ML061390295

Final Status Survey Plan for the Plum Brook Reactor Facility, Revision 0, May 2006, Book 1

ML061390295
Person / Time
Site:	Plum Brook
Issue date:	05/12/2006
From:	US National Aeronautics & Space Admin (NASA), John H. Glenn Research Ctr at Lewis Field
To:	Document Control Desk, NRC/FSME
References
Download: ML061390295 (79)

Text

MISSION AND SAFETY SAFETY AND MISSION ASSURANCE DIRECTORATE Plum Brook Reactor Facility BOOK 1

SAFETY AND MISSION ASSURANCE DIRECTORATE Plum Brook Reactor Facility Final Status Survey Plan for the Plum Brook Reactor Facility (Rev 0)

May 2006 BOOK 1

Document

Title:

Final Status Survey Plan for the Plum Brook Reactor Facility Document Number: N/A Revision Number: 0 ROUTING Signature Date Procedure Originator W. Stoner / l,_ il> ^046 Review and Concurrence:_ _

NASA Project RSO W. Stoner 2 -4A, s/t_

NASA Project Safety Officer HI. Bayes /

NASA Project Construction Manager C. Fellhauer I NASA Project Environmental P K Manager P.Kolb 12 Ad o NASA Project QA Manager P. Kolb tI Al t4.

NASA Senior Project Engineer K Peecookt ( {2 rAK C,'

IMPLEMENTING APPROVAL:  ? S -

Project Safety Committee Chairman

/I -M f Date I

EFFECTIVE DATE:

IS MAY ora

NASA PBRF DECOMMISSIONING PROJECT CHANGE/CANCELLATION RECORD DOCUMENT TITLE: DOCUMENT NO: N/A REVISION NO: 0 Final Status Survey Plan for the Plum (May 2006)

Brook Reactor Facility Revision 0: Initial issue of Document

LIST OF EFFECTIVE PAGES (Page 1)

DOCUMENT NO: N/A (FSSP) REVISION NO: 0 Page No. Revision Level Page No. Revision Level Page No. Revision Level Cover Page 0 Routing Page 0 Change Page 0 LOEP 0 i through xi 0 1-1 0 2-1 through2-13 0 3-1 through 3-7 0 _

4-1 0 5-1 through 5-6 0 6-1 through 6-6 0 7-1 through 7-6 0 8-1 through 8-5 0 9-1 through 9-3 0 10-1 throughiO-5 0 11- Ithrough 11-2 0 12-1 through 12-7 0 13-1 through 13-2 0 Attachment A A-1 through A-27 Attachment A-Addendum 1 =

7-Pages _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Attachment A-Addendum 2 0 16-Pages _

Attachment A-Addendum 3 3-Pages 0 Attachment A-Addendum 4 2-P ages __ _ _ _ _ _ _ __ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _

Attachment B B-1 through B-26 0 Attachment C C-1 through C-16 Attachment C-Addendum 1 0 Cl-i through C1-256

LIST OF EFFECTIVE PAGES (Page 2)

DOCUMENT NO: N/A (FSSP) REVISION NO: 0 Page No. Revision Level Page No. Revision Level Page No. Revision Level Attachment C - Addendum 2.

C2-1 through C2-14 0 Attachment C - Addendum 3 CM- through 0-51 Attachment D_ _

D 1 through D-33 0

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 TABLE OF CONTENTS

1.0 INTRODUCTION

1-1 1.1 Purpose l-l 1.2 Final Status Survey Plan Overview 1 2.0 CLASSIFICATION OF AREAS.2-1 2.1 hnpacted and Non-Impacted Areas -- 2-1 2.2 PBRF Classifications 2-2 2.2.1 Changes in Classification 2 3.0 RADIOLOGICAL CONTAMINANTS AND DCGL.3-1 3.1 Site-Specific DCGL Values for Soil . 3-1 3.2 Site-Specific DCGL Values for Structure 3-1 3.3 Embedded Piping DCGL Values .3-2 3.4 DCGL for Other Media 3-3 3.5 Radionuclide Mixture for FSS- 3-3 3.6 Surrogate DCGL, Gross Beta Activity and the Unity Rule, 3-4 3.6.1 Surrogate Equation 3-5 3.6.2 Gross Beta Equation 3-5 3.6.3 Unity Rule Equation 6 3.7 Area Factors 3-6 4.0 ESTABLISHING 4~~~~~.0 SURVEY UNITS SALSIGSRE NT,,..............................,,......... .......... , 4-1......... 4-1 4.1 Survey Unit, ,4-1 4.1.1 Survey Unit Size.-4-1 4.1.2 Site Reference Coordinate System (Reference Grid) .. . 4-1 5.0 SURVEY DESIGN 5-1 5.1 Scan Survey Coverage . 5-1 5.2 Sample Size Determination ,,,,,,, ,5-1 5.2.1 Determining Which Test To Be Used, 5-1 5.2.2 Establish Decision Errors 5-2 5.2.3 Relative Shift 5-2 i

Final Plan Survey Plan Status Survey Revision 0 Facility Plum Brook ReactorFacility FinalStatus Revision 0 TABLE OF CONTENTS (continued) 5.2.4 Number of Measurements-- 5-3 5.2.5 Elevated Measurement Comparison (EMC) Sample Size Adjustment- - 5-3 5.3 Background Reference Areas, -- 5-3 5.4 Measurement Locations 5-4 5.4.1 Class 3 Survey Units 4 5.4.2 Class 1 and 2 Survey Units 5-4 5.5 FSS Design Package. 5-5 6.0 RADIOLOGICAL INSTRUMENTATION..6-1 6.1 Instrumentation Selection 6-1 6.2 Instrumentation Calibration 6-2 6.3 Response Checks .6-3 6.4 Minimum Detectable Concentration (MDC) 6-3 6.5 Detection Sensitivity .6-5 6.5.1 Structure Static Measurements 6-5 6.5.2 Structure Scan Measurements. 6-5 6.5.3 Soil Scan Measurements 6-5 6.5.4 Volumetric Materials 6-5 6.5.5 Embedded and Buried Piping 6-6 7.0 RADIOLOGICAL SURVEY METHODS... 7-1 7.1 Structures 7-1 7.1.1 Surface Scans 7-1 7.1.2 Static Measurements 7-1 7.1.3 Activated Concrete 7-2 7.1.4 Volumetric Concrete Measurements 7-2 7.1.5 Loose Surface Contamination Surveys..7-2 7.1.6 Cracks, Crevices, Wall-Floor Interfaces and Small Holes 7-3 7.1.7 Paint Covered Surfaces 7-3 7.2 Surface Soil 7-3 ii

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 TABLE OF CONTENTS (continued) 7.2.1 Surface Scans 7-3 7.2.2 Surface Sampling-,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,7-4

-7.23 7.3 Subsrface Subsurface Soil Samp il.,,,,g.......

---

77-44 7.3.1 Subsurface Sampling-----------------------------------------------------------------------------7-4 7.4 Stored Excavated Soil 7-5 7.5 Embedded and Buried Piping ,,,,,--- 7-5 7.6 Pavement Covered Areas 7-5 8.0 INVESTIGATION LEVELS AND ELEVATED AREAS TEST. . 8-1 8.1 Investigation Levels-------------------------------------------------------------------------------------------8-1 8.2 Investigation Process------------------------------------------------------------------------------------------8-1 8.3 Elevated Measurement Comparison (EMC) .................................................... 8-2 8.4 Remediation and Reclassification 8-4 9.0 DATA COLLECTION AND PROCESSING .. 9-1 9.1 Sample Handling and Record Keeping--,,,,,,,,, 9-1 9.2 Data Management ,,,,,,,,,,,,,,.,,-- ,,,,,,,-- ,,,,,,,,,--,, 1 9.3 Data Verification and Validation 9-1 9.4 Graphical Data Review. ,,,,,,,,,,,,,,,,,,,,,,,,,,, , 9-2 9.4.1 Posting Plots 9-2 9.4.2 Frequency Plots 9-3 10.0 DATA ASSESSMENT AND COMPLIANCE0...1-1 10.1 Data Assessment Including Statistical Analysis. 10-1 10.1.1 Interpretation of Sample Measurement Results 10-1 10.1.2 Wilcoxon Rank Sum Test 10-2 10.1.3 Sign Test. ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,-,,10-3 10.1.4 Unity Rule.---------------------------------------------------------------------------------------------10-4 10.2 Data Conclusions 10-4 10.3 Compliance 10-5 iii

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 TABLE OF CONTENTS (continued) 11.0 REPORTING FORMAT 11-1 11.1 History File 11-1 11.2 Survey Unit Release Record 11-1 11.3 Final Status Survey Report 11-1 12.0 FSS QUALITY ASSURANCE & QUALITY CONTROL (QA/QC).................. 12-1 12.1 Project Management and Organization. 12-1 12.1.1 NASA Project Radiation Safety Officer (RSO)- 12-1 12.1.2 FSS/Characterization Manager 1 12.1.3 FSS/Characterization Engineers 2 12.1.4 FSS/Characterization Supervisors 2 12.1.5 Radiological Laboratory Manager 12-2 12.1.6 FSS/Characterization Technicians 12-2 12.2 Training. 12-3 12.3 Written Procedures. 12-3 12.4 Access Control of Surveyed Areas and Systems . 12-3 12.5 Chain-of-Custody .12-4 12.6 Instrumentation Selection, Calibration and Operation, 12-4 12.7 Quality Control Surveys and Samples .12-4 12.7.1 Replicate Field Measurements 12-4 12.7.2 Replicate Sample Analyses. . 12-5 12.7.3 Control of Vendor-Supplied Services 12-6 12.7.4 Database Control 12-6 12.7.5 Assessment and Oversight ... 12-6 12.7.5.1 Assessments 12-6 12.7.5.2 Corrective Action Process .12-6 12.7.6 Data Validation .12-6 12.7.7 NRC Confirmatory Measurements 12-7

13.0 REFERENCES

.13-1 iv

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 TABLE OF CONTENTS (continued)

ATTACHMENTA 1.0 Purpose .......................................................................................................... ,.,,,,,,,,,,,A-l 2.0 References A-i 3.0 Assumptions and General Information.................................................................... A-1 3.1 Nuclides A-1 3.2 RESRAD and RESRAD-BUILD A-1 4.0 Discussion A-1 5.0 Historical Overview A-2 5.1 Assessment of Historical and Operational Factors A-4 6.0 Sample Collection and Analysis,.................................................................................A-5 6.1 Smear Samples of Structural Surfaces ,,,--,,,,,,,,,,,,,- ,,,,,,,,,,,,,,-,,,,, A-5 6.2 Concrete Core Sampling-- ,,,,,,,,,,,,,,,,,,, ,,,,,,,,,,-------------A-7 6.3 Soil Sampling- ,,,,,,,,..,,,.. ,,,,,,,,,,,,,,,,,..,,,,,,,,,.A-16 7.0 Evaluation of Sample Results and Elimination of Insignificant RadionuclidesA-19 7.1 Minimum Detectable Activity (MDA) A-19 7.2 Mean Distribution Evaluation A-19 7.3 Application of the 10% Rule-A-23 8.0 Conclusion.-A-26 9.0 Addendums.A-27 Addendum 1, Result of Isotopic Analysis of Smear Samples (7 Pages)

Addendum 2. Preliminary Analysis of Concrete Core "Puck" Samples by the On-Site Laboratory (16 Pages)

Addendum 3 Results of Isotopic Analysis of Concrete Core Samples (3 Pages)

Addendum 4 Results of Isotopic Analysis of Surface Soil Samples (2 Pages)

ATTACHMENTB 1.0 Purpose...........................................................................................................................B-i 2.0 References B-1 3.0 DCGL Approach.B-2 4.0 DCGL Methodology .................................................................................................... B-3 V

Survey Plan Revision 0 Plum Plum Brook Reactor Facilily Reactor Facility Final Status Final Status Survey Plan Revision 0

%00 TABLE OF CONTENTS (continued) 5.0 DCGL Bases.B-4 5.1 Residual Contamination in Surface Soils B4 5.2 Residual Contamination in Buildings ... ,,....,,,..,,, .B-ll 5.3 Residual Contamination in Subsurface Structures B-18 6.0 6.0 Area Factors.B-25 Area Factors ~~~~~~~~~~~~~~~~~........ ,,,,,,,,,,,,,,,,,,,,,,,................

,,,,,,,,,,,,,-2 ATTACHMENT C 1.0 Purpose .. C-1 2.0 2.0 Refrences, References..C-1 ,C-3.0 Discussion. .C-1 3.1 Industrial Safety.. , C-2 3.2 Waste Generation C-2 3.3 Cost C-2 4.0 Embedded Piping Versus Buried Piping................................................C-3 4.1 Embedded Piping DCGL Values.... ,,,,,,,,,,,,,,,,,,, ,,, .C-3 4.2 Buried Piping DCGL Values.... ,, ,,,,, ,,,,, ... C-3 5.0 Embedded Piping Conceptual Dose Model..............................................................C-4 6.0 Dose Calculations for 9-Pipe Cluster.C.................................,,,,,,.,C-5 7.0 Embedded Piping DCGL Values ............................... ,...,,,.................. C-7 8.0 Evaluation of Alternative Exposure Scenario ...................................................... C-8 9.0 Area Factors.. C-10 Addendum 1, MicroShield Case Run Reports for Canal F- Pump Room Walkway Pipe Cluster (256 Pages)

Addendum 2, MicroShield Case Run Reports for Pump Room Trench Pipe Cluster (14 Pages)

Addendum 3, MicroShield Case Run Reports for Area Factor Calculations (51 Pages)

ATTACHMENT D Illustrationsof Survey Area Classification(33 Pages) vi

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 LIST OF ACRONYMS

- a Alpha ALARA As Low As Reasonable Achievable AF Area Factor AMCG Average Member of the Critical Group P Beta CFR Code of Federal Regulations CoC Chain-of-Custody cm Centimeters cm2 Square Centimeters cpm Counts per Minute DCGL Derived Concentration Guideline Levels DCGLEmc DCGL for small areas of elevated activity, used with the Elevated Measurement Comparison (EMC)

DCGLES Effective Surface DCGL DCGLFSS Final Status Survey DCGL DCGLv Volumetric DCGL DCGLw DCGL for average concentrations over a large area, used with statistical tests EMC Elevated Measurement Comparison EP Embedded Pipe EPA Environmental Protection Agency ERB Emergency Retention Basin FSS Final Status Survey FSSP Final Status Survey Plan FSSR Final Status Survey Report ft Feet Gamma GRC Glenn Research Center HTD Hard To Detect Kd Distribution Coefficient LBGR Lower Boundary of the Gray Region M2 Square Meters MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDC Minimum Detectable Concentration mrem Millirem NASA National Aeronautics and Space Administration NRC Nuclear Regulatory Commission ODNR Ohio Department of Natural Resources OEPA State of Ohio Environmental Protection Agency NIST National Institute of Standards and Technology PBRF Plum Brook Reactor Facility PBS Plum Brook Station pCi/g Picocuries per Gram PDF Parameter Distribution Function PRCC Partial Rank Correlation Coefficient QA Quality Assurance QC Quality Control x

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 LIST OF ACRONYMS (continued)

ROLB Reactor Office and Laboratory Building RPD Relative Percent Difference SEB Reactor Service Equipment Building TBD Technical Basis Document TEDE Total Effective Dose Equivalent WEMS Water Effluent Monitoring Station WRS Wilcoxon Rank Sum Test xi

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0

1.0 INTRODUCTION

1.1 Purpose The purpose of this Final Status Survey Plan (FSSP) is to describe the methodology and approach to performing the Final Status Survey of the Plum Brook Reactor Facility. This plan replaces the description of the FSS methodology provided in the Decommissioning Plan and provides the description of the planned final radiation surveys as required by IOCFR50.82(b)(4)(iii). The survey program described by this plan will demonstrate that the facility has been decontaminated to the levels required for release of the facility for unrestricted use as prescribed by IOCFR20, subpart E. The report of Final Status Survey results will provide the basis for requesting termination of NRC Licenses TR-3 and R-93 in accordance with IOCFR50.82(b)(6).

The FSSP is based on guidance provided in NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSIM); NUREG 1757, "Consolidated NMSS Decommissioning Guidance"; NUREG-1727, "NMSS Decommissioning Standard Review Plan"; and DG-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination". The graded approach described in NUTREG-1575 (MARSSIM) is followed to ensure that survey efforts are maximized in those areas having the greatest potential for residual contamination. The FSSP was developed to work in conjunction with the programmatic plans that are currently in place to safely and effectively decontaminate and dismantle the Plum Brook Reactor Facility (PBRF) as described in the Decommissioning Plan.

1.2 Final Status Survey Plan Overview This plan describes the scope and methodology of the PBRF final status survey (FSS) process. It includes specifics for classification of areas, DCGL development, establishment of survey units, survey design, radiological instrumentation, survey methods, investigation levels, data collection and processing, data assessment and compliance, reporting, and quality control.

Specific details of this plan will be incorporated into site procedures for implementation of the FSS.

1-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 2.0 CLASSIFICATION OF AREAS 2.1 Impacted and Non-Impacted Areas Area classification ensures that the number of measurements and the scan coverage are commensurate with the potential for residual contamination to exceed the unrestricted use criterion. Areas that have no reasonable potential for residual contamination from site operations are classified as non-impacted areas.

These areas are not required to be surveyed beyond what has already been completed as a part of site characterization to confirm the area's non-impacted classification. Areas with some potential for residual contamination are classified as impacted areas. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2, or Class 3 designations as described below:

• Class 1 Areas - Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiation surveys) above the DCGLW.

Examples of Class 1 areas include: 1) site areas previously subjected to remedial actions, 2) locations where leaks or spills are known to have occurred, 3) former burial or disposal sites, 4) waste storage sites, and 5) areas with contaminants in discrete solid pieces of material and high specific activity.

• Class 2 Areas - Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLW. To justify changing the classification from Class 1 to Class 2, there should be measurement data that provides a high degree of confidence that no individual measurement would exceed the DCGLw.

Examples of areas that might be classified as Class 2 for the final status survey include: 1) locations where radioactive materials were present in an unsealed form, 2) potentially contaminated transport routes, 3) areas downwind from stack release points, 4) upper walls and ceilings of buildings or rooms subjected to airborne radioactivity, 5) areas handling low concentrations of radioactive materials, and 6) areas on the perimeter of former contamination control areas.

• Class 3 Areas - Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGLw, based on site operating history and previous radiation surveys. Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with very low potential for residual contamination but insufficient information to justify a non-impacted classification.

2-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 2.2 PBRF Classifications Section 2.2 of the PBRF Decommissioning Plan describes the operational history and radiological status of the PBRF. The plan also discusses routine and non-routine events that occurred during the PBRF's operational history that contributed to facility radioactivity and contamination levels as well as presenting information from a characterization survey conducted in 1985 and confirmatory survey conducted in 1998. This information was used to develop initial PBRF classifications which are listed in Table 4-2 of the Decommissioning Plan. After approval of the Decommissioning Plan, additional characterization data was collected throughout the site. This information revealed that some site areas were assigned overly conservative classifications and some areas were under classified.

The current classifications and summary of characterization results are provided in Tables 2-1 and 2-2. Illustrations of structure and land area impacted areas are provided in Attachment D.

2.2.1 Changes in Classification Changes in classification are based on survey data and other relevant information that indicates a different area classification is more appropriate. Changes in classifications which decrease an area classification (i.e., impacted to non-impacted; Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3) will not be performed without prior NRC approval per license condition requirements.

2-2

c Plum Brook Reactor Facility c (

FinalStatus Survey Plan Revision 0 Table 2 Survey Area Classifications Floor Estimated Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area Ame rfe) Area Surface Classification Classification Total Beta Total Beta (a) (m) Area (b) (c) (dpm(tOO cm3) (dpml100 cm2)

Reactor Building (Building 1111):

Inside Containment Vessel

• Area 2- Work Space Inside CV 0' 4,019 374 2,244 Class 1 Class 1 <5,000 <5,000

• Area 17 - Test Area Inside CV at -25' 1,350 126 756 Class I (C O 5Cx) Class 1 88,284 100,832 1 1

• Area 18 - Canal E (C1110 111Cx) 1,281 120 720 Class t Class 1 23,235 297,000

• Area 23 - Quad A (C1110 107Cx) 803 75 450 Class 1 Class 1 18,644 116,000

• Area 24 - Quad B (C1110 108Cx) 813 76 _ .456 Class 1

• Area 25 - Quad C (C 1110 109Cx) Class 1 4733 14,000 803 75 450 Class 1 Class 1 48,406

• Area 26 - Quad D (C1110 110Cx) 337,000 637 60 360 Class 1 Class 1 13,811

• Area 29 - Sub-Pile Room (C1110 101Cx) 60,000 157 15 90 Class 1 Class 1 38,108

• Area 30 - Elevator Pit and Stair Well 78,775

- 202 2 ls ls ,8 (C1110 102Cx/103Cx104Cx) 210 20 120 Class

,5 Class 3,286 813 10,073 941 5,646 Reactor Building (Building 1111):

Outside Containment Vessel

• Area 1- General Floor Area @grade 9,449 878 5,268 1'Class1/

(C 111 207Cx) 2 <15,000 300,000

• Area 3 - Canals __ _ _

-Canal F (C111 104Cx) 382 36 216 Class 1 Class 1 117,040 995,000

• Canal G (C1111 105Cx) 5801 54 324 Class 1 Class I 50,738 426,000

• Canal H (C 1 103Cx) 455j 43 J 258.1 Class 1 Class 1 27,500 _ 39,000

• Ara 4 - ShopArea (C1 11 203Cx) -

Rooms 102 &103, Exp Ctri Annex 806 75 450 Class 3 Class 2

• Room 104, Office 400 J 381 228 Class 3 Class 2 144

• Room 105, Office 551 441J 411 2461 Class 3 Class 2

• Count Room (C111 204Cx) _ _

• Room 5, Experiment Control Room 1,156 108 648 Class 2 _ Class 2 _

• Room 6, Counting 80 8 48 Class 2 Class 2

• Room 7, Janitor Closet 124 12 _72 Class 2 Class 2___35

_195

• Room 8, Restroomr 498 47 282 Class 2 Class 2 195

• Room 9, Locker 197 19 114 Class 2

• Room 10, Personnel Decontamination Class 2 120 12 72 Class 2 Class 2 _

• Area 11 - Office Area Mezzanine (Cl11 208Cx) _ ___ ___

• Room 15, Reactor Control Room 872 l Class 2 Class 2

• Room 16,Restroom,+12 Elevation 80 8 48 Class 2 Class2 2

• Room28,Storage 87 9 54 Class 2 Class 2 l 2,957 17,749 o Rooms 202/203, Elec Equip & Parts Storage 806 7 5 450 Class 2 Class 2

- Room 204, Office 400 38 228 Class 2 Class 2 2-3

r (, (

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Table 2 Survey Area Classifications Estimated Floor Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area Area Area Surface Classification Classification Total Beta Total Beta r(c)

(a) (i) Area (dpm/100 cm) (dpml1OO cm)

(m2)

• Room 205, Office 44 41 246 Class 2 Class 2 Room 206, Office 171 16 96 Class 2 Class 2 e Room 207, Office 171 16 96 Class 2 Class 2 2,957 17,749

• Room 208, Office 171 16 96 Class 2 Class 2

• Room 209, Office 171 16 96 Class 2 Class 2

• Pipe Chase 46 5 30 Class 2 Class 2

• Area 20 - Work Area 0 -15' (C1 1 202Cx) 7,790 724 4,344 Class 1 Class 2 850 38,000

• Reactor Bldg Sub-basement (ClI1 201Cx) eArea 21 - Work Area @ -25' 3,013 280 1,680 Class 1 Class 2 6,280

• Area 22 - Pump Room Area + Roof 25,058 127 762 Class 1/2 Class 2

• Penthouse Areas (Cl11 209Cx) _ (a)

• Room 31, East Penthouse ( N/A N/A N/A N/A Class 2

• Room 32, West Penthouse N/A NA N/A N/A ( Class 2 N/A<N/A

• Area 33 - Air Lock I & 2 (Cl11i1 206Cx) 121 12 72 Class 1 Class 1 <1,000 <1,000 30,840 2,878 17,268

• Building Exterior (Cl1i1 301Cx)

'Building Roof 28,084 l 2,609 2,609 Class 2

• Building Exterior Walls (e) 658 5,188 13,110 l 1,218 l 1,218 i Class 3 (e)

Additional Information

• Reactor Bldg Loose and Fixed Equipment (C1111 101Cx) 1,061 148,117

-I,

• Rx Bldg & Canal E Loose and Fixed Equip (Cliii 106Cx) 252,936 13,079,207

(

Hot Laboratory (Bulding 1112):

82,107 7,646 26,741 L

• _Hot Cells (C1112 107Cx)

• Room 1 Hot Cell #1 342 32 192 Class I (e)

'Room2,HotCell #2 138 13 78 Class 1 (e)

• Room 3, Hot Cell #3 96 9 54 Class I (e)

Room 4, Hot Cell #4 64 6 36 Class 1 e 1,547,828 11,664,262

• Room 5,Hot Cell #5 64 6 36 Class 1 (e)

• Room 6,Hot Ceil #6 64 6 36 Class 1 (e)

'Room 7, Hot Cell #7 64 6 36 Class 1 (e)

• Mezzanine above Hot Cells 1-2 1,000 93 556 Class 1 (C 1 12 1O9Cx) Class 1/2 151,979 597,660

___ __ ___ s_Cla1C

/ 9 , 465

• bv eznno el 71Q05 94 564 Class 1 (C l1 12 1 1OC x)_ Class 1/2 989,411 4,462,502 L Cold Work Areas (Ci 112 203Cx) _ _ _ _ _ _ _ _

2-4

c r (

Plum Brook ReactorFacility Final Status Survey Plan Revision 0 Table 2 Survey Area Classifications Estimated Floor Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area Area Area Surface Classification Classification Total Beta 2 Total Beta (a) (Mi) Area (b) (c) (dpm/100 cm ) (dpm/100 cm')

(a) (m)

• Room 8, Cold Work Area + Air Lock 2,219 207 1,242 Class 1 Class 2

• Room SA, Vestibule 44 5 30 Class 2 Class 2 11,398 288,801 eRoom 9, Manipulator Repair Shop 409 38 228 Class 1 Class 210,006 43,834

• Room IO, Office 117 11 66 Class 2 Class 2

• Lavatory Areas (C1112 205Cx)

• Room 1n1, LockerArea 217 21 126 Class 1 (e)

• Room 12, Cold Lavatory 142 14 84 Class 1 (eJ

• Area 13 - Pipe Space 61 6 36 Class 1 (e) 11,834 78,234

• Room 14, Change Room 183 18 108 Class I (e) 1

• Room 15, Hot Lavatory 279 26 156 - Class 1 (e)

• Room 15A, Closet 24 3 18 Class 1 (e)

'Room 16, Hot Work Area (C1112 202Cx) 2,155 201 1,206 Class 1 (e) 628,257 1,667,274

'Hot Handling Area (C1112 106Cx)

• Room 17, Hot Handling Room 692 l 65 390 l Class l Class112 628,257 1,667,274

• Area 26 - Valve Pit 184 18 108 Class 1 (e)

'Area 18 - Canal J (C1112 105Cx) 435 411 246 Class 1 Class 1 175,662 1,192,000

'Hot Dry Storage (C1112 103Cx)

• Room 19, Hot Dry Storage 900 84 504 Class 1 (e)

• Room 19A, Off-Gas Clean-up 174 17 102 Class 1 (e) 4,400 2,160,000

'Room 19, Removable Slabs & area above 79I 7 44 Class 1 (e)

HDS (C1112 111Cx) 793 7 444 C 1_ (e)

' Area 20 - Canal K (Cl 112 104Cx) 799 75 450 Class 1 Class 1 156,519 2,298,000

' Decontamination Areas (Cl 112 204Cx)

• Room 21, Storage + Roof 246 23 138 Class 1 Class 1/2

• Room 22, Repair Shop 646 61 366 Class 1 Class 1/2

• Room 23, Decontamination Room 387 36 216 Class 1 Class 1/2 396,111 1,602,360

• Room 24, Storage 591 55 330 Class 1 Class 1/2 I

' Hot Pipe Tunnel (Cl 1112 108Cx) o Area 25A - Hot Pipe Tunnel 1,413 l 132 792 Class1 Class1 I

• Area 25B - Hot Pipe Tunnel 1,474 j 137 822 J Class 1 l Class 1 30,000 3,000,000

• Area 25C -Hot Pipe Tunnel 1,112 1041 624 Class 1 l Class 1 l l

'Room 27, Mezzanine (C1112 112Cx) 1,412 132 l 792 Class1 l Class 1/2 2,000 5,000

• Building Exterior (C1112 301Cx)

• BuildingRoof 14,134 1,313 1,313 Class2 ) 360 917 o Building Exterior 14,047 1,305 1,305 Class33 Additional Information

' Hot Cells Loose Equipment & Material 10,691,780 61,910,893 (C 112 101 Cx) I 2-5

( C (

Plum Brook Reactor Facility Final Status Survey Plan Revision 0

._ Table 2 Survey Area Classifications Floor Estimated Floo Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area Area Area Surface Classification Classification Total Beta Total Beta (a) (m) Area (b) (c) (dpmnlOO cm) (dpm/100 cm)

(ma)

Hot Lab Loose Equipment & Material 412,081 33,401,830 (C1112 201Cx) _ 60,008 205,752 48,126 l 4,487 l 13,832 I Reactor Service Equipment Building (1131)

• SEB First Floor Areas (D1 131 403Cx)

• Area 1 -Shop & Equipment 2,218 207 1,242 Class 1/3 Non-Impacted o Room 1, Chem. Lab & Record 394 37 222 Class 2/3 Non-Impacted XRoom 2, Counting 143 14 84 Class 2/3 Non-Impacted

• Room 3, Office 173 17 102 Class 2/3 Non-Impacted

• Room 4, Corridor 104 10 60 Class 2/3 Non-Impacted

• Room 5, Office 87 9 54 Class 2/3 Non-Impacted

• Room 6, Locker 162 16 96 Class 2/3 Non-Impacted Room 7, Janitors Closet 28 3 18 Class 2/3 Non-impacted

• Room 8, Restroom Vestibule 16 2 12 Class 2/3 Non-Impacted

• Room 9, Vestibule 52 5 30 Class 2/3 Non-Impacted 1,134

• Room 10, Corridor 195 19 114 Class 2/3 Non-lmpacted

• Room 11, Control Room 1,083 101 606 Class 2/3 Non-Impacted

• Room 12, Chlorine 105 10 60 Class 2/3 Non-Impacted

• Room 13, Water Treatment 3,752 349 2,094 Class 2/3 Non-Impacted

• Room 15, Boiler 1667 155 930 Class 2/3 Non-Impacted mRoom 16, Parts and Storage 290 27 162 Class 2/3 Non-Impacted

• Room 17, Vertical Lift 149 14 84 Class 2/3 Non-Impacted mRoom 18, Engine Room 5,372 500 3,000 Class 2/3 Non-Impacted

• SEB Mezzanine Areas (D1 131 404Cx)

LRoom 1A, Instrument Repair 221 21 126 Class 3 Non-Impacted _

Room 2A, Instrument Repair 270 26 156 Class 3 Non-Impacted

• Room 3A, Instrument Repair 320 30 180 Class 3 Non-Impacted 1,564 4,859 Area 14 - Mezzanine 358 34 204 Class 2/3 Non-Impacted _

mRoom 19, Chemical Storage 1,526 142 852 Class 2/3 Non-Impacted

• Room 19A, Chemical Storage 393 37 222 Class 1/3 Non-Impacted 12,791 91,375 Cold Pipe Tunnel (D1131 406Cx) .

v Area 23 - CPT (under SEB_ 1,890 176 1,056 Class 1 Class 3

• Area 23 - CPT (from 1131 to 1152) 1,098 103 618 Class 1 Class 3 e Area 23 - CPT (under 1152) 570 53 318 Class 1 Class 3 Area23-CPT (from1151 to 1152) 366 35 210 Class I Class 3 21,824 92,897

• Area 23 - CPT (under 1151) 111 11 66 Class 1 Class 3

• Area 23 - CPT (from 1 1 to 1131) 2,347 219 1,314 Class I Class 3

• SEB Basement Areas (D1 131 402Cx)

Room 20, Auxiliary Equipment 1,020 95 570 Class 3 Non-Impacted 501 1,021 2-6

(I Plum Brook Reactor Facility C (

Final Status Survey Plan Revision 0 Table 2 Survey Area Classifications Estimated Ar Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area rea Area Surface Classification Classification Total Beta Total Beta (a) (M2) Area (b) (c) (dpmr1OO cm2) (dpmllO0 cm2)

___(ma) _____

o Room 21, Water Treatment Pump 1.543 144 864 Class 3 Non-Impacted o Area 22- Clear Well 423 40 240 Class 3 Non-Impacted o Area 24 - Generator/Compressor Pit 460 43 258 Class 3 Non-Impacted

• Room 25. Electrical & Batteries 914 85 510 Class 2/3 Non-Impacted 1,0

• Area 26 - Pit 88 9 54 Class 3 Non-Impacted

• Closet, Basement 40 4 24 Class 3 Non-Impacted

• Building Exterior (DI 131 405Cx) o Building Roof 16,985 1,578 1,578 Class 2 (e) 467 699 Building Exterior Walls 4,123 383 383 Class 3 {e)

Additional Information _

• SEB Loose Equipment & Material (D1131 40OCx) 294 74,070 75 2,071 51,056 l 4,763 l 18,773 l Fan House Building (1132):

• First Floor Operating Area (Cl 132 201 Cx) _

• Room 1, Operating Area 2,772 258 1,548 Class 1 Class 1/2

• Room 2, Restroom _ 94 9 54 Class 1 Class 1/2

• Room 3 Janitors Closet 55 6 36 Class 1 Class 1/2 10,868 108,396 o Room 4, Delonizer 155 15 90 Class 1 Class 1/2

• RoomS, Office 64 6 36 Class 1 Class 1/2

• Basement Areas (Cl 132 102Cx) _____ _

• Room 6, Air Monitor 174 17 102 Class 1 Class 1

• Room 7, Sump Room 217 21 126 Class 1 Class 1

• Room 8, Equipment Room 2,906 270 1,620 _ Class 1 Class 1

• Area 9, Tunnel to Rx Bldg 11 11 420 40 240 Class 1 Class 1 35,363 652,565

• Area 10, Stairwell 96 9 54 Class 1 Class 1

• Resin P10 136 3 78 Class 1 Class 1

• Pipe Trench 183 18 108 Class 1 Class 1

• Building Exterior (Cl 132 301Cx) _

• Building Roof 3,897 3621 3621 Class 2 e

• Building Exterior Walls 99 5I0 4,445 413 413 Class 2 (e) 15,614 1,457 4,867 Waste Handling Building (1133): ,

• Operating Areas (Cl133 104Cx)

• Room 1, Operating Area 1,951 182 1,092 Class 1 Class 1/2 oRoom 2, Decontamination 414] 39 1 234 i Class 1 Class 1/2 3,062 26,221 oArea16-StairHousing 97 10I 60 Class (e?

1 l

• Laundry Area (C1133201Cx) oRoom 3, Laundry 222 21 126 Class 1 Class 1/2 2,957 17,749 2-7

c Plum Brook Reactor Facility C

FinalStatus Survey Plan

(

Revision 0 Table 2 Survev Area Classifications

-l-o - __

Floor Floor Total FSS Initial Mean Net Facility or Survey Area Maximum Net A Area Surface Classification Classification Total Beta Total Beta (a) Area (b) (dpm/lOO cm2)

(a)(i)

(c) (dpm/100 cm2)

A e Room 4, Clean Clothing 171 16 96 Class 1 Class 1/2

• Room 5, Storage 85 8 48 Class 1 Class 1/2 eRoom 6, Laboratory 155 15 90 Class 1 Class 1/2 Room 7, Office 150 14 84 Class 1 Class 1/2 e Room 8, Corridor 121 12 72 Class 1 Class 1/2

. Room 9, Janitores Closet 19 2 12 Class 1 e Room 10, Closet Class 1/2 15 2 12 Class 1 Class 1/2 eRoom 11 Contaminated Channe 2,957 17,749 122 12 72 Class I Class 1/2 eRoom 12, Clean Restroorn 120 12 72 Class 1 Class 1/2

• Room 13, Corridor 127 12 72 Class 1 Class 1/2 ePipe Chase 21 2 112 Class 1 Class 1/2

• Room 14 Boiler Room (1133 202Cx) 411 39 234 Class 1 Class 1/2

• Basement Work Areas (Cl 133 102Cx)

• Room 15, Access & Work Area 1,563 146 876 Class 1_ Class116092674 54,900

• Room 19, Access Tunnel 347 33 198 Class 1 Class 1 5 ,

• Evaporator Work Areas (Cl 133 103Cx)

• Room 17, Preparation 406 38 228 Class 1

• Room 18, Evaporator Class 1 406 38 228 Class 1 Class 1 7,582

• Mezzanine (Cl 133 105Cx) 201,885 1,350 126 756 Class 1 (e)

• Building Exterior (Cl 133 301 Cx)

• Building Roof 4,726 439 1 439 Class 2 (e3

• Buildinrg Exterior Walls 3,821 355 3355 Class 2 e 14,869 1,391 1 4,376 Primary Pump House (1134)

• Interior Rooms (Cl 134 201Cx)

Room 1 Pump 171 16 96 Class 1 Class 2

• Room 2, Pump 171 16 96 Class 1 Class 2

• Room 3, Pump 171 16 96 Class 1 Class 2 1,052,432 5,288,932

• Room 4, Heat Exchanger 1,073 100 600 Class 1 Class 2 _ 1,046,611 4,986,792

• Room 5 149 14 84 Class 1 Class 2 o Room 6, Bypass Deionizer 257 24 144 Class 1 Class 2

• Exterior Rooms (Cl 134 202Cx)

Room 7 636 60 360 Class 1 Class2 Room 8 990 92 552 Class 1 Class 2 6,998 275,028

°Mezzanine 544 51 306 Class 1 Class 2

• Resin Pit (Cl 134 203Cx) 136 13 78 Class 1 Class 2

• Building Exterior (Cl 134 301 Cx)

Building Roof 5,220 485 485 Class 1 (e) 115 713 Building ExteriorWaiis 4,316 406 406 1 Class 1 (e) I 2-8

C r (

Plum Brook ReactorFacility Final Status Survey Plan Revision 0 Table 2 Surve Area Classifications Floo Estimated FAroeoar Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area (tt2 Area Surface Classification Classification Total Beta Total Beta (a) (mi) Area (b) (c) (dpm/100 cm 2) (dpm/1 00 cm2 )

13_834 1,293 3,303 1 Reactor Office and Laboratory Building (1141)

• Area 1, Northeast Stair Hall (Cl 141 203Cx)

• Elevator Areas (C1141 202Cx) 291 28 168 Class 2 Class 2 T 818 1,575 o Area 2, Elevator & Shaft 60 6 36 Class 2 Class 2 e Area 3, Elevator Machinery 143 14 84 Class 2 Class 2 249 941

• Vault Areas (Cl 141 206Cx) _ _

• Area 4, Cold Test Area 2,491 232 1,392 Class 1 Class 2

• Area 5, Airway 700 66 396 Class 1 Class 2 oArea 8, Vestibule 99 10 60 Class 1 Class 2

• Area 9, Restroom 82 8 48 Class 1 Class 2 Area 10, Janitors Closet 38 4 24 Class 1 Class 2 3,204 38,865 e Area 11, Building Service Equipment 1,767 165 990 Class 2 Class 2 586 2,155

• Area 14, Tool Crib 265 25 150 Class 2 Class 2

.Area 15, General Shop Area 813 76 456 Class 2 Class 2

' Area 16, Telephone Exchange 303 29 174 Class 2 Class 2 eArea 17, Vault 188 18 108 Class 1 Class 2

• Area 6, Sump (C1141 208Cx) 530 50 300 Class 1 Class 2 209 781

• Area 7, Crawl Space (C1141 21OCx) 3,782 352 2,112 Class 3 Class 2 450 608

• Instrument Areas (C1141 207Cx)

'Area 12, Counting Rooms 357 34 204 Class 2 Class 2

'Area 13, Electronic Shop 317 30  : 180 Class 2 Class 2 715 2,589

- Area 13A, Calibration Room

• Utility Tunnel (C1141 209Cx) 177 17 102 [ Class 2 Class 2 2,064 192 192 Class 1 (e) 1,500 1,700

• First Floor Offices (C1141 211Cx)

'Lobby 505 47 282 Class 3 Class 2

' Room 100, Office 214 20 120 Class 3 Class 2

°Room101,Office 214 20 120 Class3 Class2

• Room 102, Office 214 20 120 Class 3 Class 2

° Room 103, Office 209 20 120 Class 3 Class 2

• Room104,Office 176 17 102 Class3 Class2

• Room105,Office 191 18 108 Class3 Class2 263 1,664

° Room 106, Office 176 17 102 Class 3 Class 2

_Room17,Office 214 20 120 Class3 Class2

'Room108,Office 197 19 114 Class3 Class2

'Room 109/110/111, Classroom 653 61 366 Class 2/3 Class 2

• Room 121, Conference 484 45 270 Class 2/3 Class 2

' Room 122, Office 300 28 168 Class 2/3 Class 2

_Room123,Offce 197 19 114 Class 2/3 Class 2 2-9

C Plum Brook Reactor Facility C

FinalStatus Survey Plan

(

Revision 0 Table 2 Survey Area Classifications Aloo Floor Estimated Total FSS Initial Mean Net Maximum Net Facility or Survey Area Are Area Surface Classification ClassificatIon Total Beta Total Beta (a) (m ) Area (b) (c) (dpm/100 cm2) (dpm/lOo cm2)

(a)_ __ _ W ) _ _ _ _ _ _

• Room124,Office 200 19 114 Class 2/3 Class 2

• Room 125 Office 200 19 114 Class 2/3 Ciass 2 Room 126, Office 212 20 120_ Class 2/3 Class 2

• Restroom, Men's, +0 Elevation 177 17 102 Class 2 Class 2

• Restroom, Women's +0 Elevation 158 15 2 Class 2 Class 2 263 1,664

• Lavatory, +0 Elevation 32 3 18 Class 2 Class 2 e Janitor's Closet, +0 Elevation 30 3 18 Class 2 Class 2

• Pipe Chase, +0 Elevation 39 4 24 Class 2 Class 2

• Hallways, +0 Elevation North/South 307 29 174 Class 3 Class 2

• First Floor Work Areas (Cl 141 212Cx)

'Room112,FirstAid 303 29 174 Class2 Class2

• Room 113, Office 100 10 60 Class 2 Class 2

• Room 114, Health & Safety Operations 406 38 228 Class 2 Class 2

• Room115 Office 341 32 192 Class 2 Class 2

• Room 116, Receiving 209 .20 120 Class 2 Class 2 528 3,026

° Room 117/118,Electronic Laboratory 768 72 432 Class 1/2 Class 2 o Room 119/120, Electronic Laboratory 642 60 360 Class 1/2 Class 2 o Hallways, +0 Elevation East/West 1,725 161 966 Class 2 Class 2

• Middle Stairwell (Cl 141 204 Cx) 196 19 114 Class 2 Class 2 497 1,527

• Southeast Stairwell (Cl 141 205Cx) 196 19 114 Class 2 Class 2 1,490 6,438

• Second Floor Offices (Cl 141 213Cx) _ _

' Room 200, Office 202 19 114 Class 3 Class 2

• Room 201, Office 214 20 120 Class 3 Class 2

• Room 202,Office 214 20 120 Class 3 Class 2

• Room 203, Office 214 20 120 Class 3 Class 2

• Room216, Office 214 20 120 Class 3. Class 2

• Room 217, Library 480 45 270 Class 3 Class 2

• Room218,Office 214 20 120 Class 3 Class 2

• Room 219, Office 214 20 120 Class 3 Class 2

'Room 220, Office 214 20 120 Class 3 Class 2 305 3,564

'Room 221, Office 214 20 120 Class 3 Class 2

' Room 222, Office 523 49 294 Class 3 Class 2

• Room 223, Office 214 20 120 Class 3 Class 2

' Restroom, Men's, +12 Elevation 134 13 78 Class 2 Class 2

' Restroom, Women's +12 Elevation 45 5 30 Class 2 Class 2

• Janitors Closet 1, +12 Elevation 28 3 18 Class 3 Class 2

' Hallways,+12 Elevation North/South 727 68 408 Class 3 Class 2 2-10

C r (

Plum Brook ReactorFacility Final Status Survey Plan Revision 0 Table 2 Survey Area Classifications Estimated Arlo Floor Total FSS Initial Mean Net Maximum Net Facility or Survey Area Area Surface Classification Classification Total Beta Total Beta (f) (m2) Area (b) (c) (dpm1iOO cm2) (dpmflOo cm2)

(a) (m2)

• Second Floor Work Areas (C1141 214Cx)

• Room 204, Materials Laboratory 34 33 198 Class 1 Class 2

• Room 205, Materials Laboratory 301 28 168 Class 1 Class 2 eRoom 206, Materials Laboratory 345 33 198 Class 1 Class 2

• Room 207, Chem. Lab Storage 352 33 198 Class 1 Class 2

• Room 207A, Counting 160 15 90 Class I Class 2

• Room 208, Office 164 16 96 Class 1 Class 2

• Room 208A, Darkroom 133 13 78 Class 1 Class 2

• Roorm 2083, Darkroom 133 13 78 Class 1 Class 2

• Room 209, Chem. Laboratory 312 29 174 Class 1 Class 2 39,099 293,254 o Room 210, Chem. Laboratory 312 29 174 Class 1 Class 2 626a 4,878a

• Room 211/212, Chem. Laboratory 695 65 390 Class 1 Class 2

• Room 213, Chem. Laboratory 349 33 198 Class 1 Class 2

• Room 213A, Chem. Laboratory 362 34 204 Class 1 Class 2 o Room 214/215, Chem. Laboratory 417 39 234 Class 1 Class 2

• Janitors Closet 2, +12 Elevation 23 3 18 Class 1 Class 2

• Pipe Chase, +12 Elevation 33 4 24 Class 1 Class 2

• Hallways, +12 Elevation East-West 855 80 480 Class 1 Class 2

• Building Exterior (C1141 215Cx)

• Building Roof 11,540 1,072 1,072 Class 2 (e) 595 1,093

• Building Exterior Walls 18,675 11,7351 1,735 Class 3 (e)

Hot Retention Area (1155)

• All Areas (Cl 155 102Cx) 4,050 377 2,262 Class 1 Class 1 22,973 41,197 ReactorSecurityBuilding (1191) _

All Areas (D1191 401Cx)

• Room 100, Visitor Control 470 44 264 Class 3 Non-impacted Room 101, Utility 28 3 18 Class 3 Non-impacted 177 523

• Room 102, Furnace 52 5 30 Class 3 Non-Impacted

• Lavatory 34 4 24 Class 3 Non-impacted

• Building Exterior (C1191 402Cx) 1,701 158 158 Class 3 Non-Impacted -12 138 2.285 214 I 494 Table 2-1 Notes:

(a) Floor Area - The majority of numerical values provided in the floor areas (f2 ) column were taken from "Plans of Buildings and Structures, Lewis Research Center, Plum Brook Station, June 1974". In locations where that document did not provide numerical values for the floor areas, drawings were used to estimate the floor surface areas.

2-11

( (7 Plum Brook ReactorFacility Final Status Survey Plan Revision 0 (b) FSS Classification - The current classification to be used for the FSS. Class X/Y designates the floors and lower walls as Class X while the upper walls and ceilings are Class Y.

(c) Initial Classification - The initial classifications derived from Table 4-2 of the Decommissioning Plan (d) Structure was demolished and removed. No FSS will be performed.

(e) Classification not specifically listed in Table 4-2 of the Decommissioning Plan 2-12

( ( (

Plum Brook Reactor Facility Final Status Survey Plan Revision 0 Table 2 Environmental Survey Area Classification Survey FSS Initial Mean Maximum PBRF Survey Area . Area ((mZ) Area - Classification Classification Isotope Measurement Measurement (a) (b) (c) (pCI/g) (pCug)

Emergency Retention Basinew 11.

'9co l 0.56406 2.1436

• All Areas (A2100 EiCx) 60 ,000 5,575 Class 1 Class 1 '10Cs l 34 37.500

"_Cs 36.023 542410 Effluent Pathways

• Drainage Systerns (Storm Sewers) NA/ACas1lss1 . WCO wCo 10.05137 0.36 02.40827 6.4

• Pentolite Ditch & E-nvirons (A2300 101Cx) 57,0 5,2 las1Cas 3Cs 37.S120 16.65 Environmental Areas

'loco 0.05172 0.08278

• South Southwest Unit (A2400 301 Cxj 125,00 11,613 Class 2/3 Class 3 6C 16 .87

.137cs 0.21873 0.44134

• South Southeast Corner (A2400 302Cx) ,WCO 0.04291 0.06650 Note: Contains the WEMS Spill area, the south ERB "3Cs 0.23082 0.36748 122,891 11,417 Class 1/2 Class 1/3 -Co .

berm and portions of the east and west ERB berm.

137 1.8708 37.151 Cs 15.327 217.90 60Co 0.04259 0.04662

• Southwest Unit (A2400 303Cx)

Note: Contains portions of the WHB Spill Area and the 37cs 19.657 150.58 125,000 11,613 Class 1/2/3 Class 1/3 C 0.09689 0.27 Waste Storage Pad.

OOCo 0.09689 0.22770

,rt 110.51 207.50

• Southeast Unit (A2400 304Cx)

Note: Contains the north ERB bermf 'Co 0.04558 0.06258 and portions of the 105,895 9,838 Class 1/2 Class 3 east and west ERB berm.

"2Cs 0.26198 0.42556

• Northwest Unit (A2400 305Cx)

Note: Contains the HRA, WHB, FH, portions of the Hot 'Co 0.05054 0.10171 108,608 10,091 Class 1/2 Class 1/3 Lab, and portions of the WHB spill area.

'3Cs 0.51620 1.5203

• Northeast Unit (A2400 306Cx)

Note: Contains the CRA's and ERB Spill Areas 60Co 0.0470 1 0.07637 117,542 10,921 Class 1/2 Class 3 '"'Cs 0.23683 0.38934 6"Co 0.03420 0.03973

• North Northwest Unit (A2400 307Cx) 37 Note: Contains the Rx Bldg, ROLB, PPH, portions of the ' Cs 026071 0.49599 74,626 6,933 Class 1/2 Class 1/3 .3 Hot Lab, and PPH resin spill area.

Co 0.03369 0.05632

"__Cs 0.05655 0.12184 2-13

r (

FinalStatus Survey Plan (C _

Revision 0 Plum Brook Reactor Facility Table 2 Environmental Survey Area Classification Survey FSS Initial Mean Maximum PBRF Survey Area Areaf) Survey Classification Classification Isotope Measurement Measurement (a) Area (mn) (b) (c) (pCIg) (pCVg)

• North Northeast Unit (A2400 308Cx) .OCo 0.04431 0.05089 Note: Contains the SEB, and portions of the Sludge 121,072 11,248 Class 1/2 Class 3 Basins and Precipitator. Cs 0.26315 0.43784

• Northem Unit (A2400 309Cx) 6°Co 0.04560 0.07090 Note: Contains portions of the Precipitator and portions 110,287 10,246 Class 1/2/3 Class 3 of the Sludge Basins. -. Cs 0.23966 0.39998

• Far Northern Unit (A2400 31OCx) .OCo 0.04903 0.06385 Note: Contains the Compressor Bldg, Gas Storage 89,817 8,345 Class 3 Class 3 3c Structure and Cryogenic and Gas Supply Farm. .'Cs 0.30677 0.61130

• Open land areas outside the PBRF Fence 00Co 0.6339 4.567 Note: Areas to serve as buffer zones to impacted areas 1,524,748 141,649 Class 1/3 (d) 2.0630 19.080 inside the fence 2_0630_ 1.080 ME M 1,216,450 113,015 3MIU Table 2-2 Notes:

(a) Survey Area - This area was estimated using facility drawings.

(b) FSS Classification - The-current classification to be used for the FSS. Refer to the illustrations in Attachment D of this plan for a classification layout.

(c) Initial Classification - The initial classifications derived from Table 4-2 of the Decommissioning Plan.

(d) Classification not specifically listed in Table 4-2 of the Decommissioning Plan 2-14

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 3.0 RADIOLOGICAL CONTAMINANTS AND DCGL The goal of the PBRF decommissioning project is to release the site for unrestricted use in compliance with the NRC's annual dose limit of 25 mrem/yr plus ALARA. The NRC dose limits apply to residual radioactivity that is distinguishable from background. The DCGL values established within this FSSP will not be increased without prior NRC approval.

Site-specific dose assessments were performed to calculate the DCGL for surface soil, structures, and embedded pipe. RESRAD 6.21, RESRAD 6.0, RESRAD-BUILD 3.22, and the Microshield code were used for these dose assessments. Model input parameters were developed and justified for each assessment.

The dose assessments and DCGL calculations for surface soil, building reuse and subsurface structures are described in detail in Attachment B, "Approach and Basis for Development of Site-Specific Derived Concentration Guideline Levels (DCGL)" of this FSSP. The dose assessment and DCGL calculations for embedded piping are described in detail in Attachment C, "Dose Assessment and DCGL Calculation for Embedded Pipe".

3.1 Site-Specific DCGL Values for Soil The surface soil DCGL values are listed in Table 3-1. The DCGL values are the volumetric activity of the first 6 inches of soil, in pCi/g, that will be used during FSS to determine compliance with the 25 mrem/yr unrestricted use criterion.

Table 3-1 DCGL Values for Surface Soil Radionuclide DCGL Co-60 3.8 Sr-90 5.4 Cs-137 14.7 3.2 Site-Specific DCGL Values for Structures The structure DCGL values are listed in Table 3-2. The DCGL values are the surface activity level in dpm/100 cm 2 that will be used during FSS to determine compliance with the 25 mrem/yr unrestricted use criterion.

3-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Table 3-2 DCGL Values for Structures Radionuclide DCGL (dpm/100 cm2 )

Co-60 11,000 Sr-90 33,100 Cs-137 40,500 Eu-154 4,500 H-3 9.1E+06 I-129 14,900 U-234 31,500 U-235 27,100 U-236 _ 33,200 _

The current decommissioning strategy for PBRF calls for the removal of activated portions of the concrete bio-shield and disposition of the activated concrete as radioactive waste. Any potentially activated concrete remaining in place will be evaluated for residual radioactive contamination volumetrically using the subsurface structure DCGL values provided in Attachment B, Table B-14.

Volumetric sampling of contaminated concrete, as opposed to surface measurements may be necessary if gross activity measurements are insufficient or subsurface activity greater than the DCGLw is suspected. The results will either be evaluated by 1) calculating the derived total gross beta dpm/100 cm2 in the sample and comparing the gross beta results directly to the gross beta DCGL or,

2) compare results to the volumetric DCGL values (Attachment B, Table B-14) for subsurface structures coupled with application of the unity rule.

3.3 Embedded Piping DCGL Values The PBRF contains a number of pipe runs that are embedded in concrete.

Embedded piping (EP) is any pipe situated below the minus three (-3) foot elevation that is totally encased in concrete or piping directly beneath building floors that may not be totally encased in concrete, but contained within the structural foundation of the building. Examples include the Containment Vessel Quadrant and Canal drain piping and remaining sections of the Primary Cooling Loop. The EP DCGL values are listed in Table 3-3. These values represent surface activity concentrations that correspond to an annual dose of 1 mrem/yr to a future building occupant. In order to utilize the EP DCGL values, piping must be grouted prior to license termination.

3-2

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 The PBRF dose goal for EP is 1 mrem/yr. However, at the discretion of PBRF, different dose goals could be applied in different areas as long as the residual contamination on the structure surface in the survey unit containing the given EP is sufficiently low to allow for the selected dose goal. For example, if the FSS results indicate that the residual contamination level in Hot Dry Storage is 0.5 times the DCGL, the dose from the two drains in this survey unit could be as high as 12.5 mrem/yr.

Table 3-3 DCGL Values for Embedded Piping Radionuclide 2 (dpm/100 cm per mrem/yr)

Co-60 2.408E+05 Eu-154 5.325E+05 Eu-152 7.352E+05 Nb-94 9.082E+05 Ag-108m 1.312E+06 Cs-137 3.785E+06 3.4 DCGL for Other Media Other media that will undergo FSS include subsurface soil, concrete foundation pads, and buried pipes. The DCGL for foundation pads and buried piping will be the same as the structure DCGL values. Buried piping is any pipe buried in soil and situated outside the structural foundation of a building, such as storm drains.

The DCGL for sediment (e.g., stream or ditch silt) will be the same as the surface soil. The DCGL for subsurface soil will be the same as surface soil.

3.5 Radionuclide Mixture for FSS The radionuclides listed in Tables 3-1, 3-2 and 3-3 were selected after a detailed evaluation of the current characterization data including concrete core samples, smears, and media samples collected for analyses. The evaluations of radionuclide mixtures for soil and structure DCGL development are contained in Attachment A, "Radionuclide Distribution Basis for DCGL Determination, and FSS of the PBRF". The evaluation of radionuclide mixtures for embedded piping is contained in Attachment C, "Dose Assessment and DCGL Calculation for Embedded Pipe". Current and representative sample data are used to determine the final radionuclide mixtures for FSS.

Per NUREG-1757, the Nuclear Regulatory Commission (NRC) staff considers radionuclides and exposure pathways that contribute no greater than 10% of the 3-3

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 25 mrem/yr dose criteria to be insignificant contributors. This 10% limit for insignificant contributors is an aggregate limitation only. That is, the sum of the dose contributions from all radionuclides and pathways considered insignificant will be no greater than 10% of the dose criteria of 25 mrem/yr per 10 CFR Part 20, Subpart E.

The dose from the insignificant radionuclides will be accounted for in demonstrating compliance, but may be eliminated from further detailed evaluations. The lowest total aggregate dose from the remaining radionuclides for PBRF structures was evaluated at 90% of the 25 mrem/yr criterion, which means that radionuclides potentially representing a dose of 2.5 mrem/yr were eliminated.

To ensure the 25 mrem/yr criterion is met, any structural survey unit where the mean of FSS results indicates that the dose may be greater than 22.5 mrem/yr (including dose from embedded pipe in the given survey unit, if applicable) shall be reviewed for compliance with the unrestricted use limit.

For PBRF soil, the doses from all radionuclides other than Cs-137, Co-60, and Sr-90 totaled 0.5 mrem/yr. Because this dose is 2% of the 25 mrem/yr limit, all radionuclides other than Cs-137, Co-60, and Sr-90 were eliminated from further consideration. To ensure the 25 mrem/yr criterion is met, any open land survey unit where the mean of FSS results indicates that the dose may be greater than 24.5 mrem/yr will be reviewed for compliance with the unrestricted use limit.

The only exposure pathway that is operative for PBRF embedded piping is direct dose from gamma-emitting radionuclides. All radionuclides that may be present in piping residual contamination are fixed in place by the grout and are not available for transport to ground water or to ingestion or inhalation pathways to a future building occupant.

3.6 Surrogate DCGL, Gross Beta Activity and the Unity Rule Hard-to-detect radionuclides (non-gamma emitters for soil and non-beta emitters for structures) will be addressed by using a surrogate relationship to another detectable radionuclide. A common example would be to relate a specific radionuclide, such as Cs-137, to one or more radionuclides of similar characteristics such as Sr-90. This directly applies to PBRF as gamma spectroscopy will be used to assess volumetric soil samples in comparison against the DCGL for soil. In such cases, to demonstrate compliance with the release criteria for the survey unit, the DCGL for the surrogate radionuclide, in this case Cs-137 must be scaled to account for the fact that it is being used as an indicator for an additional radionuclide, Sr-90. The result is referred to as the surrogate DCGL.

For structural surfaces, the final DCGL for FSS design and implementation will be a gross beta DCGL that represents the unrestricted use criterion of 25 mrem/yr.

There are two steps required to determine the gross beta DCGL; 1) perform a surrogate calculation to account for radionuclides that cannot be readily measured 3-4

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 by a beta detector (e.g., plastic beta scintillation, gas proportional) with typical efficiency; and, 2) perform a gross activity DCGL calculation on the surrogate DCGL values and other beta emitting radionuclides determined to be present in significant fractions. The equations to be used were derived from NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys" and are listed below.

3.6.1 Surrogate Equation The surrogate DCGL is computed based on the distribution ratio between the hard-to-detect radionuclides and the easy-to-detect radionuclides. The surrogate DCGL is calculated using the following equation:

Surrogate,.,[( lrR~( ~ 9

[(DCGLSr)+(DCGL 2) (DCGL,) JDCGL.j]

Where: DCGLSur = Surrogate radionuclide DCGL DCGL2, 3.. .n = DCGL for radionuclides to be represented by the surrogate R, = Ratio of concentration (or nuclide mixture fraction) of radionuclide "n" to surrogate radionuclide 3.6.2 Gross Beta Equation Where multiple radionuclides are present, the gross beta DCGL may be developed. This approach enables field measurements of gross activity rather than the determination of individual radionuclide activity for comparison to the radionuclide-specific DCGL. The gross beta DCGL is calculated using the following equation:

DCGLGB

['..DCGL, DCGL2 ) DCGL3 . tDCGLhj Where: DCGLGB = gross beta DCGL fn = mixture fraction of radionuclide "n" and DCGLI, = DCGL of radionuclide "n".

Note 1: The gross beta equation may also be used to calculate a gross alpha DCGL for application in areas where DCGL values are established for alpha emitters.

Note 2: If a surrogate radionuclide is used, the "fn" is equal to the surrogate radionuclide fraction.

3-5

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Note 3: The value of 1 in the numerator is replaced by the actual fraction of beta (or alpha) emitters if less than 100% of the mixture.

3.6.3 Unity Rule Equation The unity rule is typically used as the first test to evaluate compliance with radiological criteria for license termination when more than one radionuclide has been determined to be potentially present. In lieu of a single gross beta (or alpha) DCGL, a unity rule calculation is used to demonstrate compliance with the structure surface unrestricted use limit.

The unity rule is typically used as the first test to demonstrate compliance with surface soil limits. A surrogate DCGL, if applicable, would be used in the unity rule calculation. The unity rule is:

CI C2 C. <1 DCGLI DCGL2 DCGL, Where: C, = concentration of radionuclide n and DCGI, = DCGL of radionuclide n.

If two adjusted gross activity DCGL values (one for alpha and one for beta) are used, the unity rule must be applied to demonstrate compliance.

In the event that the unity rule test fails, i.e., the calculated sum of fractions is > 1, then evaluation of individual radionuclide concentrations against their DCGL values may be performed.

3.7 Area Factors The area factor is the multiple of the DCGL that is permitted in the area of elevated residual radioactivity without requiring remediation. The area factor is related to the size of the area over which the elevated residual radioactivity is distributed. That area, denoted AEMC, is generally bordered by levels of residual radioactivity below the DCGL, and is determined by the investigation. The area factor is the ratio of dose per unit area or volume for the default surface area for the applicable dose modeling scenario to that generated using the area of elevated residual radioactivity, AEMC. Area factors for surface soil, building reuse (structures), and embedded pipe are provided in Tables 3-4, 3-5, and 3-6, respectively. Area factor assumptions and calculations are provided in Attachments B and C.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Table 3-4 Surface Soil Area Factors Elevated Area 1 2 3 5 10 15 25 100 250 2,000 (m2)

Area Factor 10.4 6.2 4.7 3.4 2.3 1.9 1.6 1.2 1.1 1 Table 3-5 Building Reuse Area Factors (Structures)

Elevated Area 0.25 0.50 1 2 4 6 8 10 15 25 50 75 Area Factor 40.2 20.8 11.1 6.2 3.6 2.8 2.4 2.1 1.7 1.4 1.1 1 Table 3-6 Embedded Pipe Area Factors Pipe Length (ft) Total Dose from Pipes of Area Factor specified Length (mrem/h) 1 3.013E-10 5.9 2 5.834E-10 3.0 5 1.197E-09 1.5 10 1.531E-09 1.2 3-7

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 4.0 ESTABLISHING SURVEY UNITS 4.1 Survey Unit Survey areas listed in Tables 2-1 and 2-2 will be divided into discrete survey units. Survey units are areas that have similar characteristics and contamination levels. Survey units are assigned only one classification. The site and facility are surveyed, evaluated, and released on a survey unit basis.

4.1.1 Survey Unit Size Survey units are typically limited in size to ensure each area is assigned an adequate number of data points. The survey unit sizes for PBRF are provided in Table 4-1. Note that the maximum survey unit size for Class 1 structures is 75 m2 , which is smaller than the 100 m2 recommended in NIREG-1575 (MARSSIM). The 75 m2 maximum was required to be consistent with the site-specific building reuse dose assessment assumptions.

Table 4-1 Recommended Survey Unit Areas for FSS Class Structures Land 1 up to 75m 2

• up to 2000 m2 2 up to 1000 m2 up to 10,000 M2 3 up to 10,000 m2 up to 100,000 in2

• Includes floor only 4.1.2 Site Reference Coordinate System (Reference Grid)

A reference coordinate system is used for impacted areas to facilitate the identification of the location of measurements and samples within the survey unit. The reference coordinate system is basically an X-Y plot of the site area referenced to an established fixed point monument. Once the reference point is established, grids may be overlaid parallel to lines of latitude and longitude.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 5.0 SURVEY DESIGN This section describes the methods and data required to determine the number and location of measurements or samples in each survey unit, the coverage fraction for scan surveys, and requirements for measurements in background reference areas. The applicable design activities described in this section will be documented in a Survey Design Package for each survey unit.

5.1 Scan Survey Coverage The area covered by scan measurement is based on the survey unit classification.

A 100% accessible area scan of Class 1 survey units will be required. The emphasis will be placed on scanning the higher risk areas of Class 2 survey units such as soils, floors and lower walls. The percentage of scan coverage in Class 2 areas will be proportional to the potential for finding contamination that is close to the DCGL. Scanning percentage of Class 3 survey units will be performed on likely areas of contamination based on the judgment of the FSS/Characterization Engineer. Minimum scan survey coverage requirements for the PBRF are provided in Table 5-1.

Table 5-1 Minimum Scan Survey Coverage Scan Measurements Class 1 Class 2 Class 3 Scan Coverage 100% 10 to 100% Minimum of 10%

5.2 Sample Size Determination This section describes the process for detennining the number of survey measurements necessary to ensure that a data set is sufficient for statistical analysis. Sample size is based on the relative shift, the Type I and II errors, and the specific statistical test used to evaluate the data.

5.2.1 Determining Which Test Will Be Used Statistical tests will be used to determine if the FSS results are below the DCGLW. The Sign Test or Wilcoxon Rank Sum (WRS) Test will be implemented using unity rules, surrogate methods, or combinations of unity rules and surrogates.

The Sign Test is expected to be the most appropriate test for PBRF FSS results because background is expected to constitute a small fraction of the DCGL. If a situation is encountered where background is a significant fraction of the DCGL, the Wilcoxon Rank Sum (WRS) Test may be used.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 5.2.2 Establish Decision Errors A decision error is the probability of making an error in the decision on a survey unit, either passing a survey unit that should fail or failing a survey unit that should pass. The first decision error, passing a survey unit that should fail, is referred to as a false positive or TYPE I decision error. The probability of making this error is denoted by "a". Setting a high value for a results in a higher risk of passing a survey unit that should fail. Setting a low value for a lowers the risk of passing a survey unit that should fail.

The a for the PBRF will be set at 0.05 and will not be increased without prior NRC approval.

The second decision error, failing a survey unit that should pass, is referred to as a false negative or TYPE II decision error and is denoted by "1". Selecting a high value for 13 results in a higher risk of failing a survey unit that should pass and subjecting it to further investigation. Selecting a low value for P lowers the risk and minimizes these investigations. The P for the PBRF will initially be set at 0.10, or 10 percent probability based on site-specific considerations. Since the P is the licensee's risk, the PBRF project may choose to vary this value from 0.25, or 25 percent to 0.05 or 5 percent after evaluating the resulting change in the number of required survey measurements and the risk of unnecessarily investigating and/or remediating survey units that are truly below the release criteria.

5.2.3 Relative Shift The number of measurements needed depends on a ratio involving the concentration to be measured relative to the variability in the concentration. This ratio is called the relative shift (A / a) and is calculated as follows:

A = (DCGLW-LBGR) 0f 0:

Where: Delta (A) = DCGLw minus the Lower Boundary of the Gray Region (LBGR)

Sigma (a) = Standard deviation of contaminant measurements collected within the survey unit LBGR = The concentration to which the survey unit must be remediated in order to have acceptable probability of passing the statistical test for meeting the site release criteria.

NOTE 1: The LBGR is initially set at 0.5 times the DCGLw, but may be adjusted to obtain an optimal value for the relative shift between 1 and 3.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 NOTE 2: Use of a relative shift greater than 3 must be approved by the FSS/Characterization Manager.

5.2.4 Number of Measurements The number of background reference area or survey unit data points N is determined using Table 5.3 or Table 5.5 (as applicable) of NUREG-1575 (MARSSIM). The tables include the recommended minimum 20%

adjustment to ensure an adequate sample size. Computer programs (i.e.,

Compass'and VSPm) may also be used to calculate N.

Special considerations may be necessary for survey units with structure surface areas less than 10 m2 or land areas less than 100 m2 . In these cases, the number of data points obtained from the statistical tests may be unnecessarily large and not appropriate for smaller survey unit areas.

Instead, professional judgment may be used to determine the specified level of survey effort.

5.2.5 Elevated Measurement Comparison (EMC) Sample Size Adjustment If the scan Minimum Detectable Concentration (MDC) in a Class 1 survey unit is greater than the DCGLw, the sample size will be calculated using the equation provided below. No adjustment is required in a Class 2 or 3 survey unit. If NEMC exceeds the statistically determined sample size N, NEmc will replace N.

N A EMC AXMC Where: NEMC = elevated measurement comparison sample size A = survey unit area AEMC = area corresponding to the area factor calculated using the MDCSCa, concentration 5.3 Background Reference Areas The residual radioactivity of a survey unit may be compared directly to the DCGL. However, the residual radioactivity may contain radionuclides that occur in background. To identify and evaluate those contributions attributable to licensed activities, representative background radionuclide concentrations are established using background reference areas. Background reference areas have similar physical, chemical, radiological, and biological characteristics as the areas to be surveyed. They are usually selected from near site non-impacted areas, but are not limited to natural areas undisturbed by human activities. Surveys will be conducted of one or more background reference areas (where appropriate), to determine background levels for comparison with specific survey units.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 5.4 Measurement Locations 5.4.1 Class 3 Survey Units Measurements locations in Class 3 survey units and background reference areas are taken in random locations. Random means that each measurement location in the survey unit has an equal probability of being selected. The random selection process uses random numbers that correspond to a survey unit's reference coordinate system to establish the measurement locations within the survey unit. The random numbers are generated using a random number generator. Measurement locations selected that do not fall within the survey unit area or cannot be surveyed due to site conditions may be replaced using the same random process.

5.4.2 Class 1 and 2 Survey Units For Class 1 and Class 2 survey units, a systematic pattern (triangular or rectangular grid) is used to establish measurement locations. After the spacing L is determined, a starting point is identified using the random selection process as described in section 5.4.1. Beginning at the random starting point, a row of measurement locations or points is identified parallel to the X axis at intervals of L. For a triangular grid, a second row of points is then developed parallel to the first row, at a distance of 0.866 x L from the first row. Points identified along that second row are midway between the points on the first row on the X-axis. For a rectangular grid, the points are identified around the random starting point in a perpendicular manner at intervals of L. The physical spacing of measurement locations are calculated as follows:

Triangular Grid V0.866N Where: A = Survey unit Area N = Number of data points required based on statistical tests 5-4

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Rectangular Grid

\N Where: A = Survey unit Area N = Number of data points required.

To simplify the designation of data points while assuring a sufficient number of data points are obtained for statistical purposes, the value of L is rounded down to the nearest distance easily measured in the field. If a point falls outside the survey unit or at locations that cannot be surveyed, additional points may be determined using the random selection process as described in section 5.4.1.

5.5 FSS Design Package A FSS Design Package is produced for each survey unit or group of survey units, at the discretion of PBRF. Instructions for package development, implementation, and closure will be described in an approved PBRF procedure.

Each FSS Design Package will contain the following information, as applicable:

• A brief overview of the survey unit history

• A description of the survey unit

• A summary of statistical tests to be used to evaluate survey results

• A description of the background reference areas and materials and justification for their selection (required only if WRS Test is selected)

• A summary of static measurements or sample data used to both evaluate the success of remediation and to estimate the survey unit variability

• A description of radiological field instruments to be used and sensitivity

• For in-situ measurements made by field instruments, a description of the instruments, calibration, operational checks, sensitivity, and sampling methods, with a demonstration that the instruments and methods have adequate sensitivity

• Survey Area Maps showing the Survey Unit(s) and Measurement/Sample locations

• Specific survey and sampling instructions 5-5

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0

• FSS investigation levels

• Justification for radionuclide mixture used to acquire gross activity and surrogate DCGL values

• DCGL values (i.e., DCGLW, DCGLEMC, etc.)

• Supporting documentation (e.g., Compass'and VSPm) 5-6

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 6.0 RADIOLOGICAL INSTRUMENTATION Proper selection and use of instrumentation will ensure that sensitivities are sufficient to detect radionuclides at the minimum detection requirements specified by the survey design which will assure the validity of the survey data. Commercially available portable and laboratory instruments and detectors are typically used to perform the three basic survey measurements: 1) surface scanning; 2) static measurements; and 3) spectroscopy of soil and other bulk materials, such as concrete. Specific implementing procedures control the issuance, use, and calibration of instrumentation.

6.1 Instrumentation Selection Radiation detection and measurement instrumentation is selected based on the type and quantity of radiation to be measured and the detection sensitivity.

Detection sensitivity requirements are specified in section 6.5.

A listing of typical radiological instrumentation that may be used for FSS are provided in Table 6-1. As the project proceeds, other measurement instruments or technologies, such as in-situ gamma spectroscopy, may be found to be more efficient or appropriate than the survey instruments proposed in this plan. The acceptability of such an instrument or technology for use in the final survey program must be justified and documented in a Technical Basis Document (TBD) or equivalent PBRF document. The TBD should include the following: (1) a description of the conditions under which the method would be used; (2) a description of the measurement method, instrumentation and criteria; (3) justification that the technique would provide acceptable MDCs; and (4) a demonstration that the method provides data that has a Type I error (falsely concluding that the survey unit is acceptable) equivalent to 5% or less.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 Table 6-1 Typical Radiological Instrumentation and Nominal MDC Nominal Detection Sensitivity Detector Model 2 Meter Model Application M I (dpm/100 cm 2) (dpml100 cm 2)

Ludlum 44-9 Ludlum 2350-1 f static & scan 2900 985 Ludlum 43-5 Ludlum 2350-1 a static & scan 150 75 Ludlum 43-68 f mode Ludlum 2350-1 P static & scan 1050 330 Ludlum 43-68 a mode Ludlum 2350-1 a static & scan 170 70 Ludlum 44-116 Ludlum 2350-1 3 static & scan 1300 415 Ludlum 43-90 Ludlum 2350-1 a static & scan 130 55 3.5 pCi/g Ludlum 44-10 Ludlum 2350-1 y scan 6OCo 6.5 N/A pCi/g 137Cs Ludlum 43-37 Ludlum 2350-1 1 scan 1000 N/A Packard Tr-Carb 2900 N/A smear N/A 40 (Tritium)

TR LSC Canberra S5XLB or N/A a and/or p N/A 18 Protean IPC-9025 smear HPGe Gamma N/Ay Analysis N/AVaries Spectroscopy System

1. Based on 1-minute count time; and default values for surface efficiencies,es, as specified in Internal Standard, ISO 7503-1.

2. Functional equivalent instrumentation may be used.

6.2 Instrumentation Calibration All instrumentation used for FSS, including the gamma spectroscopy system, will be calibrated for normal use under typical field conditions and for the radiation types and energies of interest at least annually not to exceed 15 months.

Calibration typically includes:

• High Voltage Calibration

• Discriminator/Threshold Calibration

• Window Calibration 6-2

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0

• Alarm Operation Verification

• Scaler Calibration Verification

• Operating Voltage Determination

• Calibration Constant Determination

• Dead Time Correction Determination Calibrations will be performed in accordance with approved procedures. If vendor services are used, the vendor QA program is subject to approval of the PBRF. Radioactive sources will be traceable to the National Institute of Standards and Technology (NIST) or equivalent standards.

Calibration labels showing the instrument identification number, calibration date and calibration due date will be attached to all portable field instruments.

Calibration records will be maintained in accordance with PBRF record retention requirements.

6.3 Response Checks Instrumentation response checks are conducted to assure proper instrument response and operation. An acceptable response for field instrumentation is an instrument reading within +/-20% of the established check source value.

Laboratory instrumentation standards will be within +/- 3 sigma and not to exceed

+ 20% as documented on a control chart. Response testing of portable radiological instrumentation and/or detectors used for FSS shall be performed prior to use and following use. If an instrument and/or detector are to be used for a continuous time frame greater than a standard work shift, the instrument and/or detector will be subjected to a response test prior to the end of that shift. If an instrument fails a response check, it is labeled appropriately and is removed from service until the problem is corrected in accordance with applicable procedures.

Data obtained with that instrument since its previous acceptable performance check, will be evaluated for acceptability. The results of this evaluation will be documented.

6.4 Minimum Detectable Concentration (MDC)

Minimum Detectable Concentration (MDC) is defined as the smallest amount or concentration of radioactive material that will yield a net positive count with a 5%

probability of falsely interpreting background responses as true activity from contamination and a 5% probability of interpreting a result at the MDC level as being background. The MDC is dependent upon the counting time, geometry, sample size, detector efficiency and background count rate. There are two different MDCs that will be used, one for direct alpha or beta surface 6-3

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 contamination measurements, MDCrtjis and one for field alpha or beta scanning of surfaces, MDCSc,. These are calculated differently and each will be incorporated into the operating procedures. The equation to be used for calculating the MDCstatic for direct alpha or beta surface measurements is as follows:

lt 3+3.29 BRts (1+ ts)

MDCstatic = X tb t*tt100 Where: MDCtatic = Minimum Detectable Concentration (dpm/100 cm2 ),

BR = Background Count Rate (cpm),

tb = Background Count Time (min),

t= Sample Count Time (min),

A = Detector Area (cm 2) and Etot = Total Efficiency.

Total Efficiency (Etot) is equal to Detector Efficiency (Es) multiplied by the applicable Surface Efficiency (Es).

The equation that will be used to calculate MDCscan for beta surface scan measurements is as follows:

d'* -. 60 MDCS..n = A i vEs VP 100 Where: MDCSC,,, = Minimum Detectable Concentration (dpm/100 cm 2) d' = Decision error taken from Table 6.5 of NUREG -1575 (MARSSIM) i = Observation counting interval (scan speed divided by the detector width) bi = Background count per observation interval es= Detector Efficiency (c/d) 6-4

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 es = Surface Efficiency (typically 25% for alpha and 50% for beta (ISO 7503-1, Table 2))

p = Surveyor Efficiency (typically 50%)

A = Detector Area (cm 2) 6.5 Detection Sensitivity The nominal detection sensitivity of some of the detectors that may be used for surface contamination surveys has been determined and is provided in Table 6-1.

Count times and scanning speed are calculated and selected to ensure that the measurements are sufficiently sensitive for the DCGL.

6.5.1 Structure Static Measurements Instruments used to perform beta and alpha static measurements of structures should be capable of detecting the radiation of concern to a MDC of less than 50% of the applicable DCGLW. Although, any value below the DCGL is acceptable.

6.5.2 Structure Scan Measurements Instruments used to perform beta and alpha scan measurements of structures should be capable of detecting the radiation of concern to a MDC of less than the DCGLw. If the MDCscan exceeds the DCGLw, additional static measurements may be required.

6.5.3 Soil Scan Measurements For scanning soil with a sodium iodide gamma detector, the MDCscan values presented in Table 6.7 of NUREG -1575 (MARSSIM) will typically be used. According to NUREG-1727, these values provide an acceptable estimate of MDCSC,2 for the survey. Individual MDCSC, values may be calculated for sodium iodide detectors configured and calibrated to detect specific radionuclides such as Cs- 137. If the MDCSCa, exceeds the DCGLW, additional soil samples may be required.

6.5.4 Volumetric Materials The MDC for gamma spectral analysis of soil, concrete, and other volumetric materials will be based on sample count times sufficient to detect 10% of the applicable DCGL values for the radionuclides of concern or best sensitivity achievable. The MDC for the beta counter for analysis of smears will be based on sample count times sufficient to detect 6-5

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 10% or less of the adjusted gross beta DCGL for the radionuclide mix appropriate for the survey area or best sensitivity achievable.

6.5.5 Embedded and Buried Piping Beta or gamma sensitive pipe probes are used for surveys of embedded and buried piping. Instruments used to perform static measurements should be capable of detecting the radiation of concern to a MDC of less than 50% of the applicable DCGLW. Although, any value below the DCGL is acceptable. Instruments used to perform scan measurements should be capable of detecting the radiation of concern to a MDC of less than the DCGLw. If the MDCscaJI exceeds the DCGLW, additional static measurements may be required.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 7.0 RADIOLOGICAL SURVEY METHODS Trained and qualified personnel will perform survey measurements and collect samples.

FSS measurements include surface scans, static measurements, gamma spectroscopy of volumetric materials, and in-situ gamma spectroscopy. The surveying and sampling techniques are specified in approved procedures.

7.1 Structures Structures will receive scan surveys, static measurements and, when necessary, volumetric sampling.

7.1.1 Surface Scans Surface scanning is performed in order to locate small areas of residual activity above the investigation level. Beta scans will be performed over accessible floor, wall and ceiling interior and exterior surfaces, asphalt, and concrete paved areas. In general, beta scans are performed with the detector held at less than 2 cm from the surface and moved at a rate such that residual radioactivity can be detected at or below the investigation level. Alpha scans may be performed in areas where alpha contamination has been observed and if a surrogate is not used. In general, alpha scans are performed with the detector held as close to the surface as possible and moved at a rate such that there is a high probability of detecting elevated residual activity.

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.

If surface conditions prevent scanning at the specified distance, the detection sensitivity for an alternate distance will be determined and the scanning technique adjusted accordingly. Whenever possible, technicians will monitor the audible response to identify locations of elevated activity that require further investigation and/or evaluation.

7.1.2 Static Measurements Static measurements are performed to detect total surface activity levels.

Static measurements are conducted by placing the detector on or very near 7-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 the surface to be counted and acquiring data over a pre-determined count time. A count time of one minute is typically used and generally provides an acceptable DCGLstatc. Measurements will be performed at designated locations and at locations of elevated activity identified by surface scans within each survey unit. Static measurements will be performed in pairs of shielded and non-shielded measurements.

7.1.3 Activated Concrete The current decommissioning strategy for PBRF calls for the removal of activated portions of the concrete bio-shield and disposition of the activated concrete as radioactive waste. Any potentially activated concrete remaining in place will be evaluated for residual radioactive contamination volumetrically using the subsurface structure DCGL values provided in Attachment B, Table B-14.

7.1.4 Volumetric Concrete Measurements Volumetric sampling of contaminated concrete, as opposed to static measurements may be necessary if gross beta measurements are insufficient or subsurface activity greater than the DCGLW is suspected.

Volumetric concrete samples will be analyzed by gamma spectroscopy and for hard to detect radionuclides, as necessary. The results will either be evaluated by 1) calculating the derived total gross beta dpm/I00 cm2 in the sample and comparing the gross beta results directly to the gross beta DCGL or, 2) compare results to the volumetric DCGL values for subsurface structures coupled with application of the unity rule. Use of the unity rule may require the use of a surrogate calculation to account for the radionuclides in the mixture not identified by gamma spectroscopy.

This will be accomplished using the appropriate nuclide mixture for the applicable survey unit.

7.1.5 Loose Surface Contamination Surveys Loose surface contamination (smear) surveys will be performed to verify contamination is 10% or less of the DCGLw consistent with assumptions made during dose modeling for structural DCGL development. A smear will be collected at each static measurement location. This process is performed by wiping 100 cm 2 of surface area with a circular cloth or paper filter using moderate pressure. Smear samples will be evaluated for gross beta activity and gross alpha for areas known to contain alpha contamination. Each survey unit with smear results exceeding the building reuse 10% modeling assumption will be evaluated on a case-by-case basis to determine if the 25 mrem/yr limit is met (see Attachment B).

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 00 Plum Brook Reactor Facility FinalStatus Survey Plan Revision 7.1.6 Cracks, Crevices, Wall-Floor Interfaces and Small Holes Surface contamination on irregular structure surfaces (e.g., cracks, crevices, and small holes) is difficult to survey directly. Where no remediation has occurred and residual activity has not been detected above background, these surface blemishes may be assumed to have the same level of residual activity as that found on adjacent surfaces. The accessible surfaces are surveyed in the same manner as other structural surfaces and no special corrections or adjustments have to be made.

In situations where remediation has taken place or where residual activity has been detected above background, representative volumetric samples of the crack, crevice, or small hole may be obtained and evaluated per section 7.1.4. Other methods of evaluation may include adjustment of the instrument efficiency (ifjustifiable) or surveying the imperfection with a smaller detector (i.e., sodium iodide detector) and estimating the surface activity.

7.1.7 Paint Covered Surfaces Surfaces painted to fix loose contamination are remediated before FSS activities begin. For other surfaces painted after site start-up, representative samples are collected in areas where it is suspected that elevated levels of residual radioactivity could have been covered over.

Detection sensitivities are adjusted or remediation is performed as directed by the analysis results. NUREG 1507, "Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions" provides additional guidance.

7.2 Surface Soil Surface soil refers to outdoor areas where the soil is, for purposes of dose modeling, considered to be uniformly contaminated from the surface down to a depth of 15 cm (6 in). These areas will be surveyed through combinations of sampling, scanning, and in-situ measurements, as appropriate.

7.2.1 Surface Scans Gamma scanning will be performed over land surfaces to identify locations of residual surface activity. The gamma emitters are used as surrogates for the hard-to-detect radionuclides. NaI gamma scintillation detectors (typically 2" x 2") will be used for these scans. Scanning is generally performed by moving the detector in a serpentine pattern within 15cm (6 in) from the surface, while advancing at a rate not to exceed 0.5 m (20 in) per second. Audible signals should be monitored.

7-3

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 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 or flagged 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.

7.2.2 Surface Sampling Samples of soil (including sediment or sludge) will be obtained from designated locations and at areas of elevated activity identified by gamma scans. An appropriate volume of soil (typically 1 liter) will be collected at each sampling location using hand trowels, bucket augers, or other suitable sampling tools.

Sample processing involves removal of extraneous material and drying of the material. The sample material is then placed into a clean counting container for gamma spectroscopy analysis. All samples are tracked from time of collection through the final analysis in accordance with a chain-of-custody program.

7.3 Subsurface Soil Subsurface soil refers to residual radioactivity that is underneath structures such as building floors/foundations or material that is covered with clean soil or some other non-contaminated layer. Subsurface soil is evaluated through sampling. In accordance with DG-4006, scanning is not applicable to sub-surface activity assessments.

The historical site assessment should be reviewed to identify those survey areas where the potential exists for sub-surface radioactivity. Such areas can include, but are not limited to: soils under buildings; soils adjacent to building foundations or components where leakage was known or suspected to have occurred in the past; and areas where known spills of radioactive materials have occurred. Data from both the historical site assessment and any pertinent characterization data should be used to establish a bounding depth profile for any potential sub-surface radioactivity.

7.3.1 Subsurface Sampling Subsurface soil will be sampled by collecting core samples from 15 cm (6 in) below grade to the required depth. The number of samples required (N) is determined for the applicable statistical test applied.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 GeoProbe, split spoon sampling or other methods may be used for the FSS unless the area has already been excavated and remediated. If remediation of the area has been conducted, and post remediation surveys and sampling indicate that the remediation is complete and the area is ready for FSS, then the excavation will be treated as surface soil and surveyed and sampled accordingly.

Samples are segmented and homogenized over each one meter of depth.

Extraneous material is removed from each segment and is adequately dried. The material is then placed into a clean counting container for gamma spectroscopy analysis. All samples are tracked from time of collection through the final analysis in accordance with a chain-of-custody program.

7.4 Stored Excavated Soil In several areas, clean overburden soils may be removed and stockpiled on site for use as backfill materials. The primary area where this is expected to occur is the excavation of the storn drain system. These areas are primarily Class 3 areas; however the excavation and stockpile of soils from Class 1 and 2 areas is possible. Prior to the use of stockpiled excavated soils as backfill, the soil will be subjected to FSS in accordance with the classification of the area from which it had originated. Scanning requirements and soil sample frequency shall also be determined in accordance with the classification of the area where the soil had originated.

7.5 Embedded and Buried Piping Designated sections of embedded and buried piping will be remediated in place and undergo FSS. Compliance with the DCGL values, as presented in section 3.3 and Attachment C, will be assessed through the acquisition of measurements utilizing "pipe-crawling" technology and may include beta/gamma surface contamination measurements, static gamma measurements, or in-situ gamma-spectroscopy. Radiological evaluations that cannot be accessed directly will be performed via measurements made at traps and other appropriate access points where the activity levels are deemed to either bound or be representative of the interior surface activity levels. Embedded piping will be filled with grout after the FSS.

7.6 Pavement Covered Areas The survey design of parking lots, roads and other paved areas will be based on soil survey unit sizes since they are outdoor areas where the exposure scenario is most similar to direct radiation from surface soil. The structure DCGL values presented in Table 3-2 are applicable to paved areas because the dose pathways and source geometry (i.e., area source) are essentially the same as those assumed 7-5

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 in the Building Reuse scenario. In addition, the agricultural pathways are not applicable. The structure DCGL values are conservative since the direct dose from a 2000 m2 outdoor area is less than the dose from the Building Reuse scenario. An outdoor occupancy time of 963 hrs is used in the Surface Soil dose assessment, whereas in the Building Reuse scenario, an occupancy time of 2340 hrs is assumed. Also, dose from the re-suspension of radionuclides from paved surfaces will be less than in the Building Reuse scenario because the effective air exchange rate in an outdoor area is expected to be much higher than that seen in buildings. Scan and static beta surveys are made as determined by the survey unit design. If sub-surface contamination is possible under paved or other covered areas, sub-surface volumetric samples should be collected.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 8.0 INVESTIGATION LEVELS AND ELEVATED AREAS TEST During survey unit measurements, levels of radioactivity may be identified by an increase in count rate or an elevated sample result that warrant investigation. Elevated measurements may result from discrete particles, a distributed source, or a change in background activity.

8.1 Investigation Levels Investigation levels are radioactivity levels that are based on the site release criteria, which if exceeded, initiate an investigation of the survey data.

Investigation levels also serve as quality control checks to guard against misclassification of a survey unit, identify deficiencies in the remediation process, identify a possible instrumentation failure or large variations in background. The Investigation levels are established for each area classification as shown in Table 8-1.

Table 8-1 PBRF Investigation Levels Static Measurement or Classification Sample Result Scan Investigation Levels Investigation Levels Class 1 >DCGLEMc >DCGLEMC Class 2 > DCGLw >DCGLw or >MDCscan if MDCscan, is greater than DCGLw Class 3 C 3 % ofteDGwgreater

>5o

>5 of the DCGLW >DCGLw or >tMCan if MDCa is than DCGLw 8.2 Investigation Process Investigation levels are provided in each Survey Design Package. Locations identified by scan or static measurements with residual radioactivity that exceeds an investigation level are marked and verified true. The area around the elevated radioactivity is investigated to determine the extent and to verify that other undiscovered elevated radioactivity does not exist. If the elevated activity was identified during scanning, representative static measurements and/or samples are collected.

Investigations should consider: (1) the assumptions made in the survey unit classification; (2) the most likely or known cause of the contamination; and (3) the possibility that other areas within the survey unit may have elevated areas of 8-1

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 activity that may have gone undetected. Depending on the results of the investigation, the survey unit may require no action, may require remediation, and/or may require reclassification and resurvey. Results of the investigation process are documented. Investigation actions are shown in Table 8-2.

Table 8-2 Investigation Actions Action if Investigation Results Exceed:

Class DCGLEMc DCGLW 0.5 DCGLw Remediate and Perform statistical testing, 1 resuveyas ane y remediate and resurvey as Acceptable resurvey as necessary nesar necessary reclassify Increase scan coverage and 2 Remediate, reclassify portions as Acceptable portions as necessary necessary

. .a Increase scan coverage and Increase scan coverage 3 Remediate, reclssify reclassify portions as and reclassify portions as portions as necessary necessary necessary 8.3 Elevated Measurement Comparison (EMC)

The elevated measurement comparison may be used for Class 1 survey units when one or more scan or static measurements exceed the investigation level if remediation is not performed. The EMC provides assurance that unusually large measurements receive the proper attention and that any area having the potential for significant dose contribution is identified. As stated in NUREG-1575 (MARSSIM), the EMC is intended to flag potential failures in the remediation process and should not be considered the primary means to identify whether or not a survey unit meets the release criterion.

Locations identified by scan with levels of residual radioactivity which exceed the DCGLEMC or static measurements with levels of residual radioactivity which exceed the DCGLEMC are subject to additional surveys to determine compliance with the elevated measurement criteria. The size of the area containing the elevated residual radioactivity and the average level of residual activity within the area are determined. The average level of activity is compared to the DCGLw based on the actual area of elevated activity. (If a background reference area is being applied to the survey unit, the mean of the background reference area activity may be subtracted before conducting the EMC).

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 The initial DCGLEMC is established during the survey design and is calculated as follows:

DCGLEMfc = Area Factorx DCGL The area factor is the multiple of the DCGL that is permitted in the area of elevated residual radioactivity without remediation. The area factor is related to the size of the area over which the elevated activity is distributed. That area is generally bordered by levels of residual radioactivity below the DCGL and is determined by the investigation process. The method for calculating the area factors is provided in Attachment B and the resulting area factors are listed in Tables 3-4, 3-5, and 3-6.

The actual area of elevated activity is determined by investigation surveys and the area factor is adjusted for the actual area of elevated activity. The product of the adjusted area factor and the DCGLw determines the actual DCGLEMC. If the DCGLEMC is exceeded, additional investigations are performed.

The results of the elevated area investigations in a given survey unit that are below the DCGLEMC limit are evaluated using the equation below. If more than one elevated area is identified in a given survey unit, the unity rule can be used to determine compliance. If the equation value is less than unity, no further elevated area testing is required and the EMC test (shown below) is satisfied:

(average concentrationin elevated area - E) < 1.0 DCGL w + (Area Factor)(DCGLO)

Where: 8 is the average residual activity in the survey unit.

When calculating 8 for use in this inequality, measurements falling within the elevated area may be excluded provided the overall average in the survey unit is less than the DCGLw.1 For embedded piping, the number of elevated areas will be limited to ensure that the total inventory remaining in each EP will be maintained at the level that would be present if the entire pipe were contaminated at the DCGL level.

Compliance with the soil DCGLEMC will be determined using the FSS gamma spectroscopy results and a unity rule approach. These general methods will also be applied to other materials where sample gamma spectroscopy is used for FSS.

The application of the unity rule to the elevated measurement comparison requires MARSSIM, NUREG-1575, Revision 1,(June 2001), Section 8.5.2, per the EPA website at www.epa. gov/radiation/marssimn/docs/revisionl.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 area factors and corresponding DCGLEMC to be calculated for Cs-137, Co-60, and any other gamma emitter identified during FSS, separately.

The methods used to calculate the nuclide specific soil area factors are presented in Attachment B. These area factors are used to determine DCGLEMC for Co-60, Cs-137, and any other identified gamma emitter, for each elevated area being evaluated during FSS. The surrogate radionuclides will be conservatively accounted for through the application of the Cs- 137 area factor to the surrogate Cs-137 DCGL since the hard-to-detect radionuclides have higher area factors than Cs-137. The DCGLEMC are used as follows to determine compliance with the elevated measurement comparison. Background could be subtracted from each radionuclide concentration if necessary:

Cs-137 Co-60 RN <

Cs - 137 DCGLEMC Co - 60 DCGLEC DCGLEMCN Where: Cs-137 and Co-60 are the gamma spectroscopy results from FSS, DCGLEmC- is calculated for the size of the elevated area being evaluated, RN is any other gamma emitter identified during FSS, and DCGLEMCN is the DCGLEMC for radionuclide N.

8.4 Remediation and Reclassification As shown in Table 8-2, for any classification (1, 2 or 3), areas of elevated residual activity above the DCGLEMc are remediated to reduce the residual radioactivity to acceptable levels. Whenever an investigation confirms activity above an action level listed in Table 8-2, an evaluation of the historical site assessment, operational history, design information, and sample results will be performed and documented. The evaluation will consider: (1) the elevated area's location, dimensions, and sample results, (2) an explanation as to the potential cause and extent of the elevated area in the survey unit, (3) the recommended extent of reclassification, if considered appropriate, and (4) any other required actions.

Areas that are reclassified as Class 1 are typically bounded by a Class 2 buffer zone to provide further assurance that the reclassified area completely bounds the elevated area. This evaluation process is established to avoid the unwarranted reclassification of an entire survey unit (which can be quite large) while at the same time requiring an assessment as to extent and reasons for the elevated area.

If remediation is performed in a Class 1 survey unit, the survey is redesigned and re-performed accordingly. If an individual survey measurement (scan, static, or material sample) in a Class 2 survey unit exceeds the DCGL, the survey unit or a 8-4

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 portion of it shall be reclassified to a Class 1 and the survey redesigned and re-performed accordingly. If an individual survey measurement in a Class 3 survey unit exceeds 0.5 DCGL, the survey unit, or portion of a survey unit, will be evaluated, and if necessary, reclassified to a Class 2 and the survey redesigned and re-performed accordingly.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 9.0 DATA COLLECTION AND PROCESSING 9.1 Sample Handling and Record Keeping A chain-of-custody record accompanies each sample from the point of collection through obtaining the final results to ensure the validity of the sample data.

Chain-of-custody records are controlled and maintained and, upon completion of the data cycle, are transferred to Document Control, in accordance with applicable procedures.

Each survey unit's Survey Design Package will be controlled in accordance with FSS implementing procedures. These procedures address the design and field implementation of the survey requirements. Survey unit records are quality records.

9.2 Data Management Measurements performed during characterization, turnover and investigation surveys can be used as FSS data if they are performed according to the same requirements as the final survey data. These requirements include: (1) the survey data is representative of the as-left survey unit condition and is not impacted by further remediation; (2) the application of isolation measures to the survey unit to prevent re-contamination and to maintain final configuration; and (3) the data collection and design were in accordance with FSS methods (e.g., MDCscan, investigation levels, survey data point number and location, statistical tests, and EMC tests).

Measurement results stored as FSS data constitute the final survey of record and are included in the data set for each survey unit used for determining compliance with the site release criteria. Measurements are recorded in units appropriate for comparison to the DCGL. The recording units for surface contamination are dpm/100 cm2 and pCi/g for activity concentrations. Numerical values, including negative numbers, are recorded.

Document Control procedures establish requirements for record keeping.

Measurement records include, at a minimum, the surveyor's name, the location of the measurement, the instrument used, measurement results, the date and time of the measurement and any surveyor comments.

9.3 Data Verification and Validation The FSS data are reviewed before data assessment to ensure that they are complete, fully documented and technically acceptable. The review criteria for data acceptability will include at a minimum, the following items:

• Compliance with survey instructions as specified in the survey package 9-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0

• The instrumentation MDC for fixed or volumetric measurements was below the DCGLW or if not, below the DCGLEMC for Class 1, below the DCGLW for Class 2 and below 0.5 DCGLW for Class 3 survey units.

• The instrument calibration was current and traceable to NIST standards.

• The field instruments were source checked with satisfactory results before and after use each day data are collected or if unsatisfactory, data obtained with that instrument since its previous acceptable performance check was evaluated for acceptability.

• The MDCs and assumptions used to develop them were appropriate for the instruments and techniques used to perform the survey.

• The survey methods used to collect data were proper for the types of radiation involved and for the media being surveyed.

• "Special methods" for data collection were properly applied for the survey unit under review.

• The chain-of-custody was tracked from the point of sample collection to the point of obtaining results.

• The data set is comprised of qualified measurement results collected in accordance with the survey design, which accurately reflects the radiological status of the facility.

• The data has been properly recorded.

If the data review criteria were not met, the discrepancy will be reviewed and the decision to accept or reject the data will be documented.

9.4 Graphical Data Review Survey data may be graphed to identify patterns, relationships or possible anomalies which might not be apparent using other methods of review. A posting plot or a frequency plot may be made. Other special graphical representations of the data will be made as needed.

9.4.1 Posting Plots Posting plots may be used to identify spatial patterns in the data. The posting plot consists of the survey unit map with the numerical data shown at the location from which it was obtained. Posting plots can reveal patches of elevated radioactivity or local areas in which the DCGL is exceeded. Posting plots can be generated for background reference areas to point out spatial trends that might adversely affect the use of the data.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 Incongruities in the background data may be the result of residual, undetected activity, or they may just reflect background variability.

9.4.2 Frequency Plots Frequency plots may be used to examine the general shape of the data distribution. Frequency plots are basically bar charts showing data points within a given range of values. Frequency plots reveal such things as skewness and bimodality (having two peaks). Skewness may be the result of a few areas of elevated activity. Multiple peaks in the data may indicate the presence of isolated areas of residual radioactivity or background variability due to soil types or differing materials of construction. Variability may also indicate the need to more carefully match background reference areas to survey units or to subdivide the survey unit by material or soil type.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 10.0 DATA ASSESSMENT AND COMPLIANCE An assessment is performed on the FSS data to ensure that they are adequate to support the determination to release the survey unit. Simple assessment methods such as comparing the survey data to the DCGL or comparing the mean value to the DCGL are first performed. The statistical tests are then applied to the final data set and conclusions are made as to whether the survey unit meets the site release criterion.

10.1 Data Assessment Including Statistical Analysis The results of the survey measurements are evaluated to determine whether the survey unit meets the release criterion. In some cases, the determination can be made without performing complex, statistical analyses.

10.1.1 Interpretation of Sample Measurement Results An assessment of the measurement results is used to quickly determine whether the survey unit passes or fails the release criterion or whether one of the statistical analyses must be performed. The evaluation matrices are presented in Tables 10-1 and 10-2.

Table 10-1 Interpretation of Sample Measurements When the WRS Test Is Used Measurement Results Conclusion Difference between maximum survey unit Survey unit meets concentration and minimum reference area concentration is less than the DCGLW Difference of survey unit average concentration and reference average Survey unit fails concentrations greater than the DCGLW Difference between any survey unit concentration and any reference area Conduct WRS test and concentration is greater than the DCGLW elevated measurements and the difference of survey unit average test concentration and reference area average concentration is less than the DCGLw Jo-I

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Table 10-2 Interpretation of Sample Measurements When the Sign Test is Used Measurement Results Conclusion All concentrations less than the DCGLW Survey unit meets release criterion Average concentration greater than the Survey unit fails DCGLw Any concentration greater than the Conduct Sign Test and DCGLw and average concentration less elevated measurements than the DCGLw test When required, one of four statistical tests will be performed on the survey data:

• WRS Test

• Sign Test

• WRS Test Unity Rule

• Sign Test Unity Rule In addition, survey data are evaluated against the EMC criteria as previously described in section 8.3 and as required by NUREG 1727. The statistical test is based on the null hypothesis that the residual radioactivity in the survey unit exceeds the DCGL. There must be sufficient survey data at or below the DCGL to reject the null hypothesis and conclude the survey unit meets the site release criterion for dose. Statistical analyses may be performed using a computer software program or, if necessary, using hand calculations.

10.1.2 Wilcoxon Rank Sum Test The WRS test, or WRS Unity Rule (NUREG-1505, Chapter 11), may be used when the radionuclide of concern is present in the background or measurements are used that are not radionuclide-specific. In addition, this test is valid only when "less than" measurement results do not exceed 40 percent of the data set.

The WRS test is applied as follows:

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0

1. The background reference area measurements are adjusted by adding the DCGLw to each background reference area measurement, Xi, Zi = Xi + DCGL

2. The number of adjusted background reference area measurements, m, and the number of survey unit measurements, n, are summed to obtain N, (N = m + n).

3. The measurements are pooled and ranked in order of increasing size from 1 to N. If several measurements have the same value, they are assigned the average rank of that group of measurements.

4. The ranks of the adjusted background reference area measurements are summed to obtain Wr.

5. The value of Wr is compared with the critical value in Table I.4 of NLJREG-1575 (MARSSIM). If W, is greater than the critical value, the survey unit meets the site release dose criterion. If Wr is less than or equal to the critical value, the survey unit fails to meet the criterion.

10.1.3 Sign Test The Sign Test and Sign Test Unity Rule are one-sample statistical tests used for situations in which the radionuclide of concern is not present in background, or is present at acceptable low fractions compared to the DCGLw. If present in background, the gross measurement is assumed to be entirely from plant activities. This option is used when it can be reasonably expected that including the background concentration will not affect the outcome of the Sign Test. The advantage of using the Sign Test is that a background reference area is not needed. The Sign Test may also be applied to net values after material and/or ambient background subtraction The Sign test is conducted as follows:

1. The survey unit measurements, Xi, i = 1, 2, 3, ...N; where N = the number of measurements, are listed.

2. Xi is subtracted from the DCGLw to obtain the difference Di = DCGLW - Xi , i = 1, 2, 3,..., N.

3. Differences where the value is exactly zero are discarded and N is reduced by the number of zero measurements.

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4. The number of positive differences is counted. The result is the test statistic S+. Note that a positive difference corresponds to a measurement below the DCGLw and contributes evidence that the survey unit meets the site release criterion.

5. The value of S+ is compared to the critical value given in Table 1.3 of NUREG-1575 (MARSSIM). The table contains critical values for given values of N and a. The value of a is set at 0.05 during survey design. If S+ is greater than the critical value given in the table, the survey unit meets the site release criterion. If S+ is less than or equal to the critical value, the survey unit fails to meet the release criterion.

10.1.4 Unity Rule The Cs-137 to Co-60 ratio will vary in the final survey soil samples, and this will be accounted for using a "unity rule" approach as described in NUREG-1505 Chapter 11. Unity Rule equivalents will be calculated for each measurement result using the surrogate adjusted Cs-137 DCGL and the adjusted Co-60 DCGL, as shown in the following equation:

Ci/DCGLi +C 2 /DCGL 2 +.... C,/DCGL, < l Where: C. = concentration of radionuclide n and DCGLn = DCGL of radionuclide n The unity rule equivalent results will be used to demonstrate compliance assuming the DCGL is equal to 1.0 using the criteria listed in Tables 10-1 and 10-2. If the application of the WRS or Sign test is necessary, these tests will be applied using the unity rule equivalent results and assuming that the DCGL is equal to 1.0. An example of a test using the unity rule is provided in NUREG-1505.

10.2 Data Conclusions The results of the statistical tests, including application of the EMC, allow one of two conclusions. The first conclusion is that the survey unit meets the unrestricted use criterion. The data provides statistically significant evidence that the level of residual radioactivity in the survey unit does not exceed the release criterion. The decision to release the survey unit is made with sufficient confidence and without further analysis.

The second conclusion that can be made is that the survey unit fails to meet the release criterion. The data are not conclusive in showing that the residual 10-4

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 radioactivity is less than the release criterion. The data are analyzed further to determine the reason for the failure.

Possible reasons include:

1. The average residual radioactivity exceeds the DCGL, or

2. The test did not have sufficient power to reject the null hypothesis (i.e., the result is due to random statistical fluctuation).

The power of the statistical test is a function of the number of measurements made and the standard deviation in measurement data. The power is determined from 1-j where f is the value for Type II errors. A retrospective power analysis will be perforned using the methods described in Appendices I.9 and I.10 of NUREG-1575 (MARSSIM). A greater number of measurements increase the probability of passing if the survey unit actually meets the release criterion. If failure was due to the presence of residual radioactivity in excess of the release criterion, the survey unit must be remediated and resurveyed.

10.3 Compliance The FSS is designed to demonstrate that licensed radioactive materials have been removed from PBRF facilities and property to the extent that residual levels of radioactive contamination are below the radiological criteria for unrestricted use as approved by the NRC. The site-specific radiological criteria presented in this plan demonstrate compliance with the criteria of 10 CFR 20.1402.

If the measurement results pass the requirements of Tables 10-1 and 10-2 of section 10.1.1, and the elevated areas evaluated per section 8.3 pass the elevated measurement comparison, then the survey unit is suitable for unrestricted release.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 11.0 REPORTING FORMAT Survey results are documented in history files, survey unit release records, and in the Final Status Survey Report (FSSR). Other reports may be generated as requested by the NRC.

11.1 History File A history file of relevant operational and decommissioning data should be compiled. The history file should consist of relevant historical site assessment information, Characterization Survey data and the classification basis. The purpose of the history file is to provide a substantive basis for the survey unit classification, and hence, the level of intensity of the FSS.

11.2 Survey Unit Release Record A separate release record is prepared for each survey unit. The survey unit release record is a stand-alone document containing the information necessary to demonstrate compliance with the site release criteria. This record includes:

• Description of the survey unit

• Survey unit design information

• Survey unit measurement locations and corresponding data

• Survey unit investigations performed and their results

• Survey unit data assessment results

• Documentation of evaluations pertaining to compliance with the unrestricted use limit of 25 mrem/yr and dose contributions from EP and radionuclides contributing 10% in aggregate of the total dose for both structural scenarios and soils A check-list may be employed as part of the implementing procedure to ensure all required information and evaluations are contained within each survey unit release record. When a survey unit release record is given final approval, it becomes a quality record.

11.3 Final Status Survey Report FSS results will be described in a written report to the NRC. The actual structures, land, or system included in each written report may vary depending on the status of ongoing decommissioning activities.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 The FSS report provides a summary of the survey results and the overall conclusions, which demonstrate that the PBRF facility and site meet the radiological criteria for unrestricted use. Information such as the number and type of measurements, basic statistical quantities, and statistical analysis results are included in the report. The level of detail is sufficient to clearly describe the FSS program and to certify the results. The format of the final report will contain the following topics as applicable:

• Overview of the results

• Discussion of changes to the FSS from what was proposed in the FSSP or other prior submittals

• FSS methodology

• Survey unit sample size

• Justification for sample size

• FSS Results

• Number of measurements taken

• Maps identifying measurement locations

• Sample concentrations

• Statistical evaluations, including power curves

• Judgmental and miscellaneous data sets

• Anomalous data

• Conclusion for each survey unit

• Any changes from initial assumptions on extent of residual activity 11-2

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 12.0 FSS QUALITY ASSURANCE & QUALITY CONTROL (QA/QC)

The objective of QA/QC implementation is to ensure that the survey data collected are of the type and quality required to demonstrate that the dose from residual contamination is below the 25 mrem/yr unrestricted use criteria and to support a decision to terminate the PBRF license. Proper application of QA/QC will ensure that: 1) the elements of the FSS plan are implemented in accordance with the approved procedures; 2) surveys are conducted by trained personnel using calibrated instrumentation; 3) the quality of the data collected is adequate; 4) all phases of survey package design and implementation are properly reviewed, and oversight is provided; and 5) corrective actions, when identified, are implemented in a timely manner and are determined to be effective.

12.1 Project Management and Organization NASA's Glenn Research Center (GRC) is responsible for overall execution of the PBRF Decommissioning Project and for public health and safety. As licensee, GRC provides oversight of the decommissioning effort and has established a team of qualified. personnel to execute decommissioning of the PBRF. Complete details regarding roles and responsibilities are provided in Section 2.0 of the Decommissioning Plan.

The decommissioning team will provide the necessary personnel, materials, and subcontractors to perform all phases of decommissioning work including performance of the FSS. Trained and experienced personnel will perform the FSS in accordance with the protocols and process presented in this plan using written survey instructions and approved procedures. Specific members that are directly involved with the FSS are listed below.

12.1.1 NASA Project Radiation Safety Officer (RSO)

The NASA Project RSO is responsible for organizing, administering, and directing the radiation protection program at the PBRF during decommissioning activities, including radiation safety, environmental health, and FSS Program. Additional information is provided in Section 2.0 of the Decommissioning Plan.

12.1.2 FSS/Characterization Manager The FSS/Characterization Manager is responsible for the organization, administration, development, and implementation of the FSS program under the FSS Plan. He/she is also responsible for technical review of Survey Design Packages and development of the Final Status Survey Report. The FSS/Characterization Manager reports to the NASA Project RSO.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 12.1.3 FSS/Characterization Engineers FSS/Characterization Engineers provide technical direction for the FSS, including development and closure of FSS Design Packages, data evaluation, procedure development, and development of survey unit release records and history files. FSS/Characterization Engineers report to the FSS/Characterization Manager.

12.1.4 FSS/Characterization Supervisors FSS/Characterization Supervisors are responsible for assisting in the control and implementation of Survey Design Packages as received from the FSS/Characterization Engineers, implementation of turnover surveys, final status surveys, survey area preparation (such as gridding and accessibility needs), implementing access controls for FSS survey areas, coordination and scheduling of FSS/Characterization Technicians to support the FSS schedule and ensuring all necessary instrumentation and other equipment is available to support survey activities.

FSS/Characterization Supervisors report to the FSS/Characterization Manager.

12.1.5 Radiological Laboratory Manager The Radiological Laboratory Manager is responsible for radiological sample analyses performed in support of FSS. These analyses include, but are not limited to, volumetric soil samples and loose surface contamination activity media. The Radiological Laboratory Manager is responsible for the operation of laboratory instrumentation in accordance with approved procedures and manufacturers' recommendations, ensuring instrument QC and MDC requirements are met which support FSS criteria. The Radiological Laboratory Manager will report to the NASA Project RSO and consult with the FSS/Characterization Manager when providing direction and support for the FSS.

12.1.6 FSS/Characterization Technicians The FSS/Characterization Technicians are responsible for performance and documentation of FSS surveys and sampling in accordance with the requirements provided in the FSS implementing procedures. Technicians are responsible for performing work according to the survey package instructions and identifying any discrepancies with performance of the survey package. Through compliance with the FSS implementing procedures, Technicians shall implement the requirements contained in this FSS Plan to ensure appropriate quality is used in the collection of data for FSS decision-making. FSS/Characterization Technicians report to the FSS/Characterization Supervisor.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 12.2 Training Training is conducted to achieve initial proficiency and to maintain that proficiency throughout the FSS process. The training ensures that personnel have sufficient knowledge to perform work activities in accordance with the requirements of this FSSP. Specific details of the FSS training program will be included in a PBRF procedure. At a minimum, personnel assigned to acquire FSS survey data will be trained on the following:

• Initial set-up, pre and post-use performance checks, and operation of the data-logger instrument and associated detectors

• Performance of static and scan measurements

• Performance of volumetric material sampling, including sample identification and chain-of-custody

• Area isolation and control

• Sample processing and analysis

• Survey documentation 12.3 Written Procedures The FSS of the PBRF will be conducted in accordance with written, approved procedures to ensure quality. Approved procedures will describe the methods and techniques used to perform the FSS. Procedures will be developed, implemented and controlled in accordance with a written PBRF procedure.

12.4 Access Control of Surveyed Areas and Systems Prior to performing the FSS, the survey unit or area is isolated and controlled to prevent changes in radiological conditions. Routine access, material storage, and worker and material transit through the area are no longer permitted without proper controls. One or more of the following administrative and physical controls will be established to minimize the possibility of introducing radioactive material from ongoing decommissioning activities in adjacent or nearby areas.

• Personnel training

• Installation of barriers to control access

• Installation of postings with access/egress requirements

• Locking or otherwise securing entrances to the area 12-3

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• Installation of tamper-evident seals or labels Isolation and control measures are implemented through approved PBRF procedures and remain in place through the FSS data collection process until license termination.

12.5 Chain-of-Custody Responsibility for custody of samples from the point of collection through the determination of the final survey results is established by implementing procedures. FSS sample identification and chain-of-custody will ensure that the validity of material samples remains intact.

Samples will be accompanied by a chain-of-custody (CoC) form. Sample identification numbers will be listed on the CoC form. When transferring possession of samples, the individuals relinquishing and receiving will sign, date, and note the time on the CoC form. The CoC form documents transfer of custody of samples from the sampler to another person, to a laboratory, or to/from a secure storage area. Secure storage will be provided for archived samples.

12.6 Instrumentation Selection, Calibration and Operation Proper selection and use of instrumentation will ensure that sensitivities are sufficient to detect radioactivity at the minimum detection requirements specified by the survey design which will assure the validity of the survey data. Instrument calibration will be performed with NIST traceable sources using approved procedures. Issuance, control and operation of the survey instruments will be conducted in accordance with approved implementing procedures.

12.7 Quality Control Surveys and Samples QC replicate field measurements and replicate sample analyses will be performed to assess precision. If a replicate result falls outside the acceptance criteria, an investigation is performed. The investigation may include verification that the data set is correct, the relevant instruments were operating properly, the survey/sample points were properly identified and located, instructions and procedures were followed, and sample chain-of-custody remained intact. When deemed appropriate, additional measurements are taken. At the conclusion of the investigation, a determination is made regarding the usability of the survey or sample data. The investigation and subsequent corrective actions are documented in accordance with PBRF procedures.

12.7.1 Replicate Field Measurements Replicate field measurements will be used to monitor survey precision in the field. A minimum 5% of survey measurements will be selected for 12-4

Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 performance of replicate static and scan surveys. Acceptance criteria for scan measurements are both the QC measurement and initial FSS measurement must reach the same conclusion (i.e., both measurements are either less than or greater than the investigation level). Acceptance criterion for static measurements is both measurements must reach the same conclusion and the relative percent difference (RPD) must be within 20%. RPD is calculated as follows:

RPD _ I(Di-D 2 )x 100 (D,+D 2 )/2 Where: DI = 1 st data result D2 = 2nd data result 12.7.2 Replicate Sample Analyses Replicate sample analyses will be used to monitor sample precision. A minimum of one out of every twenty (5%) FSS samples collected will be reanalyzed or homogenized, split and submitted for independent analyses.

Similar guidance contained in NRC Inspection Manual 84750 (IP 84750) will be used for evaluation of replicate sample analyses. This evaluation is performed as follows:

1. Determine the resolution for each known nuclide concentration by dividing the initial activity by its corresponding I a uncertainty.

2. Determine the ratio of each nuclide concentration by dividing the duplicate sample result by the initial sample result.

3. The results are acceptable if the agreement ratio falls within the values given in Table 12-1 for the corresponding resolution.

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Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 Table 12-1 NRC Criteria for Accepting Sample Measurements Resolution Acceptable Ratio

<4 0.4-2.5 4-7 0.5-2.0 8-15 0.6-1.66 16-50 0.75-1.33 51-200 0.80-1.25

>200 0.85-1.18 12.7.3 Control of Vendor-Supplied Services Vendor-supplied services, such as instrument calibration and laboratory sample analysis, will be procured from appropriate vendors in accordance with approved quality and procurement procedures.

12.7.4 Database Control Software used for data reduction, storage or evaluation will be fully documented and certified. The software will be tested prior to use by an appropriate test data set. Programs developed to assist in calculating FSS data (i.e., Excel spreadsheets) shall also be tested to verify they are correct.

12.7.5 Assessment and Oversight 12.7.5.1 Assessments FSS assessments will be conducted in accordance with approved procedures. The findings will be tracked and trended in accordance with these procedures.

12.7.5.2 Corrective Action Process The corrective action process will be applied to all aspects of the FSS. FSS deficiencies will be fully investigated and documented in accordance with the applicable corrective action procedure.

12.7.6 Data Validation Survey data will be reviewed by FSS/Characterization management for completeness, procedural compliance, and the presence of outliers.

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0 Comparisons to investigation levels will be made and measurements exceeding the investigation levels will be evaluated. Procedurally verified data will be subjected to the Sign test, the Wilcoxon Rank Sum (WRS) test, Sign Unity or WRS Unity test as appropriate.

12.7.7 NRC Confirmatory Measurements PBRF anticipates that both the NRC and OEPA may choose to conduct confirmatory measurements in accordance with applicable laws and regulations. The NRC may perform confirmatory measurements to determine if FSS and associated documentation demonstrate that the facility and site are suitable for release in accordance with the criteria for decommissioning in 10 CFR Part 20, subpart E. Timely and frequent communications with these agencies will ensure that they are afforded sufficient opportunity for these confirmatory measurements prior to PBRF implementing any irreversible decommissioning actions that have an impact on FSS (e.g. backfilling open excavations, etc.).

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Plum Brook ReactorFacility FinalStatus Survey Plan Revision 0

13.0 REFERENCES

Appendix A Technical Specifications for the License No. TR-3, National Aeronautics and Space Administration Plum Brook Station, Sandusky, Ohio, Docket No. 50-30, 1998 NUREG-1575, "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSIM), Revision 1 (June 2001)

Title 10 Code of Federal Regulations, Part 20.1402, "Radiological Criteria for Unrestricted Use" Title 10 Code of Federal Regulations, Part 20, Subpart E, "Radiological Criteria for License Termination" NUREG-1727, "NMSS Decommissioning Standard Review Plan," September 15, 2000 NUJREG-1757, "Consolidated NMSS Decommissioning Guidance", September 2003 ANL/EAD/TM-89 "Probabilistic dose analysis using parameter distributions developed for RESRAD and RESRAD-BUILD codes", Argonne National Laboratory, 2000 ANL/EAD/TM-98 "Development of probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 computer codes", Argonne National Laboratory, 2000 "Decommissioning Plan for the Plum Brook Reactor", Revision 2, October 2001 Title 10 Code of Federal Regulations, Part 50.82, "Termination of License" Draft Regulatory Guide DG-4006, "Demonstrating Compliance with the Radiological Criteria for License Termination" NUR-EG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys," Rev.1, June 1998 draft Tetra Tech, Inc., "Final Environmental Baseline Survey Report for the Plum Brook Reactor Facility Decommissioning Project", February 2001 PBRF-CS-016, "Material Background Determination" ISO 7503-1, "International Standard, Evaluation of Surface Contamination" MW-PL-02-008, "Team Quality Control Plan" "Plans of Buildings and Structures, Lewis Research Center, Plum Brook Station, Sandusky, OH", June 1974 13-1

Plum Brook Reactor Facility FinalStatus Survey Plan Revision 0 EPA/SW-846, Test Methods for Evaluating Solid Waste, Volume IA: Laboratory Manual Physical/Chemical Methods, 1986 updated 1992 United States Nuclear Regulatory Commission Inspection Manual, Inspection Procedure 84750 13-2