 NASA Safety and Mission Assurance Directorate, FinalStatus Survey Planfor the Plum Brook Reactor Facility,Revision 1, February 2007.

NASA 2008 NASA Safety and Mission Assurance Directorate, DecommissioningPlanfor the Plum Brook Reactor Facility, Revision 6, July 2008.

ORISE 2011 Oak Ridge Institute for Science and Education, Independent Environmental Assessment and Verification Program, Final- Independent ConfirmatorySurvey Reportfor the Reactor Building,Hot Laboratory,Primary Pump House, and Land Areas at the Plum Brook Reactor Facility,Sandusky, Ohio, October 2011.

PBRF 2004 Plum Brook Reactor Facility, NASA Safety and Mission Assurance Directorate, Supplemental CharacterizationReportfor the Plum Brook Reactor Facility, December 16, 2004.

PBRF 2005 MWH Constructors, Inc, for US Army Corps of Engineers, Investigation of Radioactivity in PBRF Concrete, PBRF-05-0745, August 2005.

PBRF 2005a NASA Plum Brook Reactor Facility Decommissioning, CharacterizationSurvey Package C9000 101H, DCGL Development- RadionuclideDistribution,September 2005.

PBRF 2005b Plum Brook Reactor Facility, Survey Request SR-4, CV Quads and Canals, (preliminaryscans to identify potentialcore bore locationsfor characterization purposes), November 14, 2005.

PBRF 2006 Plum Brook Reactor Facility, Survey Request SR-15, CV Quads and Canals- Core Boring, March 31, 2006.

PBRF 2007 Plum Brook Reactor Facility Technical Basis Document, Adjusted Gross DCGLsfor StructuralSurfaces, PBRF-TBD-07-001, June 2007.

PBRF 2007a Plum Brook Reactor Facility Technical Basis Document, Efficiency Correction Factor,PBRF-TBD-07-004, November 2007.

PBRF 2009 Plum Brook Reactor Facility, Memorandum to Project File, J. L. Crooks, Don Young, FinalFSS Report Background -Reactor Building (1111), December 10, 2009.

PBRF 2009a Plum Brook Reactor Facility Technical Basis Document, 44-10 NaI Detector MDC Scan Values for Various Survey Conditions, PBRF-TBD-09-002, June 2009.

47

Plum Brook Reactor Facility FSSR Attachment 12, Rev. 0 PBRF 2009b Plum Brook Reactor Facility Technical Basis Document, RadionuclideDistributions and Adjusted DCGLsfor Site Soils, PBRF-TBD-09-001, June 2009.

PBRF 2012 Plum Brook Reactor Facility Decommissioning Project Office, Memorandum to Project File, EngineeringRecordfor FinalStatus Survey Report Calculations-Reactor Building Update. January 19, 2012.

PBRF 2012a Plum Brook Reactor Facility Technical Basis Document, Re-evaluation of Structure DCGLs and Uranium Activity Fractions, PBRF-TBD- 11-002, January 2012.

PNL 2010 Battelle Pacific Northwest Laboratories (PNL), Visual sample Plan, Version 5.9, 2010.

SNL 1999 Sandia National Laboratories (SNL), for US Nuclear Regulatory Commission, Residual Radioactive ContaminationFrom Decommissioning,ParameterAnalysis, NUREG/CR-5512, Vol.3, Oct. 1999.

TELE 1987 Teledyne Isotopes, An Evaluationof the Plum Brook Reactor Facilityand Documentationof Existing Conditions, Prepared for NASA Lewis Research Center, December 1987.

USEPA 2002 Memorandum of Understanding, US Environmental Protection Agency and US Nuclear Regulatory Commission, Consultation and Finality on Decommissioning and Decontaminationof ContaminatedSites, October 9, 2002.

USNRC 2000 US Nuclear Regulatory Commission, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575, Rev.1, August 2000.

USNRC 2006 US Nuclear Regulatory Commission, ConsolidatedDecommissioning Guidance, Characterization,Survey and Determinationof Radiological Criteria,NUREG 1757, Vol. 2, Rev. 1, September 2006.

7.0 Appendices Appendix A - Exhibits Appendix B - Survey Unit Maps and Tables Showing Measurement Locations and Results Appendix C - QC Measurements Appendix D - Evaluation of Revised DCGLs 48

Final Status Survey Report Attachment 12 Reactor Building Revision 0 Appendix A Exhibits

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 2 of 20 List of Exhibits Exhibit 1, Reactor Building Viewed from the West (from Radar Tower, Jan 2011) ..................... 3 Exhibit 2, Construction Photo Showing Reactor Building Framing Viewed from Northwest ..... 3 Exhibit 3, Construction Photo View from SW Showing Exterior Nearing Completion ................ 4 Exhibit 4, Reactor Building First Floor - Plan View ....................................................................... 5 Exhibit 5, Reactor Building Mezzanine - Plan View .................................................................... 6 Exhibit 6, Reactor Building - 15 ft. Elevation - Plan View ............................................................. 7 Exhibit 7, Reactor Building - 25 ft. Elevation - Plan View ........................................................... 8 Exhibit 8, View of Mockup Reactor in Canal H circa 2000 ........................................................... 9 Exhibit 9, Spent Fuel Racks in Canal G circa 2000 ......................................................................... 9 Exhibit 10, RB 0 ft. Elevation NW Area Looking South circa 2000 ............................................. 10 Exhibit 11 V iews of C anal F circa 2010 ............................................................................................ 11 Exhibit 12, Views of Canals G and H circa 2010 ......................................................................... 12 E xhibit 13, RB M ezzanine ................................................................................................................. 13 Exhibit 14, Mezzanine Pipe Chase Floor and Lower Walls (RB-2-2) ......................................... 14 Exhibit 15, PPP Room Floor and Pipe Trench (RB-4-20) ............................................................. 14 Exhibit 16 Main Pipe Trench in -25 ft. Elevation Sub-basement (RB-4-13) ................................ 15 Exhibit 17, -25 ft. Elevation Wall Section # 8 (RB-4-22) ............................................................. 15 Exhibit 18, View from -25 ft. Sub-Basement Looking NW toward Upper Basement (RB-4-29) .... 16 Exhibit 19 -25 ft. Ceiling Section #4 Facing East (RB-4-37) ........................................................ 16 Exhibit 20, SMTA - Rough Remediated Concrete in - 25 ft. Trench (RB-4-12) ........................ 17 Exhibit 21, UCM -25 ft. Wall Penetrations (RB-4-27) .................................................................. 17 Exhibit 22, UCM- 25 ft. Pipe Flanges (RB-4-28) ...................................................................... 18 Exhibit 23, UCM -25 ft. Column Joints (RB-4-29) ...................................................................... 18 Exhibit 24, UCM -25 ft. Pump Room Concrete Block with Exposed Wire Mesh (RB-4-3 1) ..... 19 Exhibit 25, UCM -25 ft. Pump Room Wall with Anchor Holes (RB-4-32) .................................. 19 Exhibit 26, UCM -25 ft. Pump Room Ceiling Pipe Stub and Welded Strips (RB-4-33) .............. 20 Exhibit 27, UCM Remediated Door Frame Between Canals E and F (RB-4-1) .......................... 20

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 3 of 20 Exhibit 1, Reactor Building Viewed from the West (from Radar Tower, Jan 2011)

Exhibit 2, Construction Photo Showing Reactor Building Framing Viewed from Northwest

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 4 of 20 Exhibit 3, Construction Photo View from SW Showing Exterior Nearing Completion

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 5 of 20 Exhibit 4, Reactor Building First Floor - Plan View EXPERIMENT CONTROL ROOM

¶L ýJ 8 19 It F 5 /

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 6 of 20 Exhibit 5, Reactor Building Mezzanine - Plan View REACTOR 1J1LIII ROOM LL3 15

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 7 of 20 Exhibit 6, Reactor Building - 15 ft. Elevation - Plan View ATS TUNNEL COLD PIPE TUNNEL

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 8 of 20 Exhibit 7, Reactor Building - 25 ft. Elevation - Plan View

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 9 of 20 Exhibit 8, View of Mockup Reactor in Canal H circa 2000 Exhibit 9, Spent Fuel Racks in Canal G circa 2000

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 10 of 20 Exhibit 10, RB 0 ft. Elevation NW Area Looking South circa 2000

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 11 of 20 Exhibit 11 Views of Canal F circa 2010 View Looking North Showing Walls and Opening to Canal E (RB-4-2)

Canal F Floor Looking toward Canal G (RB-4-1)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 12 of 20 Exhibit 12, Views of Canals G and H circa 2010 Canal G SW Lower Wall Showing Extensive Concrete Shaving (RB-4-7)

Canal H Upper Wall East End (RB-4-8)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 13 of 20 Exhibit 13, RB Mezzanine Room 207 Upper Walls and Ceiling (RB-2-1)

Room 15 (Reactor Control Room) Facing Southwest (RB-2-2)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 14 of 20 Exhibit 14, Mezzanine Pipe Chase Floor and Lower Walls (RB-2-2)

Exhibit 15, PPP Room Floor and Pipe Trench (RB-4-20)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 15 of 20 Exhibit 16 Main Pipe Trench in -25 ft. Elevation Sub-basement (RB-4-13)

Exhibit 17, -25 ft. Elevation Wall Section # 8 (RB-4-22)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 16 of 20 Exhibit 18, View from -25 ft. Sub-Basement Looking NW toward Upper Basement (RB-4-29)

Exhibit 19 -25 ft. Ceiling Section #4 Facing East (RB-4-37)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 17 of 20 Exhibit 20, SMTA - Rough Remediated Concrete in - 25 ft. Trench (RB-4-12)

Exhibit 21, UCM -25 ft. Wall Penetrations (RB-4-27)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 18 of 20 Exhibit 22, UCM - 25 ft. Pipe Flanges (RB-4-28)

Exhibit 23, UCM -25 ft. Column Joints (RB-4-29)

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 19 of 20 Exhibit 24, UCM -25 ft. Pump Room Concrete Block with Exposed Wire Mesh (RB-4-31)

Exhibit 25, UCM -25 ft. Pump Room Wall with Anchor Holes (RB-4-32) i j

Plum Brook Reactor Facility FSSR, Attachment 12 Appendix A, Rev. 0, Page 20 of 20 Exhibit 26, UCM -25 ft. Pump Room Ceiling Pipe Stub and Welded Strips (RB-4-33)

Exhibit 27, UCM Remediated Door Frame Between Canals E and F (RB-4-1)