ML15299A385

AJBRF-FSS-01, Revision B, Final Status Survey Plan Aj Blotcky Reactor Facility Decommissioning Project

ML15299A385
Person / Time
Site:	05000131
Issue date:	10/14/2015
From:	Brautigam T Enercon Federal Services, NorthStar Federal Services
To:	Office of Nuclear Material Safety and Safeguards, US Dept of Veterans Affairs, Nebraska-Western Iowa Health Care System
Shared Package
ML15299A433	List: ML15299A384 ML15299A385
References
AJBRF-FSS-01, Rev. B
Download: ML15299A385 (97)

Text

North Star Federal Services Final Status Survey Plan AJ Blotcky Reactor Facility Decommissioning Project.......

AJBRF-FSS-0 1 Rev. B Approved October 5, 2015 Prepared, by"...

NorthStar Feder:al Services, Inc.

1992"'Saint Street Richiand, WA 99354

• and

.:Enercon Federal Services, Inc.

,,679 Emery Valley Road, Suite A "Oak Ridge, TN 37830

~Prepared for

,,,..*:FUnited States Department of Veterans Affairs Nebraska-Western Iowa Health Care System Contract Number VA701-C-15-0005 NorthStar Project Number 1554005

AJ Blotcky Reactor Facility Final Status Survey Plan AJ Blotcky Reactor Facility Decommissioning Project Final Status Survey Plan AJBRF-FSS-01 Rev. B Approved October 5, 2015 Prepared by:

Reviewed by:

Reviewed by:

Approved by:

Approved by:

Todd S. Brautigam

?

,I2 Radiation Safety Officer Billy Reid NorthStar Program Manager ENERCON Program Manager*

Nelson Langub Q,*l NorthStar Project Manager ">.

Chanda Joshi;'VA'COR-:

Department of Veterans Affairs ii

AJ Blotcky Reactor Facility Final Status Survey Plan Summary of Changes Revisions to this Final Status Survey Plan will be tracked, and revisions or addenda will be issued as needed. The Project Manager maintains the signed original of the FSSP; no controlled copies are issued.

The end user is responsible to verify' with the Project Manager that any hardcopy being referenced is the current revision.

A summary description of each revision or addenda will be noted in the following table.,,

Revision Number Date Comments° Rev. A Draft July 28, 201 5 Original Issue "i

Rev. B Draft October 5, 2015 Addresses VA/AECOM comments Rev. B Approved October 14, 2015 Approved by VA

°°.

AJ Blotcky Reactor Facility Final Status Survey Plan Table of Contents PAGE 1.0 TNTRODUCTION................................................................................ 1 2.0 RESPONSIBILITIES............................................................................ 2 3.0 FINAL STATUS SURVEY PLAN............................................................. 3 3.1 Purpose............................................................................................. 3 3.2 Overview 3

3.2.1 Survey Preparation................................................................. i'2........3 3.2.2 Survey Design.......................................................

7:....... :......

.i......

4 3.2.3 D ata Collection 4.....................

3.2.4 Data Assessment and Evaluation..............

4 3.2.5 Documentation of Survey Results.......

5 3.3 Implementation..........................5 3.4 Radionuclides of Concern.............i...,...................................................... 6 3.5 Area Classification..........i.... ?..."........................................................... 6 3.5.1 Non-Impacted Areas..... :2,..i.l............................................................ 7 3.5.2 Impacted Areas.....,...,-i....................................................... 7 3.5.3 Changes in Classification................................................................... 9 3.5.4 Redefining Survey Boundaries.............................................................. 9 3.6 Establishling Survey Units........................................................................ 9 3.6.1 SurveyUnit................................................................................. 10 3.7 Survey Design.................................................................................... 12 3.7.1 Scan Coverage.............................................................................. 12 3.7.2 Sample Size Determination................................................................ 13 3.7.3 Background Reference Areas.............................................................. 18 3,7.4 Sample Grid and Measurement Location................................................. 19 iv

AJ Blotcky Reactor Facility Final Status Survey Plan 3.7.5 Survey Package Design Process........................................................... 19 3.8 Types of Surveys................................................................................. 23 3.8.1 Scan Surveys................................................................................ 23 3.8.2 Direct Measurements....................................................................... 24 3.8.3 Exposure Rate............................................................................... 24 3.8.4 Removable Activity.................................................................. iii:.....25 3.8.5 LSC Analysis.............................................................

ii...;..

....... i..25 3.8.6 Volumetric Samples.................................................;....ii..*

............. 25 3.9 Survey Methods.........................................................

,,:........................ 25 3.9.1 Buildings, Equipment, and Components...........

,... :,................................. 26 3.10 Instrumentation................................................................................... 26 3.10.1 Selection.......................

,..... :..,.................................................... 27 3.10.2 Calibration and Maintenance...i..... 2....:................................................ 27 3.10.3 Response Checks..............

............................................................. 28 3.10.4 Minimum Detectable Concentration for Direct Measurements........................ 28 3.10.5 Scan Measurement MDC;............................... Error! Bookmark not defined.

3.11 Investigation Levels and.El'evated Areas Test................................................. 28 3.11.1 Investigation Levels........................................................................ 29 3.1,1.2 Investigation Process....................................................................... 29 3.11.3 Elevated Measurement Comparison (EMC)............................................. 30 3.11.4 Remediation and Reclassification......................................................... 31 3.1 1.5 Reclassification and Resurvey............................................................. 32 3.1I2 Data Collection and Processing................................................................ 32 3.12.1 Sample Handling and Record Keeping................................................... 32 3.12.2 Data Management.......................................................................... 33 v

AJ Blotcky Reactor Facility Final Status Survey Plan 3.12.3 Data Verification and Validation.......................................................... 33 3.12.4 Graphical Data Review.................................................................... 34 3.13 Data Assessment and Compliance.............................................................. 34 3.13.1 Data Evaluation............................................................................. 34 3.14 Statistical Conclusions........................................................................... 37 3.15 Reporting Format.............................................................. "':""'"....... ::... 38 3.15.1 Survey Unit Release Record...............................................

.,.i............38 3.15.2 Final Status Survey Report............................. i........... !........................ 39 4.0 SURVEY INSTRUMENTATION................

40 4.1 Overview........................................................................................... 40 4.2 Instrument Statistics...................... :..:.................................................... 41 4.2.1 Total Detection Efficiency..........

i...... i................................................ 43 4.2.2 Minimum Detectable Concentration...................................................... 44 4.3 Instrumnentation Overview....................................................................... 47 4.3.1 Radiological Instruments................................................................... 48 4.3.2 Radiation Detectors..:..................................................................... 49 4.4 Calibration and Maintenance................................................................... 49 4.4.1 Response Checks........................................................................... 50 5.0 QUALITY ASSURANCE..................................................................... 51 5.1 Project Description and Schedule.............................................................. 51 5.2 Quality Objectives and Measurement Criteria................................................. 51 5.2.1 Training and Qualification................................................................. 51 5.2.2 Survey Documentation..................................................................... 51 5.3 Measurement/Data Acquisition................................................................. 51 5.3.1 Survey Design.............................................................................. 51 vi

AJ Blotcky Reactor Facility Final Status Survey Plan 5.3.2 Written Procedures......................................................................... 52 5.3.3 Sampling Methods.......................................................................... 52 5.3.4 Chain of Custody........................................................................... 52 5.4 QA/QC Surveys and Samples................................................................... 52 5.4.1 Replicate Measurements................................................................... 53 5.4.2 Volumetric Analyses......................53 5.5 Instrument Selection, Calibration and Operation..................................

.. :;.......53 5.6 Control of Consumables............................................................

.............. 54 5.7 Control of Vendor-Supplied Services.................

i...... '..........

........ 54 5.8 Software Control....................................... :.......i.................................. 54 5.9 Data Management......................

.......................................................... 54 5.10 Assessment and Oversight..................-..... ::.......................

• ........................ 54 5.10.1 Assessments.............

.................................................................. 54 5.10.2 Independent Review of Survey Res'ults................................................... 54 5.10.3 Corrective Action Process.................................................................. 54 5.10.4 Reports to Management.................................................................... 55 5.11 Data Validation..........................................

55 5.12 ConfirmatorS Measurements.................................................................... 55

6.0 REFERENCES

..........................56 vii

AJ Blotcky Reactor Facility Final Status Survey Plan Index of Figures Figure 3-1: Interior Space Coordinate Example....................................................... 12 Index of Tables Table 3-1: Radionuclides of Concern..........................................................

• i.......... 6 Table 3-2: AJBRF MARSSIM Classifications for FSS...............................,i..i.......

Table 3-3: Survey Unit Size.....................................................

.:.............. 10 Table 3-4: Minimum Scan Coverage............................................. !.............. :........13 Table 3-5: MARSSIM Table 5.2..............................,........,..................................1!4 Table 3-6: MARSSIM Table 5.4.........................

2.......

............................. 1!6 Table 3-7: MVARSSIM Table 5.1..................

..................................................... 17 Table 3-8: Measurement Result InvestigatiOn Levfels...,............................................... 29 Table 4-1: FSSP Instrumentation.....

..... i.:...i.i...................................................... 41 Table 4-2: Typical Instrument Operational Paramneters............................................... 42 Table 4-3: Required Minimum Detectable Concentrations........................................... 47

,:,,,i

AJ Blotcky Reactor Facility Final Status Survey Plan Acronyms Acronym AJBRF AVL CFR COG DCGLernc DCGLw DP DPM DQO EMC ENERCON FSS FSSP GPS HTD LBGR MARSSIM MDA MDC MDCSoaIn NaI NIST.

PM QA/QC ROC SO TEDE USNRC VA VA-PM WRS Description A.J. Blotcky Reactor Facility Approved Vendor List Code of Federal Regulations Chain Of Custody Derived Concentration Guideline Level, Elevated Measurement Criteria Derived Concentration Guideline Limit, Wilcoxon Rank Sum Decommissioning Plan",*

Disintegrations Per Minute Data Quality Objectives Elevated Measurement Criteria ENERCON Services, Inc.

Final Status Survey Final Status Survey Plan Global Positioning System Hard To Detect Lower Boundary of the Gray Region NUREG-1575 'Multi-Agency Radiation SUrvey and SiteI1nvestigation Manual' Minimum Detectable Activity Minimum Detectable Concentration Minimum Detectable Concentration.via Scan ---

Sodium Iodide National Institute for Standards and Testing ENERCON Project Manager Quality Assurance/Quality Control Radionucldeis of Concern Safety Officer i

Total Effective Dose Equivalent United States Nuclear Regulatory Commission Veterans Administration Veterans Administration Project Manager Wilcoxon Rank Sum ix

• ili*i Alan J. Blotcky ti*

Reactor Facility

1.0 INTRODUCTION

The Veterans Health Administration (VA) is in the process of decommissioning of the Alan J.

Bloteky Reactor Facility (AJBRF). This plan is associated with control measures that are intended to enable the Final Status Survey (FSS) to be applied to the AJBRF building surfaces and any associated soils beneath or surrounding the structure. Decommissioning Planning for the AJBRF utilizes NUREG-1575, Revision 1 'Multi-Agency Radiation Survey and Site Investigation Manual' (MARSSIM) and other guidance documents. Additional resource documents are listed in Section 6.0, References.

This document describes the Final Status Survey Plan (FSSP) in terms of the process that will be used.

The goal of implementing the FSSP is to document the data collected by radiological surveys of the facility's final radiological status in support: 0f terminating the applicable radioactive materials license. These 'Final Status Surveys' thierefore provide the inputs feeding the MARS SIM statistical evaluation process. For the remainder-of this document, 'Final Status Survey' or 'FSS,' is synonymous with 'MARSSIM SUrVey.'

References are made in this text to site-or process-sPecific information such as characterization surveys, background studies, operational procedures, instrumentation, and other items. Further details of these specific issues are disculssed ini the appendices or other plans or reports currently in process. When applicable, references to these documents will be supplied or the documents attached to complete the chain of informhation.

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• i~i Alan J. Bloteky

~Reactor Facility 2.0 RESPONSIBILITIES Northstar will be responsible for all facets of the FSS including development of applicable plans and procedures governing performance of the FSS. ENERCON has provided two individuals knowledgeable in the decommissioning process. These individuals will function as the Project Radiation Safety Officer and Health and Safety Officer. During the project at least one of these individuals will be on-site during decommissioning activities. When only one is on-site, he will act as both the Radiation Safety Officer and Health and Safety Officer and is collectively referred to as the Safety Officer (SO) throughout the remainder of the FSSP.

The SO is responsible for preparing, implementing, and managing the FSS prograim.The SO has ultimate responsibility for program direction, technical content, andi, ensuring :tlie program complies with applicable United States Nuclear Regulatory Commnission (USNRC) regulations and guidance. The SO is responsible for resolving issues or concerns raisedt by USNRC, other regulatory agencies, or stakeholders, as well as programmatic issues raised by facility management. The SO reviews and approves the qualification and selection of FSS personnel and approves the content of training to FSS personnel and other personnel on FSS topics. The SO approves reports of FSS results.

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Alan J. Blotcky Reactor Facility 3.0 FINAL STATUS SURVEY PLAN 3.1 Purpose The purpose of the FSSP is to describe the survey process that will be used to demonstrate that the AJBRF facility and site comply with the USNRC annual dose limit criteria of 25-mrem/yr Total Effective Dose Estimate (TEDE), by meeting the radiological release criteria fr'om Section 6.0 of the approved Decommissioning Plan (DP) for the facility. The ultimate goal is to support license termination for the facility.

ii 3.2 Overview The FSS includes remaining structures, land, and reactor systems that are identified as contaminated or potentially contaminated as a result of licensed activities." There are 5 major steps in the final survey process:

1.

Survey preparation

2.

Survey design

3.

Data collection

4.

Data assessment and evaluation

5.

Documentation of survey results 3.2.1 Survey Preparation Survey preparation is the first step in the final survey process and occurs after remediation is completed. In areas where remediation was required, a turnover, or post-remediation, survey is performed to confirm that remediation was successful prior to initiating final survey activities. A turnover surveymay be performed using the same process and controls as FSS so that data from the turnover survey can be used as final survey data. In order for tumnover survey data to be used for FSS, it must have been designed and collected to meet quality assurance and quality control (QA/QC) criteria described in Section 5.0, Quality Assurance, and the area must be controlled in accordance with that section. Following the turnover surveys, the FSS is performed.

The area to be surveyed is isolated and/or controlled to ensure that radioactive material is not reintroduced into the area fr'om ongoing demolition or remediation activities nearby, and to maintain the final configuration of the area.

Tools, equipment, and materials not needed to support survey activities are removed, unless authorized by the SO. Routine access, material storage, and worker transit through the area are not allowed, unless authorized by the SO.

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• 'ii'*t..........

Al an J. Bloteky

• m

• • Reactor Facility However, survey areas may, with proper approval, be used for staging of materials and equipment providing; Staging does not interfere with survey performance.

Material or equipment is free of surface contamination or radioactive materials.

Survey personnel safety is not jeopardized.

The SO conducts an inspection of the area to ensure that work is complete and the area is ready for FSS.

Control of activities in that area is then transferred for the performance-of FSS.

Approved procedures provide isolation and control measures until the area is released for unrestricted use.

I 3.2.2 Survey Design The survey design process establishes the methods and performance criteria used to conduct the survey. Survey design assumptions are documented in "Survey Packages" in accordance with approved procedures. Structures and systems are organized into, survey areas and classified by contamination potential of the area. Survey unit size-is based on the assumnptions in the dose assessment models in accordance with the guidance provid~ed in the MARSSIM.

The percent coverage for scan surveys is discussed in Section,3.7.1.

The number and location of structural surface measurements (and/or structural volumetric samples) and soil samples (if necessary) are established in accordance with Section 3.7.2. Investigation levels are also established in accordance with Section 3.10.6.

Replicate measurements are performed as part of the quality process established to identify",

assess, and control errors and the uncertainty associated with sampling, survey, and/or analytical activities. This quality control measure, described in Section 5.4.1, provides assurance that the survey data meets the accuracy and reliability requirements necessary to support the decision to release or not release a survey unit.

3.2.3 Data Collection After preparation of a survey package, the FSS data are collected. Measurements are performed using calibrated instruments in accordance with approved procedures and instructions contained in the survey package.

3.2.4 Data Assessment and Evaluation Survey data assessment is performed to verify that the data are sufficient to demonstrate that the survey unit meets the unrestricted use criterion (i.e., the Null Hypothesis may be rejected.).

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Alan J. Bloteky

• tli~m*Reactor Facility Statistical analyses are performed on the data and the data are compared to investigation levels.

Depending on the results of an investigation, the survey unit may require further remediation, reclassification, and/or resurvey. Graphical representations of the data, such as posting plots or histograms, may be generated to provide qualitative information from the survey and to verify' the assumptions in the statistical tests, such as spatial independence, symmetmy,, data variance and statistical power. The assumptions and requirements in the survey package are reviewed by the SO. Additional data needs, if required, are identified during this review.

3.2.5 Documentation of Survey Results Survey results are documented by Survey Area in "Survey Packages." Each£'iSS package may contain the data from the several Survey Units that are within a given SurveyArea. :'The data is reviewed, analyzed, and processed, and the results documented in. a "'ReleaSe Record."

The Release Record provides the necessary information to support the decision to release the survey units for unrestricted use. A Final Survey Report is prepared that provides the necessary data and analyses from the Survey Packages and Release Records;, and is submitted by the licensee to the USNRC.-"

3.2.5.1 Contamination Event Resurvey::

If a contamination event occurs in an area that has received FSS, a resurvey of the area, or the affected portion, is required to demonstrate compliance with the release criteria. The resurvey needs to address the affected area with a reasonable surrounding buffer area to ensure the entire affected area has been evaluated. Thie data collection needs to be of sufficient quality and quantity to demonstrate compliance.

The survey quantity requirements, minimum detection levels, and investigation levels appropriate to a Class 1 area as listed in Section 3.10 apply, regardless of the original classification of the area.

The number of direct measurements or volumetric samples (as appropriate) is determined as discussed in Section 3.7.2, and is applied as necessary.

The survey data collected following remediation is appended to, or replaces when appropriate, the original FSS data. The entire data set is re-evaluated as per Section 3.12 to determine compliance with the release criteria. The results of the resurvey are included in the FSS report for the survey area.

3.3

_implementation The VA anticipates that the USNRC and other regulatomy agencies will choose to conduct confirmatory measurements in accordance with applicable laws and regulations. These agencies may take confirmatory measurements to make a determination in accordance with 10 CFR 5

Alan J. Bloteky I*':*!*Reactor Facility 50.82(a)(1 1) that the final radiation survey and associated documentation demonstrate that the facility and site are suitable for release in accordance with the criteria for decommissioning established in 10 CFR Part 20, Subpart E. The VA will comply with the 25 mrem/yr criteria of 10 CFR Part 20, Subpart E by complying with the radiological criteria described in the decommissioning planning document prepared for the facility. The confirmatoiy, measurements collected by the agencies are based upon the same release criteria.

Timely and frequent communications by the VA with these agencies ensure sufficient opportunity to,perform confirmatory measurements prior to the VA implementing difficult to reverse decommissioning actions such as filling an open excavation or" new construction."

3.4 Radionuclides of Concern and Release Criteria The approved Decommissioning Plan (DP) provides a list of Radionuciides of Concern (ROC) for use during implementation of the FSSP. Table 3-1 lists the ROC along with the approved release criteria originating in NUREG-1757 Appendix B. The release criteria listed below serves as Derived Concentration Guideline Levels (DCGL) input-parameters for the FSS calculations.

Table 3-1: Radionuclides of Concern and Release Criteria Building Surfaces Volumetric Radionuclide Release Criteria*

Release Criteria H-3 1.2E+08 DPM/100cm 2 110 pCi/g C-14 3.7E+06 DPM/100cm 2 12 pCi/g Fe-55 4.5E+06 DPM/l00cm2 10,000 pCi/g Co-60 7.1E+03 DPM/100cm 2 3.8 pCi/g Ni-63 1.8E+06 DPM/100cm 2 2,100 pCi/g Cs-137 2.8E+04 DPM!100cm 2 11 pCi/g Eu-15.2 None 6.9 pCi/g Eu-154 None 8 pCi/g 3.5,Area Classification Prior to beginning FSS, a thorough characterization of the radiological status and history of the site is completed. Initial classifications have been determined based on characterization surveys.

The structures and open land areas are classified following the guidance in Section 4.4 of the MARSSIM. Area classification ensures that the number of measurements and the scan coverage is commensurate with the potential for residual contamination to exceed the approved release criteria. Additional data from operational surveys performed in support of decommissioning, routine surveillances or similar applicable surveys may be used to change the initial classification of an area up to the time of initiation of the FSS, as long as the classification reflects the levels of 6

• ii*Alan J. Blotcky

,*Reactor Facility residual radioactivity that existed prior to any remediation activities. When the FSS of a given survey unit begins, the basis for any reclassification is documented, requiring a redesign of the survey unit package and the initiation of a new survey using the redesigned survey package. If, during the conduct of an FSS survey, sufficient evidence is accumulated to warrant an investigation and reclassification of the survey unit, the survey may be terminated without completing the survey unit package.

Initial classifications of the AJBRF are provided in Table 13.1 of the Decommissioning Plan.

Current classifications are shown in Table 3-2 below. All areas are subject to :changes in classification based on encountered radiological conditions.

Table 3-2: AJBRF MARSSIM Classifications forFSS MARSSIM Survey Area Classification Reactor tank wall Portions not removed 1

Reactor tank pit Exposed concrete and/or soil 1

Reactor water cooling system vault Floors and walls 1

Rooms B526, B535, B535A, B537, B540, and FalooS 12mtr B540AWalls

>2 fiheters and ceilings 3

Roos 533 ad B33A Floors and walls <2 meters 2

RomsB33 ndB33AWalls>2 meters and ceilings 3

Rooms B522 and B522A

-Floors only 3

Hall outside Room B526

'Floors only 2

Stairs on south side of Room B526 Floors only 2

Outside areas All Non-impacted 3.5.1 Non-Impacted Areas Non-Impacted areas have no reasonable potential for residual contamination because there was no known impact from ste operations. Non-impacted areas will not be required to be surveyed beyofid what is completed as a part of site characterization to confirm the area's non-impacted classification.-

3.5.2 Impacted Areas hnpacted areas may contain residual radioactivity from licensed activities. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2 or Class 3 designations. Class 1 areas have the greatest potential for residual activity while Class 3 areas have the least potential for impacted areas. Each classification will typically be bounded by areas classified one step lower to provide a buffer zone around the higher class. Exceptions occur 7

• iJ*':.....Alan J. Bloteky Reactor Facility when an area is surrounded by a significant physical barrier that would make transport of residual activity unlikely firom one area to the adjacent area. In such cases, each area will be classified solely on its own merit using the most reliable information available.

The class definitions provided below are fr'om Section 4.4 of the MARSSIM.

Class 1 "Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys). 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 high specific activity.

Note that areas containing contamination in excess of the DCGLw prior to remediation should be classified as Class I areas."

Class 2 "These areas have, or had prior to remedial/ion,.a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLw. To justify changing an area's classification fr'om Class i to Class 2, the existing data (from the HSA, scoping surveys, or characterization surveys) should provide a high degree of confidence that no individual measurement would exceed the DCGLw." Other justifications for this change in an area's classification may be. appropriate based on the outcome of the DQO process.

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

Class 3 "Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fr'action of the DCGLw, based on site operating history and previous radiological surveys.

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• f*J

• Alan J. Blotcky t*::!*Reactor Facility Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with veiy low potential for residual contamination but insufficient information to justify a

non-impacted classification."

3.5.3 Changes in Classification Data from operational surveys performed in support of decommissioning, routine surveillance and any other applicable survey data may be used to change the initial classification of an~area up to the time of FSS commencement as long as the classification reflects the levels of r~esidual radioactivity that existed prior to remediation. Once the FSS of a given survey uniti beginf*," the basis for reclassification will be documented. If, during the conduct of an FSS,'survey sufficient evidence is accumulated to warrant an investigation and reclassification of the survey unit, the survey may be terminated without completing the survey unit package. A reclassification of a survey unit because of detected radioactivity may only result in,a more stringent classification, unless demolition or remediation activities result in the entire survey unit being eliminated, such as a building being demolished, rather than surveyed for release.

3.5.4 Redefining Survey Boundaries Survey units that are adjacent to one another may be combined prior to the survey design to simplify the performance and evaluation of the survey data, provided the new unit meets all the parameters applicable to the classification. If two or more areas of different classifications are combined, the new area will be classified at the most stringent classification applicable.

Discussion of the redefined survey boundaries will be included in the FSS documentation of the survey activities. Survey areas which have their classification increased as a result of boundary changes do not fall under th~e investigation requirements for reclassifying a survey area described in Sect~ion 3.10.10.

3.6 Establishing Survey Units Land areas, structures, and systems are made up of at least one smaller area defined as a survey unit.

Each land area, structure, or system may have multiple survey units of differing classification since data acquisition, analysis, and reporting are done on a survey unit basis. The survey unit release records applicable to a larger survey area may be combined into one report for submission to USNRC.

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• J01*:Alan J. Bloteky

~Reactor Facility 3.61 Survey Unit Survey areas may be divided into smaller survey units. Survey units are areas that have similar characteristics and contamination levels. Survey units must be contiguous and will be assigned only one classification. Survey areas may include survey units of differing classifications since the site and facility are surveyed, evaluated, and released on a survey unit basis.

3.6.1.1 Survey Unit Size Section 4.6 of the MVARSSIM provides suggested sizes for survey units. However, as stated in the MARSSIM, the suggested survey unit sizes were based on a finding of reasonable Sample density and consistency.with commonly used dose modeling codes.

Table 3-2 lists the recommended survey unit size for each applicable classification.

Table 3-3: Survey Unit Size Minimum" / Maximumt Classification Buildings Open Land Class 1 10-rn2 / 100-mn2J

'100-m 2/ 2,000-in 2 Class 2 100-rn 2 / 1,000-rn 2:,

2,000-rn 2 / 10,000-rn 2 Class 3 1,000-m2 / No limit 10,000-rn 2!/ No limit For standing buildings, the MARSSIM recommennds 100-rn 2 of floor surface in a Class 1 area as a survey unit size based on the dose model assumption that a 100-rn 2 office space would be occupied. The source term attributed to the total area in this case is essentially the 100-rn 2 floor surface which equates to 180-rn 2 if the lower walls are included. If there is potential residual contamination on the upper Surfaces, i.e. the ceiling and walls above 2-rn, the area may be either broken into multiple survey units or the number of required data locations may be adjusted to account for the increased area.

A concerted effort should be made to define survey units of the sizes listed in Table 3-2; however there may be situations that would result in multiple survey units significantly smaller than those sizes.' In such a case, a single larger survey unit may remain provided the survey density is proportionally increased to account for the additional area. This is accomplished by dividing the

• Recommended minimum size SFrom MARSSIM Section 4.6 SFloor area of 100-rn 2 plus the lower walls (2-rn high) for a total area of 180-rn 2 10

Nii!!m i21*Alan J. Bloteky Reactor Facility maximum value for the classification listed in Table 3-2 by the number of required data points developed in the survey design package and proportionally increasing the sample density to account for the additional area.

Section 4.6 in the MARSSIM states

"...special considerations may be necessary for survey units with structure surface areas less than 10 m2 or land areas less than 100 in2.

In this case, the number of data points obtained from the statistical tests is unnecessarily large and not appropriate for smaller survey unit areas. Instead, some specified level of survey effort should be determined based on the DQO process and with the*

concurrence of the responsible regulatory agency. The data generated from thlese,

smaller survey units should be obtained based on judgment, ratherbthaan* on systematic or random design, and compared individually to the DCGLS."'

Consequently, the recommended minimum survey unit sizes in Table 3;2 are deri*'ed from this particular guidance.

If such a situation arises, the small survey area will be surveyed at a proportionate density developed in the same manner as additional samples in oversize areas were developed above.

3.6.1.2 Reference Coordinate System A reference coordinate grid system is used as an' aid' in the identification of survey locations within a survey unit. Software is used to develop the actual survey locations to the nearest foot.

Survey areas within buildings use a grid system based on the defined 'site north'. The grid patterns use a standard Cartesian grid system in the format of X,Y where X represents the east-west coordinates and Y represents tlh north-south coordinates. Each grid value represents one foot of linear distance in the applicable direction. Each survey unit has an origin indicated on the survey map as an aid in determihing orientation for conduct of the survey. Each survey unit may incorporate combinations of floors, walls, columns, ceilings, or special features within each unit as determined by th~e contamination potential and the need for survey. Local coordinates for each surface extend from the point of intersection of the Cartesian grid system applicable to that surface. Floors use the southwest corner as the point of intersection, while walls use the bottom left corner. Ceiling surfaces utilize a superimposition of the floor grid system to readily identify' survey locations.

For example, Grid Location X2,Y1 on the ceiling is directly above Grid Location X2,Y1 on the floor. Figure 3-1 shows an example of the splayed map of an interior space grid coordinate system for a typical room.

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LI Alan J. Bloteky Reactor Facility Figure 3-1: Interior Space Coordinate Example B540 Ceiling N

3, 5 11,5 Wall 1 Wall 8 T,

Wall 2 Floor 6,[ 5 Wal-W-1 0, 51 11ll7 Wall 5 Wl~

4, 5 3.7 Survey Design Surveys to demonstrate compliance with release criteria will be performed using design parameters from MARS SIM.

3.7.1 Scan Coverage The area covered by scan measurements is based on the survey unit classification as described in Table 2 of the MARSSIM and is summarized in Table 3-4 below. A 100% accessible area scan of Class 1 survey units will be required. The emphasis is placed on scanning the higher risk areas 12

Alan J. Bloteky

• !}*

Reactor Facility of Class 2 survey units such as soils, floors and lower walls. Scanning in Class 3 survey units will focus on likely areas of contamination based on the judgment of the SO and will generally consist of a minimum of the 1-mn2 area around the selected direct measurement location.

Table 3-4: Minimum Scan Coverage Classification Required Minimum Scan Coverage Class 1 100% of accessible areas Class 2

Ž10% of accessible areas Class 3 Judgmental, but generally a 1-im2 area around each data location" 3.7.2 Sample Size Determination MARSSIM recommends that FSS be performed using a relative shift between 1 and 3. A relative shift of 1 provides for a more static survey locations and is therefore more conservative while a relative shift of 3 requires fewer static survey locations. Existing characterization data for the AJBRF support facilities (i.e. rooms and hallways outside the reactor cavity) consistently demonstrates minimal residual radioactivity, therefore the number of static survey locations within those survey units will be based on a relative shift of 3unless remediation support surveys indicate the presence of elevated concentrations greater than 50 percent of the release criteria.

3.7.2.1 Statistical Test Determination Appropriate tests will be used for the statistical evaluation of survey data. Tests such as the Sign test and Wilcoxon Rank Sum (WRS) test will be implemented using unity rules, surrogate methodologies, or combinations, of unity rules and surrogate methodologies, as described in MARSSIM and NUREG-1 505 chapters 11 and 12.

If the contaminant is not in the background or constitutes a small fraction of the DCGLw, the Sign test will be used. It is anticipated that the sign test will be the statistical test applied to the collected data because of the small fraction of the DCGLw that background radionuclides will contribute. If it is determined that background radionuclides contribute a significant fraction of the DCGLw, the Wilcoxon Rank Sum (WRS) test will be used.

3.7.2.2 Establish Decision Errors The probability of making decision errors is controlled by hypothesis testing. The survey results will be used to select between one condition of the environment (the null hypothesis) and an alternate condition (the alternative hypothesis).

These hypotheses, chosen from MARSSIM Scenario A, are defined as follows:

13

• '*Alan J. Blotcky

• t*

m*Reactor Facility Null Hypothesis (H0): The survey unit does not meet the release criteria.

Alternate Hypothesis (Ha): The survey unit does meet the release criteria.

A Type I decision error would result in the release of a survey unit containing residual radioactivity above the release criteria. It occurs when the null hypothesis is rejected, but in reality is true. The probability of making this error is designated as "a."

A Type II decision error would result in the failure to release a survey unit when the residual radioactivity is below the release criteria. This occurs when the Null Hypothesis is accepted when it is not true. The probability of making this error is designated as "3" Appendix E of NUREG-1727 recommends using a Type I error probability (a) of 0.05 and States that any value for the Type II error probability (13) is acceptable. Following the guidance, a* will be set at 0.05 A 13 of 0.05 will initially be selected based ona site-specific considerations. The Type I and Type II error values are used as lookup values in MARSSIM Table 5.2 to provide Zia and Zni-I values for use in the sample size formula. In this case, Type I and Type II error probabilities of 0.05 result in a value of 1.645 for Zl_* and Z*_p when referencing MARSSIM Table 5.2. The 13 may be modified, as necessarY, after weighing the resulting change in the number of required survey measurements against the risk of unnecessarily investigating and/or remediating survey units that are truly.below the release criteria.

Table 3-5: MARSSIM Table 5.2 Error Percentile 0.005 2.576 0.01 2.326 S

0.015 2.241 0.025 1.96 0.05 1.645 0.1 1.282 0.15 1.036 0.2 0.842

• .0.25 0.674

0.3 0.524 3.7.2.3 Relative Shift The relative shift (A/cs) is a calculated value. Delta (A) is equal to the DCGLW minus the Lower Boundaiy of the Gray Region (LBGR). The sigma used for the relative shift calculation may be recalculated based on the most current data obtained from post-remediation or post-demolition 14

i*

Alan J. Blotcky

• • !*!*Reactor Facility surveys; or from background reference areas, as appropriate.

The LBGR may be adjusted to obtain an optimal value for the relative shift, normally between 1.0 and 3.0. Administratively, the relative shift will have a maximum value of 3.0.

DCGL -

LBGR relative shift =

cr 3.7.2.4 Lower Boundary of the Gray Region The Lower Boundary of the Gray Region (LBGR) is the point at which the Type II (f3) error applies. The default value of the LBGR is initially set at the mean of the post-remediation'survey results, if available, or at 50 percent of the DCGLw, whichever is higher. If the relative shift is greater than 3.0, then the number of data points, N, listed for the relative shift, values of 3.0 from Table 5.5 or Table 5.3 in the MARSSIM, will normally be used, ats the~ minimum 'sample size.

Use of a relative shift greater than 3.0 requires approval by the SO.

3.7.2.5 Sigma Sigma values (the estimate of the standard deviation of the measured values inl a survey unit, and/or reference area) will be initially calculated from characterization samples/survey data with activity at or below the approved DCGLWv since thais value is what is expected to remain after necessary remediation has been completed. *These sigma values can be used in FSS design or more current post-remediation sigma values Can be used.

3.7.2.1 Sian Test Sample Size; The number of data points to be collected may be determined from Table 5.5 in the MARSSIM and includes the recommended 20% adjustment to ensure an adequate sample size.

An alternative method is to use Formula 5-2 in the MARSSIM.

Calculating the number of data points by using the formula will require an adjustment for a recommended 20% surplus of data pointS.

MARSSIM Formula 5-2: N = (-

+ z-)

4(SignP - 0.5)2 Where:

ZI*=desired Type I error from MARS SIM Table 5.2 ZI*=desired Type II error fr'om MARSSIM Table 5.2 SignP =value from MARS SIM Table 5.4 (see below) associated with the relative shift 15

• Lm*J*Alan J. Blotcky

• l*

Reactor Facility Table 3-6: MARSSIM Table 5.4 Relative Shift ISign p IRelative Shift ISign p 0.1 0.539828 1.2 0.88493 0.2 0.57926 1.3 0.903199 0.3 0.617911 1.4 0.919243 0.4 0.655422 1.5 0.933193 0.5 0.691462 1.6 0.945201 0.6 0.725747 1.7 0.955435 0.7 0.758036 1.8 0.96407 0.8 0.788145 1.9 0.971284 0.9 0.81594 2

0.97725 1

0.841345 2.5 0.99379 1.1 0.864334 3

0.99865 If relative shift > 3.0, use SignP = 1.0 When using the Sign test and assuming a relative shift of 3-as discussed in Section 3.7.2, the required sample size (rounded up) is 11. MARS SIM recommends an additional 20 percent which results in a total of 14 sample locations.

(1.645 + 1.645)2 N4(0.99865 - 0.5)2 10.824 4

• 0.249 10.824 N=-

0.996 N=II

~N = 11 plus 20% =14 3.7.2.2 Wilcoxon Rank Sum (WRS) Test Sample Size While it is anticipated that the Sign test will be used for the FSS data evaluation, the Wilcoxon Rank Sum (WRS) test is an alternate statistical test that may be used if background radionuclides contribute significant radioactivity. The number of data points to be obtained from each reference area/survey unit pair may be determined using the Table 5.3 in the MARSSIM.

The table includes the recommended 20% adjustment to ensure an adequate sample size. An alternative method is to use Formula 5-1 to directly calculate the required number of data points. Using the formula will require an adjustment for a recommended 20% surplus of data points.

16

Alan J, Blotcky Reactor Facility MARSSIM Formula 5-1 :

(Z1 _* + Z1 ~)2 3(P - 0.5)2 Where:

Z1_, = desired Type I error Z1_* = desired Type II error Pr = probability value from MARSSIM Table 5.1 (see below) associated with the relative shift Table 3-7: MARSSIM Table 5.1

... i,>*,

Relative Shift Pr Relative Shift [ Pr 0.1 0.528182 1.4 0.838864 0.2 0.556223 1.5 0.855.541 0.3 0.583985 1.6 0.871014 0.4 0.611335 1.7 0.885299 0.5 0.638143

.1.8 0.89842 0.6 0.66429

\\1.9,,

0.910413 0.7 0.689,665 2,*'..."20 0.921319 0.8 0.714167 2.3 0.944167 0.9 0.73771 2.5 0.961428 1.0 0.760217 2.8 0.974067 1.1 0.78 1627 3.0 0.983039 1.2.

0.801892 3.5 0.993329

.1.3 0.820978 4.0 0.997658 If relative shift > 4.0, use Pr = 1.0 When using the WIRS test and assuming a relative shift of 3 as discussed in Section 3.7.2 and including an additional 20 percent as recommended by MARSSJIM, the sample size required both in the survey unit and in the background reference area is 20.

(1.645 + 1.645)2 N3(0.983039 - 0.5)2 10.824 3

• 0.233 10.824 0.699 17

• ,*m:*

NAlan J. Blotcky W*IaI*NReactor Facility N =16 N =16 plus 20% =20 3.7.2.3 Elevated Measurement Comparison (EMC) Sample Size Adiustment If the scan MDC (MiDCso*) is greater than the DCGLW, the sample size will be calculated using the equation provided below. If Ne,,, exceeds the previously determined sample size, Neine will replace N.

Am emc Where:

Nemc is the elevated measurement comparison sample size,-

A is the survey unit area Aeimo is the area corresponding to the area factor calculated using the MDCSC8I concentration.

3.7.3 Background Reference Areas Background reference area measurements, are required when the WRS test is used, and background subtraction may be used with thie Sign test, under certain conditions such as those described in Chapter 12 of NUREG-1505. The reference area measurements are collected using the methods and procedures required for Class 3 FSS. For survey units that contain a variety of materials with markedly different backgrounds, a reference area that has similar materials will be selected. If one material is predominant or if there is not much variation in background among materials, a background from a reference area containing only a single material is appropriate when it is demonstrated that the selected reference area will not result in underestimating the residual radioactivity in the survey unit.

Background reference areas should have physical characteristics (including soil type and rock formation). simtilar to the site and shall not be contaminated by site activities. In general, VA commits to using background reference areas, when possible, that are offsite.

If non-contaminated onsite areas are to be used, then VA will verify and justify the use by appropriate comparison with samples fr'om appropriate off-site locations.

Should significant variations in background reference areas be encountered, appropriate evaluations will be performed to define the background concentration.

As noted in NUREG-1757, Appendix A, Section 3.4, the Kruskal-Wallis test can be conducted in such circumstances

Alan J. Blotcky Nj~in~i*Reactor Facility to determine that there are no significant differences in the mean background concentrations among potential reference areas.

VA will consider this and other statistical guidance in the evaluation of apparent significant variations in background reference areas.

3.7.4 Sample Grid and Measurement Location Sample location is a function of the number of measurements required, the survey unit classification, and the contaminant variability.

3.7.4.1 Sample Grid The reference grid is primarily used for reference purposes and is illustrated on sample maps.

Physical marking of the reference grid lines in the survey unit will only b~e performed When necessary.

For the sample grid in Class 1 and 2 survey units, a randomly selected sample start point will be identified and sample locations will generally be laid out ina triangular grid pattern at distance, L, fr'om the start point. A rectangular grid pattern may be used at the discretion of the survey plan designer. The sample and reference grids aire illustraited on sample maps and may be physically marked in the field. For Class 3 survey.uhits, sample locations may be randomly selected based on the reference grid or may use a systematic-grid.

3.7.4.2 Measurement Location Computer software will be used to determine measurement locations within a survey unit.

Measurement locations within the survey unit will be clearly identified and documented for purposes of reproducibility. Ana idenftification code will match a survey location to a particular survey unit.

Sample points for Class.1 and Class 2 survey units will be determnined using a systematic triangular grid pattern with a random start location. Sample locations in a Class 3 survey unit will be determined randomly.

Measurement locations selected using either a random selection process or a random-start systematic pattern that do not fall within the survey unit or that cannot be surveyed due to site conditions will be replaced with other measurement locations as determined by the SO.

3.7.5 Survey Package Design Process A Survey Package is produced for each survey area. The survey package is a collection of documentation detailing survey design, implementation and data evaluation for FSS of a survey area.

19

~Alan J. Blotcky Reactor Facility The VA intends to apply the 10 CER 50, App. B requirements for field and laboratory counting equipment, as well as the corrective action process to address data or programmatic discrepancies. Using the existing Part 50, Appendix B program precludes developing redundant measures for FSS activities.

3.7.5.1 Survey Package Initiation Each survey area and package is assigned a unique identification number. To allow continuity of area identification, the protocol used for identifying survey areas during the characterization survey will be used, as appropriate....

3.7.5.2 Review of Characterization Surveys The SO gathers and reviews historical data applicable to the survey area. Inf'ormation used for survey design is filed in the survey package. Sources of data include:

Characterization Surveys, Classification basis 10 CFR 50.75(g) files...

Operational Survey Records 3.7.5.3 Survey Area Walkdown The survey designer performs a walkdown to gather information about the physical characteristics of the survey area.

The, walkdown provides the designer an opportunity to determine if any physical or safety related interferences are present that may affect survey design or survey implementation, and to determine any support activities necessary to implement surveys.

The walkdown is documented and filed in the survey package.

Following the walkdown, representati~ce maps of the survey area are prepared.

• 3.7.5.4 Survey Desi~n Survey Design is the process of determining the number, type and location of survey measurements or samples required for each survey unit within a survey area. The various aspects of survey design are documented and filed in the survey package. The survey unit design process is controlled by approved procedures.

The size and number of suorvey units for a survey area are determined based on area classification, modeling assumptions used to develop DCGLW, and the layout of the survey area. The design will divide the area into discrete survey units as appropriate.

Each survey unit is numbered sequentially. The design provides a description of each survey unit including survey unit size, 20

• Alan J. Bloteky Reactor Facility classification and location. The types of material (i.e. soil, concrete, etc.) found in the survey unit and survey measurement and/or sampling methods are identified.

The design provides the number of measurements or samples required for each survey unit in accordance with MARSSIM. Count rate equivalent investigation levels for survey measurements based on the instrument detection efficiency may also be provided.

Table 3-8 provides measurement result investigation levels.

The design determines measurement/sample locations based on the classification of the survey unit and in accordance with the MARS SIM. A survey map is prepared of each survey unit. A sample and/or reference grid is superimposed on the map to provide an (X, Yjc'oodi~dnate system.

Each measurement/sample location is assigned a unique identification code which identifies the measurement/sample by Survey Unit and sequential number.

The design indicates the appropriate instruments and detectors, instrument operating modes and survey methods to be used to collect and analyze data.

Written survey instructions that incorporate the requirements set forth in the survey design may be provided. Direction includes selecting instruments,, cotPnt times, instrument modes, survey methods, required documentation, background requirements, and other appropriate instructions.

In conjunction with the survey instructions, survey, data forms, indicating desired measurements, are prepared to assist in survey documentation..

3.7.5.5 Survey Area Turnover Prior to performing FSS, the SO coordinates with the appropriate personnel to ensure decommissioning activities, area remediation and housekeeping are complete. The SO may direct surveys to be performed to verify that the area meets the radiological criteria for performance of the FSS. These post-remediation surveys, if performed, provide conclusive proof that an area is acceptable for FSS initiation. When satisfied the area is acceptable, the SO will direct the area to be posted to indicate that the area is controlled for the performance of FSS. Access controls will be implemented to prevent contamination of areas during and following FSS.

3.7.5.6 Survey Implementation Survey areas and/or locations are identified by gridding, markings, or flags as appropriate.

Instruments and equipment as indicated in the survey instructions are checked for proper operation and surveys are performed in accordance with the appropriate procedures. All survey results will be documented and a chain of custody for any collected samples will be maintained.

21

Reactor Facility On completion of the FSS, instruments will receive post-use source and background checks, and any collected samples will be prepared for analysis either on-site or at the selected off-site laboratory.

Survey instruments will be prepared in accordance with appropriate procedures and the survey instructions. Instruments are performance checked prior to and following surveys. Survey data will be reviewed and placed into the survey package. The data will be examined for any results that exceed investigation criteria so that appropriate investigation surveys and/or remediation may be performed in a timely fashion.

Several quality control measures and features have been developed for the implemnentation phase of the FSS program. These measures include:

Pre-implementation area walkdowns Survey location verification Instrument source checks prior to the start andupo completion of an FSS survey Conduct of scan surveys in the peak trap mode,,th~ereby providing a record of the maximum scan value for any scan grid

=.

3.7.5.7 Data Evaluation The survey data will be reviewed to verify completeness, legibility and compliance with the following requirements:

Convert data to standardize~d units (if necessary)

Calculate mean, median and range of the data set Review the data for outliers Calculate the standard deviation of the data set Verify MDC for each survey type performed meets requirements These steps may be completed by a software package or performed manually using tools of choice.

The SO reviews and verifies the statistical calculations, verifies the integrity and usefulness of the data set and determines the need for further data. Investigations of suspect data will be performed as necessary, Once satisfied that all data are valid, the appropriate statistical test will be performed and a decision will be made on the radiological status of each survey unit.

The data evaluation process is documented and filed in the survey package.

3.7.5.8 Quality Control Surveys Following completion of FSS, the need for QC surveys (replicate surveys, sample recounts, etc.)

22

o*?s*;:Alan J. Blotcky Reactor Facility will be determined. If necessary, a QC survey package will be developed and modeled after the original survey. QC measurement results are compared to the original measurement results. If QC results do not agree with the original survey, an investigation is performed.

Following investigation, the SO will decide data validity. Additional discussion of the performance of quality assurance and quality control surveys is provided in Section 5.4.1.

3.7.5.9 Release Record Following data evaluation, a Release Record will be prepared. The Release Record summarizes survey results and data evaluation. The Release Record is reviewed and approved by the alternate 3.8 Types of Surveys Survey measurements and sample collection are performed by personnel trained and qualified in:

accordance with the applicable procedure. The techniques for* performing survey measurements or collecting samples are specified in approved procedures. FSS measurements include surface scans, direct surface measurements, and gamma spectroscopy of volumetric materials. In-situ gamma spectroscopy or other methods not specifically described may also be used for FSS.

If required, a technical basis documnent will be created by the SO. Upon the documents acceptance, VA will give the USNRC a 30 day notice to review the technical basis document prior to implementation.

On-site and off-site lab facilities are used* for gamma spectroscopy, liquid scintillation and gas proportional counting in accordance *with applicable procedures. Regardless which facilities are used, analytical methods will use an administrative level of 50% the applicable DCGLw value for detection of radioactivity."

3.8.1.

Scan Surveys Scanning is performed in order to locate small areas of residual activity above the investigation level.

Structures receive scan surveys, direct measurements and, when necessary, volumetric sampling. The percent of scan measurement coverage is based on the survey unit classification and is provided in Table 3-4.

3.8.1.1 Two-Staae Scanning Method The two-stage scanning method is one where a surveyor begins the scan at a pre-determined speed, e.g. 10-cm per second, until they detect an elevated count rate. At such time, they return to a location immediately before the elevated detection and repeat the scan at a slower rate to 23

Alan J. Blotcky

i*

Reactor Facility determine the maximum count rate in the area. When the count rate has returned to expected levels, the scan speed is returned to normal. This method relies on the ability of the surveyor to reliably detect elevated count rates. The 'Surveyor Efficiency' (p) is detailed in NUREG-1507.

A variable accounting for this efficiency (a value from 0.5 to 0.75) is included in the formulae used in the MARSSIM. Lower values for p increase the MDCSCa,n indicating a smaller probability of detecting elevated count rates. The MDCscan equations used throughout this document use a value for p of 0.5.

Additional evaluations may be made by VA on the effectiveness of eliminating p from the MDCscan equation and utilizing the alarm functions of the selected instrumentation.

All scans performed in support of the FSS use the two-stage scan1 nethod to assure residual radioactive material is sufficiently quantified.

r 3.8.1.2 Beta-gamma Surface Scans Surface scans for beta-gamma activity on structures and selected systems will be performed at a scan rate capable of meeting a pre-determined MIDCsoan, applicab~le to the survey unit classification. Surface scans should have a probe to surface distance as close as practical, not to exceed 1-cm (-1/2/

inch). Situations where the maximum detector to surface distance cannot be met may require an alternate scan method, a detector to surface distance correction factor, or justification for not completing the scan. For Class 1 areas, the MDCSCan will be no greater than the DCGLemC. Class 2 and Class 3 areas require a more stringent MDCscan of no greater than the DCGLW unless scan coverage of 1 00% is specified in the survey design when a higher MDCscan equal to the DCGLe~nC may be used.

Minimum scan coverage is detailed in Table 3-4 by classification.

3.8.2 Direct Measurements Direct measurements are performed to detect surface activity levels. Direct measurements are conducted by placing the detector on or very near the surface to be measured and acquiring data over a pre-determined count time. A six (6) second count time will be used for direct surface measurements and provides detection levels well below the release criteria however the count time may be varied to ensure the required detection level is achieved. The MIDC for a static count may be calculated using the formula contained in Section 3.10.4. Direct measurements may be collected anywhere within the grid block, but are generally collected in the center.

3.8.3 Exposure Rate Gamma exposure rate measurements will be taken at each floor location approximately 1-rn from the surface.

24

gL*;

Alan J. Bloteky 1**

Reactor Facility 3.8.4 Removable Activity Removable contamination surveys, while not required as part of the MARSSIM, may be collected to assess the removable activity fraction for selected structural and system surfaces or as part of routine post-FSS verification surveys. When possible, a wipe sample will comprise 100-cm 2 of surface area. When a 100-cm2 area is not obtainable, the wiped area will be documented and the analyzed result adjusted accordingly.

Wipe samples will be analyzed on-site utilizing a wipe sample counter. Investigation levels will be 10% of the applicable building surface DCGLW. The origin of removable activity in excess of this investigation level should be determined and the area should receive additional remediation prior to FSS.

4.

3.8.5 LSC Analysis All direct measurement survey locations will have a removable activity wipe collected for tritium and C-14 analysis on-site using a liquid scintillation counter (LSC).

3.8.6 Volumetric Samples Volumetric sampling of media, as opposed to direct measurements may be necessary if the validity of direct measurements is questionable.*i Volumetric samples will be analyzed by appropriate analytical methods for the radionuclides of interest applicable to the subject area.

The results will be evaluated by one of the following:

Calculating the derived total gross beta or gross alpha DPM/1 00-cm 2 in the sample and comparing the gross results directly to the applicable DCGLw Using the radionuclide specific results to derive the surface activity equivalent and determine compliance using the unity rule.

Use of the unity' rule will require the use of a surrogate calculation to account for the radionuclides in the; mixture that are not identified by gamma spectroscopy.

This will be accomplished using the nuclide mixture established during the applicable characterization.

Sample preparation will be performed by the off-site contracting laboratory. Separate containers will be used for each sample and each sample will be tracked through the analysis process using a chain-of-custody record. Samples will be split as directed in Section 5.4.2.

3.9 Survey Methods Survey methods are applied differently depending on the data requirements of a survey area. For example, removable activity measurements provide little, if any, benefit when attempting to 25

Alan J. Bloteky Reactor Facility assess the radiological conditions in an excavation. Conversely, assessing a building surface via volumetric sampling would provide the necessary data, but at great costs of time and money.

This section will discuss the steps necessary to strike a reasonable balance between data needs and ease of survey performance based on the data needs of the survey area.

3.9.1 Buildin~s, Eq~uipment* and Components Buildings, equipment, and components that are destined to remain after license termination require the following surveys to demonstrate they meet the appropriate release criteria."

3.9.1.1 Scans Buildings, equipment, and components require two-stage scan measurements as part of the FSS process at coverage rates and speeds as directed in Table 3-4 and Section 3.8.1; *respectively. The two-stage scan method has been described in Section 3.8.1.1.

Gross beta and/or gross alpha measurements are utilized as appropriate to the potential contaminatiodn.

The measurements typically are performed at a distance of 1-cm or less from the surface. *Adjustments to scan speed and distance may be made in accordance with approved procedures.

3.9.1.2 Direct Measurements Direct measurements are required for, buildings, equipment, and components as part of the FSS process. The required quantity of direct measurements is a calculated value. The calculation is described in Section 3.7.2. Direct measurement data for buildings, equipment and components is collected with a gas proportiona! detector. As much as practical, the detector is of an appropriate size to maintain the surface to detector distance of no greater than the calibrated distance +0.5-cm.

3.10 Instrumentation Radiation detection and measurement instrumentation for the FSS are selected to provide both reliable operation and adequate detection sensitivity of the final list of the radionuclides of concern as identified during the characterization evolution.

Detector selection is based on detection sensitivity, operating characteristics and expected performance in the field.

The instrulmentation, to the extent practicable, is capable of data logging operations.

Commercially available portable and laboratory instruments and detectors are typically used to perform the three basic survey measurements:

1. Surface scanning 26

Alan J. Bloteky Reactor Facility

2.

Direct surface contamination measurements

3. Spectroscopy of soil and other bulk materials, such as concrete 3.10.1 Selection Radiation instruments and detectors are selected based on the type and quantity of radiation to be measured. The instruments used for direct measurements are capable of detecting the radiation of concern to a Minimum Detectable Concentration (MIDC) of less than 50% of the applicable DCGLw. The use of 50% of the DCGLw is an administrative limit only. Any value below the DCGLw is acceptable in impacted survey units. MDCs of less than 50% of the DCGLi* allow detection of residual activity in Class 3 survey units at an investigation level. of,0.5 times the DCGLw. Instruments used for scan measurements in Class 1 areas are required to ble capable of detecting radioactive material at less than or" equal to the DCGLemc.'

Specific instrument selection and performance capabilities (e.g. detection efficiency, MDCs, data logging, etc.) are discussed in Section 4.0 "Survey Instrumentation'. VA will generally follow the instrument manufacturers' recommendations and/or supporting basis documents for considerations such as temperature dependency and other operational parameters.

A description of the conditions under which themethod would be used.

A description of the measurement method, instrumentation, and criteria.

Justification that the technique WVould p5rovide equivalent scan coverage for the given survey unit classification and that the M\\DCscan is adequate when compared to the DCGLemc.

A demonstration that the methlod provides data that has a Type 1 error (falsely concluding that the survey unit, is acceptable) equivalent to 5% or less and provides sufficient confidence that the DCGLeinc criterion is satisfied.

3.10.2 Calibration and Maintenance Instruments and detectors are calibrated for the radiation types and energies of interest at the site.

The calibration sources for beta survey instruments are Tc99or Sr90. The alpha calibration sources are Pu239 or Th23° which have an appropriate alpha energy for plant-specific alpha emitting nuclides. Gamma scintillation detectors are calibrated using Cs 137. Calibration sources other than those listed may be used provided they demonstrate appropriately conservative detection efficiency for the radionuclides of interest. In all cases, the surface efficiency as determined appropriate for the weighted mean energy of the radionuclides of concern will be utilized regardless of the energy of the calibration source.

Instrumentation used for FSS is calibrated at least annually and maintained in accordance with the 27

• • =*

=i~i*AlIan J. Blote ky r~s*

Reactor Facility Instrumentation Program procedure. Radioactive sources used for calibration are traceable to the National Institute of Standards and Technology (NIST) and use, to the extent practical, geometries designed to mimic the type of samples being counted. If vendor services are used, these are obtained in accordance with purchasing requirements for quality related services to ensure an appropriate level of quality.

3.10.3 Response Checks Instrument response checks are conducted to assure proper instrument response and operation.

An acceptable response for ratemeter field instrumentation is an instrument reading +I10%

of the check source value established during or immediately after calibration.

Scaler and benchtop counter instrumentation standards are +/-3 sigma as documented on a control chart. Response checks are performed daily before instrument use and again at the enad of-use. 'Check sources are appropriate for the type of radiation as that being measured 'in the field and are, to the extent practical, held in fixed geometly jigs for reproducibility. If anidnstrument fails a response check, it will be clearly labeled "~Do Not Use" and will be removed fr'om service until the problem is corrected in accordance with applicable procedures.'

Measurements made between the last acceptable check and the failed check will be evaluated to determine if they may remain in the data set.

3.10.4 Minimum Detectable Concentration The MDC for static and scan measurements is determined for the instruments and techniques used for FSS.

The static measurement MDC is the concentration of radioactivity that an instrument is expected to reliably (i.e. 95 percent certainty) detect when used in an integrated count mode, i.e. number of counts per unit time, while MVDC~an is the concentration of radioactivity that an instrument is expected to reliably detect in ratemeter mode. Detailed discussion regarding calculation of instrument MDC and MDCscn is included in Section 4.2.2 3,10.5 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 which warrants investigation. Elevated measurements may result from discrete particles, a distributed source, or a change in background activity. In any case, investigative actions should be implemented. Depending on the investigation results, the survey unit may require:

No action 28

'*:Alan J. Bloteky

• ii;i,

• Reactor Facility Remediation Reclassification and resurvey 3.10.6 Investigation Levels Table 5.8 in the MARSSIM provides guidance on investigation levels for scan surveys.

In addition to investigation levels for scan surveys, direct measurement survey investigation levels may be used.

These additional investigation levels include a conservative value for Class 3 survey units and are provided in Error! Reference source not found....

Table 3-8: Measurement Result Investigation Levels Classification Investigation Level Class 1 Result >DCGLemc0 Class 2 Result >DCGLw Class 3 Result >50% DCGLw :

3.10.7 Investigation Process Technicians respond to all audibly detected elevated count rates while surveying.

Upon determining an elevated count rate, the technician stops and resurveys the suspect area to verify the count rate elevation and determine the areal extents of the elevated count rate. Technicians are cautioned, in training, about the. iportance of the elevated count rate and the verification survey. They are given specific. direction regarding the extent and scan speed of the verification survey. If the elevated count rate is verified, the technician marks the area. Each marked area will receive an additional documented survey which requires a re-scan of the area and one or more direct measurements.

Field gamma spectroscopy measurement and collection of soil samples may also be, dictated. Results of each investigation are discussed and reported in the survey: unit Release,Record.

The size and average activity level in the elevated area will be defined to determine compliance with the area factors. If any location in a Class 2 area exceeds the DCGLw, scanning coverage is increased in order to detennine the extent and level of the elevated reading(s). If any location in a Class 2 area exceeds the DCGLemC, the scan coverage is increased and the area may be reclassified, if necessaly*. If the elevated reading occurs in a Class 3 area, the scanning coverage is increased and the area may be reclassified, if necessary.

Investigations should consider:

29

im*

Alan J. Blotcky Reactor Facility The assumptions made in the survey unit classification The most likely or known cause of the contamination The possibility that other areas within the survey unit may have elevated areas of activity that may have gone undetected Depending on the results of the investigation, a portion of the survey unit may be reclassified if there is sufficient justification. The results of the investigation process are documented in the survey area Release Record.

See Section 3.5.3 for additional discussion regarding potential reclassification of the survey unit.

3.10.8 Elevated Measurement Comparison (EMC)

...i 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 the MARSSIM, the EMC is intended to flag potential failur~es in the remediation process and should not be considered the primary means to identify wheth~er 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 EMC. The size of the area containing the elevated residual radioactivity and the average level of residual activity within the survey unit are determined. The initial DCGLeimo is established during the survey design and is calculated as follows:

DCGLeinc determination:

DCG c = A-* DCGJ Where:

AF = Area Factor corresponding to the size of the elevated area DCGLw =Derived Concentration Guideline Limit The area factor is a multiple of the DCGLW that is permitted for 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 DCGLw and is determined by the investigation process. Area factors are determined during the DCGLW development phase of the project.

30

• m....Alan J. Bloteky Reactor Facility 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 DCGLemnc. If the DCGLemo is exceeded, the area is remediated and resurveyed.

The results of the elevated area investigations in a given survey unit that are below the DCGLeImC 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 formula result is less than unity, no further elevated area testing is required and the EMC test is satisfied.,

Elevated area evaluation:

8 a~vg DCGLW A)(CL.

Where:

6 = average residual activity in the survey unit....

Cavg = average concentration of the elevated area AF = Area Factor corresponding to the size of the elevated area When calculating 6 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.

Compliance with the soil DCGLemoc is determined using gamma spectroscopy results and a unity rule approach. These general methods are also applied to other materials where sample gamma spectroscopy is used for FSS. *The application of the unity rule to the EMC requires that area factors and a corresponding DCGLernc be calculated separately for any gamma emitters identified during FSS.

3.10.9 Remediation and Reclassification Areas of elevated residual activity above the DCGLermc within any classification are remediated to reduce the residual radioactivity to acceptable levels.

Whenever an investigation confirms activity above an action level applicable to the classification, an evaluation of the operational history, design information, and sample results is performed to assure the area was classified properly. The evaluation considers:

The elevated area location, dimensions, and sample results.

An explanation of the potential cause and extent of the elevated area in the survey unit.

The recommended extent of reclassification, if considered appropriate.

Any other required actions.

31

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

If an individual scan or static location measurement within a survey unit exceeds the applicable investigation level listed in Table 3-8Error! Reference source not found., the survey unit or a portion of it may be reclassified and the survey redesigned and re-performed> accordingly.

Instrument performance, background fluctuation, surveyor performance, ambient radiological conditions, and other variables should be considered to avoid unnecessary reclassification.

3.10.10 Reclassification and Resurvey:

Following an investigation, if a survey unit is reclassified or* if remediation activities occur, a resurvey will be performed. If the average value of Class 2 direct survey measurements was less than the DCGLw, the MDCscan was sensitive enough to.detect the DCGLemc and there were no areas greater than the DCGLemnc, the survey redesign may be limited to obtaining a 100% scan without having to re-perform the direct measurements, This condition assumes that the sample density meets the requirements for a Class 1 area. If the Class 2 area had contamination greater than the DCGLw, but the MDCSCan was not sensitive enough to detect the DCGLemc, the affected area is reclassified as Class 1 and resurveyed with the sample density determined for the new classification.

Class 3 areas are treated in a similar manner, using 50% DCGLw as the investigation limit. If a C2lass 3 area had activity in excess of 50% DCGLw, but less than the DCGLw and the MDCscan was sensitive enough to detect the DCGLeinc, then the expansion of scan survey coverage to 100% will be sufficient. If activity is detected above the DCGLw, or the MDCscan was not sensitive enough, the area is increased to the appropriate classification as determined by the activity detected and the survey redesigned and performed as directed by the SO. *Reclassification of a survey area requires notification be made to the SO. A more detailed investigation of the reason for the improper classification will be performed at the discretion of the SO.

3.11 Data Collection and Processin2 3.11.1 Sample Handlini* and Record Keeping A chain-of-custody (COC) record will accompany each sample from the collection point through obtaining the final results to ensure the validity of the sample data. COC records are controlled 32

GAlan J. Blotcky

• i

!iReactor Facility and maintained in accordance with applicable procedures.

Each survey unit has an associated document package which covers the design and field implementation of the survey requirements. Survey unit records are considered quality records.

3.11.2 Data Management Survey data are collected fr'om several sources during the data life cycle and evaluated.

QC replicate measurements are not used as final status survey data.

See the MAR-QAP for design and evaluation of QC replicate measurements.

Measurements performed during turnover and investigation surveys may be used as FSS data if they were performed according to the same requirements as the FSS data.,The requirements include:

s The survey data reflects the as-left survey unit condition and is untouched by further remediation.

Isolation measures are put into effect for the survey unit to prevent re-contamination and to maintain final configuration.

The data collection and design were in accordance with FSS methods, e.g., scan coveageMDC requirements, inyestigation levels, survey data point quantity and location, statistical tests, and EMC tests.

• Measurement results intended as FSS data constitute the final survey of record and are included with 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 DCGLw.

The recording units for surface,contamination are DPM/100-crn 2 and pCi/g for activity concentrations. Numerical values, even negative numbers, should be 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,.rneasurement results, the date and time of the measurement and any surveyor comments.

3.11.3 Data Verification and Validation The FSS data will be reviewed prior to 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:

The instrumentation MDC for fixed or volumetric measurements was below the DCGLw or if not, it was below the DCGLeinc for Class 1, below the DCGLw for Class 2 and below 50% DCGLw for Class 3 survey units.

33

EI*

Alan J. Blotcky Reactor Facility The instrument calibration was current and traceable to INIST standards.

The field instruments were source checked with satisfactoiy results before and after use each day data were collected or data were evaluated by the SO if instruments did not pass a source check in accordance with Section 3.10.

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

o The survey methods used were proper for the types of radiation involved and for the media being surveyed.

• The COG 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 reflect the radiological status of the~area. ",-

The data have 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, reviewed, and approved by the SO.

3.11.4 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. This is an optional task which is intended to aid the SO in situations where the decision.of whether an *area or unit meets the applicable criteria is not readily apparent. Visual representations may include any combination of posting plots, frequency plots, histograms, contour maps, or 3-D surface plots.

3.12 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 DCGLW or comparing the mean value to the DCGLw 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.

3.12.1 Data Evaluation 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.

An assessment of the measurement results is used to quickly determine whether the survey unit passes or fails the release criterion or whether statistical analyses must be performed.

34

• t?

• Alan J. Blotcky Reactor Facility If all concentrations within the survey unit are less than the DCGLW, the unit meets the criterion and no statistical tests are necessary. If the average concentration is greater than the DCGLw, the survey unit does not meet the release criterion and additional remediation may be necessary. If the average concentration is less than the DCGLw, but one or more individual measurements exceed the DCGLw, the sign test or WRS test, and elevated measurement comparison tests should be conducted to determine the disposition of the survey unit.

When required, one of four statistical tests is performed on the survey data:

1. WRS Test-,

2.

Sign Test i..."'

3. WRS Test Unity Rule

4.

Sign Test Unity Rule It is not anticipated that the WRS tests will be conducted,. but it will be discussed in this document for topic completeness.

In addition, survey data are evaluated against the EM~c criteria as previously described in Section 3.10.8.

The statistical test is based on the null h*ypothesis (Ho) that the residual radioactivity in the survey unit exceeds the DCGLw. There must be sufficient survey data at or below the DCGLw to reject the null hypothesis and conclude the survey unit meets the site release criterion for dose. Statistical analyses are performed using a specially designed software package or, if necessary, hand calculations.

3.12.1.1 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 acceptably low fractions compared to the DCGLw. Instrument background is subtracted from the gross survey result to determine a net result appropriate for each survey location. Should any of the radionuclides of concern be present in background, the measurement net result (i.e. gross result minus instrument background) is assumed to be entirely from plant activities. This option is used when it can be reasonably expected that including background activity 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 used with background subtraction in accordance with Chapter 12 of NUREG-1 505.

35

U, Alan J. Blotcky N;'*f*,*Reactor Facility The sign test is conducted as follows:

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

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

N.

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

• The numbeir 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.

• The value of S+ is compared to the critical value given in Table 1.3 of thle MARSSIM.

The table contains critical values for given values of N and ox. The value of cc is set at 0.05 during survey design. If S+ is greater than the critical value given inl 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.

3.12.1.2 Wilcoxon Rank Sum Test The WRS test, or WRS unity rule, described in detail in NUREG-1505 Chapter 11, may be used when the radionuclide of concern is present in the backgroun4, or if measurements are used that are not radionuclide-specific. In addition, this test is Valid only when measurement results less than MDC values do not exceed 40 percent of the data set.

The WRS test is applied as follows:

The background reference area measurements (Xi) are adjusted by adding the DCGLW to each background reference area measurement to obtain the value for Zi (Zi = X1 +

DCGLw).

• The number of adjusted background reference area measurements (in) and the number of survey unit measurements (n) are added to obtain N (N = m + n).

• 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.

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

The value Of Wr is compared with the critical value in Table 1.4 of the MARSSIM. If Wr 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.

3.12.1.3 Unity Rule The radionuclides of concern 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 36

• ,*Alan J. Bloteky

• ,;**:**Reactor Facility values, also called weighted sum, will be calculated as shown in the following equation.

Unity calculation:

UNITY-C

+

C

+

C.+

DCGI1 DCGL2 DCGI*

Where:

Cx= radionuclide concentration The unity calculation results are used to demonstrate compliance by defining the DCGL as 1.0 and using the decision criteria listed in Section 3.12.1. If the application of the WRS~or sign test is necessary, these tests will be applied using the unity calculation results and ~defining the DCGLw as 1.0. An example of a WRS test using the unity rule is provided in the MARSSIM Appendix I, Section 11.4.

If the WRS test is used, or background' subtraction is used in conjunction with the Sign test, background concentrations will also be converted to their unity equivalents prior to performing the test. The sign test is used.without background subtraction if background values are not considered a significant fraction of the DCGLw.

The unity rule method described above will be applied* as necessary when multiple radionuclides or emission types are being evaluated as opposed. to a single radionuclide, a surrogate nuclide that incorporates a nuclide fraction, gross alpha measuirements, or gross beta-gamma measurements.

3.13 Statistical Conclusions The results of the statistical tests, including application of the EMC, allow one of two conclusions to be made. The first conclusion-is that the survey unit meets the site release dose criterion. The data provide 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 possible conclusion is that the survey unit fails to meet the release criterion. The data are not conclusive in showing that the residual radioactivity is less than the release criterion. In this case, the data are analyzed further to determine the reason for the failure.

Possible reasons are:

The average residual radioactivity exceeds the DCGLw.

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 collected measurements and the 37

E~i1:,Alan J. Blotcky Reactor Facility standard deviation of the measurement data. The power is determined from 1-f3 where fI is the value for Type II errors. A retrospective power analysis may be performed using the methods described in Appendices 1.9 and 1.10 of the MARS SIM. If the power of the test is insufficient due to the number of measurements, additional samples may be collected and the data re-evaluated. Additional 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.

3.13.1 Compliance The FSS is designed to demonstrate that licensed radioactive materials have been removed. from the VA facilities and property to the extent that residual levels of radioactive contamination are below the radiological criteria defined by the selected and approved dose-based scenario.

If the measurement results pass the requirements of Section 3.13, and the elevated areas evaluated per Section 3.10.8 pass the elevated measurement comparison, then the survey unit is determined to meet the criteria for license termination.

3.14 Reporting Format' Survey results are documented in history files, survey unit release records, and in the final status survey report. Other reports may be generated as requested by the USNRC.

3.14.1 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 When a survey unit release record is given final approval it becomes a quality record.

38

AlanJ. Bloteky

• (*,!*Reactor Facility 3.14.2 Final Status Survey Report Survey results will be described in written reports to the USNRC. The subject areas included in each written report will vary depending on the status of ongoing decommissioning activities.

Other reports may be written and submitted to USNRC as necessary, or" upon request.

The FSS report provides a summary of the survey results and the overall conclusions which demonstrate that the VA facility and site meet the radiological criteria for release. 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 basic outline of the final reports will be:."

1.0 Overview of the Results 2.0 Discussion of Changes to FSS 3.0 Final Status Survey Methodology Survey unit sample size Justification for sample size 4.0 Final Status Survey Results

• Number of measurements taken.

Survey maps Sample concentrations Statistical evaluations, including power curves Judgmental and miscellaneous data sets 5.0 Anomalous Data 6.0 Discussion of the Cross-Contamination Prevention and Monitoring Plan Inplementation 7.0 Conclusion for each survey unit 39

• '::'Alan J, Bloteky

• i*

Reactor Facility 4.0 SURVEY JINSTRUMENTATION The FSSP presents an outline of the MARSSIM survey process for the purpose of terminating a United States Nuclear Regulatory Commission (USNRC) radioactive materials license. This section discusses common radiological survcey instrument selection, their capabilities, general operation, and calibration.

The purpose of this section is to:

• Present a selection of instruments that may be used in support of the MIARSSIM activities at the AJBRIF-Provide methods for evaluation of the suitability of various instruments for their intended roles Discuss statistical formulae demonstrating detection efficiencies, minimum detectable concentrations, and expected surface efficiencies

• . Establish general calibration requirements Set Quality Assurance/Quality Control (QA/QC) requirements for the instrument program 4.1 Overview Measurement data for surface measurements are recorded as counts per unit time per detector area. These are normally converted into standardized units for evaluation purposes which eliminate instrument specific parameters. Common standardized units within the United States are disintegrations per minute per 100-cm 2 (DPM/100cm2) which will be used in this document.

Volumetric analyses are typically, reported in units of radioactivity per unit mass. PicoCuries per gram is the most common standardized unit for volumetric measurements for decommissioning activities and is used in this document.

Radiological survey instruments are made up of a meter and a detector. Collectively, this combination is referred to as an instrument. Pairings of a meter and a detector are controlled by calibration procedures.

The list of instrumentation currently being considered for use in support of the FSSP is provided in Table 4-I.

40

Alan J. Bloteky Reactor Facility Table 4-1: FSSP Instrumentation Instrument! I Detector Measurement Detector Detector Manufacturer Output Type Type Area and Model jUnits Surface Scintillation 100cm Ludlum model CPM alpha/beta-2360 with 43-gamma 9

Surface beta-G-M Tube 20-cm2 Ludlum model CPM gamma 2

3 (or 12) with mgcm 2 44-9 Gamma NaI l"x 1" Ludlum model pR/hr' exposure rate 19 Gamma scan NaI 2"x2" Ludlum model

CPM, 2221 with 44-

____10 Removable Scintillation 20-cmz Ludlum model DPM alpha/beta-0.8

,,3030E with 43-gamma mg/cm2 1.0-1 4.2 Instrument Statistics Radiological instrumentation will be selected to" provide both reliable operation and adequate detection sensitivity of the final list of the radionuclides of concern as identified for the Derived Concentration Guideline Limits (DCGLw§). Detector selection is based on detection sensitivity, operating characteristics and expected performance in the field. The instrumentation will, to the extent practical, be capable of data logging operations and have digital displays. Data logging provides the ability to electronically record the collected data, thereby reducing potential transcription errors. Digital displays reduce the observer interpretation of the measurement allowing for more accurate readings.

Commercially available portable and laboratory instruments and detectors are typically used to perform the basic survey measurements:

• The "w" in DCGLw stands for Wilcoxon Rank Sum test, which is the statistical test recommended in MARSSIM for demonstrating compliance when the contaminant is present in background. The Sign test recommended for demonstrating compliance when the contaminant is not present in background also uses the DCGLw.

41

• • '*Alan J. Bloteky Reactor Facility Surface scanning Direct surface contamination measurements Removable activity analysis Spectroscopy of soil and other bulk materials, such as concrete Instrument use and survey procedures detail the issuance, use, and calibration of instrumentation.

Records supporting the Instrumentation Program are maintained by the PM or designee in accordance with document control procedures.

Radiation instruments and the associated detectors are selected based on the type and quantity of radiation to be measured. The instruments used for direct measurements are generally capable of detecting the radiation of concern to a Minimum Detectable Concentration. (MDC) of less than 50% of the applicable DCGLwV. The use of 50% of the DCGL is an administrative limit only. Any value below the DCGLW is acceptable in Class t or 2 survey units. MDCs" of less than 50% of the DCGLw allow detection of residual activity in Class 3 survey units at an investigation level of 0.5 times the DCGLw. Instruments used for scan measurements are required to be capable of detecting radioactive material at less than or equal to the DCGLemnc.

A list of some common instrument brand and models is presented in Table 4-2 along with their nominal operational parameters e.g., background rate, probe area, total efficiency. A final determination of the specific brands and models will precede implementation of the FSS program.

VA will generally follow the instrument manufacturers' recommendations and/or supporting basis documents for considerations such as temperature dependency and other operational parameters.

Table 4-2: Typical Instrument Operational Parameters

Background

Probe Area Detection Instrument Model Range (em 2)

Efficiency %)

Scaler/Ratemeter Zinc 2 CPM 100 13.0 (4-t)

Sulfide. Scintillator (alpha)

Scaler/Ratemeter Geiger 50 CPM 17.5 19.0 (4-t)

Mueller Detector (beta/gamma)

Scaler/Ratemeter Gas 80 CPM 100 30.0 (4-7t)

Proportional Detector (beta/gamma)

MicroR Gamma 10-15 micro R/hr N/A N/A Scintillator 42

• im

.Alan J. Bloteky

!i*

Reactor Facility I Background Probe Area { Detection Instrument Model j

Range (cm 2)

{ Efficiency (%)

Removable Activity T 1 DPM/sample N/A 39.0 Counter (alpha)

I________________

Removable Activity

{

220 N/A

~

38.8 Counter (beta)

DPM/samnpleI As the project proceeds, other measurement instruments or technologies, such as in-situ gamma spectroscopy or continuous data collection scan devices, may be found to be more efficient than the survey instruments currently under consideration. The acceptability of sudh an inistrument or technology for use in the final survey program would be justified in a technical basis document.

The technical basis document would include, among other things, the following:

A description of the conditions under which the method or equipment would be used A description of the measurement method, instrumentation, and criteria Justification that the technique would provide equivalent scan coverage for the given survey unit classification and that the MDCscan' is adequate when compared to the DCGLeinc A demonstration that the method provides data that has a Type 1 error (falsely concluding that the survey unit is acceptable) equivalent to 5% or less and provides sufficient confidence that the DCGLe~nC criterion is satisfied 4.2.1 Total Detection Efficiency The total instrument efficiency (st) is* a calculation of the percentage of activity present in or on a surface that an instrument *detects. The *t is a product of two components, the instrument efficiency (si) and the surface efficiency (s*) as shown in the formula:

Equation 4.1 - Total Detection Efficiency

=t**

4.2.1.1 Instrument Efficiency The instrument efficiency (si) is a measurement of the surface emissions that interact with the detector elements with an instrument. To determine this, the instrument detector is exposed to a source of a known emission rate for a specified time period and the number of interactions is recorded. The efficiency percent is then calculated by:

43

• q!*m*Alan J. Blotcky

~Reactor Facility Eqiuation 4.2 - Instrument Efficiency (C5 -C)

Where:

Cs= Measured interaction count per one minute Cb = Measured background interaction count per one minute S -- known source value in DPM The known source value should be that of the hemispherical area (2-7) exposed to the detector as opposed to the total emission of the sphere around the source (4-vt). The surface efficiency (*s),

discussed below, accounts for the remaining half of the emission sphere.

4.2.1.2 Surface Efficiency Surface efficiency (*s) is an estimation of the affect the media surface has on the interaction of residual radioactive material with a detector and is a function of the surface condition, i.e.

smoothness, and the relative emission energy of the i'adionuclide. The *s for potentially contaminated structures and systems will follow recommendations contained in NUREG-1507 and ISO 7503-1 for the energies applicable to the radionuclides of concern developed in the characterization study. Beta-gamma detection instruments will use an *s of 0.5 and alpha detection instruments will use an

• of 0.25.

The methods for determining efficiency in NUREG-1507 were specifically developed to address situations when the source, in this case concrete, affects radiation emission rate due to self-attenuationa, backscatter, thin coverings, etc.

Media-specific* r. may he developed as necessary. These new surface efficiencies will subsequently override NUREG-1507 recommendations upon acceptance.

The condition of the surface being measured has an effect on the s* as well. For direct surface and scan measurements, the surface area beneath the detector should not have variability of depth greater than 0.5-cm more than the source to detector distance used for the instrument calibration.

According to NUREG-1507, instrument efficiency drops considerable when the source to detector distance increases more than 0.5-cm.

4.2.2 Minimum Detectable Concentration Minimum Detectable Concentration (MDC) and Minimum Detectable Activity (MDA) are 44

Alan J. Bloteky

• .'**:i*Reactor Facility estimations of the lowest level of concentration or activity the subject instrument can detect 95%

of the time. For purposes of this document MDC and MDA will be used interchangeably. Factors that directly affect the MDC are the total instrument efficiency (discussed in Section 4.2.1),

background rate, and count duration. The MDC is calculated individually for direct (static) measurements, ratemeter count rates, and scans. Static measurement count times will be adjusted to achieve the required MDC values.

4.2.2.1 Minimum Detectable Concentration for Direct Measurements Direct, or static, measurements are measurements where the detector is placed finto a fixed position and the instrument records individual unique counts for a specified, *period of* time.

Measurement data for direct surface measurements are recorded as counts per unit time per detector area prior to conversion to standardized units.

For" static (direct) surface measurements, with conventional detectors, the MDC will be calculated using Formula 3-10 in NUREG-1507:

Equation 4.3 - Minimum Detectable Concentration Direct measurement MDC:

MDC 3

.2 -

b(T)(1.STb Where:

MDC =Minimum detectable concentration (DPM/100-cm2)

Rb= Background count rate (CPM)

Tb

- Background count time (minute)

Td = Sample run time (minute) l's= Sample count time (minute)

,*=Counting system efficiency (decimal)

Direct measurements require an MDC less than the DCGLw, but an administrative limit of<*50%

DCGLw will be used to assure adequate sensitivity for the investigation levels applicable to lower unit classifications (Class 2 and Class 3).

4.2.2.2 Scan Measurement MDC Scan measurements are measurements taken with the detector in a steady motion over an area larger than the surface area of the detector and at a predetermined surface to detector distance.

Scan rates and surface to detector distances are specified for the type of emission (alpha, beta, 45

• • I*I*Alan J. Blotcky Reactor Facility etc.) in the Survey Design section of the MSP.

Scan measurement MDC (MDCscan).calculations depend on the emissions of the radionuclides of concern. Beta-gamma MDCscan is discussed in Section 6.7.2.1 of the MARSSIM. The desired MDCscaii for an area is a function of its classification. Because of the lower activity anticipated in lower class areas (Classes 2 and 3), a more stringent MDCsca, is recommended. These recommendations are presented in Table 2-3 of the MARSSIM Survey Plan.

Alpha MDCscan is calculated as a probability of detection as outlined in Section 6.7.2.2 of the MARS SIM. A more detailed discussion appears in Appendix J of that document.

4.2.2.2.1 Beta-Gamma Scan MDC The MDCscan for beta-gamma measurements may be calculated by first deterniining the Minimum Detectable Count Rate (MDCR). The MDCR is calculated by first defining the minimum detectable net source counts (si) using Formula 6-8 from the MARSSIM as below.

Equation 4.4 - Minimum Detectable Source Counts Si = d' b-*

Where:

d'= value taken from Table 6.5 in the MARSSIM for applicable true and false positive rates bi= Number of background counts in a given time interval The MDCR is then calculated fr'om Formula 6-9 in the MARSSIM:

Equation 4.5 - Minimum Detectable Count Rate 60 MDCR = S* *-

Where:.

i = Observed time interval in seconds Finally, applying the detection efficiency correction results in an MDCscan in standardized units (DPM/100-cm 2) fr'om Formula 6-9 in NUREG-1507:

46

Alan J. Blotcky Reactor Facility Equation 4.6 - Scan MDC MDCR MlC...

=

probearea

/P**i~s*

100cm 2 Where:

p3 = Surveyor efficiency (value from a range between 0.5 and 0.75)

• i= Instrument efficiency as= Surface efficiency The value for p has been developed in Draft NU7REG/CR-6364 and NUREG-1507 and, is a percentage estimate of the likelihood a surveyor will reliably detect an elevated count rate.

Currently, the equations presented here use a value for p of 0.5.

Table 4-3: Required Minimum Detectable Concentrations 1:Required Static fRequired Instrument Model Release Criteria J MDC*

MDCscan Scaler/Ratemeter Zinc 1,000 Sulfide Scintillator (alpha)

DPM/100cm2 500 DPM/100cm2 N/A Scaler/Ratemeter Zinc 7,100 3,550 7,100 Sulfide Scintillator DP/0c2 DM10m2DM/0c2 (beta/gamma)DM/00mDM10c 2

DM10m MicroR Gamma Scintillator N/A N/A N/A Removable Activity*

N/A 100 DPM/1 00cm2 N/A Counter (alpha)

Removable Activity 710 DPM/100cm2 355 DPM/100cm2 N/A Counter (beta) 4.3 Instrumentation Overview An instrument is a combination meter and detector. The meter houses the power supply and electronics which record the count data. Meters may have specialized abilities, such as data logging, multiple detectors, digital display, multiple modes (scaler, ratemeter, etc.), GPS output,

• • Required static measurement MDC values are approximately 50 percent of the most restrictive release criteria 47

B*

Alan J. Blotcky Reactor Facility etc.

4.3.1 Radiological Instruments 4.3.1.1 Ludlum Model 2360 The Ludlum Model 2360 (L23 60) is a digital/analog display instrument capable of scaler and ratemeter/scan operations with one or two detection channels (dual discriminator). The L2360 has data logging capabilities and may be calibrated with one detector type at a time without adjusting operating parameters (voltage, threshold, etc). Common applications for the L2360 are to *match it with a Ludlum Model 43-44 air proportional or Model 43-68 gas proportional..detector and calibrate it for alpha and/or beta emitters. In dual discriminator mode, it has thle ability to distinguish between alpha and beta on separate channels. The drawback to dual discriminator mode is lower detection efficiencies when compared to single channel calibration.

4.3.1.2 Ludlum Model 2221 The Ludlum Model 2221 (L2221) is a single-channel scaler/ratemeter with digital and analog display. The L2221I may be fitted with a GPS connection so that textual output of the count data may be sent to an external GPS unit for synchronization to location data. This meter may also be used as a single-channel analyzer with the proper setup of operating parameters (voltage, threshold, window, etc) which allows it to be used as a very capable screening tool for known isotopes. Common detector pairings are with a L43-68 for alpha or beta detection or a Ludlum Model 44-10 (L44-10) Sodium iodide scintillator for gross gamma capabilities. ENERCON routinely uses the L2221 with a Ludlum Model 44-10 detector connected to a Trimble ProXR GPS system for open land gross gamma scan surveys.

4.3.1.3 Ludlum Model 3/12 The Ludlum Model 3 and Model 12 (L3/12) are nearly identical instruments with the only appreciable difference being the displays. Both are ratemeter-only instruments with analog displays. This limits their usefulness in final status survey applications to limited scanning or elevated area delineation.

4.3.1.4 Ludlum Model 19 The Ludlum Model 19 (L19) is a MicroR meter with a 1"xl" Nal crystal. The L19 has an analog display and is capable of detection ranges up to 5,000 microR per hour. The low display range and steady response make this a good instrument for ambient exposure rate surveys. While the detection response may vaiy¢ by isotope, as a hand-held exposure rate meter it is quite capable.

48

_, L Alan J. Blotcky

~Reactor Facility etc.

4.3.1 Radiologieial Instruments 4.3.1.1 Ludlum Model 2360 The Ludlurn Model 2360 (L2360) is a digital/analog display instrument capable of scaler and ratemeter/scan operations with one or" two detection channels (dual discriminator). The L2360 has data logging capabilities and may be calibrated with one detector type at a time without adjusting operating parameters (voltage, threshold, etc). Common applications for the L2360 are to match it with a Ludlum Model 43-44 air proportional or Model 43-68 gas proportional. detector and calibrate it for alpha and/or beta emitters. In dual discriminator mode, it has the ability to distinguish between alpha and beta on separate channels. The drawback to dual discriminator mode is lower detection efficiencies when compared to single channel calibration.

4.3.1.2 Ludlum Model 2221 The Ludlum Model 2221 (L2221) is a single-channel scaler/ratemeter with digital and analog display. The L2221 may be fitted with a GPS connection so that textual output of the count data may be sent to an external GPS unit for synchronization to location data. This meter may also be used as a single-channel analyzer with the proper setup of operating parameters (voltage, threshold, window, etc) which allows it to be used as a very capable screening tool for known isotopes. Common detector pairings are with a L43-68 for alpha or beta detection or a Ludlum Model 44-10 (L44-l0) Sodium iodide scintillator for gross gamma capabilities. ENERCON routinely uses the L2221 with a Ludlum Model 44-10 detector connected to a Trimble ProXR GPS system for open land gross gamma scan surveys.

4.3.1.3 Ludlum Model 3/12 The Ludlum Model 3 and Model 12 (L3/12) are nearly identical instruments with the only appreciable difference being the displays. Both are ratemeter-only instxniments with analog displays. This limits their usefulness in final status survey applications to limited scanning or elevated area delineation.

4.3.1.4 Ludlum Model 19 The Ludlum Model 19 (L19) is a MicroR meter with a 1"x1" NaI ciy¢stal. The L19 has an analog display and is capable of detection ranges up to 5,000 microR per hour. The low display range and steady response make this a good instrument for ambient exposure rate surveys. While the detection response may vary by isotope, as a hand-held exposure rate meter it is quite capable.

48

[*;

Alan J. Blotcky

)!*

Reactor Facility Mor'e accurate measurements may be obtained with an instrument such as one of the Reuter-Stokes Pressurized Ion Chamber models, but for portable operational needs, the LI19 is sufficient.

4.3.2 Radiation Detectors 4.3.2.1 Ludlum Model 43-93 The Ludlum Model 43-93 (L43-93) is a Zinc Sulfide (ZnS) alpha and beta scintillation with a Mylar window. The probe active surface area is 1 00-cm 2. Typical 4-2t detection efficiency si is

• 9%-1 5% depending on the meter voltage and other calibration parameters. This detector is most commonly used on hard, relatively smooth surfaces, such as concrete. Care must be taken iniareas where sharp edges are present to avoid puncturing the Mylar window or breaking an anode wire.

4.3.2.2 Ludlum Model 44-9 The Ludlum 44-9 (L44-9) Geiger-Mueller tube detector is Useful for alpha, beta, and gamma detection for small, discreet areas. It is ideally suited for very small areas where relatively high levels of residual contamination are possible. While it has fairly good typical detection efficiencies from 5-32% (4-3z), the small detection area severely limits its' suitability for FSS.

4.3.2.3 Ludlum Model 44-10 The Ludlum Model 44-10 (L44-10) detector is a gamma scintillator with a 2" x 2" Sodium Iodide (NaI) crystal. It is commnonly used for open area gross gamma surveys. Typical count response is approximately 900 counts per minute per microR per hour for Cs137 energy (662 keV).

4.4 Calibration and Maintenance Instrumnentation used for FSS are calibrated and maintained in accordance with the Instrumentation Program procedures. Radioactive sources used for calibration are traceable to the National Institute of Standards and Technology (NIST) and use, to the extent practical, geometries designed to mimic the type of samples being counted. If vendor services are used, these are obtained in accordance with purchasing requirements for quality related services to ensure an appropriate level of quality. The calibration source isotopes for beta survey instruments are Tc 99 and Sr90 which have an appropriate energy for site-specific beta emitting nuclides. The alpha calibration sources are Th23° and Pu239 which have an appropriate alpha energy for plant-specific alpha emitting nuclides. Gamma scintillation detectors generally are calibrated using Cs'37. Calibration sources other than those listed may be used provided they demonstrate appropriately conservative detection efficiency for the radionuclides of interest. In all cases, the surface efficiency as determined for the energies of the radionuclides of concern is utilized 49

Jm Alan J. Blotcky

~Reactor Facility regardless of the energy of the calibration source.

4.4.1 Response Checks Instrument response checks are conducted to assure proper instrument response and operation.

Appropriate response values are discussed in the applicable instrument program procedures.

Response checks are performed daily before instrument use and again at the end of use. Check sources are appropriate for the type of radiation as that being measured in the field and are, to the extent practical, held in fixed geometry jigs for reproducibility, if an instrument fails a response check, it will be clearly labeled "Do Not Use" and will be removed from service until the problem is corrected in accordance with applicable procedures. Measurements made between the last acceptable check and the failed check will be evaluated to determine if they may remain in the data set.

50

m*

Reactor Facility 5.0 QUALITY ASSURANCE 5.1 Project Description and Schedule Each area of the site is divided into survey areas and units, and is classified as listed in Section 3.5. The survey measurements for each survey unit are determined during the survey design phase.

Portions of the FSS are performed during deconstruction activities as areas become available for survey.

Non-impacted and Class 3 areas may be evaluated for release prior to significant decommissioning activities taking place.

5.2 Quality Objectives and Measurement Criteria Type I errors are established at 0.05 unless another value is authorized by the USNRC. Type II errors will initially be set at 0.05, but may be increased with concurrence by the SO.':::

5.2.1 Trainina and Qualification i

Personnel performing FSS measurements will be trained and qualifiedi At a minimum, training will include the following topics:

Procedures governing handling FSS~, data such as, but not limited to, document control, records retention, and ~chain,of custody.

Operating field and laboratory instrumentation used for FSS.

Performing FSS measurements and collecting samples.

The extent of training and qualification will be commensurate with the education, experience and proficiency of the individual, and the.scope, complexity and nature of the activity.

Training records will be maintained as quality records.

5.2.2 Survey Document~ation Each FSS measuirement will be identified by date, instrument, location, type of measurement, and mode of operation.-: Generation, handling and storage of the original FSS design and data packages will be controlled. The FSS records have been designated as quality documents and will be maintained in accordance with document control procedures.

5.3 Measurement/D~ata Acq uisition 5.3.1 Survey Desiin The site will be divided into survey areas. Each survey area will contain one or more survey units. A survey package specifies the type and number of measurements required for a survey unit based on the classification and known characterization data results. Each survey area will 51

• ,UE*

Alan J. Blotcky Reactor Facility have one or more survey packages.

5.3.2 Written Procedures Sampling and survey tasks must be performed properly and consistently in order to assure the quality of the FSS results.

The measurements are performed in accordance with approved, written procedures. Approved procedures describe the methods and techniques used for the FSS measurements.

Each procedure written for the purpose of directing FSS data collection or evaluation shall include a section describing the QA/QC goals and methods specific to that procedure.

5.3.3 Sampling Methods" Samples are collected and placed into new containers using either new tools or tools that have been thoroughly decontaminated and double-rinsed with clean, water from two sources, i.e. two separate dip tanks. Surface abrasion of the tools may be necessary to dislodge adhered media from previous samples. This may entail using a stiff-bristled brush in the first dip tank. Tools will be air-or towel-dried prior to reuse.

5.3.4 Chain of Custody Responsibility for custody of samples from the point of collection through the determination of the final survey results is established by procedure.

When custody is transferred, a COC will accompany the sample for tracking purposes. Secure storage is provided for archived samples until such time it is determined to no longer be necessary, i.e. license termination for samples utilized for FSS.

5.4 OA/QC Surveys and Samples If replicate QA/QC measurements or sample analyses fall outside of their respective acceptance criteria, a documented investigation is performed in accordance with approved procedures; and the Corrective Action Process described in Section 5.10.3 will be implemented. The investigation typicalliyinvlmv*ves verification that the proper data sets were compared, the relevant instruments were operating properly and the survey/sample points were properly identified and located.

Relevant personnel are interviewed, as appropriate, to determine if proper instructions and procedures were provided and followed, and proper measurement and handling techniques were used, including COC, where applicable. When appropriate, additional measurements are taken.

Following the investigation, a documented determination is made regarding the usability of the survey data and if the impact of the discrepancy adversely affects the decision on the radiological 52

F;2*

Alan J. Blotcky

~Reactor Facility status of the survey unit.

5.4.1 Replicate Measurements Quality assurance of the survey process is evaluated using one or more of the following methods and is performed on at least 5% of the collected data. Replicate measurements generally involve a full redesign of the survey package. When a full replicate survey redesign is not desired or practical, replicate scan or direct measurements may be performed to satisfy the 5% minimum requirement for QA/QC evaluation.

QA/QC replicate measurements will be performed once a survey unit FSS has,been completed.

At a minimum, one replicate measurement will be performed for each survey iunit ais a standard practice. This will be done to provide an early indication of potential problems With final status V,--7 survey procedures/guidelines, instrumentation, and/or possible cross-c*ontamination from ongoing site decontamination activities. All replicate measurements wilt be perfoirmed in accordance with the FSSP. Verification of final status surveys will include the finvestigation of documentation, survey protocol, instrument calibration, and any practice 'associated with the collection of survey data. Replicate measurements may be completed by a differ~ent technician than performed the original survey, the SO, or SO.

5.4.2 Volumetric Analyses,,

For volumetric samples, Quality Control will consist of requiring the vendor analytical laboratory to be NVLAP accredited. However, as an additional quality measure, randomly selected samples are subject to blank sample, blind duplicate, split, recount, or third party analyses.

The acceptance criterion for blank saimples is that no plant-derived radionuclides are detected to the required MDA. Some sample media, such as asphalt, will only be subject to third party analyses or a sample recount due to the lack of homogeneity. The criterion for blind duplicates, split, recount, and third'-party analyses is that the two measurements are within +20% of each other.

5.5 Instrument Selection. Calibration and Operation Proper selection and use of instrumentation ensure sensitivities are sufficient to detect radionuclides at specified MDC. These requirements help assure the validity of the collected survey data.

Instrument calibrations are performed with sources traceable to the National institute for Standards and Testing (NIST) using approved procedures. Issuance, control and operation of the survey instruments are conducted in accordance with the Instrumentation Control and Issue procedure.

53

F tUL Alan J. Blotcky Reactor Facility 5.6 Control of Consumables In order to ensure the quality of data obtained from FSS surveys and samples, new sample containers will be used for each sample taken. Tools used to collect samples will be cleaned to remove contamination prior to taking additional samples. Tools will be decontaminated after each sample collection and surveyed for contamination.

5.7 Control of Vendor-Supplied Services Vendor-supplied services, such as instrument calibration and laboratory sample analysis,, will be procured from vendors on the site Approved Vendor List (AVL) in accordance with approved quality and procurement procedures.

5.8 Software Control Software used for data reduction, storage, or evaluation will be fully documented and certified by the vendor. The software will be tested prior to the initial use by an appropriate test data set.

5.9 Data Manaaement Survey data control from the time of collection through evaluation is specified by procedure. All survey and data analysis records pertaining to the final radiological status are considered quality records and are maintained in accordance with applicable document control procedures.

5.10 Assessment and Oversight 5.10.1 Assessments Periodic FSS self-assessments are conducted in accordance with approved procedures.

The findings are tracked and trended in accordance with these procedures.

5.10.2 Independent Review of Survey Results A minaimum of 5% of randomly selected survey packages from completed survey units are independentlyreviewed by QA to ensure that the survey measurements are taken and documented in accordance with approved procedures.

5.10.3 Corrective Action Process The corrective action process, already established as part of the site 10 CFR Part 50 Appendix B Quality Assurance Program, is applied to FSS for the documentation, evaluation, and implementation of corrective actions.

The process will be conducted in accordance with approved procedures which describe the methods used to initiate Condition Reports (CR) and 54

!Jm*

Alan J. Blotcky Reactor Facility resolve self-assessment and corrective action issues related to FSS. The CR evaluation effort is commensurate with the classification of the CR and could include root cause determination, barrier screening and extent of condition reviews.

5.10.4 Reports to Management Reports of audits, assessments, and trend data are reported to management upon completion.

5.11 Data Validation Survey data are reviewed prior to evaluation or analysis for completeness and for the presence of outliers.

Comparisons to investigation levels are made and measurements :exceeding the investigation levels are evaluated. Verified data are subjected to the Sign test,* theWRS test, Sign Unity test, or WRS Unity test as determined appropriate by the assignied SO.

5.12 Confirmatory Measurements

{:

It is anticipated that the USNRC and other regulatory ageicies:' will choose to conduct confirmatory measurements in accordance with applicable laws and regulations.

The USNRC may take confirmatory measurements to make a determination in accordance with 10 CFR 50.82(a)(1 1) that the final radiation survey 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. VA will comply with the 25-mrem/yr criteria of 10 CFR Part 20, Subpart E by demonstrating that measurement results meet the DCGLw developed using the approved dose-based scenario. Therefore, the confirmatory measurements taken by the USNRC and other regulatory agencies are based on the same DCGLw.

Timely and frequent communications with these agencies ensure that they are afforded sufficient opportunity for these confirmatory measurements prior to any difficult to reverse decommissioning actions, e.g. new construction or lexcavation backfill.

55

Alan J. Blotcky

• lj Reactor Facility

6.0 REFERENCES

• Alan J. Blotcky Reactor Facility Decommissioning Plan (Revised), May 7, 2013

• 10CFR20.1402, "Radiological Criteria for Unrestricted Use"

• 10CFR50.82, "Termination of License"

• NUREG-15 75, "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSLM), Revision 1, August 2002

• NUREG-1507, "Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Field Conditions," December 1997" NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys," Rev. 1, June 1998 draft NUREG-1549, "Using Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination," Ju1i* 1998 draft NUREG-1757, "Consolidated Decommissioning Guidanace," September 2006 NUREG-1727, "NMSS Decommissioning Standard Review Plan", September 2000

• 1O 7503-1, "Evaluation of Surface Co nta~inao",18 inaton",198 56

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2 Survey Unit5 Area: B540A X Coord V Coord Z Coord

...Type Surface LX LY Ref/Surv 13 42 7

Random Ceiling 7

2 Survey Unit 5 Area: B540 X Coorcd V coor d z Coord Type Surface LX LY Ref/Surv 23497 anom Celng6 Srvy ni 230 52 7

Random Ceiling 13 8

Survey Unit 5 Area: 8522 X Coord Y Coord Z Coord Type Surface LX LY Ref/Surv 42 63 0

Random Floor 2

6 Survey Unit 5 Area: B522A_______

X Coord Y Coord Z Coord Type Surface LX LY Ref/Surv 411 49 0 Random Floor 1

4 Survey UnitS5 P a ge 23 1 31

FSSP Survey Maps-Omaha Blotcky Reactor Decommissioning Area: B526 X Coord Y Coord Z Coord Type Surface LX LY Ref/Surv 44 35 7

Random Ceiling 13 35 Survey Unit 5 38 53 7

Random Ceiling 7

52 Survey Unit 5 35 18 7

Random Ceiling 4

18 Survey Unit 5 47 36 7

Random Ceiling 16 36 Survey Unit 5 53 24 7

Random Ceiling 22 24 Survey Unit 5 34 42 7

Random Ceiling 2

41 Survey Unit 5 46 13 7

Random Ceiling 14 12 Survey Unit 5 403 7Rndm Ciln 83 ure Ui 37 30 7

Random Ceiling 5

20 Survey Unit 5 49 20 7

Random Ceiling 17 20 Survey Unit 5 43 38 7

Random Ceiling 11 38 Survey Unit 5 32 26 7

Random Ceiling "0

26 Survey Unit 5 44 44 7

Random Ceiling 12 43 Survey Unit 5 38 15 7

Random Ceiling 6 -14 Survey Unit 5 B533AA P a ge 24 I 31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning BS33A 8535153?

P a ge 25 I31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning B535A Pad ge 261I31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning B540A Pai ge 271I31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning B540 i

P a ge 281j 31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning 8522 P a ge 291I31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning P a ge 30 I 31

FSSP Survey Maps - Omaha Blotcky Reactor Decommissioning B526 P ~ ge 31 I31

North Star Federal Services Final Status Survey Plan AJ Blotcky Reactor Facility Decommissioning Project.......

AJBRF-FSS-0 1 Rev. B Approved October 5, 2015 Prepared, by"...

NorthStar Feder:al Services, Inc.

1992"'Saint Street Richiand, WA 99354

• and

.:Enercon Federal Services, Inc.

,,679 Emery Valley Road, Suite A "Oak Ridge, TN 37830

~Prepared for

,,,..*:FUnited States Department of Veterans Affairs Nebraska-Western Iowa Health Care System Contract Number VA701-C-15-0005 NorthStar Project Number 1554005

AJ Blotcky Reactor Facility Final Status Survey Plan AJ Blotcky Reactor Facility Decommissioning Project Final Status Survey Plan AJBRF-FSS-01 Rev. B Approved October 5, 2015 Prepared by:

Reviewed by:

Reviewed by:

Approved by:

Approved by:

Todd S. Brautigam

?

,I2 Radiation Safety Officer Billy Reid NorthStar Program Manager ENERCON Program Manager*

Nelson Langub Q,*l NorthStar Project Manager ">.

Chanda Joshi;'VA'COR-:

Department of Veterans Affairs ii

AJ Blotcky Reactor Facility Final Status Survey Plan Summary of Changes Revisions to this Final Status Survey Plan will be tracked, and revisions or addenda will be issued as needed. The Project Manager maintains the signed original of the FSSP; no controlled copies are issued.

The end user is responsible to verify' with the Project Manager that any hardcopy being referenced is the current revision.

A summary description of each revision or addenda will be noted in the following table.,,

Revision Number Date Comments° Rev. A Draft July 28, 201 5 Original Issue "i

Rev. B Draft October 5, 2015 Addresses VA/AECOM comments Rev. B Approved October 14, 2015 Approved by VA

°°.

AJ Blotcky Reactor Facility Final Status Survey Plan Table of Contents PAGE 1.0 TNTRODUCTION................................................................................ 1 2.0 RESPONSIBILITIES............................................................................ 2 3.0 FINAL STATUS SURVEY PLAN............................................................. 3 3.1 Purpose............................................................................................. 3 3.2 Overview 3

3.2.1 Survey Preparation................................................................. i'2........3 3.2.2 Survey Design.......................................................

7:....... :......

.i......

4 3.2.3 D ata Collection 4.....................

3.2.4 Data Assessment and Evaluation..............

4 3.2.5 Documentation of Survey Results.......

5 3.3 Implementation..........................5 3.4 Radionuclides of Concern.............i...,...................................................... 6 3.5 Area Classification..........i.... ?..."........................................................... 6 3.5.1 Non-Impacted Areas..... :2,..i.l............................................................ 7 3.5.2 Impacted Areas.....,...,-i....................................................... 7 3.5.3 Changes in Classification................................................................... 9 3.5.4 Redefining Survey Boundaries.............................................................. 9 3.6 Establishling Survey Units........................................................................ 9 3.6.1 SurveyUnit................................................................................. 10 3.7 Survey Design.................................................................................... 12 3.7.1 Scan Coverage.............................................................................. 12 3.7.2 Sample Size Determination................................................................ 13 3.7.3 Background Reference Areas.............................................................. 18 3,7.4 Sample Grid and Measurement Location................................................. 19 iv

AJ Blotcky Reactor Facility Final Status Survey Plan 3.7.5 Survey Package Design Process........................................................... 19 3.8 Types of Surveys................................................................................. 23 3.8.1 Scan Surveys................................................................................ 23 3.8.2 Direct Measurements....................................................................... 24 3.8.3 Exposure Rate............................................................................... 24 3.8.4 Removable Activity.................................................................. iii:.....25 3.8.5 LSC Analysis.............................................................

ii...;..

....... i..25 3.8.6 Volumetric Samples.................................................;....ii..*

............. 25 3.9 Survey Methods.........................................................

,,:........................ 25 3.9.1 Buildings, Equipment, and Components...........

,... :,................................. 26 3.10 Instrumentation................................................................................... 26 3.10.1 Selection.......................

,..... :..,.................................................... 27 3.10.2 Calibration and Maintenance...i..... 2....:................................................ 27 3.10.3 Response Checks..............

............................................................. 28 3.10.4 Minimum Detectable Concentration for Direct Measurements........................ 28 3.10.5 Scan Measurement MDC;............................... Error! Bookmark not defined.

3.11 Investigation Levels and.El'evated Areas Test................................................. 28 3.11.1 Investigation Levels........................................................................ 29 3.1,1.2 Investigation Process....................................................................... 29 3.11.3 Elevated Measurement Comparison (EMC)............................................. 30 3.11.4 Remediation and Reclassification......................................................... 31 3.1 1.5 Reclassification and Resurvey............................................................. 32 3.1I2 Data Collection and Processing................................................................ 32 3.12.1 Sample Handling and Record Keeping................................................... 32 3.12.2 Data Management.......................................................................... 33 v

AJ Blotcky Reactor Facility Final Status Survey Plan 3.12.3 Data Verification and Validation.......................................................... 33 3.12.4 Graphical Data Review.................................................................... 34 3.13 Data Assessment and Compliance.............................................................. 34 3.13.1 Data Evaluation............................................................................. 34 3.14 Statistical Conclusions........................................................................... 37 3.15 Reporting Format.............................................................. "':""'"....... ::... 38 3.15.1 Survey Unit Release Record...............................................

.,.i............38 3.15.2 Final Status Survey Report............................. i........... !........................ 39 4.0 SURVEY INSTRUMENTATION................

40 4.1 Overview........................................................................................... 40 4.2 Instrument Statistics...................... :..:.................................................... 41 4.2.1 Total Detection Efficiency..........

i...... i................................................ 43 4.2.2 Minimum Detectable Concentration...................................................... 44 4.3 Instrumnentation Overview....................................................................... 47 4.3.1 Radiological Instruments................................................................... 48 4.3.2 Radiation Detectors..:..................................................................... 49 4.4 Calibration and Maintenance................................................................... 49 4.4.1 Response Checks........................................................................... 50 5.0 QUALITY ASSURANCE..................................................................... 51 5.1 Project Description and Schedule.............................................................. 51 5.2 Quality Objectives and Measurement Criteria................................................. 51 5.2.1 Training and Qualification................................................................. 51 5.2.2 Survey Documentation..................................................................... 51 5.3 Measurement/Data Acquisition................................................................. 51 5.3.1 Survey Design.............................................................................. 51 vi

AJ Blotcky Reactor Facility Final Status Survey Plan 5.3.2 Written Procedures......................................................................... 52 5.3.3 Sampling Methods.......................................................................... 52 5.3.4 Chain of Custody........................................................................... 52 5.4 QA/QC Surveys and Samples................................................................... 52 5.4.1 Replicate Measurements................................................................... 53 5.4.2 Volumetric Analyses......................53 5.5 Instrument Selection, Calibration and Operation..................................

.. :;.......53 5.6 Control of Consumables............................................................

.............. 54 5.7 Control of Vendor-Supplied Services.................

i...... '..........

........ 54 5.8 Software Control....................................... :.......i.................................. 54 5.9 Data Management......................

.......................................................... 54 5.10 Assessment and Oversight..................-..... ::.......................

• ........................ 54 5.10.1 Assessments.............

.................................................................. 54 5.10.2 Independent Review of Survey Res'ults................................................... 54 5.10.3 Corrective Action Process.................................................................. 54 5.10.4 Reports to Management.................................................................... 55 5.11 Data Validation..........................................

55 5.12 ConfirmatorS Measurements.................................................................... 55

6.0 REFERENCES

..........................56 vii

AJ Blotcky Reactor Facility Final Status Survey Plan Index of Figures Figure 3-1: Interior Space Coordinate Example....................................................... 12 Index of Tables Table 3-1: Radionuclides of Concern..........................................................

• i.......... 6 Table 3-2: AJBRF MARSSIM Classifications for FSS...............................,i..i.......

Table 3-3: Survey Unit Size.....................................................

.:.............. 10 Table 3-4: Minimum Scan Coverage............................................. !.............. :........13 Table 3-5: MARSSIM Table 5.2..............................,........,..................................1!4 Table 3-6: MARSSIM Table 5.4.........................

2.......

............................. 1!6 Table 3-7: MVARSSIM Table 5.1..................

..................................................... 17 Table 3-8: Measurement Result InvestigatiOn Levfels...,............................................... 29 Table 4-1: FSSP Instrumentation.....

..... i.:...i.i...................................................... 41 Table 4-2: Typical Instrument Operational Paramneters............................................... 42 Table 4-3: Required Minimum Detectable Concentrations........................................... 47

,:,,,i

AJ Blotcky Reactor Facility Final Status Survey Plan Acronyms Acronym AJBRF AVL CFR COG DCGLernc DCGLw DP DPM DQO EMC ENERCON FSS FSSP GPS HTD LBGR MARSSIM MDA MDC MDCSoaIn NaI NIST.

PM QA/QC ROC SO TEDE USNRC VA VA-PM WRS Description A.J. Blotcky Reactor Facility Approved Vendor List Code of Federal Regulations Chain Of Custody Derived Concentration Guideline Level, Elevated Measurement Criteria Derived Concentration Guideline Limit, Wilcoxon Rank Sum Decommissioning Plan",*

Disintegrations Per Minute Data Quality Objectives Elevated Measurement Criteria ENERCON Services, Inc.

Final Status Survey Final Status Survey Plan Global Positioning System Hard To Detect Lower Boundary of the Gray Region NUREG-1575 'Multi-Agency Radiation SUrvey and SiteI1nvestigation Manual' Minimum Detectable Activity Minimum Detectable Concentration Minimum Detectable Concentration.via Scan ---

Sodium Iodide National Institute for Standards and Testing ENERCON Project Manager Quality Assurance/Quality Control Radionucldeis of Concern Safety Officer i

Total Effective Dose Equivalent United States Nuclear Regulatory Commission Veterans Administration Veterans Administration Project Manager Wilcoxon Rank Sum ix

• ili*i Alan J. Blotcky ti*

Reactor Facility

1.0 INTRODUCTION

The Veterans Health Administration (VA) is in the process of decommissioning of the Alan J.

Bloteky Reactor Facility (AJBRF). This plan is associated with control measures that are intended to enable the Final Status Survey (FSS) to be applied to the AJBRF building surfaces and any associated soils beneath or surrounding the structure. Decommissioning Planning for the AJBRF utilizes NUREG-1575, Revision 1 'Multi-Agency Radiation Survey and Site Investigation Manual' (MARSSIM) and other guidance documents. Additional resource documents are listed in Section 6.0, References.

This document describes the Final Status Survey Plan (FSSP) in terms of the process that will be used.

The goal of implementing the FSSP is to document the data collected by radiological surveys of the facility's final radiological status in support: 0f terminating the applicable radioactive materials license. These 'Final Status Surveys' thierefore provide the inputs feeding the MARS SIM statistical evaluation process. For the remainder-of this document, 'Final Status Survey' or 'FSS,' is synonymous with 'MARSSIM SUrVey.'

References are made in this text to site-or process-sPecific information such as characterization surveys, background studies, operational procedures, instrumentation, and other items. Further details of these specific issues are disculssed ini the appendices or other plans or reports currently in process. When applicable, references to these documents will be supplied or the documents attached to complete the chain of informhation.

1

• i~i Alan J. Bloteky

~Reactor Facility 2.0 RESPONSIBILITIES Northstar will be responsible for all facets of the FSS including development of applicable plans and procedures governing performance of the FSS. ENERCON has provided two individuals knowledgeable in the decommissioning process. These individuals will function as the Project Radiation Safety Officer and Health and Safety Officer. During the project at least one of these individuals will be on-site during decommissioning activities. When only one is on-site, he will act as both the Radiation Safety Officer and Health and Safety Officer and is collectively referred to as the Safety Officer (SO) throughout the remainder of the FSSP.

The SO is responsible for preparing, implementing, and managing the FSS prograim.The SO has ultimate responsibility for program direction, technical content, andi, ensuring :tlie program complies with applicable United States Nuclear Regulatory Commnission (USNRC) regulations and guidance. The SO is responsible for resolving issues or concerns raisedt by USNRC, other regulatory agencies, or stakeholders, as well as programmatic issues raised by facility management. The SO reviews and approves the qualification and selection of FSS personnel and approves the content of training to FSS personnel and other personnel on FSS topics. The SO approves reports of FSS results.

2

• i*

Alan J. Blotcky Reactor Facility 3.0 FINAL STATUS SURVEY PLAN 3.1 Purpose The purpose of the FSSP is to describe the survey process that will be used to demonstrate that the AJBRF facility and site comply with the USNRC annual dose limit criteria of 25-mrem/yr Total Effective Dose Estimate (TEDE), by meeting the radiological release criteria fr'om Section 6.0 of the approved Decommissioning Plan (DP) for the facility. The ultimate goal is to support license termination for the facility.

ii 3.2 Overview The FSS includes remaining structures, land, and reactor systems that are identified as contaminated or potentially contaminated as a result of licensed activities." There are 5 major steps in the final survey process:

1.

Survey preparation

2.

Survey design

3.

Data collection

4.

Data assessment and evaluation

5.

Documentation of survey results 3.2.1 Survey Preparation Survey preparation is the first step in the final survey process and occurs after remediation is completed. In areas where remediation was required, a turnover, or post-remediation, survey is performed to confirm that remediation was successful prior to initiating final survey activities. A turnover surveymay be performed using the same process and controls as FSS so that data from the turnover survey can be used as final survey data. In order for tumnover survey data to be used for FSS, it must have been designed and collected to meet quality assurance and quality control (QA/QC) criteria described in Section 5.0, Quality Assurance, and the area must be controlled in accordance with that section. Following the turnover surveys, the FSS is performed.

The area to be surveyed is isolated and/or controlled to ensure that radioactive material is not reintroduced into the area fr'om ongoing demolition or remediation activities nearby, and to maintain the final configuration of the area.

Tools, equipment, and materials not needed to support survey activities are removed, unless authorized by the SO. Routine access, material storage, and worker transit through the area are not allowed, unless authorized by the SO.

3

• 'ii'*t..........

Al an J. Bloteky

• m

• • Reactor Facility However, survey areas may, with proper approval, be used for staging of materials and equipment providing; Staging does not interfere with survey performance.

Material or equipment is free of surface contamination or radioactive materials.

Survey personnel safety is not jeopardized.

The SO conducts an inspection of the area to ensure that work is complete and the area is ready for FSS.

Control of activities in that area is then transferred for the performance-of FSS.

Approved procedures provide isolation and control measures until the area is released for unrestricted use.

I 3.2.2 Survey Design The survey design process establishes the methods and performance criteria used to conduct the survey. Survey design assumptions are documented in "Survey Packages" in accordance with approved procedures. Structures and systems are organized into, survey areas and classified by contamination potential of the area. Survey unit size-is based on the assumnptions in the dose assessment models in accordance with the guidance provid~ed in the MARSSIM.

The percent coverage for scan surveys is discussed in Section,3.7.1.

The number and location of structural surface measurements (and/or structural volumetric samples) and soil samples (if necessary) are established in accordance with Section 3.7.2. Investigation levels are also established in accordance with Section 3.10.6.

Replicate measurements are performed as part of the quality process established to identify",

assess, and control errors and the uncertainty associated with sampling, survey, and/or analytical activities. This quality control measure, described in Section 5.4.1, provides assurance that the survey data meets the accuracy and reliability requirements necessary to support the decision to release or not release a survey unit.

3.2.3 Data Collection After preparation of a survey package, the FSS data are collected. Measurements are performed using calibrated instruments in accordance with approved procedures and instructions contained in the survey package.

3.2.4 Data Assessment and Evaluation Survey data assessment is performed to verify that the data are sufficient to demonstrate that the survey unit meets the unrestricted use criterion (i.e., the Null Hypothesis may be rejected.).

4

• i*

Alan J. Bloteky

• tli~m*Reactor Facility Statistical analyses are performed on the data and the data are compared to investigation levels.

Depending on the results of an investigation, the survey unit may require further remediation, reclassification, and/or resurvey. Graphical representations of the data, such as posting plots or histograms, may be generated to provide qualitative information from the survey and to verify' the assumptions in the statistical tests, such as spatial independence, symmetmy,, data variance and statistical power. The assumptions and requirements in the survey package are reviewed by the SO. Additional data needs, if required, are identified during this review.

3.2.5 Documentation of Survey Results Survey results are documented by Survey Area in "Survey Packages." Each£'iSS package may contain the data from the several Survey Units that are within a given SurveyArea. :'The data is reviewed, analyzed, and processed, and the results documented in. a "'ReleaSe Record."

The Release Record provides the necessary information to support the decision to release the survey units for unrestricted use. A Final Survey Report is prepared that provides the necessary data and analyses from the Survey Packages and Release Records;, and is submitted by the licensee to the USNRC.-"

3.2.5.1 Contamination Event Resurvey::

If a contamination event occurs in an area that has received FSS, a resurvey of the area, or the affected portion, is required to demonstrate compliance with the release criteria. The resurvey needs to address the affected area with a reasonable surrounding buffer area to ensure the entire affected area has been evaluated. Thie data collection needs to be of sufficient quality and quantity to demonstrate compliance.

The survey quantity requirements, minimum detection levels, and investigation levels appropriate to a Class 1 area as listed in Section 3.10 apply, regardless of the original classification of the area.

The number of direct measurements or volumetric samples (as appropriate) is determined as discussed in Section 3.7.2, and is applied as necessary.

The survey data collected following remediation is appended to, or replaces when appropriate, the original FSS data. The entire data set is re-evaluated as per Section 3.12 to determine compliance with the release criteria. The results of the resurvey are included in the FSS report for the survey area.

3.3

_implementation The VA anticipates that the USNRC and other regulatomy agencies will choose to conduct confirmatory measurements in accordance with applicable laws and regulations. These agencies may take confirmatory measurements to make a determination in accordance with 10 CFR 5

Alan J. Bloteky I*':*!*Reactor Facility 50.82(a)(1 1) that the final radiation survey and associated documentation demonstrate that the facility and site are suitable for release in accordance with the criteria for decommissioning established in 10 CFR Part 20, Subpart E. The VA will comply with the 25 mrem/yr criteria of 10 CFR Part 20, Subpart E by complying with the radiological criteria described in the decommissioning planning document prepared for the facility. The confirmatoiy, measurements collected by the agencies are based upon the same release criteria.

Timely and frequent communications by the VA with these agencies ensure sufficient opportunity to,perform confirmatory measurements prior to the VA implementing difficult to reverse decommissioning actions such as filling an open excavation or" new construction."

3.4 Radionuclides of Concern and Release Criteria The approved Decommissioning Plan (DP) provides a list of Radionuciides of Concern (ROC) for use during implementation of the FSSP. Table 3-1 lists the ROC along with the approved release criteria originating in NUREG-1757 Appendix B. The release criteria listed below serves as Derived Concentration Guideline Levels (DCGL) input-parameters for the FSS calculations.

Table 3-1: Radionuclides of Concern and Release Criteria Building Surfaces Volumetric Radionuclide Release Criteria*

Release Criteria H-3 1.2E+08 DPM/100cm 2 110 pCi/g C-14 3.7E+06 DPM/100cm 2 12 pCi/g Fe-55 4.5E+06 DPM/l00cm2 10,000 pCi/g Co-60 7.1E+03 DPM/100cm 2 3.8 pCi/g Ni-63 1.8E+06 DPM/100cm 2 2,100 pCi/g Cs-137 2.8E+04 DPM!100cm 2 11 pCi/g Eu-15.2 None 6.9 pCi/g Eu-154 None 8 pCi/g 3.5,Area Classification Prior to beginning FSS, a thorough characterization of the radiological status and history of the site is completed. Initial classifications have been determined based on characterization surveys.

The structures and open land areas are classified following the guidance in Section 4.4 of the MARSSIM. Area classification ensures that the number of measurements and the scan coverage is commensurate with the potential for residual contamination to exceed the approved release criteria. Additional data from operational surveys performed in support of decommissioning, routine surveillances or similar applicable surveys may be used to change the initial classification of an area up to the time of initiation of the FSS, as long as the classification reflects the levels of 6

• ii*Alan J. Blotcky

,*Reactor Facility residual radioactivity that existed prior to any remediation activities. When the FSS of a given survey unit begins, the basis for any reclassification is documented, requiring a redesign of the survey unit package and the initiation of a new survey using the redesigned survey package. If, during the conduct of an FSS survey, sufficient evidence is accumulated to warrant an investigation and reclassification of the survey unit, the survey may be terminated without completing the survey unit package.

Initial classifications of the AJBRF are provided in Table 13.1 of the Decommissioning Plan.

Current classifications are shown in Table 3-2 below. All areas are subject to :changes in classification based on encountered radiological conditions.

Table 3-2: AJBRF MARSSIM Classifications forFSS MARSSIM Survey Area Classification Reactor tank wall Portions not removed 1

Reactor tank pit Exposed concrete and/or soil 1

Reactor water cooling system vault Floors and walls 1

Rooms B526, B535, B535A, B537, B540, and FalooS 12mtr B540AWalls

>2 fiheters and ceilings 3

Roos 533 ad B33A Floors and walls <2 meters 2

RomsB33 ndB33AWalls>2 meters and ceilings 3

Rooms B522 and B522A

-Floors only 3

Hall outside Room B526

'Floors only 2

Stairs on south side of Room B526 Floors only 2

Outside areas All Non-impacted 3.5.1 Non-Impacted Areas Non-Impacted areas have no reasonable potential for residual contamination because there was no known impact from ste operations. Non-impacted areas will not be required to be surveyed beyofid what is completed as a part of site characterization to confirm the area's non-impacted classification.-

3.5.2 Impacted Areas hnpacted areas may contain residual radioactivity from licensed activities. Based on the levels of residual radioactivity present, impacted areas are further divided into Class 1, Class 2 or Class 3 designations. Class 1 areas have the greatest potential for residual activity while Class 3 areas have the least potential for impacted areas. Each classification will typically be bounded by areas classified one step lower to provide a buffer zone around the higher class. Exceptions occur 7

• iJ*':.....Alan J. Bloteky Reactor Facility when an area is surrounded by a significant physical barrier that would make transport of residual activity unlikely firom one area to the adjacent area. In such cases, each area will be classified solely on its own merit using the most reliable information available.

The class definitions provided below are fr'om Section 4.4 of the MARSSIM.

Class 1 "Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiological surveys). 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 high specific activity.

Note that areas containing contamination in excess of the DCGLw prior to remediation should be classified as Class I areas."

Class 2 "These areas have, or had prior to remedial/ion,.a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGLw. To justify changing an area's classification fr'om Class i to Class 2, the existing data (from the HSA, scoping surveys, or characterization surveys) should provide a high degree of confidence that no individual measurement would exceed the DCGLw." Other justifications for this change in an area's classification may be. appropriate based on the outcome of the DQO process.

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

Class 3 "Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fr'action of the DCGLw, based on site operating history and previous radiological surveys.

8

• f*J

• Alan J. Blotcky t*::!*Reactor Facility Examples of areas that might be classified as Class 3 include buffer zones around Class 1 or Class 2 areas, and areas with veiy low potential for residual contamination but insufficient information to justify a

non-impacted classification."

3.5.3 Changes in Classification Data from operational surveys performed in support of decommissioning, routine surveillance and any other applicable survey data may be used to change the initial classification of an~area up to the time of FSS commencement as long as the classification reflects the levels of r~esidual radioactivity that existed prior to remediation. Once the FSS of a given survey uniti beginf*," the basis for reclassification will be documented. If, during the conduct of an FSS,'survey sufficient evidence is accumulated to warrant an investigation and reclassification of the survey unit, the survey may be terminated without completing the survey unit package. A reclassification of a survey unit because of detected radioactivity may only result in,a more stringent classification, unless demolition or remediation activities result in the entire survey unit being eliminated, such as a building being demolished, rather than surveyed for release.

3.5.4 Redefining Survey Boundaries Survey units that are adjacent to one another may be combined prior to the survey design to simplify the performance and evaluation of the survey data, provided the new unit meets all the parameters applicable to the classification. If two or more areas of different classifications are combined, the new area will be classified at the most stringent classification applicable.

Discussion of the redefined survey boundaries will be included in the FSS documentation of the survey activities. Survey areas which have their classification increased as a result of boundary changes do not fall under th~e investigation requirements for reclassifying a survey area described in Sect~ion 3.10.10.

3.6 Establishing Survey Units Land areas, structures, and systems are made up of at least one smaller area defined as a survey unit.

Each land area, structure, or system may have multiple survey units of differing classification since data acquisition, analysis, and reporting are done on a survey unit basis. The survey unit release records applicable to a larger survey area may be combined into one report for submission to USNRC.

9

• J01*:Alan J. Bloteky

~Reactor Facility 3.61 Survey Unit Survey areas may be divided into smaller survey units. Survey units are areas that have similar characteristics and contamination levels. Survey units must be contiguous and will be assigned only one classification. Survey areas may include survey units of differing classifications since the site and facility are surveyed, evaluated, and released on a survey unit basis.

3.6.1.1 Survey Unit Size Section 4.6 of the MVARSSIM provides suggested sizes for survey units. However, as stated in the MARSSIM, the suggested survey unit sizes were based on a finding of reasonable Sample density and consistency.with commonly used dose modeling codes.

Table 3-2 lists the recommended survey unit size for each applicable classification.

Table 3-3: Survey Unit Size Minimum" / Maximumt Classification Buildings Open Land Class 1 10-rn2 / 100-mn2J

'100-m 2/ 2,000-in 2 Class 2 100-rn 2 / 1,000-rn 2:,

2,000-rn 2 / 10,000-rn 2 Class 3 1,000-m2 / No limit 10,000-rn 2!/ No limit For standing buildings, the MARSSIM recommennds 100-rn 2 of floor surface in a Class 1 area as a survey unit size based on the dose model assumption that a 100-rn 2 office space would be occupied. The source term attributed to the total area in this case is essentially the 100-rn 2 floor surface which equates to 180-rn 2 if the lower walls are included. If there is potential residual contamination on the upper Surfaces, i.e. the ceiling and walls above 2-rn, the area may be either broken into multiple survey units or the number of required data locations may be adjusted to account for the increased area.

A concerted effort should be made to define survey units of the sizes listed in Table 3-2; however there may be situations that would result in multiple survey units significantly smaller than those sizes.' In such a case, a single larger survey unit may remain provided the survey density is proportionally increased to account for the additional area. This is accomplished by dividing the

• Recommended minimum size SFrom MARSSIM Section 4.6 SFloor area of 100-rn 2 plus the lower walls (2-rn high) for a total area of 180-rn 2 10

Nii!!m i21*Alan J. Bloteky Reactor Facility maximum value for the classification listed in Table 3-2 by the number of required data points developed in the survey design package and proportionally increasing the sample density to account for the additional area.

Section 4.6 in the MARSSIM states

"...special considerations may be necessary for survey units with structure surface areas less than 10 m2 or land areas less than 100 in2.

In this case, the number of data points obtained from the statistical tests is unnecessarily large and not appropriate for smaller survey unit areas. Instead, some specified level of survey effort should be determined based on the DQO process and with the*

concurrence of the responsible regulatory agency. The data generated from thlese,

smaller survey units should be obtained based on judgment, ratherbthaan* on systematic or random design, and compared individually to the DCGLS."'

Consequently, the recommended minimum survey unit sizes in Table 3;2 are deri*'ed from this particular guidance.

If such a situation arises, the small survey area will be surveyed at a proportionate density developed in the same manner as additional samples in oversize areas were developed above.

3.6.1.2 Reference Coordinate System A reference coordinate grid system is used as an' aid' in the identification of survey locations within a survey unit. Software is used to develop the actual survey locations to the nearest foot.

Survey areas within buildings use a grid system based on the defined 'site north'. The grid patterns use a standard Cartesian grid system in the format of X,Y where X represents the east-west coordinates and Y represents tlh north-south coordinates. Each grid value represents one foot of linear distance in the applicable direction. Each survey unit has an origin indicated on the survey map as an aid in determihing orientation for conduct of the survey. Each survey unit may incorporate combinations of floors, walls, columns, ceilings, or special features within each unit as determined by th~e contamination potential and the need for survey. Local coordinates for each surface extend from the point of intersection of the Cartesian grid system applicable to that surface. Floors use the southwest corner as the point of intersection, while walls use the bottom left corner. Ceiling surfaces utilize a superimposition of the floor grid system to readily identify' survey locations.

For example, Grid Location X2,Y1 on the ceiling is directly above Grid Location X2,Y1 on the floor. Figure 3-1 shows an example of the splayed map of an interior space grid coordinate system for a typical room.

11

LI Alan J. Bloteky Reactor Facility Figure 3-1: Interior Space Coordinate Example B540 Ceiling N

3, 5 11,5 Wall 1 Wall 8 T,

Wall 2 Floor 6,[ 5 Wal-W-1 0, 51 11ll7 Wall 5 Wl~

4, 5 3.7 Survey Design Surveys to demonstrate compliance with release criteria will be performed using design parameters from MARS SIM.

3.7.1 Scan Coverage The area covered by scan measurements is based on the survey unit classification as described in Table 2 of the MARSSIM and is summarized in Table 3-4 below. A 100% accessible area scan of Class 1 survey units will be required. The emphasis is placed on scanning the higher risk areas 12

Alan J. Bloteky

• !}*

Reactor Facility of Class 2 survey units such as soils, floors and lower walls. Scanning in Class 3 survey units will focus on likely areas of contamination based on the judgment of the SO and will generally consist of a minimum of the 1-mn2 area around the selected direct measurement location.

Table 3-4: Minimum Scan Coverage Classification Required Minimum Scan Coverage Class 1 100% of accessible areas Class 2

Ž10% of accessible areas Class 3 Judgmental, but generally a 1-im2 area around each data location" 3.7.2 Sample Size Determination MARSSIM recommends that FSS be performed using a relative shift between 1 and 3. A relative shift of 1 provides for a more static survey locations and is therefore more conservative while a relative shift of 3 requires fewer static survey locations. Existing characterization data for the AJBRF support facilities (i.e. rooms and hallways outside the reactor cavity) consistently demonstrates minimal residual radioactivity, therefore the number of static survey locations within those survey units will be based on a relative shift of 3unless remediation support surveys indicate the presence of elevated concentrations greater than 50 percent of the release criteria.

3.7.2.1 Statistical Test Determination Appropriate tests will be used for the statistical evaluation of survey data. Tests such as the Sign test and Wilcoxon Rank Sum (WRS) test will be implemented using unity rules, surrogate methodologies, or combinations, of unity rules and surrogate methodologies, as described in MARSSIM and NUREG-1 505 chapters 11 and 12.

If the contaminant is not in the background or constitutes a small fraction of the DCGLw, the Sign test will be used. It is anticipated that the sign test will be the statistical test applied to the collected data because of the small fraction of the DCGLw that background radionuclides will contribute. If it is determined that background radionuclides contribute a significant fraction of the DCGLw, the Wilcoxon Rank Sum (WRS) test will be used.

3.7.2.2 Establish Decision Errors The probability of making decision errors is controlled by hypothesis testing. The survey results will be used to select between one condition of the environment (the null hypothesis) and an alternate condition (the alternative hypothesis).

These hypotheses, chosen from MARSSIM Scenario A, are defined as follows:

13

• '*Alan J. Blotcky

• t*

m*Reactor Facility Null Hypothesis (H0): The survey unit does not meet the release criteria.

Alternate Hypothesis (Ha): The survey unit does meet the release criteria.

A Type I decision error would result in the release of a survey unit containing residual radioactivity above the release criteria. It occurs when the null hypothesis is rejected, but in reality is true. The probability of making this error is designated as "a."

A Type II decision error would result in the failure to release a survey unit when the residual radioactivity is below the release criteria. This occurs when the Null Hypothesis is accepted when it is not true. The probability of making this error is designated as "3" Appendix E of NUREG-1727 recommends using a Type I error probability (a) of 0.05 and States that any value for the Type II error probability (13) is acceptable. Following the guidance, a* will be set at 0.05 A 13 of 0.05 will initially be selected based ona site-specific considerations. The Type I and Type II error values are used as lookup values in MARSSIM Table 5.2 to provide Zia and Zni-I values for use in the sample size formula. In this case, Type I and Type II error probabilities of 0.05 result in a value of 1.645 for Zl_* and Z*_p when referencing MARSSIM Table 5.2. The 13 may be modified, as necessarY, after weighing the resulting change in the number of required survey measurements against the risk of unnecessarily investigating and/or remediating survey units that are truly.below the release criteria.

Table 3-5: MARSSIM Table 5.2 Error Percentile 0.005 2.576 0.01 2.326 S

0.015 2.241 0.025 1.96 0.05 1.645 0.1 1.282 0.15 1.036 0.2 0.842

• .0.25 0.674

0.3 0.524 3.7.2.3 Relative Shift The relative shift (A/cs) is a calculated value. Delta (A) is equal to the DCGLW minus the Lower Boundaiy of the Gray Region (LBGR). The sigma used for the relative shift calculation may be recalculated based on the most current data obtained from post-remediation or post-demolition 14

i*

Alan J. Blotcky

• • !*!*Reactor Facility surveys; or from background reference areas, as appropriate.

The LBGR may be adjusted to obtain an optimal value for the relative shift, normally between 1.0 and 3.0. Administratively, the relative shift will have a maximum value of 3.0.

DCGL -

LBGR relative shift =

cr 3.7.2.4 Lower Boundary of the Gray Region The Lower Boundary of the Gray Region (LBGR) is the point at which the Type II (f3) error applies. The default value of the LBGR is initially set at the mean of the post-remediation'survey results, if available, or at 50 percent of the DCGLw, whichever is higher. If the relative shift is greater than 3.0, then the number of data points, N, listed for the relative shift, values of 3.0 from Table 5.5 or Table 5.3 in the MARSSIM, will normally be used, ats the~ minimum 'sample size.

Use of a relative shift greater than 3.0 requires approval by the SO.

3.7.2.5 Sigma Sigma values (the estimate of the standard deviation of the measured values inl a survey unit, and/or reference area) will be initially calculated from characterization samples/survey data with activity at or below the approved DCGLWv since thais value is what is expected to remain after necessary remediation has been completed. *These sigma values can be used in FSS design or more current post-remediation sigma values Can be used.

3.7.2.1 Sian Test Sample Size; The number of data points to be collected may be determined from Table 5.5 in the MARSSIM and includes the recommended 20% adjustment to ensure an adequate sample size.

An alternative method is to use Formula 5-2 in the MARSSIM.

Calculating the number of data points by using the formula will require an adjustment for a recommended 20% surplus of data pointS.

MARSSIM Formula 5-2: N = (-

+ z-)

4(SignP - 0.5)2 Where:

ZI*=desired Type I error from MARS SIM Table 5.2 ZI*=desired Type II error fr'om MARSSIM Table 5.2 SignP =value from MARS SIM Table 5.4 (see below) associated with the relative shift 15

• Lm*J*Alan J. Blotcky

• l*

Reactor Facility Table 3-6: MARSSIM Table 5.4 Relative Shift ISign p IRelative Shift ISign p 0.1 0.539828 1.2 0.88493 0.2 0.57926 1.3 0.903199 0.3 0.617911 1.4 0.919243 0.4 0.655422 1.5 0.933193 0.5 0.691462 1.6 0.945201 0.6 0.725747 1.7 0.955435 0.7 0.758036 1.8 0.96407 0.8 0.788145 1.9 0.971284 0.9 0.81594 2

0.97725 1

0.841345 2.5 0.99379 1.1 0.864334 3

0.99865 If relative shift > 3.0, use SignP = 1.0 When using the Sign test and assuming a relative shift of 3-as discussed in Section 3.7.2, the required sample size (rounded up) is 11. MARS SIM recommends an additional 20 percent which results in a total of 14 sample locations.

(1.645 + 1.645)2 N4(0.99865 - 0.5)2 10.824 4

• 0.249 10.824 N=-

0.996 N=II

~N = 11 plus 20% =14 3.7.2.2 Wilcoxon Rank Sum (WRS) Test Sample Size While it is anticipated that the Sign test will be used for the FSS data evaluation, the Wilcoxon Rank Sum (WRS) test is an alternate statistical test that may be used if background radionuclides contribute significant radioactivity. The number of data points to be obtained from each reference area/survey unit pair may be determined using the Table 5.3 in the MARSSIM.

The table includes the recommended 20% adjustment to ensure an adequate sample size. An alternative method is to use Formula 5-1 to directly calculate the required number of data points. Using the formula will require an adjustment for a recommended 20% surplus of data points.

16

Alan J, Blotcky Reactor Facility MARSSIM Formula 5-1 :

(Z1 _* + Z1 ~)2 3(P - 0.5)2 Where:

Z1_, = desired Type I error Z1_* = desired Type II error Pr = probability value from MARSSIM Table 5.1 (see below) associated with the relative shift Table 3-7: MARSSIM Table 5.1

... i,>*,

Relative Shift Pr Relative Shift [ Pr 0.1 0.528182 1.4 0.838864 0.2 0.556223 1.5 0.855.541 0.3 0.583985 1.6 0.871014 0.4 0.611335 1.7 0.885299 0.5 0.638143

.1.8 0.89842 0.6 0.66429

\\1.9,,

0.910413 0.7 0.689,665 2,*'..."20 0.921319 0.8 0.714167 2.3 0.944167 0.9 0.73771 2.5 0.961428 1.0 0.760217 2.8 0.974067 1.1 0.78 1627 3.0 0.983039 1.2.

0.801892 3.5 0.993329

.1.3 0.820978 4.0 0.997658 If relative shift > 4.0, use Pr = 1.0 When using the WIRS test and assuming a relative shift of 3 as discussed in Section 3.7.2 and including an additional 20 percent as recommended by MARSSJIM, the sample size required both in the survey unit and in the background reference area is 20.

(1.645 + 1.645)2 N3(0.983039 - 0.5)2 10.824 3

• 0.233 10.824 0.699 17

• ,*m:*

NAlan J. Blotcky W*IaI*NReactor Facility N =16 N =16 plus 20% =20 3.7.2.3 Elevated Measurement Comparison (EMC) Sample Size Adiustment If the scan MDC (MiDCso*) is greater than the DCGLW, the sample size will be calculated using the equation provided below. If Ne,,, exceeds the previously determined sample size, Neine will replace N.

Am emc Where:

Nemc is the elevated measurement comparison sample size,-

A is the survey unit area Aeimo is the area corresponding to the area factor calculated using the MDCSC8I concentration.

3.7.3 Background Reference Areas Background reference area measurements, are required when the WRS test is used, and background subtraction may be used with thie Sign test, under certain conditions such as those described in Chapter 12 of NUREG-1505. The reference area measurements are collected using the methods and procedures required for Class 3 FSS. For survey units that contain a variety of materials with markedly different backgrounds, a reference area that has similar materials will be selected. If one material is predominant or if there is not much variation in background among materials, a background from a reference area containing only a single material is appropriate when it is demonstrated that the selected reference area will not result in underestimating the residual radioactivity in the survey unit.

Background reference areas should have physical characteristics (including soil type and rock formation). simtilar to the site and shall not be contaminated by site activities. In general, VA commits to using background reference areas, when possible, that are offsite.

If non-contaminated onsite areas are to be used, then VA will verify and justify the use by appropriate comparison with samples fr'om appropriate off-site locations.

Should significant variations in background reference areas be encountered, appropriate evaluations will be performed to define the background concentration.

As noted in NUREG-1757, Appendix A, Section 3.4, the Kruskal-Wallis test can be conducted in such circumstances

Alan J. Blotcky Nj~in~i*Reactor Facility to determine that there are no significant differences in the mean background concentrations among potential reference areas.

VA will consider this and other statistical guidance in the evaluation of apparent significant variations in background reference areas.

3.7.4 Sample Grid and Measurement Location Sample location is a function of the number of measurements required, the survey unit classification, and the contaminant variability.

3.7.4.1 Sample Grid The reference grid is primarily used for reference purposes and is illustrated on sample maps.

Physical marking of the reference grid lines in the survey unit will only b~e performed When necessary.

For the sample grid in Class 1 and 2 survey units, a randomly selected sample start point will be identified and sample locations will generally be laid out ina triangular grid pattern at distance, L, fr'om the start point. A rectangular grid pattern may be used at the discretion of the survey plan designer. The sample and reference grids aire illustraited on sample maps and may be physically marked in the field. For Class 3 survey.uhits, sample locations may be randomly selected based on the reference grid or may use a systematic-grid.

3.7.4.2 Measurement Location Computer software will be used to determine measurement locations within a survey unit.

Measurement locations within the survey unit will be clearly identified and documented for purposes of reproducibility. Ana idenftification code will match a survey location to a particular survey unit.

Sample points for Class.1 and Class 2 survey units will be determnined using a systematic triangular grid pattern with a random start location. Sample locations in a Class 3 survey unit will be determined randomly.

Measurement locations selected using either a random selection process or a random-start systematic pattern that do not fall within the survey unit or that cannot be surveyed due to site conditions will be replaced with other measurement locations as determined by the SO.

3.7.5 Survey Package Design Process A Survey Package is produced for each survey area. The survey package is a collection of documentation detailing survey design, implementation and data evaluation for FSS of a survey area.

19

~Alan J. Blotcky Reactor Facility The VA intends to apply the 10 CER 50, App. B requirements for field and laboratory counting equipment, as well as the corrective action process to address data or programmatic discrepancies. Using the existing Part 50, Appendix B program precludes developing redundant measures for FSS activities.

3.7.5.1 Survey Package Initiation Each survey area and package is assigned a unique identification number. To allow continuity of area identification, the protocol used for identifying survey areas during the characterization survey will be used, as appropriate....

3.7.5.2 Review of Characterization Surveys The SO gathers and reviews historical data applicable to the survey area. Inf'ormation used for survey design is filed in the survey package. Sources of data include:

Characterization Surveys, Classification basis 10 CFR 50.75(g) files...

Operational Survey Records 3.7.5.3 Survey Area Walkdown The survey designer performs a walkdown to gather information about the physical characteristics of the survey area.

The, walkdown provides the designer an opportunity to determine if any physical or safety related interferences are present that may affect survey design or survey implementation, and to determine any support activities necessary to implement surveys.

The walkdown is documented and filed in the survey package.

Following the walkdown, representati~ce maps of the survey area are prepared.

• 3.7.5.4 Survey Desi~n Survey Design is the process of determining the number, type and location of survey measurements or samples required for each survey unit within a survey area. The various aspects of survey design are documented and filed in the survey package. The survey unit design process is controlled by approved procedures.

The size and number of suorvey units for a survey area are determined based on area classification, modeling assumptions used to develop DCGLW, and the layout of the survey area. The design will divide the area into discrete survey units as appropriate.

Each survey unit is numbered sequentially. The design provides a description of each survey unit including survey unit size, 20

• Alan J. Bloteky Reactor Facility classification and location. The types of material (i.e. soil, concrete, etc.) found in the survey unit and survey measurement and/or sampling methods are identified.

The design provides the number of measurements or samples required for each survey unit in accordance with MARSSIM. Count rate equivalent investigation levels for survey measurements based on the instrument detection efficiency may also be provided.

Table 3-8 provides measurement result investigation levels.

The design determines measurement/sample locations based on the classification of the survey unit and in accordance with the MARS SIM. A survey map is prepared of each survey unit. A sample and/or reference grid is superimposed on the map to provide an (X, Yjc'oodi~dnate system.

Each measurement/sample location is assigned a unique identification code which identifies the measurement/sample by Survey Unit and sequential number.

The design indicates the appropriate instruments and detectors, instrument operating modes and survey methods to be used to collect and analyze data.

Written survey instructions that incorporate the requirements set forth in the survey design may be provided. Direction includes selecting instruments,, cotPnt times, instrument modes, survey methods, required documentation, background requirements, and other appropriate instructions.

In conjunction with the survey instructions, survey, data forms, indicating desired measurements, are prepared to assist in survey documentation..

3.7.5.5 Survey Area Turnover Prior to performing FSS, the SO coordinates with the appropriate personnel to ensure decommissioning activities, area remediation and housekeeping are complete. The SO may direct surveys to be performed to verify that the area meets the radiological criteria for performance of the FSS. These post-remediation surveys, if performed, provide conclusive proof that an area is acceptable for FSS initiation. When satisfied the area is acceptable, the SO will direct the area to be posted to indicate that the area is controlled for the performance of FSS. Access controls will be implemented to prevent contamination of areas during and following FSS.

3.7.5.6 Survey Implementation Survey areas and/or locations are identified by gridding, markings, or flags as appropriate.

Instruments and equipment as indicated in the survey instructions are checked for proper operation and surveys are performed in accordance with the appropriate procedures. All survey results will be documented and a chain of custody for any collected samples will be maintained.

21

Reactor Facility On completion of the FSS, instruments will receive post-use source and background checks, and any collected samples will be prepared for analysis either on-site or at the selected off-site laboratory.

Survey instruments will be prepared in accordance with appropriate procedures and the survey instructions. Instruments are performance checked prior to and following surveys. Survey data will be reviewed and placed into the survey package. The data will be examined for any results that exceed investigation criteria so that appropriate investigation surveys and/or remediation may be performed in a timely fashion.

Several quality control measures and features have been developed for the implemnentation phase of the FSS program. These measures include:

Pre-implementation area walkdowns Survey location verification Instrument source checks prior to the start andupo completion of an FSS survey Conduct of scan surveys in the peak trap mode,,th~ereby providing a record of the maximum scan value for any scan grid

=.

3.7.5.7 Data Evaluation The survey data will be reviewed to verify completeness, legibility and compliance with the following requirements:

Convert data to standardize~d units (if necessary)

Calculate mean, median and range of the data set Review the data for outliers Calculate the standard deviation of the data set Verify MDC for each survey type performed meets requirements These steps may be completed by a software package or performed manually using tools of choice.

The SO reviews and verifies the statistical calculations, verifies the integrity and usefulness of the data set and determines the need for further data. Investigations of suspect data will be performed as necessary, Once satisfied that all data are valid, the appropriate statistical test will be performed and a decision will be made on the radiological status of each survey unit.

The data evaluation process is documented and filed in the survey package.

3.7.5.8 Quality Control Surveys Following completion of FSS, the need for QC surveys (replicate surveys, sample recounts, etc.)

22

o*?s*;:Alan J. Blotcky Reactor Facility will be determined. If necessary, a QC survey package will be developed and modeled after the original survey. QC measurement results are compared to the original measurement results. If QC results do not agree with the original survey, an investigation is performed.

Following investigation, the SO will decide data validity. Additional discussion of the performance of quality assurance and quality control surveys is provided in Section 5.4.1.

3.7.5.9 Release Record Following data evaluation, a Release Record will be prepared. The Release Record summarizes survey results and data evaluation. The Release Record is reviewed and approved by the alternate 3.8 Types of Surveys Survey measurements and sample collection are performed by personnel trained and qualified in:

accordance with the applicable procedure. The techniques for* performing survey measurements or collecting samples are specified in approved procedures. FSS measurements include surface scans, direct surface measurements, and gamma spectroscopy of volumetric materials. In-situ gamma spectroscopy or other methods not specifically described may also be used for FSS.

If required, a technical basis documnent will be created by the SO. Upon the documents acceptance, VA will give the USNRC a 30 day notice to review the technical basis document prior to implementation.

On-site and off-site lab facilities are used* for gamma spectroscopy, liquid scintillation and gas proportional counting in accordance *with applicable procedures. Regardless which facilities are used, analytical methods will use an administrative level of 50% the applicable DCGLw value for detection of radioactivity."

3.8.1.

Scan Surveys Scanning is performed in order to locate small areas of residual activity above the investigation level.

Structures receive scan surveys, direct measurements and, when necessary, volumetric sampling. The percent of scan measurement coverage is based on the survey unit classification and is provided in Table 3-4.

3.8.1.1 Two-Staae Scanning Method The two-stage scanning method is one where a surveyor begins the scan at a pre-determined speed, e.g. 10-cm per second, until they detect an elevated count rate. At such time, they return to a location immediately before the elevated detection and repeat the scan at a slower rate to 23

Alan J. Blotcky

i*

Reactor Facility determine the maximum count rate in the area. When the count rate has returned to expected levels, the scan speed is returned to normal. This method relies on the ability of the surveyor to reliably detect elevated count rates. The 'Surveyor Efficiency' (p) is detailed in NUREG-1507.

A variable accounting for this efficiency (a value from 0.5 to 0.75) is included in the formulae used in the MARSSIM. Lower values for p increase the MDCSCa,n indicating a smaller probability of detecting elevated count rates. The MDCscan equations used throughout this document use a value for p of 0.5.

Additional evaluations may be made by VA on the effectiveness of eliminating p from the MDCscan equation and utilizing the alarm functions of the selected instrumentation.

All scans performed in support of the FSS use the two-stage scan1 nethod to assure residual radioactive material is sufficiently quantified.

r 3.8.1.2 Beta-gamma Surface Scans Surface scans for beta-gamma activity on structures and selected systems will be performed at a scan rate capable of meeting a pre-determined MIDCsoan, applicab~le to the survey unit classification. Surface scans should have a probe to surface distance as close as practical, not to exceed 1-cm (-1/2/

inch). Situations where the maximum detector to surface distance cannot be met may require an alternate scan method, a detector to surface distance correction factor, or justification for not completing the scan. For Class 1 areas, the MDCSCan will be no greater than the DCGLemC. Class 2 and Class 3 areas require a more stringent MDCscan of no greater than the DCGLW unless scan coverage of 1 00% is specified in the survey design when a higher MDCscan equal to the DCGLe~nC may be used.

Minimum scan coverage is detailed in Table 3-4 by classification.

3.8.2 Direct Measurements Direct measurements are performed to detect surface activity levels. Direct measurements are conducted by placing the detector on or very near the surface to be measured and acquiring data over a pre-determined count time. A six (6) second count time will be used for direct surface measurements and provides detection levels well below the release criteria however the count time may be varied to ensure the required detection level is achieved. The MIDC for a static count may be calculated using the formula contained in Section 3.10.4. Direct measurements may be collected anywhere within the grid block, but are generally collected in the center.

3.8.3 Exposure Rate Gamma exposure rate measurements will be taken at each floor location approximately 1-rn from the surface.

24

gL*;

Alan J. Bloteky 1**

Reactor Facility 3.8.4 Removable Activity Removable contamination surveys, while not required as part of the MARSSIM, may be collected to assess the removable activity fraction for selected structural and system surfaces or as part of routine post-FSS verification surveys. When possible, a wipe sample will comprise 100-cm 2 of surface area. When a 100-cm2 area is not obtainable, the wiped area will be documented and the analyzed result adjusted accordingly.

Wipe samples will be analyzed on-site utilizing a wipe sample counter. Investigation levels will be 10% of the applicable building surface DCGLW. The origin of removable activity in excess of this investigation level should be determined and the area should receive additional remediation prior to FSS.

4.

3.8.5 LSC Analysis All direct measurement survey locations will have a removable activity wipe collected for tritium and C-14 analysis on-site using a liquid scintillation counter (LSC).

3.8.6 Volumetric Samples Volumetric sampling of media, as opposed to direct measurements may be necessary if the validity of direct measurements is questionable.*i Volumetric samples will be analyzed by appropriate analytical methods for the radionuclides of interest applicable to the subject area.

The results will be evaluated by one of the following:

Calculating the derived total gross beta or gross alpha DPM/1 00-cm 2 in the sample and comparing the gross results directly to the applicable DCGLw Using the radionuclide specific results to derive the surface activity equivalent and determine compliance using the unity rule.

Use of the unity' rule will require the use of a surrogate calculation to account for the radionuclides in the; mixture that are not identified by gamma spectroscopy.

This will be accomplished using the nuclide mixture established during the applicable characterization.

Sample preparation will be performed by the off-site contracting laboratory. Separate containers will be used for each sample and each sample will be tracked through the analysis process using a chain-of-custody record. Samples will be split as directed in Section 5.4.2.

3.9 Survey Methods Survey methods are applied differently depending on the data requirements of a survey area. For example, removable activity measurements provide little, if any, benefit when attempting to 25

Alan J. Bloteky Reactor Facility assess the radiological conditions in an excavation. Conversely, assessing a building surface via volumetric sampling would provide the necessary data, but at great costs of time and money.

This section will discuss the steps necessary to strike a reasonable balance between data needs and ease of survey performance based on the data needs of the survey area.

3.9.1 Buildin~s, Eq~uipment* and Components Buildings, equipment, and components that are destined to remain after license termination require the following surveys to demonstrate they meet the appropriate release criteria."

3.9.1.1 Scans Buildings, equipment, and components require two-stage scan measurements as part of the FSS process at coverage rates and speeds as directed in Table 3-4 and Section 3.8.1; *respectively. The two-stage scan method has been described in Section 3.8.1.1.

Gross beta and/or gross alpha measurements are utilized as appropriate to the potential contaminatiodn.

The measurements typically are performed at a distance of 1-cm or less from the surface. *Adjustments to scan speed and distance may be made in accordance with approved procedures.

3.9.1.2 Direct Measurements Direct measurements are required for, buildings, equipment, and components as part of the FSS process. The required quantity of direct measurements is a calculated value. The calculation is described in Section 3.7.2. Direct measurement data for buildings, equipment and components is collected with a gas proportiona! detector. As much as practical, the detector is of an appropriate size to maintain the surface to detector distance of no greater than the calibrated distance +0.5-cm.

3.10 Instrumentation Radiation detection and measurement instrumentation for the FSS are selected to provide both reliable operation and adequate detection sensitivity of the final list of the radionuclides of concern as identified during the characterization evolution.

Detector selection is based on detection sensitivity, operating characteristics and expected performance in the field.

The instrulmentation, to the extent practicable, is capable of data logging operations.

Commercially available portable and laboratory instruments and detectors are typically used to perform the three basic survey measurements:

1. Surface scanning 26

Alan J. Bloteky Reactor Facility

2.

Direct surface contamination measurements

3. Spectroscopy of soil and other bulk materials, such as concrete 3.10.1 Selection Radiation instruments and detectors are selected based on the type and quantity of radiation to be measured. The instruments used for direct measurements are capable of detecting the radiation of concern to a Minimum Detectable Concentration (MIDC) of less than 50% of the applicable DCGLw. The use of 50% of the DCGLw is an administrative limit only. Any value below the DCGLw is acceptable in impacted survey units. MDCs of less than 50% of the DCGLi* allow detection of residual activity in Class 3 survey units at an investigation level. of,0.5 times the DCGLw. Instruments used for scan measurements in Class 1 areas are required to ble capable of detecting radioactive material at less than or" equal to the DCGLemc.'

Specific instrument selection and performance capabilities (e.g. detection efficiency, MDCs, data logging, etc.) are discussed in Section 4.0 "Survey Instrumentation'. VA will generally follow the instrument manufacturers' recommendations and/or supporting basis documents for considerations such as temperature dependency and other operational parameters.

A description of the conditions under which themethod would be used.

A description of the measurement method, instrumentation, and criteria.

Justification that the technique WVould p5rovide equivalent scan coverage for the given survey unit classification and that the M\\DCscan is adequate when compared to the DCGLemc.

A demonstration that the methlod provides data that has a Type 1 error (falsely concluding that the survey unit, is acceptable) equivalent to 5% or less and provides sufficient confidence that the DCGLeinc criterion is satisfied.

3.10.2 Calibration and Maintenance Instruments and detectors are calibrated for the radiation types and energies of interest at the site.

The calibration sources for beta survey instruments are Tc99or Sr90. The alpha calibration sources are Pu239 or Th23° which have an appropriate alpha energy for plant-specific alpha emitting nuclides. Gamma scintillation detectors are calibrated using Cs 137. Calibration sources other than those listed may be used provided they demonstrate appropriately conservative detection efficiency for the radionuclides of interest. In all cases, the surface efficiency as determined appropriate for the weighted mean energy of the radionuclides of concern will be utilized regardless of the energy of the calibration source.

Instrumentation used for FSS is calibrated at least annually and maintained in accordance with the 27

• • =*

=i~i*AlIan J. Blote ky r~s*

Reactor Facility Instrumentation Program procedure. Radioactive sources used for calibration are traceable to the National Institute of Standards and Technology (NIST) and use, to the extent practical, geometries designed to mimic the type of samples being counted. If vendor services are used, these are obtained in accordance with purchasing requirements for quality related services to ensure an appropriate level of quality.

3.10.3 Response Checks Instrument response checks are conducted to assure proper instrument response and operation.

An acceptable response for ratemeter field instrumentation is an instrument reading +I10%

of the check source value established during or immediately after calibration.

Scaler and benchtop counter instrumentation standards are +/-3 sigma as documented on a control chart. Response checks are performed daily before instrument use and again at the enad of-use. 'Check sources are appropriate for the type of radiation as that being measured 'in the field and are, to the extent practical, held in fixed geometly jigs for reproducibility. If anidnstrument fails a response check, it will be clearly labeled "~Do Not Use" and will be removed fr'om service until the problem is corrected in accordance with applicable procedures.'

Measurements made between the last acceptable check and the failed check will be evaluated to determine if they may remain in the data set.

3.10.4 Minimum Detectable Concentration The MDC for static and scan measurements is determined for the instruments and techniques used for FSS.

The static measurement MDC is the concentration of radioactivity that an instrument is expected to reliably (i.e. 95 percent certainty) detect when used in an integrated count mode, i.e. number of counts per unit time, while MVDC~an is the concentration of radioactivity that an instrument is expected to reliably detect in ratemeter mode. Detailed discussion regarding calculation of instrument MDC and MDCscn is included in Section 4.2.2 3,10.5 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 which warrants investigation. Elevated measurements may result from discrete particles, a distributed source, or a change in background activity. In any case, investigative actions should be implemented. Depending on the investigation results, the survey unit may require:

No action 28

'*:Alan J. Bloteky

• ii;i,

• Reactor Facility Remediation Reclassification and resurvey 3.10.6 Investigation Levels Table 5.8 in the MARSSIM provides guidance on investigation levels for scan surveys.

In addition to investigation levels for scan surveys, direct measurement survey investigation levels may be used.

These additional investigation levels include a conservative value for Class 3 survey units and are provided in Error! Reference source not found....

Table 3-8: Measurement Result Investigation Levels Classification Investigation Level Class 1 Result >DCGLemc0 Class 2 Result >DCGLw Class 3 Result >50% DCGLw :

3.10.7 Investigation Process Technicians respond to all audibly detected elevated count rates while surveying.

Upon determining an elevated count rate, the technician stops and resurveys the suspect area to verify the count rate elevation and determine the areal extents of the elevated count rate. Technicians are cautioned, in training, about the. iportance of the elevated count rate and the verification survey. They are given specific. direction regarding the extent and scan speed of the verification survey. If the elevated count rate is verified, the technician marks the area. Each marked area will receive an additional documented survey which requires a re-scan of the area and one or more direct measurements.

Field gamma spectroscopy measurement and collection of soil samples may also be, dictated. Results of each investigation are discussed and reported in the survey: unit Release,Record.

The size and average activity level in the elevated area will be defined to determine compliance with the area factors. If any location in a Class 2 area exceeds the DCGLw, scanning coverage is increased in order to detennine the extent and level of the elevated reading(s). If any location in a Class 2 area exceeds the DCGLemC, the scan coverage is increased and the area may be reclassified, if necessaly*. If the elevated reading occurs in a Class 3 area, the scanning coverage is increased and the area may be reclassified, if necessary.

Investigations should consider:

29

im*

Alan J. Blotcky Reactor Facility The assumptions made in the survey unit classification The most likely or known cause of the contamination The possibility that other areas within the survey unit may have elevated areas of activity that may have gone undetected Depending on the results of the investigation, a portion of the survey unit may be reclassified if there is sufficient justification. The results of the investigation process are documented in the survey area Release Record.

See Section 3.5.3 for additional discussion regarding potential reclassification of the survey unit.

3.10.8 Elevated Measurement Comparison (EMC)

...i 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 the MARSSIM, the EMC is intended to flag potential failur~es in the remediation process and should not be considered the primary means to identify wheth~er 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 EMC. The size of the area containing the elevated residual radioactivity and the average level of residual activity within the survey unit are determined. The initial DCGLeimo is established during the survey design and is calculated as follows:

DCGLeinc determination:

DCG c = A-* DCGJ Where:

AF = Area Factor corresponding to the size of the elevated area DCGLw =Derived Concentration Guideline Limit The area factor is a multiple of the DCGLW that is permitted for 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 DCGLw and is determined by the investigation process. Area factors are determined during the DCGLW development phase of the project.

30

• m....Alan J. Bloteky Reactor Facility 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 DCGLemnc. If the DCGLemo is exceeded, the area is remediated and resurveyed.

The results of the elevated area investigations in a given survey unit that are below the DCGLeImC 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 formula result is less than unity, no further elevated area testing is required and the EMC test is satisfied.,

Elevated area evaluation:

8 a~vg DCGLW A)(CL.

Where:

6 = average residual activity in the survey unit....

Cavg = average concentration of the elevated area AF = Area Factor corresponding to the size of the elevated area When calculating 6 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.

Compliance with the soil DCGLemoc is determined using gamma spectroscopy results and a unity rule approach. These general methods are also applied to other materials where sample gamma spectroscopy is used for FSS. *The application of the unity rule to the EMC requires that area factors and a corresponding DCGLernc be calculated separately for any gamma emitters identified during FSS.

3.10.9 Remediation and Reclassification Areas of elevated residual activity above the DCGLermc within any classification are remediated to reduce the residual radioactivity to acceptable levels.

Whenever an investigation confirms activity above an action level applicable to the classification, an evaluation of the operational history, design information, and sample results is performed to assure the area was classified properly. The evaluation considers:

The elevated area location, dimensions, and sample results.

An explanation of the potential cause and extent of the elevated area in the survey unit.

The recommended extent of reclassification, if considered appropriate.

Any other required actions.

31

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

If an individual scan or static location measurement within a survey unit exceeds the applicable investigation level listed in Table 3-8Error! Reference source not found., the survey unit or a portion of it may be reclassified and the survey redesigned and re-performed> accordingly.

Instrument performance, background fluctuation, surveyor performance, ambient radiological conditions, and other variables should be considered to avoid unnecessary reclassification.

3.10.10 Reclassification and Resurvey:

Following an investigation, if a survey unit is reclassified or* if remediation activities occur, a resurvey will be performed. If the average value of Class 2 direct survey measurements was less than the DCGLw, the MDCscan was sensitive enough to.detect the DCGLemc and there were no areas greater than the DCGLemnc, the survey redesign may be limited to obtaining a 100% scan without having to re-perform the direct measurements, This condition assumes that the sample density meets the requirements for a Class 1 area. If the Class 2 area had contamination greater than the DCGLw, but the MDCSCan was not sensitive enough to detect the DCGLemc, the affected area is reclassified as Class 1 and resurveyed with the sample density determined for the new classification.

Class 3 areas are treated in a similar manner, using 50% DCGLw as the investigation limit. If a C2lass 3 area had activity in excess of 50% DCGLw, but less than the DCGLw and the MDCscan was sensitive enough to detect the DCGLeinc, then the expansion of scan survey coverage to 100% will be sufficient. If activity is detected above the DCGLw, or the MDCscan was not sensitive enough, the area is increased to the appropriate classification as determined by the activity detected and the survey redesigned and performed as directed by the SO. *Reclassification of a survey area requires notification be made to the SO. A more detailed investigation of the reason for the improper classification will be performed at the discretion of the SO.

3.11 Data Collection and Processin2 3.11.1 Sample Handlini* and Record Keeping A chain-of-custody (COC) record will accompany each sample from the collection point through obtaining the final results to ensure the validity of the sample data. COC records are controlled 32

GAlan J. Blotcky

• i

!iReactor Facility and maintained in accordance with applicable procedures.

Each survey unit has an associated document package which covers the design and field implementation of the survey requirements. Survey unit records are considered quality records.

3.11.2 Data Management Survey data are collected fr'om several sources during the data life cycle and evaluated.

QC replicate measurements are not used as final status survey data.

See the MAR-QAP for design and evaluation of QC replicate measurements.

Measurements performed during turnover and investigation surveys may be used as FSS data if they were performed according to the same requirements as the FSS data.,The requirements include:

s The survey data reflects the as-left survey unit condition and is untouched by further remediation.

Isolation measures are put into effect for the survey unit to prevent re-contamination and to maintain final configuration.

The data collection and design were in accordance with FSS methods, e.g., scan coveageMDC requirements, inyestigation levels, survey data point quantity and location, statistical tests, and EMC tests.

• Measurement results intended as FSS data constitute the final survey of record and are included with 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 DCGLw.

The recording units for surface,contamination are DPM/100-crn 2 and pCi/g for activity concentrations. Numerical values, even negative numbers, should be 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,.rneasurement results, the date and time of the measurement and any surveyor comments.

3.11.3 Data Verification and Validation The FSS data will be reviewed prior to 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:

The instrumentation MDC for fixed or volumetric measurements was below the DCGLw or if not, it was below the DCGLeinc for Class 1, below the DCGLw for Class 2 and below 50% DCGLw for Class 3 survey units.

33

EI*

Alan J. Blotcky Reactor Facility The instrument calibration was current and traceable to INIST standards.

The field instruments were source checked with satisfactoiy results before and after use each day data were collected or data were evaluated by the SO if instruments did not pass a source check in accordance with Section 3.10.

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

o The survey methods used were proper for the types of radiation involved and for the media being surveyed.

• The COG 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 reflect the radiological status of the~area. ",-

The data have 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, reviewed, and approved by the SO.

3.11.4 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. This is an optional task which is intended to aid the SO in situations where the decision.of whether an *area or unit meets the applicable criteria is not readily apparent. Visual representations may include any combination of posting plots, frequency plots, histograms, contour maps, or 3-D surface plots.

3.12 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 DCGLW or comparing the mean value to the DCGLw 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.

3.12.1 Data Evaluation 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.

An assessment of the measurement results is used to quickly determine whether the survey unit passes or fails the release criterion or whether statistical analyses must be performed.

34

• t?

• Alan J. Blotcky Reactor Facility If all concentrations within the survey unit are less than the DCGLW, the unit meets the criterion and no statistical tests are necessary. If the average concentration is greater than the DCGLw, the survey unit does not meet the release criterion and additional remediation may be necessary. If the average concentration is less than the DCGLw, but one or more individual measurements exceed the DCGLw, the sign test or WRS test, and elevated measurement comparison tests should be conducted to determine the disposition of the survey unit.

When required, one of four statistical tests is performed on the survey data:

1. WRS Test-,

2.

Sign Test i..."'

3. WRS Test Unity Rule

4.

Sign Test Unity Rule It is not anticipated that the WRS tests will be conducted,. but it will be discussed in this document for topic completeness.

In addition, survey data are evaluated against the EM~c criteria as previously described in Section 3.10.8.

The statistical test is based on the null h*ypothesis (Ho) that the residual radioactivity in the survey unit exceeds the DCGLw. There must be sufficient survey data at or below the DCGLw to reject the null hypothesis and conclude the survey unit meets the site release criterion for dose. Statistical analyses are performed using a specially designed software package or, if necessary, hand calculations.

3.12.1.1 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 acceptably low fractions compared to the DCGLw. Instrument background is subtracted from the gross survey result to determine a net result appropriate for each survey location. Should any of the radionuclides of concern be present in background, the measurement net result (i.e. gross result minus instrument background) is assumed to be entirely from plant activities. This option is used when it can be reasonably expected that including background activity 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 used with background subtraction in accordance with Chapter 12 of NUREG-1 505.

35

U, Alan J. Blotcky N;'*f*,*Reactor Facility The sign test is conducted as follows:

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

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

N.

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

• The numbeir 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.

• The value of S+ is compared to the critical value given in Table 1.3 of thle MARSSIM.

The table contains critical values for given values of N and ox. The value of cc is set at 0.05 during survey design. If S+ is greater than the critical value given inl 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.

3.12.1.2 Wilcoxon Rank Sum Test The WRS test, or WRS unity rule, described in detail in NUREG-1505 Chapter 11, may be used when the radionuclide of concern is present in the backgroun4, or if measurements are used that are not radionuclide-specific. In addition, this test is Valid only when measurement results less than MDC values do not exceed 40 percent of the data set.

The WRS test is applied as follows:

The background reference area measurements (Xi) are adjusted by adding the DCGLW to each background reference area measurement to obtain the value for Zi (Zi = X1 +

DCGLw).

• The number of adjusted background reference area measurements (in) and the number of survey unit measurements (n) are added to obtain N (N = m + n).

• 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.

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

The value Of Wr is compared with the critical value in Table 1.4 of the MARSSIM. If Wr 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.

3.12.1.3 Unity Rule The radionuclides of concern 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 36

• ,*Alan J. Bloteky

• ,;**:**Reactor Facility values, also called weighted sum, will be calculated as shown in the following equation.

Unity calculation:

UNITY-C

+

C

+

C.+

DCGI1 DCGL2 DCGI*

Where:

Cx= radionuclide concentration The unity calculation results are used to demonstrate compliance by defining the DCGL as 1.0 and using the decision criteria listed in Section 3.12.1. If the application of the WRS~or sign test is necessary, these tests will be applied using the unity calculation results and ~defining the DCGLw as 1.0. An example of a WRS test using the unity rule is provided in the MARSSIM Appendix I, Section 11.4.

If the WRS test is used, or background' subtraction is used in conjunction with the Sign test, background concentrations will also be converted to their unity equivalents prior to performing the test. The sign test is used.without background subtraction if background values are not considered a significant fraction of the DCGLw.

The unity rule method described above will be applied* as necessary when multiple radionuclides or emission types are being evaluated as opposed. to a single radionuclide, a surrogate nuclide that incorporates a nuclide fraction, gross alpha measuirements, or gross beta-gamma measurements.

3.13 Statistical Conclusions The results of the statistical tests, including application of the EMC, allow one of two conclusions to be made. The first conclusion-is that the survey unit meets the site release dose criterion. The data provide 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 possible conclusion is that the survey unit fails to meet the release criterion. The data are not conclusive in showing that the residual radioactivity is less than the release criterion. In this case, the data are analyzed further to determine the reason for the failure.

Possible reasons are:

The average residual radioactivity exceeds the DCGLw.

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 collected measurements and the 37

E~i1:,Alan J. Blotcky Reactor Facility standard deviation of the measurement data. The power is determined from 1-f3 where fI is the value for Type II errors. A retrospective power analysis may be performed using the methods described in Appendices 1.9 and 1.10 of the MARS SIM. If the power of the test is insufficient due to the number of measurements, additional samples may be collected and the data re-evaluated. Additional 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.

3.13.1 Compliance The FSS is designed to demonstrate that licensed radioactive materials have been removed. from the VA facilities and property to the extent that residual levels of radioactive contamination are below the radiological criteria defined by the selected and approved dose-based scenario.

If the measurement results pass the requirements of Section 3.13, and the elevated areas evaluated per Section 3.10.8 pass the elevated measurement comparison, then the survey unit is determined to meet the criteria for license termination.

3.14 Reporting Format' Survey results are documented in history files, survey unit release records, and in the final status survey report. Other reports may be generated as requested by the USNRC.

3.14.1 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 When a survey unit release record is given final approval it becomes a quality record.

38

AlanJ. Bloteky

• (*,!*Reactor Facility 3.14.2 Final Status Survey Report Survey results will be described in written reports to the USNRC. The subject areas included in each written report will vary depending on the status of ongoing decommissioning activities.

Other reports may be written and submitted to USNRC as necessary, or" upon request.

The FSS report provides a summary of the survey results and the overall conclusions which demonstrate that the VA facility and site meet the radiological criteria for release. 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 basic outline of the final reports will be:."

1.0 Overview of the Results 2.0 Discussion of Changes to FSS 3.0 Final Status Survey Methodology Survey unit sample size Justification for sample size 4.0 Final Status Survey Results

• Number of measurements taken.

Survey maps Sample concentrations Statistical evaluations, including power curves Judgmental and miscellaneous data sets 5.0 Anomalous Data 6.0 Discussion of the Cross-Contamination Prevention and Monitoring Plan Inplementation 7.0 Conclusion for each survey unit 39

• '::'Alan J, Bloteky

• i*

Reactor Facility 4.0 SURVEY JINSTRUMENTATION The FSSP presents an outline of the MARSSIM survey process for the purpose of terminating a United States Nuclear Regulatory Commission (USNRC) radioactive materials license. This section discusses common radiological survcey instrument selection, their capabilities, general operation, and calibration.

The purpose of this section is to:

• Present a selection of instruments that may be used in support of the MIARSSIM activities at the AJBRIF-Provide methods for evaluation of the suitability of various instruments for their intended roles Discuss statistical formulae demonstrating detection efficiencies, minimum detectable concentrations, and expected surface efficiencies

• . Establish general calibration requirements Set Quality Assurance/Quality Control (QA/QC) requirements for the instrument program 4.1 Overview Measurement data for surface measurements are recorded as counts per unit time per detector area. These are normally converted into standardized units for evaluation purposes which eliminate instrument specific parameters. Common standardized units within the United States are disintegrations per minute per 100-cm 2 (DPM/100cm2) which will be used in this document.

Volumetric analyses are typically, reported in units of radioactivity per unit mass. PicoCuries per gram is the most common standardized unit for volumetric measurements for decommissioning activities and is used in this document.

Radiological survey instruments are made up of a meter and a detector. Collectively, this combination is referred to as an instrument. Pairings of a meter and a detector are controlled by calibration procedures.

The list of instrumentation currently being considered for use in support of the FSSP is provided in Table 4-I.

40

Alan J. Bloteky Reactor Facility Table 4-1: FSSP Instrumentation Instrument! I Detector Measurement Detector Detector Manufacturer Output Type Type Area and Model jUnits Surface Scintillation 100cm Ludlum model CPM alpha/beta-2360 with 43-gamma 9

Surface beta-G-M Tube 20-cm2 Ludlum model CPM gamma 2

3 (or 12) with mgcm 2 44-9 Gamma NaI l"x 1" Ludlum model pR/hr' exposure rate 19 Gamma scan NaI 2"x2" Ludlum model

CPM, 2221 with 44-

____10 Removable Scintillation 20-cmz Ludlum model DPM alpha/beta-0.8

,,3030E with 43-gamma mg/cm2 1.0-1 4.2 Instrument Statistics Radiological instrumentation will be selected to" provide both reliable operation and adequate detection sensitivity of the final list of the radionuclides of concern as identified for the Derived Concentration Guideline Limits (DCGLw§). Detector selection is based on detection sensitivity, operating characteristics and expected performance in the field. The instrumentation will, to the extent practical, be capable of data logging operations and have digital displays. Data logging provides the ability to electronically record the collected data, thereby reducing potential transcription errors. Digital displays reduce the observer interpretation of the measurement allowing for more accurate readings.

Commercially available portable and laboratory instruments and detectors are typically used to perform the basic survey measurements:

• The "w" in DCGLw stands for Wilcoxon Rank Sum test, which is the statistical test recommended in MARSSIM for demonstrating compliance when the contaminant is present in background. The Sign test recommended for demonstrating compliance when the contaminant is not present in background also uses the DCGLw.

41

• • '*Alan J. Bloteky Reactor Facility Surface scanning Direct surface contamination measurements Removable activity analysis Spectroscopy of soil and other bulk materials, such as concrete Instrument use and survey procedures detail the issuance, use, and calibration of instrumentation.

Records supporting the Instrumentation Program are maintained by the PM or designee in accordance with document control procedures.

Radiation instruments and the associated detectors are selected based on the type and quantity of radiation to be measured. The instruments used for direct measurements are generally capable of detecting the radiation of concern to a Minimum Detectable Concentration. (MDC) of less than 50% of the applicable DCGLwV. The use of 50% of the DCGL is an administrative limit only. Any value below the DCGLW is acceptable in Class t or 2 survey units. MDCs" of less than 50% of the DCGLw allow detection of residual activity in Class 3 survey units at an investigation level of 0.5 times the DCGLw. Instruments used for scan measurements are required to be capable of detecting radioactive material at less than or equal to the DCGLemnc.

A list of some common instrument brand and models is presented in Table 4-2 along with their nominal operational parameters e.g., background rate, probe area, total efficiency. A final determination of the specific brands and models will precede implementation of the FSS program.

VA will generally follow the instrument manufacturers' recommendations and/or supporting basis documents for considerations such as temperature dependency and other operational parameters.

Table 4-2: Typical Instrument Operational Parameters

Background

Probe Area Detection Instrument Model Range (em 2)

Efficiency %)

Scaler/Ratemeter Zinc 2 CPM 100 13.0 (4-t)

Sulfide. Scintillator (alpha)

Scaler/Ratemeter Geiger 50 CPM 17.5 19.0 (4-t)

Mueller Detector (beta/gamma)

Scaler/Ratemeter Gas 80 CPM 100 30.0 (4-7t)

Proportional Detector (beta/gamma)

MicroR Gamma 10-15 micro R/hr N/A N/A Scintillator 42

• im

.Alan J. Bloteky

!i*

Reactor Facility I Background Probe Area { Detection Instrument Model j

Range (cm 2)

{ Efficiency (%)

Removable Activity T 1 DPM/sample N/A 39.0 Counter (alpha)

I________________

Removable Activity

{

220 N/A

~

38.8 Counter (beta)

DPM/samnpleI As the project proceeds, other measurement instruments or technologies, such as in-situ gamma spectroscopy or continuous data collection scan devices, may be found to be more efficient than the survey instruments currently under consideration. The acceptability of sudh an inistrument or technology for use in the final survey program would be justified in a technical basis document.

The technical basis document would include, among other things, the following:

A description of the conditions under which the method or equipment would be used A description of the measurement method, instrumentation, and criteria Justification that the technique would provide equivalent scan coverage for the given survey unit classification and that the MDCscan' is adequate when compared to the DCGLeinc A demonstration that the method provides data that has a Type 1 error (falsely concluding that the survey unit is acceptable) equivalent to 5% or less and provides sufficient confidence that the DCGLe~nC criterion is satisfied 4.2.1 Total Detection Efficiency The total instrument efficiency (st) is* a calculation of the percentage of activity present in or on a surface that an instrument *detects. The *t is a product of two components, the instrument efficiency (si) and the surface efficiency (s*) as shown in the formula:

Equation 4.1 - Total Detection Efficiency

=t**

4.2.1.1 Instrument Efficiency The instrument efficiency (si) is a measurement of the surface emissions that interact with the detector elements with an instrument. To determine this, the instrument detector is exposed to a source of a known emission rate for a specified time period and the number of interactions is recorded. The efficiency percent is then calculated by:

43

• q!*m*Alan J. Blotcky

~Reactor Facility Eqiuation 4.2 - Instrument Efficiency (C5 -C)

Where:

Cs= Measured interaction count per one minute Cb = Measured background interaction count per one minute S -- known source value in DPM The known source value should be that of the hemispherical area (2-7) exposed to the detector as opposed to the total emission of the sphere around the source (4-vt). The surface efficiency (*s),

discussed below, accounts for the remaining half of the emission sphere.

4.2.1.2 Surface Efficiency Surface efficiency (*s) is an estimation of the affect the media surface has on the interaction of residual radioactive material with a detector and is a function of the surface condition, i.e.

smoothness, and the relative emission energy of the i'adionuclide. The *s for potentially contaminated structures and systems will follow recommendations contained in NUREG-1507 and ISO 7503-1 for the energies applicable to the radionuclides of concern developed in the characterization study. Beta-gamma detection instruments will use an *s of 0.5 and alpha detection instruments will use an

• of 0.25.

The methods for determining efficiency in NUREG-1507 were specifically developed to address situations when the source, in this case concrete, affects radiation emission rate due to self-attenuationa, backscatter, thin coverings, etc.

Media-specific* r. may he developed as necessary. These new surface efficiencies will subsequently override NUREG-1507 recommendations upon acceptance.

The condition of the surface being measured has an effect on the s* as well. For direct surface and scan measurements, the surface area beneath the detector should not have variability of depth greater than 0.5-cm more than the source to detector distance used for the instrument calibration.

According to NUREG-1507, instrument efficiency drops considerable when the source to detector distance increases more than 0.5-cm.

4.2.2 Minimum Detectable Concentration Minimum Detectable Concentration (MDC) and Minimum Detectable Activity (MDA) are 44

Alan J. Bloteky

• .'**:i*Reactor Facility estimations of the lowest level of concentration or activity the subject instrument can detect 95%

of the time. For purposes of this document MDC and MDA will be used interchangeably. Factors that directly affect the MDC are the total instrument efficiency (discussed in Section 4.2.1),

background rate, and count duration. The MDC is calculated individually for direct (static) measurements, ratemeter count rates, and scans. Static measurement count times will be adjusted to achieve the required MDC values.

4.2.2.1 Minimum Detectable Concentration for Direct Measurements Direct, or static, measurements are measurements where the detector is placed finto a fixed position and the instrument records individual unique counts for a specified, *period of* time.

Measurement data for direct surface measurements are recorded as counts per unit time per detector area prior to conversion to standardized units.

For" static (direct) surface measurements, with conventional detectors, the MDC will be calculated using Formula 3-10 in NUREG-1507:

Equation 4.3 - Minimum Detectable Concentration Direct measurement MDC:

MDC 3

.2 -

b(T)(1.STb Where:

MDC =Minimum detectable concentration (DPM/100-cm2)

Rb= Background count rate (CPM)

Tb

- Background count time (minute)

Td = Sample run time (minute) l's= Sample count time (minute)

,*=Counting system efficiency (decimal)

Direct measurements require an MDC less than the DCGLw, but an administrative limit of<*50%

DCGLw will be used to assure adequate sensitivity for the investigation levels applicable to lower unit classifications (Class 2 and Class 3).

4.2.2.2 Scan Measurement MDC Scan measurements are measurements taken with the detector in a steady motion over an area larger than the surface area of the detector and at a predetermined surface to detector distance.

Scan rates and surface to detector distances are specified for the type of emission (alpha, beta, 45

• • I*I*Alan J. Blotcky Reactor Facility etc.) in the Survey Design section of the MSP.

Scan measurement MDC (MDCscan).calculations depend on the emissions of the radionuclides of concern. Beta-gamma MDCscan is discussed in Section 6.7.2.1 of the MARSSIM. The desired MDCscaii for an area is a function of its classification. Because of the lower activity anticipated in lower class areas (Classes 2 and 3), a more stringent MDCsca, is recommended. These recommendations are presented in Table 2-3 of the MARSSIM Survey Plan.

Alpha MDCscan is calculated as a probability of detection as outlined in Section 6.7.2.2 of the MARS SIM. A more detailed discussion appears in Appendix J of that document.

4.2.2.2.1 Beta-Gamma Scan MDC The MDCscan for beta-gamma measurements may be calculated by first deterniining the Minimum Detectable Count Rate (MDCR). The MDCR is calculated by first defining the minimum detectable net source counts (si) using Formula 6-8 from the MARSSIM as below.

Equation 4.4 - Minimum Detectable Source Counts Si = d' b-*

Where:

d'= value taken from Table 6.5 in the MARSSIM for applicable true and false positive rates bi= Number of background counts in a given time interval The MDCR is then calculated fr'om Formula 6-9 in the MARSSIM:

Equation 4.5 - Minimum Detectable Count Rate 60 MDCR = S* *-

Where:.

i = Observed time interval in seconds Finally, applying the detection efficiency correction results in an MDCscan in standardized units (DPM/100-cm 2) fr'om Formula 6-9 in NUREG-1507:

46

Alan J. Blotcky Reactor Facility Equation 4.6 - Scan MDC MDCR MlC...

=

probearea

/P**i~s*

100cm 2 Where:

p3 = Surveyor efficiency (value from a range between 0.5 and 0.75)

• i= Instrument efficiency as= Surface efficiency The value for p has been developed in Draft NU7REG/CR-6364 and NUREG-1507 and, is a percentage estimate of the likelihood a surveyor will reliably detect an elevated count rate.

Currently, the equations presented here use a value for p of 0.5.

Table 4-3: Required Minimum Detectable Concentrations 1:Required Static fRequired Instrument Model Release Criteria J MDC*

MDCscan Scaler/Ratemeter Zinc 1,000 Sulfide Scintillator (alpha)

DPM/100cm2 500 DPM/100cm2 N/A Scaler/Ratemeter Zinc 7,100 3,550 7,100 Sulfide Scintillator DP/0c2 DM10m2DM/0c2 (beta/gamma)DM/00mDM10c 2

DM10m MicroR Gamma Scintillator N/A N/A N/A Removable Activity*

N/A 100 DPM/1 00cm2 N/A Counter (alpha)

Removable Activity 710 DPM/100cm2 355 DPM/100cm2 N/A Counter (beta) 4.3 Instrumentation Overview An instrument is a combination meter and detector. The meter houses the power supply and electronics which record the count data. Meters may have specialized abilities, such as data logging, multiple detectors, digital display, multiple modes (scaler, ratemeter, etc.), GPS output,

• • Required static measurement MDC values are approximately 50 percent of the most restrictive release criteria 47

B*

Alan J. Blotcky Reactor Facility etc.

4.3.1 Radiological Instruments 4.3.1.1 Ludlum Model 2360 The Ludlum Model 2360 (L23 60) is a digital/analog display instrument capable of scaler and ratemeter/scan operations with one or two detection channels (dual discriminator). The L2360 has data logging capabilities and may be calibrated with one detector type at a time without adjusting operating parameters (voltage, threshold, etc). Common applications for the L2360 are to *match it with a Ludlum Model 43-44 air proportional or Model 43-68 gas proportional..detector and calibrate it for alpha and/or beta emitters. In dual discriminator mode, it has thle ability to distinguish between alpha and beta on separate channels. The drawback to dual discriminator mode is lower detection efficiencies when compared to single channel calibration.

4.3.1.2 Ludlum Model 2221 The Ludlum Model 2221 (L2221) is a single-channel scaler/ratemeter with digital and analog display. The L2221I may be fitted with a GPS connection so that textual output of the count data may be sent to an external GPS unit for synchronization to location data. This meter may also be used as a single-channel analyzer with the proper setup of operating parameters (voltage, threshold, window, etc) which allows it to be used as a very capable screening tool for known isotopes. Common detector pairings are with a L43-68 for alpha or beta detection or a Ludlum Model 44-10 (L44-10) Sodium iodide scintillator for gross gamma capabilities. ENERCON routinely uses the L2221 with a Ludlum Model 44-10 detector connected to a Trimble ProXR GPS system for open land gross gamma scan surveys.

4.3.1.3 Ludlum Model 3/12 The Ludlum Model 3 and Model 12 (L3/12) are nearly identical instruments with the only appreciable difference being the displays. Both are ratemeter-only instruments with analog displays. This limits their usefulness in final status survey applications to limited scanning or elevated area delineation.

4.3.1.4 Ludlum Model 19 The Ludlum Model 19 (L19) is a MicroR meter with a 1"xl" Nal crystal. The L19 has an analog display and is capable of detection ranges up to 5,000 microR per hour. The low display range and steady response make this a good instrument for ambient exposure rate surveys. While the detection response may vaiy¢ by isotope, as a hand-held exposure rate meter it is quite capable.

48

_, L Alan J. Blotcky

~Reactor Facility etc.

4.3.1 Radiologieial Instruments 4.3.1.1 Ludlum Model 2360 The Ludlurn Model 2360 (L2360) is a digital/analog display instrument capable of scaler and ratemeter/scan operations with one or" two detection channels (dual discriminator). The L2360 has data logging capabilities and may be calibrated with one detector type at a time without adjusting operating parameters (voltage, threshold, etc). Common applications for the L2360 are to match it with a Ludlum Model 43-44 air proportional or Model 43-68 gas proportional. detector and calibrate it for alpha and/or beta emitters. In dual discriminator mode, it has the ability to distinguish between alpha and beta on separate channels. The drawback to dual discriminator mode is lower detection efficiencies when compared to single channel calibration.

4.3.1.2 Ludlum Model 2221 The Ludlum Model 2221 (L2221) is a single-channel scaler/ratemeter with digital and analog display. The L2221 may be fitted with a GPS connection so that textual output of the count data may be sent to an external GPS unit for synchronization to location data. This meter may also be used as a single-channel analyzer with the proper setup of operating parameters (voltage, threshold, window, etc) which allows it to be used as a very capable screening tool for known isotopes. Common detector pairings are with a L43-68 for alpha or beta detection or a Ludlum Model 44-10 (L44-l0) Sodium iodide scintillator for gross gamma capabilities. ENERCON routinely uses the L2221 with a Ludlum Model 44-10 detector connected to a Trimble ProXR GPS system for open land gross gamma scan surveys.

4.3.1.3 Ludlum Model 3/12 The Ludlum Model 3 and Model 12 (L3/12) are nearly identical instruments with the only appreciable difference being the displays. Both are ratemeter-only instxniments with analog displays. This limits their usefulness in final status survey applications to limited scanning or elevated area delineation.

4.3.1.4 Ludlum Model 19 The Ludlum Model 19 (L19) is a MicroR meter with a 1"x1" NaI ciy¢stal. The L19 has an analog display and is capable of detection ranges up to 5,000 microR per hour. The low display range and steady response make this a good instrument for ambient exposure rate surveys. While the detection response may vary by isotope, as a hand-held exposure rate meter it is quite capable.

48

[*;

Alan J. Blotcky

)!*

Reactor Facility Mor'e accurate measurements may be obtained with an instrument such as one of the Reuter-Stokes Pressurized Ion Chamber models, but for portable operational needs, the LI19 is sufficient.

4.3.2 Radiation Detectors 4.3.2.1 Ludlum Model 43-93 The Ludlum Model 43-93 (L43-93) is a Zinc Sulfide (ZnS) alpha and beta scintillation with a Mylar window. The probe active surface area is 1 00-cm 2. Typical 4-2t detection efficiency si is

• 9%-1 5% depending on the meter voltage and other calibration parameters. This detector is most commonly used on hard, relatively smooth surfaces, such as concrete. Care must be taken iniareas where sharp edges are present to avoid puncturing the Mylar window or breaking an anode wire.

4.3.2.2 Ludlum Model 44-9 The Ludlum 44-9 (L44-9) Geiger-Mueller tube detector is Useful for alpha, beta, and gamma detection for small, discreet areas. It is ideally suited for very small areas where relatively high levels of residual contamination are possible. While it has fairly good typical detection efficiencies from 5-32% (4-3z), the small detection area severely limits its' suitability for FSS.

4.3.2.3 Ludlum Model 44-10 The Ludlum Model 44-10 (L44-10) detector is a gamma scintillator with a 2" x 2" Sodium Iodide (NaI) crystal. It is commnonly used for open area gross gamma surveys. Typical count response is approximately 900 counts per minute per microR per hour for Cs137 energy (662 keV).

4.4 Calibration and Maintenance Instrumnentation used for FSS are calibrated and maintained in accordance with the Instrumentation Program procedures. Radioactive sources used for calibration are traceable to the National Institute of Standards and Technology (NIST) and use, to the extent practical, geometries designed to mimic the type of samples being counted. If vendor services are used, these are obtained in accordance with purchasing requirements for quality related services to ensure an appropriate level of quality. The calibration source isotopes for beta survey instruments are Tc 99 and Sr90 which have an appropriate energy for site-specific beta emitting nuclides. The alpha calibration sources are Th23° and Pu239 which have an appropriate alpha energy for plant-specific alpha emitting nuclides. Gamma scintillation detectors generally are calibrated using Cs'37. Calibration sources other than those listed may be used provided they demonstrate appropriately conservative detection efficiency for the radionuclides of interest. In all cases, the surface efficiency as determined for the energies of the radionuclides of concern is utilized 49

Jm Alan J. Blotcky

~Reactor Facility regardless of the energy of the calibration source.

4.4.1 Response Checks Instrument response checks are conducted to assure proper instrument response and operation.

Appropriate response values are discussed in the applicable instrument program procedures.

Response checks are performed daily before instrument use and again at the end of use. Check sources are appropriate for the type of radiation as that being measured in the field and are, to the extent practical, held in fixed geometry jigs for reproducibility, if an instrument fails a response check, it will be clearly labeled "Do Not Use" and will be removed from service until the problem is corrected in accordance with applicable procedures. Measurements made between the last acceptable check and the failed check will be evaluated to determine if they may remain in the data set.

50

m*

Reactor Facility 5.0 QUALITY ASSURANCE 5.1 Project Description and Schedule Each area of the site is divided into survey areas and units, and is classified as listed in Section 3.5. The survey measurements for each survey unit are determined during the survey design phase.

Portions of the FSS are performed during deconstruction activities as areas become available for survey.

Non-impacted and Class 3 areas may be evaluated for release prior to significant decommissioning activities taking place.

5.2 Quality Objectives and Measurement Criteria Type I errors are established at 0.05 unless another value is authorized by the USNRC. Type II errors will initially be set at 0.05, but may be increased with concurrence by the SO.':::

5.2.1 Trainina and Qualification i

Personnel performing FSS measurements will be trained and qualifiedi At a minimum, training will include the following topics:

Procedures governing handling FSS~, data such as, but not limited to, document control, records retention, and ~chain,of custody.

Operating field and laboratory instrumentation used for FSS.

Performing FSS measurements and collecting samples.

The extent of training and qualification will be commensurate with the education, experience and proficiency of the individual, and the.scope, complexity and nature of the activity.

Training records will be maintained as quality records.

5.2.2 Survey Document~ation Each FSS measuirement will be identified by date, instrument, location, type of measurement, and mode of operation.-: Generation, handling and storage of the original FSS design and data packages will be controlled. The FSS records have been designated as quality documents and will be maintained in accordance with document control procedures.

5.3 Measurement/D~ata Acq uisition 5.3.1 Survey Desiin The site will be divided into survey areas. Each survey area will contain one or more survey units. A survey package specifies the type and number of measurements required for a survey unit based on the classification and known characterization data results. Each survey area will 51

• ,UE*

Alan J. Blotcky Reactor Facility have one or more survey packages.

5.3.2 Written Procedures Sampling and survey tasks must be performed properly and consistently in order to assure the quality of the FSS results.

The measurements are performed in accordance with approved, written procedures. Approved procedures describe the methods and techniques used for the FSS measurements.

Each procedure written for the purpose of directing FSS data collection or evaluation shall include a section describing the QA/QC goals and methods specific to that procedure.

5.3.3 Sampling Methods" Samples are collected and placed into new containers using either new tools or tools that have been thoroughly decontaminated and double-rinsed with clean, water from two sources, i.e. two separate dip tanks. Surface abrasion of the tools may be necessary to dislodge adhered media from previous samples. This may entail using a stiff-bristled brush in the first dip tank. Tools will be air-or towel-dried prior to reuse.

5.3.4 Chain of Custody Responsibility for custody of samples from the point of collection through the determination of the final survey results is established by procedure.

When custody is transferred, a COC will accompany the sample for tracking purposes. Secure storage is provided for archived samples until such time it is determined to no longer be necessary, i.e. license termination for samples utilized for FSS.

5.4 OA/QC Surveys and Samples If replicate QA/QC measurements or sample analyses fall outside of their respective acceptance criteria, a documented investigation is performed in accordance with approved procedures; and the Corrective Action Process described in Section 5.10.3 will be implemented. The investigation typicalliyinvlmv*ves verification that the proper data sets were compared, the relevant instruments were operating properly and the survey/sample points were properly identified and located.

Relevant personnel are interviewed, as appropriate, to determine if proper instructions and procedures were provided and followed, and proper measurement and handling techniques were used, including COC, where applicable. When appropriate, additional measurements are taken.

Following the investigation, a documented determination is made regarding the usability of the survey data and if the impact of the discrepancy adversely affects the decision on the radiological 52

F;2*

Alan J. Blotcky

~Reactor Facility status of the survey unit.

5.4.1 Replicate Measurements Quality assurance of the survey process is evaluated using one or more of the following methods and is performed on at least 5% of the collected data. Replicate measurements generally involve a full redesign of the survey package. When a full replicate survey redesign is not desired or practical, replicate scan or direct measurements may be performed to satisfy the 5% minimum requirement for QA/QC evaluation.

QA/QC replicate measurements will be performed once a survey unit FSS has,been completed.

At a minimum, one replicate measurement will be performed for each survey iunit ais a standard practice. This will be done to provide an early indication of potential problems With final status V,--7 survey procedures/guidelines, instrumentation, and/or possible cross-c*ontamination from ongoing site decontamination activities. All replicate measurements wilt be perfoirmed in accordance with the FSSP. Verification of final status surveys will include the finvestigation of documentation, survey protocol, instrument calibration, and any practice 'associated with the collection of survey data. Replicate measurements may be completed by a differ~ent technician than performed the original survey, the SO, or SO.

5.4.2 Volumetric Analyses,,

For volumetric samples, Quality Control will consist of requiring the vendor analytical laboratory to be NVLAP accredited. However, as an additional quality measure, randomly selected samples are subject to blank sample, blind duplicate, split, recount, or third party analyses.

The acceptance criterion for blank saimples is that no plant-derived radionuclides are detected to the required MDA. Some sample media, such as asphalt, will only be subject to third party analyses or a sample recount due to the lack of homogeneity. The criterion for blind duplicates, split, recount, and third'-party analyses is that the two measurements are within +20% of each other.

5.5 Instrument Selection. Calibration and Operation Proper selection and use of instrumentation ensure sensitivities are sufficient to detect radionuclides at specified MDC. These requirements help assure the validity of the collected survey data.

Instrument calibrations are performed with sources traceable to the National institute for Standards and Testing (NIST) using approved procedures. Issuance, control and operation of the survey instruments are conducted in accordance with the Instrumentation Control and Issue procedure.

53

F tUL Alan J. Blotcky Reactor Facility 5.6 Control of Consumables In order to ensure the quality of data obtained from FSS surveys and samples, new sample containers will be used for each sample taken. Tools used to collect samples will be cleaned to remove contamination prior to taking additional samples. Tools will be decontaminated after each sample collection and surveyed for contamination.

5.7 Control of Vendor-Supplied Services Vendor-supplied services, such as instrument calibration and laboratory sample analysis,, will be procured from vendors on the site Approved Vendor List (AVL) in accordance with approved quality and procurement procedures.

5.8 Software Control Software used for data reduction, storage, or evaluation will be fully documented and certified by the vendor. The software will be tested prior to the initial use by an appropriate test data set.

5.9 Data Manaaement Survey data control from the time of collection through evaluation is specified by procedure. All survey and data analysis records pertaining to the final radiological status are considered quality records and are maintained in accordance with applicable document control procedures.

5.10 Assessment and Oversight 5.10.1 Assessments Periodic FSS self-assessments are conducted in accordance with approved procedures.

The findings are tracked and trended in accordance with these procedures.

5.10.2 Independent Review of Survey Results A minaimum of 5% of randomly selected survey packages from completed survey units are independentlyreviewed by QA to ensure that the survey measurements are taken and documented in accordance with approved procedures.

5.10.3 Corrective Action Process The corrective action process, already established as part of the site 10 CFR Part 50 Appendix B Quality Assurance Program, is applied to FSS for the documentation, evaluation, and implementation of corrective actions.

The process will be conducted in accordance with approved procedures which describe the methods used to initiate Condition Reports (CR) and 54

!Jm*

Alan J. Blotcky Reactor Facility resolve self-assessment and corrective action issues related to FSS. The CR evaluation effort is commensurate with the classification of the CR and could include root cause determination, barrier screening and extent of condition reviews.

5.10.4 Reports to Management Reports of audits, assessments, and trend data are reported to management upon completion.

5.11 Data Validation Survey data are reviewed prior to evaluation or analysis for completeness and for the presence of outliers.

Comparisons to investigation levels are made and measurements :exceeding the investigation levels are evaluated. Verified data are subjected to the Sign test,* theWRS test, Sign Unity test, or WRS Unity test as determined appropriate by the assignied SO.

5.12 Confirmatory Measurements

{:

It is anticipated that the USNRC and other regulatory ageicies:' will choose to conduct confirmatory measurements in accordance with applicable laws and regulations.

The USNRC may take confirmatory measurements to make a determination in accordance with 10 CFR 50.82(a)(1 1) that the final radiation survey 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. VA will comply with the 25-mrem/yr criteria of 10 CFR Part 20, Subpart E by demonstrating that measurement results meet the DCGLw developed using the approved dose-based scenario. Therefore, the confirmatory measurements taken by the USNRC and other regulatory agencies are based on the same DCGLw.

Timely and frequent communications with these agencies ensure that they are afforded sufficient opportunity for these confirmatory measurements prior to any difficult to reverse decommissioning actions, e.g. new construction or lexcavation backfill.

55

Alan J. Blotcky

• lj Reactor Facility

6.0 REFERENCES

• Alan J. Blotcky Reactor Facility Decommissioning Plan (Revised), May 7, 2013

• 10CFR20.1402, "Radiological Criteria for Unrestricted Use"

• 10CFR50.82, "Termination of License"

• NUREG-15 75, "Multi-Agency Radiation Survey and Site Investigation Manual" (MARSSLM), Revision 1, August 2002

• NUREG-1507, "Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Field Conditions," December 1997" NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys," Rev. 1, June 1998 draft NUREG-1549, "Using Decision Methods for Dose Assessment to Comply with Radiological Criteria for License Termination," Ju1i* 1998 draft NUREG-1757, "Consolidated Decommissioning Guidanace," September 2006 NUREG-1727, "NMSS Decommissioning Standard Review Plan", September 2000

• 1O 7503-1, "Evaluation of Surface Co nta~inao",18 inaton",198 56

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