ML11262A285

Plum Brook Reactor Facility Final Status Survey Report Attachment 11, Revision 0, Cover Page Through Appendix a

ML11262A285
Person / Time
Site:	Plum Brook
Issue date:	09/15/2011
From:	Mann B US National Aeronautics & Space Admin (NASA)
To:	NRC/FSME
References
Download: ML11262A285 (70)

Text

Plum Brook Reactor Facility Final Status Survey Report Attachment 11 Revision 0 Reactor Containment Vessel

FINAL STATUS SURVEY REPORT ROUTING AND APPROVAL SHEET Document

Title:

Final Status Survey Report, Attachment 11I Reactor Containment Vessel Revision Number: 0 ROUTING Y

SIGNATW DATE Prepared By B. Mann ........ It (5t L Prepared By N/A REVIEW & CONCURRE~NCE Independent Technical Reviewer R. Case Other Reviewer, QA Manager J. Thomas - IR" \

Other Reviewer N/A FSS/Charactrization Manager W. Stoner 9 /i //

NASA Project Radiation Safety Officer W. Stoner P /

ii I

NASA PBRF DECOMMISSIONING PROJECT CHANGE/CANCELLATION RECORD DOCUMENT TITLE: Final Status DOCUMENT NO: N/A REVISION NO: 0 Survey Report, Attachment 11, Reactor Containment Vessel Revision 0: Initial issue of Report IAD-01/31 Rev 1 iii

Plum Brook Reactor Facility FSSR Attachment 11, 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, 37 pages 0 Appendix A 0 24 pages Appendix B 0 96 pages 4 4. 4. 4 4 4. 4. 4 I

~1 *6~ 4 I

4. *1~ 4 I I U I 7~7 L~~iJ AD-C 1/5 iv

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 TABLE OF CONTENTS 1.0 Introduction ........................................................................................................................... 1 2.0 Reactor C ontainm ent V essel D escription ...................................................................... 2 2.1 CV Layout and Construction ........................................................................................... 2 2.2 System s and Services ....................................................................................................... 4 2.3 Containm ent Vessel M odifications .................................................................................. 6 2.4 Final Configuration and Scope ...................................................................................... 6 3.0 H istory of O perations ....................................................................................................... 7 3.1 Chronology .......................................................................................................................... 7 3.2 Startup and Operations ..................................................................................................... 8 3.3 Radioactive M aterials in the CV .................................................................................... 8 3.4 Post-Shutdown Materials Disposition and Characterization ......................................... 11 3.5 Decom m issioning ......................................................................................................... 12 4.0 Survey D esign and Im plem entation ............................................................................... 13 4.1 FSS Plan Requirem ents ................................................................................................... 14 4.2 Area Classification and Survey Unit Breakdown .......................................................... 15 4.2 Num ber of Measurem ents .............................................................................................. 18 4.4 Instrum entation and M easurem ent Sensitivity ............................................................. 22 5.0 CV Survey R esults ......................................................................................................... 24 5.1 Scan Surveys ...................................................................................................................... 24 5.2 Systematic Measurem ents and Tests .............................................................................. 27 5.3 Investigative M easurem ents and Tests ........................................................................ 29 5.4 Q C M easurem ents .......................................................................................................... 31 5.5 A LARA Evaluation ....................................................................................................... 33 5.6 Comparison w ith EPA Trigger Levels ........................................................................... 34 5.7 Conclusions ........................................................................................................................ 35 6.0 References ............................................................................................................................. 35 7.0 Appendices ........................................................................................................................... 37 A ppendix A - Exhibits .....................................................................................................................

Appendix B - Survey Unit Maps and Tables Showing Measurement Locations and Results ........

v

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 LIST OF TABLES Table 1, Unusual Incident Reports Involving Radioactive Materials in the CV (1960-1973) ...... 10 Table 2, CV Radionuclide Activity Fractions and Gross Activity DCGLs ..................................... 14 Table 3, Class-Based Survey Scan Coverage and Action Level Requirements .............................. 15 Table 4, CV Survey U nits for FSS .................................................................................................. 15 Table 5, CV Survey Unit Breakdown by Major Elevation ............................................................ 18 Table 6, CV Survey Design Summ ary ........................................................................................... 20 Table 7, Sensitivity Analysis for CV FSS Design .......................................................................... 21 Table 8, Typical Detection Sensitivities of Field Instruments ........................................................ 23 Table 9, Scan Survey Results ......................................................................................................... 24 Table 10, CV Total Surface Beta Activity Measurement Summary and Test Results .................... 27 Table 11, CV Investigative Static Measurements .......................................................................... 30 Table 12, CV Total Surface Activity QC Measurements ............................................................... 31 Table 13, Screening Level Values for CV and Radionuclide Activity Fractions ........................... 34 LIST OF FIGURES Figure 1, PBRF NW Area Showing the CV, Reactor Building and Other Support Buildings ...... 2 vi

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 LIST OF ACRONYMS & SYMBOLS A Detector open area (cm 2)

(X 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 13 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 cfm cubic feet per minute cm centimeters cm 2 square Centimeters cpm counts per Minute CV Reactor Containment Vessel (Part of Reactor Building, Building 1111)

CRB Cold Retention Basin, Building 1154 A delta, DCGLw - LBGR d' 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 Ft Total efficiency EMC Elevated Measurement Comparison EMT Elevated measurement test EP embedded pipe EPA US Environmental Protection Agency F Fahrenheit FSS Final Status Survey FSSP Final Status Survey Plan FSSR Final Status Survey Report ft. feet

, gamma, denotes gamma radiation g gram gpm gallons per minute h hour HRA Hot Retention Area, Building 1155 HSOO Health Safety Operations Office HVAC Heating, Ventilation and Air Conditioning i observation counting interval during scan surveys in. inch vii

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 LIST OF ACRONYMS & SYMBOLS, Continued LMI Ludlum Measurements, Inc.

LBGR Lower Bound of the Gray Region m2 square meters mrad millirad, unit of absorbed dose mrem millirem, unit of effective dose equivalent MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC Minimum Detectable Concentration MDCscan Minimum Detectable Concentration for scanning surveys MDCstatic Minimum Detectable Concentration for static surface activity measurements MDCR Minimum Detectable Count Rate MPC Maximum Permissible Concentration MOU Memorandum of Understanding MTR Materials Test Reactor MW Megawatt NASA National Aeronautics and Space Administration N Number of FSS measurements or samples established in a survey design N/A Not Applicable NRC US Nuclear Regulatory Commission PBRF Plum Brook Reactor Facility PCW Primary Cooling Water PNL Pacific Northwest Laboratory Standard normal distribution function p surveyor efficiency for scan surveys psig pounds per square inch - gauge pCi/g picocuries per gram PPH Primary Pump House, Building 1134 PPP process piping pump percent 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 SEB Service Equipment Building, Building 1131 SNL Sandia National Laboratory SR Survey Request viii

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 LIST OF ACRONYMS & SYMBOLS, Continued tb background count time ts sample count time TBD Technical Basis Document P Mean activity concentration UCM Unusual Condition Measurement UL Upper limit of the confidence interval about the mean itCi/cc microcuries per cubic centimeter gtCi/ml microcuries per milliliter VSP Visual Sample Plan y year Z l-. 100(1-a) percentile of the normal distribution ZI-1 100(1 -3) percentile of the normal distribution 00 Mathematical symbol for infinity ix

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 1.0 Introduction This report presents the results of the final status radiological survey of the Plum Brook Reactor Facility (PBRF) Reactor Containment Vessel (CV). It is Attachment 11 of the PBRF Final Status Survey Report (FSSR). 1 This attachment describes the CV, its operational history and final condition 2 for the final status survey (FSS). It describes the methods used in the FSS and presents the results.

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 (DCGL) have been established for residual surface contamination in the CV. Considering the radionuclide mixtures established for areas within the Reactor Containment Vessel, gross beta DCGLs range from 11,563 to 37,235 dpm/100-cm2 .

The FSS measurement results and supporting information are presented to demonstrate that residual contamination levels in each survey unit of the CV are well below the respective DCGLs. It is also shown that residual contamination has been reduced to levels that are consistent with the ALARA requirement. Therefore, the Reactor Containment Vessel meets the criteria for unrestricted release.

Section 2.0 of the report provides a description of the CV. The CV layout, relation to the Reactor Building and other PBRF facilities, design and materials of construction, contents and use, systems and services, building modifications, final configuration and scope of the FSS are described.

In Section 3.0, a chronology of significant milestones is followed by a 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 CV. 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 presented.

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

Supporting information is contained in Appendices. Appendix A contains drawings and photos 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.

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

2 While the CV is within the Reactor Building (RB), due to their combined size and complexity, the two structures were administratively separated for conduct and reporting of the FSS. The Reactor Building FSS is reported in Attachment 12 of the FSS Report.

I

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 2.0 Reactor Containment Vessel Description The CV is a large cylindrical steel shell, or Tank which encloses the 60 MW Plum Brook Test Reactor. It was designed to prevent release of radioactive materials to the environment in the event of an accident involving the Reactor or other experimental facilities within the CV [NACA 1956]. It is 100 ft. in diameter and 111 ft. in height at the center, extending from 56 ft. below grade to 55 ft.

above grade. It surrounds the Reactor Tank, Bioshield and Reactor experiment-test areas. The CV layout, history of construction, systems and services and configuration for FSS are described below.

2.1 CV Layout and Construction Figure 1, a map of the PBRF main facilities area, shows the Reactor Building, CV and other principal support buildings. A view of the Reactor Building and CV exterior in late 2010 is shown in Exhibit 1 of Appendix A.

Figure 1, PBRF NW Area Showing the CV, Reactor Building and Other Support Buildings RD 77'

," "-7/" "..

[SB RB SECURITY BUILDING REACTOR BUILDING CV CONTAINMENT VESSEL S HL HOT LAB PPH PR.IMAR.Y PUMP HOUSE

-DEMOLISHED BUILDINGS 1771 ED SERVICES EQUIPMENT BUILDING FH FFAN HOUSE WHB WASTE HANDLING BUILDING

.I.ROW REACTOR OFFICE - LABORATORY BUILDING HRA HOT RETENTION AREA

"*THESE BUILDINGS WERE DEMOLISHED TO -3 ft.

ELEVATION AND BACKFILLED.

2

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 The CV housed the 60 MW Test Reactor, including the Reactor core and associated experiment hardware. The Reactor pressure vessel, or Reactor Tank, a vertically mounted steel cylinder was located at the center of the CV. It was of varying wall thickness (one to two in.) with a hemispherical lower end and a flanged elliptical upper head. It was of dimensions 9 ft. inside diameter by 32 ft. in height. The Tank was constructed with an interior cladding of 304 low-cobalt stainless steel of thickness up to 10 per cent of the Tank wall thickness. The cladding varied with location (wall section, head and bottom). Three removable 20-ton shrapnel shields were positioned over the Tank during Reactor operations. Thick high density concrete shielding (the Reactor Bioshield) surrounded the Reactor Tank circumference. The top of the Bioshield structure extended beyond the main shield radius to provide a work area (called the Lily Pad) for refueling and other in-vessel work. The Reactor core center was located 21 ft. below grade. Exhibit 2 of Appendix A, a construction era photo, shows the Reactor Tank, Bioshield and Lily Pad and the surrounding quadrants.

As shown in Exhibit 3 of Appendix A, a plan view of the CV 0 ft. elevation, the experiment support facilities surrounding the Reactor were housed in concentric rings. The inner ring was comprised of four quadrants (A, B, C and D). The circular concrete quadrant outer wall was about 73 feet in diameter and roughly 3-ft. thick. The dividing walls that separated the 25-foot deep quadrants were about 2 ft. thick. The outer ring, 13 ft. wide at the floor (-25 ft.

elevation), was located between the quadrant outer wall and the CV shell. The southern one-third of this annulus, a structure called Canal E, was connected to Quadrants A and C on the inside and to Canal F outside the CV. Canal E was connected to Quadrants A and C and Canal F by 9 x12 foot vertical lift doors. The section comprising the northern two-thirds of the CV outer annulus was called the "Dry Annulus", or the "Experimental Area".

Quadrant B and Canal E were dry during Reactor operations. Quadrant B housed a stairway from grade level to the bottom at the -25 ft. elevation for access to three thermal column beam tubes. Quadrant C contained a fuel transfer pit in the north floor area and quadrant D contained an underwater beam room on the west floor area. The dry northern part of the annular ring opposite Canal E contained experimental facilities, an elevator and stair wells connecting the Sub-pile Room, -25 ft. and 0 ft. levels within the CV. It also contained hydraulic support equipment for the Reactor control regulating rods, the poison injection system, two hot caves (one on the north side and one on the east side), a beam catcher in line with the underwater beam room in Quad D and a hot sump. There were two metal cat-walkways at different levels on the Dry Annulus walls to provide access to equipment.

The CV wall and upper dome are constructed of 3/4/4-in. steel plate while the bottom dish is constructed of welded 3/8-in.steel plate. The free volume of the CV is approximately one half million cubic feet with all quadrants and Canal E empty and 450, 000 cubic feet with those areas full of water. The elliptical CV dome was insulated to maintain the CV minimum temperature above 32 degrees F. Exhibit 4 of Appendix A is a construction era photo of the above grade portion of the CV. See the photo in Exhibit 5 of Appendix A for a view of the CV bottom shell under construction. It was constructed in-place over crushed limestone fill and pressure-sealed with grout to provide full contact support and to prevent leakage into the surrounding bedrock (and groundwater table).

3

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 The Subpile Room is located at the -44 ft. elevation at the bottom of a concrete stairwell. It was a shielded watertight room designed for 150 psig located directly below the Reactor core.

It housed the control rod drives and four instrumentation thimbles. The room was steel lined with a dogged door, normally closed during power operations to restrict entry. The floor was grated with a hot sump underneath. A large lead donut (about 10 foot in diameter and one foot thick) embedded in concrete provided biological shielding between the Reactor and the Subpile Room.

There are approximately 80 electrical penetrations and 110 piping penetrations in the CV walls, floors and dome. During licensed operations, the maximum allowable leak rate, including penetrations, was 450 cubic feet per day at standard temperature and pressure at 0.3 psig which is the maximum overpressure associated with the design basis, or maximum credible accident [NACA 1956].

The physical arrangement of the CV within the surrounding Reactor Building is illustrated by Exhibits 6 and 7 of Appendix A. They are E-W and N-S cross section views of the combined structures, respectively. Personnel access to the CV was provided through two airlock doors:

one on the northwest side and one on the southeast side. They provided access from the Reactor Building at the 0 ft. elevation to the CV. There was an experiment airlock connecting the Reactor Building to the 0 ft. elevation in the CV above Canal E.

2.2 Systems and Services Basic CV and 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 Service Equipment Building (SEB). Of special note for the CV are the 60 MW Test Reactor, the Primary Cooling Water (PCW) System and associated experiment and test facilities. Also of interest are radioactive material handling, process waste handling and monitoring systems associated with the CV.

These include HVAC systems, the Hot Drain System and the Remote Area Monitoring System (RAMS).

The Plum Brook 60 MW Test Reactor was light-water cooled and moderated. The core consisted of 27 fuel elements, each containing 18 curved plates of highly enriched uranium-aluminum alloy clad with aluminum. The core design was based on the Materials Test Reactor (MTR) located at the Idaho National Engineering Laboratory in Arco, Idaho. The Reactor was cooled by demineralized light water, which also served as the neutron moderator.

The primary neutron reflector was constructed of beryllium plates and primary cooling water surrounding the core provided additional reflection. The average core fast flux (E > 0.1 MeV) and average core thermal flux were 4.2xl 014 and 3.5x 10 5neutrons/cm2 per second, respectively [NACA 1956].3 Primary cooling water was circulated between the Reactor Tank and the Primary Pump House (PPH) through 24 in. supply and return piping. The PPH housed PCW processing and makeup 3 Many systems and services are common to the CV, the Reactor building and to other PBRF buildings. Those highlighted here were unique to the CV and the 60 MW Test Reactor, were radioactive process systems or were essential to operations within the CV.

4

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 equipment and the secondary cooling system. Exhibit 8 of Appendix A shows a cross section schematic of the PCW piping layout between the PPH and the CV.

The experimental facilities included two horizontal through-holes (one passing through the south beryllium reflector and one passing adjacent to the north side of the core), six horizontal beam holes (including a large thermal column), two vertical test holes, and forty-five in-core test locations. The photo in Exhibit 9 of Appendix A shows the arrangement of the Reactor Tank internals including the fuel box, beryllium reflector assembly, experimental irradiation through tubes and beam tubes. The beam tubes and through-tubes provided access to the Reactor core for high neutron flux experiments. Cryogenic loop transfer tables were attached to the floors of Quadrants A, C and D. Quadrant D also contained an underwater beam room and shutdown cooling water equipment on the floor level. Quadrant D also provided access to three large horizontal beam tubes. Quadrant A contained two hydraulic "rabbit" insertion and removal systems.

Fill water for the quadrants and canals was supplied from the PBRF deionized water system located in the SEB (via the Upper-elevated Water Storage Tank). The Quadrant Recirculation System circulated quadrant water through filters and deionizers to maintain water quality.

Two 350 gpm recirculation pumps located in the CV Dry Annulus at the - 25 ft. elevation circulated water from the quadrants through filters and deionizers located in the Fan House.

The Quadrant and Canal Pump-out system provided the principal method to remove water from the quadrants and canals. Two 1500 gpm pumps, also located in the CV Dry Annulus, pumped water from Quads A, C, D and Canal E. The water could be discharged either to the Cold Retention Basins (CRBs) or the Hot Retention Area (HRA).

Water from all CV floor drains was collected in hot sumps. The water was discharged through the Reactor Building Process Piping Pump (PPP) room sump to the Fan House for cleanup and/or storage in the CRBs or the HRA Tanks. Primary Cooling Water system drainage also flowed through the PPP Room to the HRA.

Ventilation for the CV was controlled through an air intake system located on the southeast side of the dome. A negative pressure of 2 inches of water was maintained so that leakage was always inward to the CV. The CV was heated (and cooled) by four 5600 cfmn units located above the main operating floor at the 12 ft. elevation. Ventilation air was purged from the CV through a 6 in. line on the west side of the CV. The exhaust air was directed to the Fan house where it was filtered, compressed, monitored in holdup Tanks and then released to the 100-foot PBRF stack.

Material handling in the CV was conducted manually on transfer carts (both dry and in the water-filled quads and Canal E). Large items were handled by cranes. Heavy hoisting services in the CV were provided by an overhead polar crane with 100 ft. bridge-rails and dual rolling hoist carriages of 20 ton and 5 ton capacity (the latter was subsequently de-rated to 4 ton capacity).

There were areas within the CV where large-sized highly radioactive and contaminated materials were handled, usually underwater. Also, numerous small shielded irradiated items (flux wires, rabbit capsules, primary water samples, etc.) were transported outside to the Hot 5

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Laboratory (HL) or the radiochemistry labs in the Reactor Office and Laboratory Building (ROLB).

The RAMS provided essential information on radiological conditions in the CV and throughout the PBRF. It provided continuous radiation monitoring information to the Health Safety Operations Office (HSOO) in the 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, gaseous 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 and gaseous activity, water activity and direct radiation were located throughout the CV. 4 2.3 Containment Vessel Modifications After the start of Reactor operations in 1963, modifications to experimental facilities, equipment and structures within the CV continued throughout the operating life of the facility.

These included:

" beryllium reflector plate replacement in the 60 MW Test Reactor,

" addition of insertion tables for Reactor experiments in Quadrants A, C and D,

" an underwater beam room was installed on the floor of Quadrant D,

" hot cells were installed on the east and north sides of the Dry Annulus at the - 25 ft.

elevation (these were small hot cells, generally referred to as the Hot Caves),

" neutron radiography capability was added to Quadrant A,

" an experimental air lock was added to the CV shell on the southwest area of the 0 ft.

elevation

" concrete partitions were installed on the Quadrant B thermal column,

" a balcony was added to Quadrant B,

• acoustic panels were added to the CV dome ceiling, and

" a balcony was added and the traveling bridge was relocated above Quadrant C.

In addition, support services were added for the expanded experimental facilities. These included: cooling water and cryogenic cooling, service air, electrical power and instrumentation.

2.4 Final Configuration and Scope The final configuration of the CV for the FSS was primarily bare structure surfaces with external piping equipment and fixtures removed where possible. The Reactor Tank and internals have been removed, as well as associated mechanical and electrical equipment. The scope of the FSS reported in this attachment includes the CV interior surfaces and the Dome exterior. Exterior CV surfaces below the RB roof are covered in the FSS Report Attachment 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.

6

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 12 (Reactor Building). 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 FSS of piping embedded in CV concrete or piping buried beneath or adpacent to the structure. These results are reported in separate attachments to the FSS Report. See Section 3.5 for additional discussion of equipment removed and preparation of the CV for the FSS.

After completion of FSS measurements, the spaces below the -25 ft. elevation were sealed by filling with grout. This includes the elevator shaft and walkway at the - 37 ft. 6 in. elevation, the Subpile Room at the - 44 ft. 2 in. elevation and the stairway between.

3.0 History of Operations A chronology of major milestones is given below. This is followed by a discussion of CV operations, post-shutdown and decommissioning activities. Emphasis is on operations with radioactive materials 6

that could affect the final building condition and final status survey.

3.1 Chronology Major PBRF and Reactor CV milestones are listed below:

1956 - September, groundbreaking for PBRF including excavation for CV.

1959 - Exterior construction of CV completed.

1961 - June, 60 MW Test Reactor critical.

1961- 1963 Reactor system testing.

1963 - Full Power 60 MW Test Reactor Operations begin.

1973- January 5th, Reactor shutdown (after 152 operating cycles).7 1973 - June 30, PBRF facilities placed in "standby condition.

1985 - Initial radiological characterization, Teledyne Isotopes Inc.

2002 - Decommissioning Plan approved.

2003-2004 - Initial equipment removal and structure decontamination.

2004 -2008 - Reactor Tank and internals and portions of Bioshield removed.

2008 - 2010 - Remediation of CV contaminated areas and preparation for FSS.

2011 - FSS measurements completed.

5 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.

6 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.

7 The length of an operating cycle was determined by fuel bum-up in the 60 MW Test Reactor. When core neutron flux levels decreased below experiment requirements, 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.

7

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 3.2 Startup and Operations From 1960-1973, significant activities took place within the CV. Startup activities included hydraulic testing of primary cooling system flow through the Reactor core, initial Reactor fueling, low power and approach to full power neutronics testing. Experimental facility and experiment installations took place; both in-core and in the quadrants.

Operations in the CV required a large number of personnel during the shutdown periods when refueling and major facility modifications took place. Most personnel were on the Monday through Friday day shift. A limited number of personnel worked the back shifts (less than 10 normal shift operations staff plus special needs personnel) seven days per week. During Reactor full power operations, the CV was entered only as needed for routine equipment monitoring and maintenance, special operations and experimental activities.

Operations activities conducted included preparation for in-Tank work such as removing the 20-ton shrapnel shields, opening the Reactor Tank hatch and unloading spent fuel from the Reactor core into Quadrant C and loading new fuel. Irradiated materials were transferred to and from the Reactor area through the quadrants to Canal E using underwater carts. Water management activities included filling quads and canals with deionized water and draining them when required. Underwater activities included using long handled tools, loading and unloading rabbit facilities for irradiation and subsequent examination. Supporting activities included operating and maintaining auxiliary equipment such as the airlocks and moving un-irradiated experiment and Reactor hardware into the CV.

3.3 Radioactive Materials in the CV The US Atomic Energy Commission (AEC) authorized Reactor operations and use of radioactive materials at the PBRF under several licenses. 8 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 Containment Vessel were those originating from PBRF tests and experiments [PBRF 2009].

Many activities in the CV involved potentially hazardous work. Some involved potential for significant radiation exposure or radioactive contamination. Refueling, for example, involved work over the open Reactor hatch with long handled tools in contaminated Primary Cooling Water; post-irradiation rabbit removal involved tools to remove them from the end station under water. These types of activities were controlled by procedures and safe work permits to ensure personnel exposure were within regulatory limits. Protective clothing was normally required for work in the CV.

Radioactive contamination in the CV was of two primary origins: 1) irradiated test specimens and associated hardware and 2) releases of airborne radioactive materials and contaminated liquids from the Reactor Tank and process systems. Most of the irradiated test specimens did 8 Authority for the PBRF Reactor and radioactive materials licenses was assumed by the US Nuclear Regulatory Commission in 1975.

8

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 not remain in the CV; they were transferred to other buildings in the PBRF, primarily the Hot Laboratory. Experiment transfers were occasionally conducted overland significantly increasing general radiation levels in the CV.

Airborne radioactivity often resulted from off gassing of Primary Cooling Water into the CV atmosphere. The resulting airborne radioactivity included particulates such as Rb-88, Cs-138, W-187, In-i 15 and Cd-i 15; gases such as Xe-133 and 135, Kr-85 and 88, and volatile radionuclides 1-131 and 1-132. Leaking valves or contaminated water dripping from handling tools led to occasional surface contamination incidents. Experimental work led to gaseous and particulate releases into the CV which included the radionuclides Ar-41, Po-210, Sb-124, Cu-64, Au-198, Na-24, Eu-154 and fission products including Sr-90 and Cs-137. Also U-235, H-3, Co-60 and Fe-59 were also associated with such activities.

Table 1 lists the types of the incidents reported in PBRF Operating Cycle Reports that involved operations within the CV. The list is not all-inclusive, but rather shows the types of events that occurred during operations. About seventy such events were reported over the 152 operating cycles of the 60 MW Test Reactor. Note that units of measure and concentration limits reported in the Table 1 event descriptions may not be consistent with those in current PBRF procedures and current regulations. These cycle reports were written in the 1960s.

9

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 1, Unusual Incident Reports Involving Radioactive Materials in the CV (1960-1973)

Report Cycle Date Description No.

3 8/18/63 A pin-hole leak in the wall between Quad C and the Dry Annulus leaked about 1/2 pint per hour and contaminated about 15 square feet of the floor area with quadrant water reading 1.6x105 gtCi/ml. Airborne radioactivity was also observed at grade level and the -25 foot level due to primary cooling water off gassing. The CV was evacuated. Gamma scans indicated W-187 readings of 5xl0-9 and 4xl0-9 gCi/ml at grade level and the - 25 ft. level respectively. The maximum permissible concentration (MPC) for W-187 is 3x10-7 gtCi/ml. Reentry to the CV was subsequently authorized. 9 4 9/7/63 Radon and daughter products increased airborne activity levels in the CV to lxl07 pCi/cc. This became a common occurrence and was not reported thereafter in Unusual Incident Reports.

8 11/3/63 During Reactor shutdown with Quad A dry, water spilled from a rabbit system tube and contaminated half of the quadrant floor creating direct radiation levels of up to 5 mrem/h. The area was decontaminated. Also during this cycle, the Quad D floor was contaminated; personnel were contaminated while securing the Canal E to Quad C door. The Quad D Hot Cave flooded when a seal to Quad D failed; and primary water leaked to floor when the Lockheed test loop was withdrawn from the core.

13 3/4/64 A small piece of spiro-helic gasket from the Primary Cooling Water system was found on the CV grade level floor. The piece, reading 25 rad/h beta and 2.5 rad/h gamma, on contact, was disposed of as radioactive waste.

15 3/26/64 The Quad D water activity monitor alarmed from in-leakage of primary water during experiment loop insertion and removal. The activity level was 5.6x10 tCi/ml gross beta-gamma. Also, during this refueling cycle, several people received minor skin contamination. The contamination was removed with repeated washing of the skin.

19 5/20/64 Three maintenance personnel were soaked with primary water while removing control rod drives in the Subpile Room. Contamination levels were up to 500 mrad/h beta-gamma on contact. The men took several showers in the RB to reduce levels to background.

24 8/28/64 Increased airborne radioactivity levels were noted on all CV RAMS units. The maximum surface radioactivity level was 150,000 dpm beta-gamma near the Quad D Hot Cave. Respiratory protection was required for CV entry. Radioactivity appeared to be Cd-1 15 and In-I15m that came through a ruptured exhaust filter during melting operation in the Hot Cave.

33 4/15/65 The Reactor Tank water level was drained to about 6 inches above the core for in-Tank work.

When radiation levels of 400 mrem/h occurred at the Reactor Tank hatch, the Lily Pad RAMS units alarmed. Also, two personnel handled a small piece of metallic gasket from the Reactor Tank reading 290 rem/h. Each received a hand exposure of about 800 mrem.

39 10/15/65 Two workers were exposed to 2 rad/h fields when an experiment capsule was withdrawn while they were checking flow switches in Quad D. The incident was caused by poor communications.

67 10/20/67 Primary cooling water leaked into Quad C from Experiment 62-06 causing direct radiation levels to increase in the CV. Water samples from Quadrant C were measured to contain 1.4x1 03 piCi/ml, gross beta activity.

98 10/21/69 A mechanic received an extremity exposure of 3800 mrem while handling an irradiated rabbit (capsule) without the use of a handling tool.

9 This was the first of 25 incidents involving Primary Cooling Water leaks and off gassing into the CV to be reported in Cycle reports. Data on individual events is available in the following PBRF Cycle Reports: 8, 17, 23, 24, 28, 29, 31, 33, 38, 44, 55, 59, 61, 79, 80, 114, 133, 135, 136, 140, 142, 145, 146 and 147.

10

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 3.4 Post-Shutdown Materials Disposition and Characterization In the period between termination of Reactor operations in January 1973 and June 30th of 1973, the CV 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 due to budget constraints, NASA was terminating all nuclear related research operations at the PBRF. The Test Reactor, Mock-up Reactor, and all associated operations were to be shutdown and placed in standby condition and the Reactor staff terminated 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 PBRF operating licenses and placing the facility in standby status. The PBRF End Condition Statements for Protected Safe Storage Mode detailed the facility final condition status goals for mid-1973, including the CV.

In accordance with the shutdown end condition report requirements the 60 MW Reactor core was defueled (some in-core experiments and reflector components were left in place). The CV was utilized to defuel the Reactor and transport Reactor and experiment equipment to the Hot Laboratory. The Primary Cooling Water System was drained and a nitrogen purge system installed to keep the Reactor internals dry. The airlocks into the CV were closed, locked and left in condition for manual operation. Radioactive materials (sealed sources) were inventoried and safely stored and radioactive waste was removed and shipped for disposal.

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

34-06706-03. These licenses authorized possession-only of the remaining radioactive materials on site. 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 objectives for the facility changed from possible restart to standby to decommissioning. In 1982, the NRC terminated BPL 34-06706-03 per NASA's request.

Licenses TR-3 and R-93 were amended to transfer licensed radioactive materials that remained on site 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 2007a].

The radiological status of the CV has been investigated several times during the period between shutdown in January 1973 and start of decommissioning in 2002. The CV 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 CV resided in the activated Reactor Tank and internals. Calculated estimates of principal radionuclides in the inventory as of June 1, 1973 (in Ci) were: Tritium, 206,000; Co-60, 5,077; Fe-55, 26,200; Zn-65, 14,700 and Ni-63, 47.

The Bioshield surrounding the Reactor Tank was investigated by core boring. Activated concrete was encountered at depths of up to 16 inches in the portion of the shield surrounding the Reactor core region and the smaller beam tubes and up to 46 inches in the concrete surrounding the thermal column. The core samples were analyzed for gross beta and alpha 11

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 activity and analyzed by gamma spectroscopy. The only reported isotope-specific results were for Co-60, with maximum concentrations of 33 pCi/g in Bioshield concrete and 325 pCi/g in activated rebar [TELE 1987].

Internal surface contamination of PCW piping and equipment was measured in an 8 in. valve in the PCW system shutdown loop. The measured fixed contamination level was 13,250 dpm/100-cm 2 beta-gamma. A scraping sample analyzed by gamma spectroscopy yielded 704 dpm/100-cm 2 Co-60 and 284 dpm/100-cm 2 Eu-152 with lesser concentrations of Eu-154 and Eu-155. Five samples of sediment-crud were collected from quadrant and canal drains and analyzed by gamma spectroscopy. The average measured Co-60 concentration was 12,000 pCi/g and the average Cs-137 concentration was 2,800 pCi/g. No samples from the CV were analyzed for Sr-90, but Sr-90 was measured in other liquid process system residues with Cs:Sr ratios of approximately 20 to 30 to one [TELE 1987].

In the Teledyne study, the accessible floor and wall surfaces of the CV were divided into 20 pie shaped grids at the 0, 10 and 20 ft. elevations above grade and at the -7, -15, -25 ft.

elevations and the Subpile Room below grade. Measurements of removable and fixed contamination and direct radiation were taken at 222 locations (including 44 in Canal E).

Measured levels of fixed beta-gamma surface contamination ranged from non-detectable to 12,000 cpm 2in the Sub-pile Room) and removable ranged from non-detectable to 9,440 dpm/100-cm (in Canal E). Fixed alpha surface contamination levels ranged from non-detectable to 206 cpm2 and removable alpha surface contamination ranged from non-detectable to 9,440 dpm/100-cm .10 A confirmatory characterization survey of the PBRF was performed by GTS-Duratek in 1998.

However, no results were reported for the CV [GTS 1998].

3.5 Decommissioning The PBRF Decommissioning Plan stated that the 60 MW Test Reactor Tank and internals, contaminated piping, operational and experiment equipment would be removed. It also suggested that the Bioshield (or portions thereof) would be removed [NASA 2007a]. The NRC approved the Decommissioning Plan in 2002. Decommissioning and supporting radiological characterization activities in the CV extended from 2003 through 2010.

The CV was cleared of removable items such as piping, pumps, valve extensions and handles, electrical boxes, cable trays, ventilation systems, electronic instrumentation, experiment hardware, etc. This included waste disposal of radioactively contaminated items. Recyclable items were segregated where possible. Hazardous materials and building fixtures, except for structural components, have been removed from the building to facilitate the FSS.

The Reactor Tank and associated support structures were cut out in sections and shipped off-site for disposal. See Exhibit 10 of Appendix A for a photo of segmentation-machining Reactor internals for removal. The region of the biological shielding surrounding the Reactor core area was removed (3 to 4 ft. above and below the core center-line) and disposed of as

'0 Under the procedures used in the Teledyne study, fixed surface contamination levels on structures (alpha and beta) were reported in units of counts per minute (cpm) [TELE 1987].

12

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 radioactive waste. The Thermal Column and surrounding concrete were removed. Beam tubes and through tubes, were also cut out and disposed of. The photo in Exhibit 11, of Appendix A, shows a view looking downward into the Bioshield interior. The Reactor Tank, surrounding hardware and Bioshield inner steel liner have been removed. In the lower part of the photo, it can be seen that the Bioshield concrete surrounding the active core region has also been removed. Exhibit 12 in Appendix A, shows the Bioshield exterior Quad B face with the concrete surrounding the former Thermal Column removed.

The Subpile Room, access hallway, stairway and connected piping systems and thimbles were grouted up to the ft. elevation. The lead shield-donut and the control rod drive housing below the Reactor Tank were also grouted in place. Exhibits 13 and 14 show views of the Subpile Room prepared for the FSS. The CV airlocks were removed. The lift-gate doors (and frames) connecting the quads and canals were also removed.

Three of the four quadrants (A, C and D) and Canal E were filled with water during Reactor operations and the surfaces of these structures were exposed to contaminated water for extended periods. After removal of attached external piping and equipment, detailed characterization surveys performed in 2005 and 2006 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 2005, PBRF 2006].

A variety of techniques were used to prepare the CV surfaces for FSS. Surface wall coatings of water-filled quads and Canal E were removed by sponge jet blasting and the underlying concrete was shaved or scabbled. Floors of the quads, Canal E and the main elevations were shaved and scabbled. Areas where elevated activity was measured and where contamination had penetrated to depths were remediated by over-coring and impact-tools.

Exhibits 15 through 26 of Appendix A show views of the main CV areas before and after preparation for FSS.

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

4.0 Survey Design and Implementation This section describes the method for determination of the number of fixed measurements and samples for the FSS of the Reactor Containment Vessel. Applicable requirements of the FSS Plan are summarized. These include the DCGLW11, the gross activity DCGL, scan survey coverage and action-investigation levels, classification of areas and breakdown of the survey units. The radiological instrumentation and their detection sensitivities are discussed.

"1 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).

13

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 4.1 FSS Plan Requirements Individual radionuclide DCGLs were obtained for PBRF structures considering exposure to future site occupants from two potential pathways. Building reuse scenario DCGLs for structures were calculated using RESRAD-BUILD Version 3.22. Volumetric DCGLs were calculated for subsurface structures using RESRAD Version 6.21 for a resident farmer scenario. 12 The volumetric DCGLs (in pCi/g) were converted to "effective surface" DCGLs using surface-to-volume ratios for the assumed volume of contaminated subsurface concrete.

The DCGL calculations are described in the FSSP, Attachment B. The smaller of the two DCGLs calculated for each of the radionuclides of concern were used as the structure DCGLs.

A gross activity DCGL is used for structural surfaces in the CV, since multiple radionuclides are potentially present in residual contamination. It can account for beta-gamma and alpha emitters and for so-called hard-to-detect 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 DCGLs for the CV are calculated using equations in the FSSP for gross beta, gross alpha and surrogate DCGLs, based on radionuclide mixtures in residual contamination. Activity fractions and gross activity DCGLs for the CV are shown in Table 2.

Table 2, CV Radionuclide Activity Fractions and Gross Activity DCGLs Radionuclides DCGLw Location H-3 Co-60 Sr-90 ] 1-129 Cs-137 Eu-154 IU-234 I U-235 (dpm/100-Activity Fractions Assigned to Reactor Containment Vessel (%) ( cm2) (2)

Area Are 2121 & 7.29 53.89 22.22 0 12.05 1.71 2.42 0.42 14,600 Subpile Rm.

Area Aread22 && 97.57 Lily Pad 0' 2.10 0.13 0 0.20 0 0 0 11,563 Area 17 88.20 0 0 0 5.30 0 6.50 0 37,235

-25 ft.

Quad A 0 74.99 0 0 25.01 0 0 0 13,450 Quads B, C, 0 33.05 0 0 66.95 0 0 0 21,566 D

Canal E 1.16 11.69 0 0 86.99 0 0.16 0 31,711 CV Dome, Default (3) 27.07 9.65 7.88 1.42 46.71 0.12 6.98 0.17 27,166 Table 2 Notes:

1. Activity profiles and gross activity DCGLs for structures are reported in the Technical Basis Document PBRF-TBD-07-001 [PBRF 2007].

2. These are the basic gross activity DCGLs. In the CV designs, the DCGL values may be adjusted to account for dose contributions from "insignificant radionuclides" and embedded piping (see Table 6 for details).

3. The default radionuclide mixture and associated DCGL in TBD-07-001 were applied to the CV Dome.

12 Potential 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.

14

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Survey designs incorporate requirements for scan coverage and investigation levels derived from the MARSSIM classification of survey units. The values applicable to the Reactor Containment Vessel are shown in Table 3.

Table 3, Class-Based Survey Scan Coverage and Action Level Requirements Static Measurement Classification Scan Survey Scan Lvestigation or Sample Result Coverage Levels Investigation Levels Class 1 100% > DCGLEMC > DCGLEMC Class 2 10 to 100% > DCGLw or > MDCs, > DCGLw if MDCscan is > DCGLw Class 3 Minimum of 10% > DCGLw or > MDCS, > 50% of the DCGLw if MDCcan is > DCGLw -

4.2 Area Classification and Survey Unit Breakdown The CV interior was divided into nine areas for initial classification and final status survey planning as shown in Table 2-1 of the FSS Plan. All areas except the CV dome exterior were classified as MARSSIM Class 1 in the FSS Plan. The Reactor Building roof, including the CV dome was classified as Class 2. As part of the FSS implementation process, individual survey units were established and their final MARSSIM classification assigned. The CV was divided into 87 survey units for the FSS (85 Class 1 and two Class 2). These are identified in Table 4. Table 5 summarizes the survey unit breakdown by major elevation.' 3 Note that the only major floor elevations inside the CV are the 0 ft. and the -25 ft. elevations. The convention for survey unit IDs in the CV is as follows:

0 CV-1, 1 st Floor, 0 ft. elevation and above 0 CV-3, -25 ft. elevation 0 CV-4, roof and exterior walls (note that the prefix CV-2 is not used).

Table 4, CV Survey Units for FSS Survey Area Survey Class in Unit (1) Class (M 2) Design SR # Description FSSP (2)

CV-1-1 1 99.9 48A 261 CV Upper Dome - Section 1 East I CV-1-2 1 99.9 48A 261 CV Upper Dome - Section 1 West 1 CV-1-3 1 99.6 48A 261 CV Upper Dome - Section 2 East 1 CV-1-4 1 99.6 48A 261 CV Upper Dome - Section 2 West I CV-1-5 1 99.7 48A 261 CV Upper Dome - Section 3 East 1 CV-1-6 1 99.7 48A 261 CV Upper Dome - Section 3 West 1 CV-1-7 1 72.8 48A 261 CV Upper Dome - Section 4 1 CV-1-8 1 76.7 48A 261 CV Upper Dome - Section5 1 CV-1-9 1 80.7 48A 261 CV Upper Dome - Section 6 1 CV-1-10 1 84.7 48A 261 CV Upper Dome - Section 7 1 13 The calculations performed in preparation of this report are documented in a memorandum to the PBRF Decommissioning Project File [PBRF 2011].

15

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 4, CV Survey Units for FSS SuvArea Survey Class in Unit U)(i Class in2 )

Are Surey D e sign SR # _ __ _ __ _ _ __ _ _2_

_ __ _ _ __ _ Description FSSP F _S SP__

__ (2)

CV-1-1 1 1 95.9 48A 261 CV Upper Dome - Top of Dome Section 1 CV-1-12 1 99.2 48B 324/345 CV Crane Support Steel - Section #1 1 CV-1-13 1 90.6 48B 324/345 CV Crane Support Steel - Section #2 1 CV-1-14 1 99.2 48B 324/345 CV Crane Support Steel - Section #3 1 CV-1-15 1 90.6 48B 324/345 CV Crane Support Steel - Section #4 1 CV-1-16 1 99.2 48B 324/345 CV Crane Support Steel - Section #5 1 CV-1-17 1 90.6 48B 324/345 CV Crane Support Steel - Section #6 1 CV-1-18 1 99.2 48B 324/345 CV Crane Support Steel - Section #7 1 CV-1-19 1 90.6 48B 324/345 CV Crane Support Steel - Section #8 1 CV-1-20 1 86.7 48B 324/345 CV Lower Dome - Om to 2m - Section #1 1 CV-1-21 1 98.8 48B 324/345 CV Lower Dome - Om to 2m - Section #2 1 CV-1-22 1 96.3 48B 324/345 CV Lower Dome - 2m to 6m - Section #1 1 CV-1-23 1 99.5 48B 324/345 CV Lower Dome - 2m to 6m - Section #2 1 CV-1-24 1 98.5 48B 324/345 CV Lower Dome - 2m to 6m - Section #3 1 CV-1-25 1 95.3 48B 324/345 CV Lower Dome - 2m to 6m - Section #4 1 CV-1-26 1 74.6 48B 324/348 CV 0' - Floor Section #1 - South Floor I CV-1-27 1 69.1 48C 324/348 CV 0' - Floor Section #2 - West Floor I CV-1-28 1 52.4 48C 324/348 CV 0' - Floor Section #3 - Northwest 1 CV-1-29 1 52.0 48C 324/348 CV 0' - Floor Section #4 - Northeast 1 CV-1-30 1 66.3 48C 324/348 CV 0' - Floor Section #5 - East Floor 1 CV-1-31 1 73.8 48C 324/348 CV 0' - Floor Section #6 - Lilly Pad 1 CV-1-32 1 81.9 48D 325/354 CV 0' - Polar Crane - 4 and 20 ton hoist trolleys and end beams CV-1-33 1 89.4 48D 325/3 54 CV 0' - Polar Crane - West Half of the North Rail CV-1-34 1 78.7 48D 325/354 CV 0' - Polar Crane - East Half of the North Rail CV-1-35 1 75.9 48D 325/354 CV 0' - Polar Crane - West Half of the South Rail CV-1-36 1 75.9 48D 325/354 CV 0' - Polar Crane - East Half of the South Rail CV-1-37 1 99.6 48D 325/354 CV 0' - Polar Crane - South Rail Walkway CV-3-1 1 51.9 19 97 Sub-Pile Room - Floor and Liner Floor 1 1 95.1 19 97 Sub-Pile Room - South Wall, Elevator 1 CV-3-2 walls, liner top/side, shielding plug vault CV-3-3 1 96.7 19 97/100 Sub-Pile Room - Ceiling. North Wall, Pipe Chase, Elevator Pit CV-3-4 1 30.5 45A 284 CV-Bio-shield- Section 1 lower I CV-3-5 1 87.0 45B 331 CV-Bio-shield- Section 2 upper 1 CV-3-6 1 60.5 43 316/338 Canal E - West Floor Section and 1 CV-3-7 1 58.9 43 316/338 Canal E - East Floor Section 1 16

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 4, CV Survey Units for FSS Survey Area Survey Class in Unit (1 Class (M2) Design SR # Description FSSP (2)

CV-3-8 1 94.6 43 316/338 Canal E - Wall Section #1 1 CV-3-9 1 99.1 43 316/338 Canal E - Wall Section #2 1 CV-3-10 1 85.0 43 316/338 Canal E - Wall Section #3 1 CV-3-11 1 81.5 43 316/338 Canal E - Wall Section #4 1 CV-3-12 1 99.0 43 316/338 Canal E - Wall Section #5 1 CV-3-13 1 99.9 43 316/338 Canal E - Wall Section #6 1 CV-3-14 1 55.3 43 316/338 CanalE- Ceiling I CV-3-15 1 73.0 64 317 CV -25' Reactor Annulus - West Floor 1 CV-3-16 1 59.8 64 317 CV -25' Reactor Annulus - North Floor 1 CV-3-17 1 58.0 64 317 CV -25' Reactor Annulus - East Floor I CV-3-18 1 26.8 64 317 CV -25' Reactor Annulus - Cave 1 CV-3-19 1 56.2 64 317 CV -25' Reactor Annulus - East Platform 1 CV-3-20 1 89.0 64 317 CV -25' Reactor Annulus - West Stairs I CV-3-21 1 87.7 64 317 CV -25' Reactor Annulus - East Stairs 1 CV-3-22 1 99.2 64 317 CV -25' Reactor Annulus - West End Wall & Inner Wall Section #1 CV-3-23 1 94.1 64 317 CV -25' Reactor Annulus - Inner Wall Section #2 CV-3-24 1 93.7 64 317 CV -25' Reactor Annulus - Inner Wall Section #3 CV-3-25 1 72.7 64 317 CV -25' Reactor Annulus - Inner Wall Section #4 CV-3-26 1 97.9 64 317 CV -25' Reactor Annulus - Outer Wall Section #1 CV-3-27 1 99.4 64 317 CV -25' Reactor Annulus - Outer Wall Section #2 CV-3-28 1 99.6 64 317 CV -25' Reactor Annulus - Outer Wall Section #3 CV-3-29 1 99.0 64 317 CV -25' Reactor Annulus - Outer Wall Section #4 CV-3-30 1 98.3 64 317 CV -25' Reactor Annulus - Outer Wall Section #5 CV-3-31 1 83.7 64 317 CV -25' Reactor Annulus - West Ceiling 1 CV-3-32 1 79.5 64 317 CV -25' Reactor Annulus - East Ceiling 1 CV-3-33 1 64.1 63 321/344 CV-25' Quad D- Floor 1 CV-3-34 1 86.5 63 321/344 CV -25' Quad D - Wall Section #1 (W 1 Wall & West 4m of N Wall)

CV-3-35 1 94.0 63 321/344 CV -25' Quad D - Wall Section #2 (North Wall)

CV-3-36 1 78.4 63 321/344 CV -25' Quad D - Wall Section #3 ( East 1 Wall & Rx Wall)

CV-3-37 1 75.8 63 319/344 CV -25' - Quad B Floor I CV-3-38 1 97.3 63 319/344 CV -25' - Quad B - West & North Walls 1 17

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 4, CV Survey Units for FSS Survey Area Survey SR # Description Fss in Unit () Class (M 2) Design SR#DescriptinFSSP (2)

CV-3-39 1 96.3 63 319/344 CV -25' - Quad B - East Wall & East 1 4.5m of South Wall CV-3-40 1 98.0 63 319/344 CV -25' - Quad B - West I Im of South 1 Wall CV-3-41 1 73.5 62 318/343 CV -25' - Quad A - Floor 1 CV-3-42 1 98.3 62 318/343 CV -25' - Quad A - North Wall, Reactor, I

& East 3m of South Wall CV-3-43 1 93.8 62 318/343 CV -25' - Quad A - South Wall & Canal 1 E Doorway CV-3-44 1 96.3 62 318/343 CV -25' - Quad A - West Wall 1 CV-3-45 1 72.0 62 320/343 CV -25' - Quad C - Floor 1 CV-3-46 1 97.3 62 320/343 CV -25' - Quad C - North Wall, Rx Wall, 1

& S. Wall west section CV-3-47 1 96.4 62 320/343 CV -25' - Quad C - East Wall 1 CV-3-48 1 98.0 62 320/343 CV -25' - Quad C - South Wall & Canal 1 E Doorway CV-4-1 2 177.4 58 296 CV Dome Exterior - 26'10' to 32'10" 2 CV-4-2 2 866.2 58 296 CV Dome Exterior - 32' 10" to 53'2" 2 Table 4 Notes:

1. The FSSP Table 2-1 identified 9 survey areas in the CV. For the FSS, the CV was divided into 87 survey units to meet FSS Plan classification-based size limits.

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

Table 5, CV Survey Unit Breakdown by Major Elevation Major No. of Surface  % of Survey  % of Surface Elevation Survey Units Area (M2) Units Area 0 ft. Elev. 37 3233.1 42.5 39.2

-25 ft. Elev. 48 3978.6 55.2 48.2 Dome Eote 2 1043.6 2.3 12.6 Exterior Total 87 8255.3 100 100 4.2 Number of Measurements 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 to be used are equivalent to a 18

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 small fraction of the applicable DCGLw. 14 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 Containment Vessel. 15 When the Sign Test is selected, the VSP software uses MARSSIM Equation 5-2 to calculate the number of measurements. Equation 5.2 is shown below:

N =.2 (ZI-a + ZI-quation 1) 4[(D( -0.5 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, 13= the type II error probability, ZI. = 100(l-a) percentile of the normal distribution (1.6449 for cr = 0.05),

Z1-p = 100(1-13) percentile of the normal distribution (1.2816 for 13= 0.1),

(D(A/a) = value of cumulative standard normal distribution over the interval

-oo, 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 a. The numbers of measurements were calculated for the CV survey units using the parameters established in 12 survey designs. Table 6 summarizes the CV survey designs and lists the values of the key VSP input parameters.

14 Background count rates for the LMII 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%).

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

19

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 6, CV Survey Design Summary Table 1

1. The data reported in this table is obtained from the Survey Design reports listed. They are maintained in the PBRF Document Control System.

2. Units are dpm/100-cm 2.

3. The DCGLw in this design is the gross activity value from TBD-07-001 [PBRF 2007].

4. The gross activity DCGL in TBD-07-001 (31,711 dpm/100-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides and 1 mrem/y from embedded piping (EP).

5. The gross activity DCGL in TBD-07-001 (11,563 dpm/100-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides and 2 mrem/y from EP. The slight differences in DCGLw and LBGR values between Designs 45A and 45B are due to round-off errors.

6. The gross activity DCGL in TBD-07-001 (11,563 dpm/l00-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides.

7. The default gross activity DCGL in TBD-07-001 (27,166 dpm/100-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides.

8. The gross activity DCGL value in TBD-07-001 (13,450 dpm/1 00-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides and 1 mrem/y from EP.

9. The gross activity DCGL in TBD-07-001 (21,566 dpm/100-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides and 1.37 mrem/y from EP.

10. The gross activity DCGL in TBD-07-001 (37,235 dpm/100-cm 2) was adjusted to account for 2.5 mrem/y contribution from "deselected" radionuclides and 1 mrem/y from EP.

20

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 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 6 that relative shift values ranging from 2.36 to 3.0 were used in the calculations for determining N in the CV survey designs.

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 report for survey unit CV-1-20 (Design 48B). 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 7 summarizes this analysis. The region of interest is for a = 0.05 (required to be fixed), 03= 0.10 and the LBGR at 70% to 90%

of the DCGL. With the LBGR at 90% of the DCGL, doubling a causes a moderate increase in N (from 11 to 14). With the LBGR at 70 to 80% of the DCGL, doubling a causes no increase in N (it remains at 11). 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 a constant at 325 dpm/100-cm 2 (and 13constant at 0.1), shows that the number of measurements is 11 or fewer in all cases. These results indicate that N = 11 represents a conservative design.

Table 7, Sensitivity Analysis for CV FSS Design Number of Samples DCGL= 10,407 a 7=0.05 (2) a=0.10 a=0.15 a=650 ( c)(3)

= 325 a = 650 a = 325 a = 650 a = 325 LBGR=90% (1)(4) 13=0.05 17 14 14 11 12 10 13=0.10 14 11 11 9 9 8 13=0.15 12 10 9 8 8 6 LBGR=80% 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 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 7 Notes:

1. Units of DCGL, a and LBGR are dpm/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 Containment Vessel survey unit.

The survey designs also contain instructions to ensure that scan survey coverage and action levels meet the FSS Plan requirements based on the MARSSIM classification listed in Table

3. 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 21

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 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 Containment Vessel 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 to be used in the FSS of each survey unit are selected in each survey design.

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 (alpha and beta) measurements are calculated using equations from Section 6.4 of the FSS Plan [NASA 2007]. The static MDC equation is:

3 + 3.29 BRtS(1+ t)

M/DCstatic -A tb (Equation 2) tsEtot 100 1--

where:

MDCsQC = 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 Et., = 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:

MDC'c" = EiEs..-§ A , (Equation 3) 100 where:

MDCS.. = Minimum Detectable Concentration (dpm/I00-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, Ei = Detector Efficiency (counts per disintegration),

Es = Surface Efficiency, typically 25% for alpha and 50% for beta per ISO 7503-1, Table 2 [ISO 1988],

22

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 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 CV is provided in Table 8.

Table 8, Typical Detection Sensitivities of Field Instruments Detector MDCSscan Net cpm MDCfiIc Detector Model Efficiency (dpm/1 00-cm 2) Equivalent to (dpm/1 00-cm 2)

(c/d) (1) (2) DCGLw (2)

LMI 44-116 3( )(4) 0.140 2,992 1108 655 LMI 44-9 5( )(6) 0.139 10,169 158 3,310 Table 8 Notes:

1. The detector efficiencies listed are total efficiency, i. e., E, = 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. 45A 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. 45A 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. 45A. 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. 45A. 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 3 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. This practice was established early in the FSS of PBRF 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. 16 Modifications to survey instructions are adjusted to account for unusual measurement conditions. Modified detection sensitivities may be applied taking into account adjustments in 16 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."

23

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 detector efficiency. Scan speeds may be reduced to ensure that required scan sensitivities are achieved. The bases for adjustments due to non-standard conditions are provided in PBRF Technical Basis Documents.1 7 Examples of areas or locations in CV survey units where special measurement conditions apply are shown in Exhibits 27 through 31 of Appendix A.

5.0 CV Survey Results Results of the CV FSS are presented in this section. This includes scan survey frequencies (% of areas covered) for each survey unit and occurrence of events where scan investigation levels were exceeded. 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 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 Containment Vessel 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 CV scan surveys are compiled in Table 9. The table shows that scan coverage requirements were satisfied for all survey units. The table also shows that investigations were performed in four survey units. These are discussed in Section 5.3 below.

Table 9, Scan Survey Results Scan Survey Survey Investigation QC Replicate Scan Survey Unit Class Coverage (%) (1) Request Performed Coverage (%) (2)(3)

No.

CV-1-1 1 100 261 No 5.0%

CV-1-2 1 100 261 No 5.0%

CV-1-3 1 100 261 No 6.8%

CV- 1-4 1 100 261 No 5.5%

CV-1-5 1 100 261 No 6.3%

CV-1-6 1 100 261 No 6.3%

CV-1-7 1 100 261 No 7.6%

CV-1-8 1 100 261 No 6.6%

CV-1-9 1 100 261 No 6.7%

CV-1-10 1 100 261 No 6.1%

CV-1-11 1 100 261 No 5.6%

CV-1-12 1 100 324/345 No 6.6%

CV-1-13 1 100 324/345 No 6.6%

'7 The PBRF-TBD-07-004 [PBRF 2007a] presents efficiency correction factors developed for the LM1 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.

24

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 9, Scan Survey Results Scan Survey Survey Investigation QC Replicate Scan Survey Unit Class Coverage (%) l Request Performed Coverage (%) 9)(3)

No.

CV-1-14 1 100 324/345 No 6.6%

CV-1-15 1 100 324/345 No 6.6%

CV-1-16 1 100 324/345 No 6.6%

CV-1-17 1 100 324/345 Yes 6.6%

CV-1-18 1 100 324/345 No 6.6%

CV-I-19 1 100 324/345 No 6.6%

CV-1-20 1 100 324/345 No 6.6%

CV-1-21 1 100 324/345 No 6.6%

CV-1-22 1 100 324/345 No 6.6%

CV-1-23 1 100 324/345 No 6.6%

CV- 1-24 1 100 324/345 No 6.6%

CV-1-25 1 100 324/345 No 6.6%

CV-1-26 1 100 324/348 No 6.6%

CV-1-27 1 100 324/348 No 6.6%

CV-1-28 1 100 324/348 No 6.6%

CV-1-29 1 100 324/348 No 6.6%

CV-1-30 1 100 324/348 No 6.6%

CV-1-31 1 100 324/348 Yes 6.6%

CV-1-32 1 100 325/354 No 6.4%

CV-1-33 1 100 325/354 No 6.4%

CV-1-34 1 100 325/354 No 6.4%

CV-1-35 1 100 325/354 No 6.4%

CV-1-36 1 100 325/354 No 6.4%

CV-1-37 1 100 325/354 No 6.4%

CV-3-1 1 100 97 No 6.2%

CV-3-2 1 100 97 No 6.2%

CV-3-3 1 100 97/100 No 6.2%

CV-3-4 1 100 284 No 6.6%

CV-3-5 1 100 331 No 8.5%

CV-3-6 1 100 316/338 No 8.1%

CV-3-7 1 100 316/338 No 8.1%

CV-3-8 1 100 316/338 No 8.1%

CV-3-9 1 100 316/338 No 8.1%

CV-3-10 1 100 316/338 No 8.1%

CV-3-11 1 100 316/338 No 8.1%

CV-3-12 1 100 316/338 No 8.1%

CV-3-13 1 100 316/338 No 8.1%

CV-3-14 1 100 316/338 No 8.1%

CV-3-15 1 100 317/349 No 6.8%

CV-3-16 1 100 317/349 No 6.8%

CV-3-17 1 100 317/349 No 6.8%

CV-3-18 1 100 317/349 No 6.8%

CV-3-19 1 100 317/349 No 6.8%

CV-3-20 1 100 317/349 No 6.8%

CV-3-21 1 100 317/349 No 6.8%

25

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 9, Scan Survey Results UitClss Suve Scan Survey Survey Investigation QC Replicate Scan SreUntCasCoverage (%) () Request Performed Coverage(%()()

No. ________________

CV-3-22 1 100 3 17/349 No 6.8%

CV-3-23 1 100 3 17/349 No 6.8%

CV-3-24 1 100 317/349 No 6.8%

CV-3-25 1 100 317/349 No 6.8%

CV-3-26 1 100 3 17/349 No 6.8%

CV-3-27 1 100 317/349 No 6.8%

CV-3-28 1 100 317/349 No 6.8%

CV-3-29 1 100 317/349 No 6.8%

CV-3-30 1 100 3 17/349 No 6.8%

CV-3-31 1 100 317/349 No 6.8%

CV-3-32 1 100 3 17/349 No 6.8%

CV-3-33 1 100 321/344 No 6.6%

CV-3-34 1 100 321/344 Yes 6.3%

CV-3-35 1_ 100 321/344 No 6.3%

CV-3-36 1 100 321/344 No 6.3%

CV-3-37 1 100 3 19/344 No 5.4%

CV-3-38 1 100 319/344 No 5.4%

CV-3-39 1 100 319/344 No 5.4%

CV-3-40 1 100 319/344 No 5.4%

CV-3-41 1 100 318/343 No 8.7%

CV-3-42 1 100 318/343 No 8.7%

CV-3-43 1 100 318/343 No 8.7%

CV-3-44 1 100 318/343 No 8.7%

CV-3-45 1 100 320/343 No 5.7%

CV-3-46 1 100 320/343 No 5.7%

CV-3-47 1 100 320/343 No 5.7%

CV-3-48 1 100 320/343 Yes 5.7%

CV-4-1 2 54 296 No 12.3%

CV-4-2 12 40 1 296 1 No 12.3%

Table 9 Notes:

1. One hundred % of the accessible surface area in Class I survey units was scanned. A fraction of the surface area of a few survey units was inaccessible for scanning. In most such survey units, it was less than a few %of the total surface area and in all cases less than 10%.

2. The %scan coverage is given as the % of the area scanned in the initial survey. Percent QC scan coverage values are rounded to the nearest whole per cent. Values reported with the first decimal, the tenths value, as 5, e. g., 5.5, are rounded downward.

3. Replicate QC scan results are reported for multiple survey units in some 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.

26

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 5.2 Systematic Measurements and Tests Results of the assessment of CV FSS total surface beta measurements are presented in Table 10 (individual measurement results for 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 10 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. The average of 967 total surface beta measurements reported in the CV release records is: 413 -

348 dpm/100-cm 2 (one standard deviation) [PBRF 20111.18 Removable surface activity measurements were also performed at each fixed activity measurement location and counted for gross alpha and gross beta activity. A review of the CV release records was conducted to ensure that all smear counting results were less than 10% of the gross activity DCGL. The requirement for PBRF laboratory smear counting instruments is that the MDAs be < 10% of the applicable gross activity DCGL 9 . Gross beta and gross alpha counts for all smears were all less than 10% of the DCGL for each CV survey unit; over 99 %

were less than the MDA of the counting instruments.

Table 10, CV Total Surface Beta Activity Measurement Summary and Test Results Survey No. of Maximum Test Result: Averae Standard Test Result:

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

Unit ID Unit ID DCGLw (1)(2) DCGLw CV-1-1 11 350 Yes 204 107 Yes CV-1-2 11 500 Yes 236 163 Yes CV-1-3 11 457 Yes 256 128 Yes CV- 1-4 11 419 Yes 282 96 Yes CV-1-5 11 557 Yes 227 181 Yes CV-1-6 11 450 Yes 188 137 Yes CV-1-7 11 481 Yes 232 143 Yes CV-1-8 11 329 Yes 195 119 Yes CV- 1-9 11 406 Yes 261 108 Yes CV-1-10 11 418 Yes 213 152 Yes CV-1-11 11 448 Yes 275 133 Yes CV-1-12 11 340 Yes 188 96 Yes CV-1-13 11 333 Yes 125 131 Yes CV-1-14 11 476 Yes 176 153 Yes CV-1-15 11 279 Yes 151 106 Yes CV-1-16 11 932 Yes 376 329 Yes CV-I-17 11 265 Yes 76 120 Yes 18 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.

19 Typical MDAs for PBRF low background smear counting instruments range from 8 to 14 dpm for alpha and 10 to 20 dpm for beta. Smears cover 100 cm 2, so these MDA values are equivalent to dpm/100-cm 2.

27

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 10, CV Total Surface Beta Activity Measurement Summary and Test Results Suvy N.of Mxmm Test .Result: Aeae Standard Test Result:

SUrvey No. of) Maximum Avrg Deviation Average <

UiID Measurements (1)~ Maximum <C~

CV-1-18 11 429 Yes 224 130 Yes CV-1-19 1 11 _ __ 429 Yes 128 133 Yes CV-1-20 11 _ __ 361 Yes 89 163 Yes CV-1-21 11 510 Yes 223 153 Yes CV-1-22 12 415 Yes 197 87 -Yes CV-1-23 11___ 456 Yes 186 133 Yes CV-1-24 12 449 Yes 182 146 Yes CV-1-25 11 _ __ 361 Yes 174 140 Yes CV-1-26 11 935 Yes 478 317 Yes CV-1-27 11 797 Yes 430 275 Yes CV-1-28 11 1021 Yes 614 246 Yes CV-1-29 11 867 Yes 484 348 Yes CV-1-30 11 784 Yes 466 211 Yes CV-1-31 11 1008 Yes 586 324 Yes CV-1-32 11 608 Yes 296 204 Yes CV-1-33 11 385 Yes 149 118 Yes CV-1-34 11 405 Yes 178 133 Yes CV-1-35 11 419 Yes 175 146 Yes CV-1-36 11 514 Yes 217 163 Yes CV-1-37 11 507 Yes 195 117 Yes CV-3-1 11 1298 Yes 671 338 Yes CV-3-2 11 1715 Yes 951 317 Yes CV-3-3 11 1167 Yes 757 276 Yes CV-3-4 11 727 Yes 441 209 Yes CV-3-5 12 2240 Yes 502 583 Yes CV-3-6 11 1020 Yes 708 203 Yes CV-3-7 11 780 Yes 505 217 Yes CV-3-8 11 2620 Yes 469 812 Yes CV-3-9 12 1250 Yes 626 327 Yes CV-3-10 11 644 Yes 315 240 Yes CV-3-11 11 1050 Yes 625 245 Yes CV-3-12 11 1190 Yes 678 302 Yes CV-3-13 11 992 Yes 720 186 Yes CV-3-14 I11 1100 Yes 634 224 Yes CV-3-15 11 1020 Yes 728 161 Yes CV-3-16 11 836 Yes 526 190 Yes CV-3-17 11 829 Yes 545 199 Yes CV-3-18 12 1743 Yes 718 404 Yes CV-3-19 11 2757 Yes 529 840 Yes CV-3-20 11 750 Yes 124 295 Yes CV-3-21 11 1078 Yes 346 341 Yes CV-3-22 11 608 Yes 318 245 Yes CV-3-23 11 662 Yes 388 173 Yes CV-3-24 I11___ 1987 Yes 780 535 Yes CV-3-25 I11 816 Yes 547 190 Yes 28

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 10, CV Total Surface Beta Activity Measurement Summary and Test Results No. of Maximum Test Result: Standard Test Result:

Survey (1) Deviation Average <

Unit ID Measurements (1) DCGLw <

Maximum (1)(2) DCGLw CV-3-26 11 676 Yes 391 234 Yes CV-3-27 11 941 Yes 487 197 Yes CV-3-28 11 914 Yes 516 203 Yes CV-3-29 11 704 Yes 522 122 Yes CV-3-30 11 796 Yes 523 214 Yes CV-3-31 11 858 Yes 484 154 Yes CV-3-32 11 615 Yes 440 121 Yes CV-3-33 11 894 Yes 623 190 Yes CV-3-34 11 1007 Yes 513 299 Yes CV-3-35 11 966 Yes 599 192 Yes CV-3-36 11 966 Yes 506 340 Yes CV-3-37 11 1240 Yes 636 255 Yes CV-3-38 11 474 Yes 219 205 Yes CV-3-39 11 725 Yes 438 208 Yes CV-3-40 11 786 Yes 438 203 Yes CV-3-41 11 1020 Yes 617 293 Yes CV-3-42 11 738 Yes 461 217 Yes CV-3-43 11 3180 Yes 655 864 Yes CV-3-44 11 691 Yes 440 156 Yes CV-3-45 11 922 Yes 654 163 Yes CV-3-46 11 3080 Yes 900 799 Yes CV-3-47 11 1070 Yes 600 266 Yes CV-3-48 11 1860 Yes 665 475 Yes CV-4-1 11 196 Yes -6 146 Yes CV-4-2 16 962 Yes 195 290 Yes Table 10 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 all the Reactor Containment Vessel structural survey units, values of a obtained from the FSS measurements are typically much less than values used in the survey designs (see Table 6). This confirms that the survey designs for the CV were 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.

5.3 Investigative Measurements and Tests Additional measurements were performed as a result of investigations initiated during scan surveys of the CV survey units. The investigations and the results of the investigative static measurements are summarized in Table 11 (investigative measurements are identified with the prefix "IM" or "EMA" followed by a number). It is noted that most of the investigations were initiated when technicians observed increased count rates during scans as opposed to observing count rates above action levels (scan survey action levels in survey instructions are discussed in Section 4.0). In one survey unit, CV- 1-17, an investigative static measurement exceed the DCGLw.

29

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 11, CV Investigative Static Measurements Static Size of Survey SR Measurement Elevated Measurement DCGLw (1)

Unit (1) Area (cm 2 ) ID CV-1-17 324 13,800 (2) 100 EMA-1 10,407 CV-1-31 324 6,788 100 IM-2 10,407 CV-1-31 324 8,061 100 IM-3 10,407 CV-3-34 321 4,779 (3) 12 IM-1 18,228 CV-3-48 320 7,990 100 IM-1 11,567 Table 11 Notes:

1. Units are dpm/100-cm 2.

2. Elevated measurement comparison (EMC) and elevated measurement test (EMT) performed.

3. Measurement performed with LMI-44-9 detector (GM Pancake probe).

As a result of the investigative static measurement in Survey Unit CV- I-17, which exceeded the DCGLw, the elevated measurement comparison (EMC) and elevated measurement test (EMT) were performed. 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 unity rule equation:

++(average concentration in elevated area)-6 1

_<1.0 [Equation 4]

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 4 is calculated for each and summed with the first term to perform the unity rule calculation for the EMT. Results of the DCGLEMc and EMT calculations are summarized:

• From Table 11, the size of the elevated area is 100 cm 2 ; per the FSS Plan Table 3-5, the area factor for elevated areas of size < 0.25 m is 40.2.

• From Table 10, the average residual activity concentration, 8, in Survey Unit CV-I-17 is 76 dpm/100-cm 2.

0 From Table 6, the applicable DCGLw is 10,407 dpm/100-cm 2.

• The DCGLEMC is calculated to be 418,361 dpm/I00-cm 2 (40.2 x 10,407 dpm/1 00-cm2 ); this is much greater than 13,800 dpm/I00-cm 2 , the measured activity in the elevated area. Thus the elevated measurement comparison is easily satisfied.

0 Calculating the unity value using Equation 4 yields:

76 + (13,800 - 76) - 0 04 which is much less than 1.0.

10,407 (40.2)(10,407)-

As a result of elevated contamination levels measured during the initial FSS scan survey of survey unit CV-3-3, Sub Pile Room and Elevator Pit, the survey unit was failed. After 30

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 additional remediation the survey unit was subjected to a "new" FSS which showed that the survey unit satisfied the FSS Plan requirements for release for unrestricted use. All systematic measurements were below the DCGLw and no scans were observed above investigation-action levels.

5.4 QC Measurements Per FSS Plan requirements, QC replicate measurements were taken for at least 5% of the CV measurements. This included scan surveys and systematic total surface activity measurements. Scan QC survey results are shown in Table 9 wherein the 5% scan QC coverage is confirmed. These surveys confirmed the results of the original scan surveys of the areas covered. No QC scan surveys identified areas of elevated activity.

Replicate total surface activity measurements were performed at selected systematic measurement locations. The 5 % requirement is satisfied in that 54 QC measurements were reported; this represents 5.6 % of the 967 systematic measurements performed in the FSS of the CV. The results for the 54 surface activity original and QC replicate measurement data pairs are shown in Table 12.

Table 12, CV Total Surface Activity QC Measurements Measurement Net Activity QC Replicate Survey Unt Location No. (dpr/m100 cm 2) (dpm/1 00 cm 2) RPD (%)

CV-1-37 5 74 316 124.1 CV-4-1 8 98 172 54.8 CV-1-4 11 111 93 17.6 CV-1-20 1 116 218 61.1 CV- 1-21 11 184 120 42.1 CV-1-36 11 196 0 200.0 CV-1-37 3 196 -99 608.2 CV-3-47 6 210 516 84.3 CV-1-23 1 211 408 63.7 CV-1-20 5 218 7 187.6 CV-1-37 10 223 164 30.5 CV-3-8 1 280 -61 311.4 CV-1-10 3 304 413 30.4 CV-1-8 8 329 297 10.2 CV-1-7 2 335 266 23.0 CV-1-21 5 347 183 61.9 CV-1-20 11 361 289 22.2 CV-1-9 2 406 44 160.9 CV-3-41 1 412 503 19.9 CV-3-41 3 440 591 29.3 CV-1-11 7 448 479 6.7 CV-1-23 5 456 239 62.4 31

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 Table 12, CV Total Surface Activity QC Measurements Survey Unit Measurement Net Activity QC Replicate RPD Location No. (dpm/100 cm2) (dpm/100 cm2)

CV-1-3 3 457 504 9.8 CV-3-10 1 477 578 19.1 CV-3-8 2 492 1043 71.8 CV-1-21 1 510 296 53.1 CV-3-37 5 515 924 56.8 CV-3-37 3 528 284 60.1 CV-1-28 3 531 882 49.7 CV-1-28 7 559 477 15.8 CV-1-28 1 601 608 1.2 CV-3-35 11 608 694 13.2 CV-3-15 6 612 811 28.0 CV-3-47 1 615 707 13.9 CV-3-18 5 618 649 4.9 CV-3-35 1 635 211 100.2 CV-3-18 4 711 304 80.2 CV-3-4 1 727 972 28.8 CV-3-6 10 773 589 27.0 CV-3-41 5 776 598 25.9 CV-3-35 6 784 728 7.4 CV-3-15 1 803 764 5.0 CV-3-15 4 836 547 41.8 CV-3-6 11 848 576 38.2 CV-3-18 8 888 1061 17.8 CV-3-15 8 921 1054 13.5 CV-4-2 1 962 622 42.9 CV-3-15 5 1020 1101 7.6 CV-1-28 5 1021 542 61.3 CV-3-18 3 1138 1054 7.7 CV-3-1 4 1298 993 26.6 CV-3-2 1 1715 1460 16.1 CV-3-18 6 1743 209 157.2 CV-3-5 12 2240 2067 8.0 Table 12 Notes:

1. Table data is sorted by increasing net activity.

2. The relative percent difference (RPD) is calculated as: absolute value [(original -

QC)/{(original + QC)/2}].

The FSS Plan (Section 12.7) specifies that the RPD between original and replicate measurements be within 20% [NASA 2007]. Table 12 shows that 35 of the 54 measurement 32

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 pairs are above the 20% RPD criterion. Each measurement pair failing to meet the criterion was individually investigated and resolved in accordance with FSS Plan requirements and implementing procedures. 20 It is noted that the measurements constituting the 54 QC replicate measurement pairs are of very low activity (all < 2250 dpmr/100-cm 2). Experience and the theory of measurement errors have shown that low activity measurements such as these are subject to variation which is high relative to the measured activity. In this situation the 20% RPD criterion is often not achieved.

5.5 ALARA Evaluation It is shown that residual contamination in the Reactor Containment Vessel 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 need to conduct an explicit analysis to meet the ALARA requirement."21 Screening level values published by the NRC for the mix of radionuclides in structural surface residual contamination potentially present in the CV are shown in Table 13. 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 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. 22 The activity fractions listed in Table 2 for the CV Dome (also shown in Table 13) were used in the calculation. The resultant screening level equivalent DCGL is calculated to be 1188 dpm/I00-cm2 .

The average (mean) total surface beta activity measured in the FSS of the CV is 413 + 348 dpm/I100-cm2 (one standard deviation). The upper limit of the 95th % confidence interval of 20 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 PBRFLicense Termination).

21 This guidance was initially published in Draft Regulatory Guide DG-4006, but has been reissued in NUREG-1757 Volume 2, Appendix N.

22 The equivalent screening level gross activity DCGL is calculated using an EXCEL template [PBRF 2011]. This template incorporates the equations in section 5.3 of the FSS Plan [NASA 2007]. Screening level equivalent DCGLs were calculated for the various radionuclide mixtures shown in Table 2, and 1188 dpm/I00-cm 2 is the smallest value obtained.

33

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 the mean is 435 dpm/I00-cm2 .23 This value is well below the screening level gross activity DCGL of 1188 dpm/100-cm 2. From this comparison, it is concluded that the ALARA criterion is satisfied.

Table 13, Screening Level Values for CV and Radionuclide Activity Fractions Screening Level Value Activity Fraction Radionuclide (dpm/100-cm2) (%) (1)

H-3 1.2 E+08 27.07 Co-60 7.1E+03 (2) 9.65 Sr-90 8.7E+03 (2) 7.88 1-129 3.5E+04(2) 1.42 Cs-137 2.8E+04 (2) 46.71 Eu-154 1.2E+0413) 0.12 U-234 9.IE+01 (3) 6.98 U-235 9.8E+01c3) 0.17 Table 13 Notes.

1. Activity fractions used to develop the DCGLw for the CV Dome per Design 58 (radionuclide mixture from TBD-07-001, also see Table 2 of this report).

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 90'h percentile values of residual surface activity corresponding to 25 mrem/y to a future building occupant.

5.6 Comparison with EPA Trigger Levels The PBRF license termination process includes a review of residual contamination levels in groundwater and soil, 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.

However, no soil survey units are included in the FSS of the CV. It is also noted that groundwater is not within the scope of the CV FSS. Therefore, comparison with EPA Trigger Levels is not applicable to CV FSS measurement results.

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

967 measurements.

34

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 5.7 Conclusions The results presented above demonstrate that the Reactor Containment Vessel satisfies all FSS Plan commitments and meets the release criteria in 10CFR20 Subpart E. The principal conclusions are:

" Scan surveys were performed in 100 % of the accessible surfaces of all 85 Class 1 Reactor Containment Vessel survey units. The two Class 2 survey units received scan coverage of 40% (CV Upper Dome Exterior) and 54% (CV Lower Dome Exterior).

" Residual surface contamination levels requiring investigation were observed in four survey units. In one of the survey units, activity above the DCGLw was measured in a small localized area; the EMC and EMT were successful.

" All fixed total surface activity measurements (established on systematic grids with random start) are less than the applicable DCGLw.

" The mean values of all survey unit systematic total surface beta activity measurements are below the DCGLw, hence no statistical tests were required.

" All removable surface activity measurements are less than 10% of the DCGLw.

" One survey unit was "failed" in the initial FSS. It was remediated, a new survey design and instructions prepared and resurveyed. The FSS resurvey was successful.

• 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 Containment Vessel was divided into 87 survey units, whereas the FSS Plan had identified 9 survey areas, not divided into survey units.

• There were no changes from initial assumptions (in the FSS Plan) regarding the extent of residual activity in the Reactor Containment Vessel. No reclassification of survey units was required as a result of FSS measurements and investigations.

6.0 References GTS 1998 GTS-Duratek, Inc., NASA/Lewis Plum Brook Facility 1998 ConfirmationSurvey, Volume I - Survey Packages and Survey Results, November 1998.

ISO 1988 International Organization for Standardization, Evaluationof Surface Contamination,Part 1: Beta Emitters andAlpha Emitters, ISO-7503-1, 1988.

NACA 1956 National Advisory Committee for Aeronautics, Lewis Flight Propulsion Laboratory, NACA Reactor FacilityHazardsSummary, October 1956.

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

35

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 NASA 2007a NASA Safety and Mission Assurance Directorate, DecommissioningPlanfor the Plum Brook Reactor Facility,Revision 6, July 2007.

PBRF 2005 Plum Brook Reactor Facility, Survey Request SR-4, CV Quads and Canals, (preliminaryscans to identify potential core 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 Containment Vessel (1112),

December 10, 2009.

PBRF 2011 Plum Brook Reactor Facility Decommissioning Project Office, Memorandum to Project File, EngineeringRecordfor Final Status Survey Report Calculation- CV Update. Sept. 14, 2011.

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 Evaluation of the Plum Brook ReactorFacility and Documentation of Existing Conditions,Prepared for NASA Lewis Research Center, December 1987.

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

USNRC 2000 US Nuclear Regulatory Commission, Multi-Agency RadiationSurvey andSite InvestigationManual (MARSSIM), NUREG-1575, Rev.1, August 2000.

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

36

Plum Brook Reactor Facility FSSR Attachment 11, Rev. 0 7.0 Appendices Appendix A - Exhibits Appendix B - Survey Unit Maps and Tables Showing Measurement Locations and Results 37

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 1 of 24 Final Status Survey Report Attachment 11 Reactor Containment Vessel Revision 0 Appendix A Exhibits

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 2 of 24 List of Exhibits Exhibit 1, Outdoor View of the Reactor Building and CV in late 2010 .......................................... 3 Exhibit 2, Construction Era Photo of Reactor Tank & Bioshield top (Lily Pad) with Quad A in Foreground ........................................................................................................................................... 4 Exhibit 3, Plan View ov Elevation 0 ft. Showing Annular Arrangement within CV ..................... 5 Exhibit 4, Construction Photo Showing Above Grade Portion of CV Outer Shell ......................... 6 Exhibit 5, Construction Photo Showing Installation of CV Bottom Liner ..................................... 7 Exhibit 6, E-W Section View Showing CV and Reactor Building Arrangement ........................... 8 Exhibit 7, N-S Section View Showing CV and Reactor Building Arrangement ............................ 9 Exhibit 8, Schematic Showing PCW Piping Layout Between CV and PPH ................................ 10 Exhibit 9, View of Defueled Reactor circa 1973 Showing Reactor Tank Internals .......................... 11 Exhibit 10, Segmentation of Reactor Tank Internals in Progress - 2005 ........................................... 11 Exhibit 11, Inside Bioshield after Reactor Tank Removal ............................................................ 12 Exhibit 12, External View of Bioshield From Quad B Showing Thermal Column Cutout .......... 12 Exhibit 13, Subpile Room Ceiling -2010 ....................................................................................... 13 Exhibit 14, Subpile Room Floor Prepared for FSS ...................................................................... 13 Exhibit 15, Quad A in 2000 Prior to D&D .................................................................................. 14 Exhibit 16, Quad D Floor Showing Extensive Remediation to Prepare for FSS .......................... 14 Exhibit 17, Quad D Facing Bioshield circa 2010 - Ready for FSS ............................................... 15 Exhibit 18, Canal E circa 2000 .................................................................................................... 16 Exhibit 19, Canal E South Wall Showing Lift Door Opening to Canal F ..................................... 17 Exhibit 20, Canal E Ceiling with View into CV Dome Above ..................................................... 17 Exhibit 21, Canal E Floor and Lower Wall Showing Extensive Remediation .............................. 18 Exhibit 22, Dry Annulus Experimental Area Prepared for FSS ................................................... 19 Exhibit 23, CV 0 ft. Elevation in 2000 ......................................................................................... 20 Exhibit 24, CV 0 ft. El Showing Dome Walls in Preparation for FSS .......................................... 20 Exhibit 25, CV Dome Interior Overhead in 2000 Showing Polar Crane and Acoustic Baffles ........ 21 Exhibit 26, CV Dom e Interior - 2011 ......................................... ;................................................ 21 Exhibit 27, Example of Remediated Concrete - Quad B Wall Penetration ................................... 22 Exhibit 28, Extensive Remediation Around Grouted Pipes - Quad B SW Comer ....................... 22 Exhibit 29, Lift Door Channel in Quad A Extensively Remediated by Flame Cuting ................. 23 Exhibit 30, STMA on Extensively Remediated Concrete Floor in Quad B Floor ....................... 24 Exhibit 31, UCM at Heavily Remediated Unistrut in Quad D North Wall .................................. 24

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 3 of 24 Exhibit 1, Outdoor View of the Reactor Building and CV in late 2010

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 4 of 24 Exhibit 2, Construction Era Photo of Reactor Tank & Bioshield top (Lily Pad) with Quad A in Foreground

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 5 of 24 Exhibit 3, Plan View ov Elevation 0 ft. Showing Annular Arrangement within CV CONTROL ROOM E7lIR9j1j 5

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 6 of 24 Exhibit 4, Construction Photo Showing Above Grade Portion of CV Outer Shell

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 7 of 24 Exhibit 5, Construction Photo Showing Installation of CV Bottom Liner

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 8 of 24 Exhibit 6, E-W Section View Showing CV and Reactor Building Arrangement CONTAINMENT VESSEL REACTOR BUILDING REACTOR BUILDING OFFICE AREA 11 SHOP .0 IL BASEMENT OUAD_'A I REACTOR QUAD "C TEST AREA

'o"- - :TANK .. -_,,-"-_..ANKT

-25*-0"

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 9 of 24 Exhibit 7, N-S Section View Showing CV and Reactor Building Arrangement CONTAINMENT VESSEL REACTOR BUILDING REACTOR BUILDING QUAD 'D"

<iý

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 10 of 24 Exhibit 8, Schematic Showing PCW Piping Layout Between CV and PPH 4 PPH ROOM 4 Ur HEAT EXCHANGERS 0

REACTOR

-6'-8"

-12'-2"

-39,-6"

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 11 of 24 Exhibit 9, View of Defueled Reactor circa 1973 Showing Reactor Tank Internals Exhibit 10, Segmentation of Reactor Tank Internals in Progress - 2005

Plum Brook Reactor Facility FSSR, Attachment II Appendix A, Rev. 0, Page 12 of 24 lxhihit 11. Inmide Rinthield after Reaetar Tank Remnval Exhibit 12, External View of Bioshield From Quad B Showing Thermal Column Cutout

Plum Brook Reactor Facility FSSR, Attachment I1 Appendix A, Rev. 0, Page 13 of 24

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 14 of 24 Exhibit 16, Quad D Floor Showing Extensive Remediation to Prepare for FSS

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 15 of 24 Exhibit 17, Quad D Facing Bioshield circa 2010 - Ready for FSS

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 16 of 24 Exhibit 18, Canal E circa 2000

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 17 of 24 Exhibit 19. Canal E South Wall Showi in Lift Door ODenine to Canal F Exhibit 20, Canal E Ceiling with View into CV Dome Above

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 18 of 24 Exhibit 21, Canal E Floor and Lower Wall Showing Extensive Remediation

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 19 of 24 Exhibit 22, Dry Annulus Experimental Area Prepared for FSS

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 20 of 24 Exhibit 24, CV 0 ft. El Showing Dome Walls in Preparation for FSS

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 21 of 24 Exhibit 25, CV Dome Interior Overhead in 2000 Showing Polar Crane and Acoustic Baffles Exhibit 26. CV Dome Interior - 2011

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 22 of 24 Exhibit 27. Examnle of Remediated Concrete - Ouad B Wail Penetration Exhibit 28, Extensive Remediation Around Grouted Pipes - Quad B SW Corner

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 23 of 24 Exhibit 29, Lift Door Channel in Quad A Extensively Remediated by Flame Cuting

Plum Brook Reactor Facility FSSR, Attachment 11 Appendix A, Rev. 0, Page 24 of 24 Exhibit 30, STMA on Extensively Remediated Concrete Floor in Quad B Floor Exhibit 31, UCM at Heavily Remediated Unistrut in Quad D North Wall