ML20296A513

Confirmatory Survey Report 5299-SR-04 - June 26, 2019

ML20296A513
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	06/26/2019
From:	Altic N Oak Ridge Institute for Science & Education
To:	Marlayna Vaaler Doell Reactor Decommissioning Branch
Marlayna V. Doell
References
DCN 5299-SR-04-0, RFTA 18-003
Download: ML20296A513 (39)

Text

100 ORAU Way

• Oak Ridge

• TN 37830

• orise.orau.gov June 26, 2019 Ms. Marlayna Doell U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Division of Decommissioning, Uranium Recovery, and Waste Programs Reactor Decommissioning Branch, Mail Stop: T8F5 11545 Rockville Pike Rockville, MD 20852

SUBJECT:

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE REACTOR BUILDING BOWL AT THE LA CROSSE BOILING WATER REACTOR, GENOA, WISCONSION DOCKET NO. 50-409; RFTA 18-003 DCN 5299-SR-04-0

Dear Ms. Doell:

The Oak Ridge Institute for Science and Education (ORISE) is pleased to provide the enclosed final report detailing the confirmatory survey activities of the Reactor Building basement at the La Crosse Boiling Water Reactor in Genoa, Wisconsin. NRCs comments have been incorporated into this final version.

Please feel free to contact me at 865.241.8793 or Erika Bailey at 865.576.6659 if you have any comments or concerns.

Sincerely, Nick A. Altic, CHP Health Physicist/Project Manager ORISE NAA:jc Attachment Electronic distribution:

K. Conway, NRC R. Edwards, NRC D. Hagemeyer, ORISE E. Bailey, ORISE File/5299 File/5271

INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE REACTOR BUILDING BOWL AT THE LA CROSSE BOILING WATER REACTOR, GENOA, WISCONSIN N. A. Altic, CHP ORISE FINAL REPORT Prepared for the U.S. Nuclear Regulatory Commission JUNE 2019 Further dissemination authorized to NRC only; other requests shall be approved by the originating facility or higher NRC programmatic authority.

ORAU provides innovative scientific and technical solutions to advance research and education, protect public health and the environment and strengthen national security. Through specialized teams of experts, unique laboratory capabilities and access to a consortium of more than 100 major Ph.D.-granting institutions, ORAU works with federal, state, local and commercial customers to advance national priorities and serve the public interest. A 501(c) (3) nonprofit corporation and federal contractor, ORAU manages the Oak Ridge Institute for Science and Education (ORISE) for the U.S. Department of Energy (DOE). Learn more about ORAU at www.orau.org.

NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge Associated Universities.

This report was prepared as an account of work sponsored by the United States Government.

Neither the United States Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

LACBWR Rx Bowl 5299-SR-04-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE REACTOR BUILDING BOWL AT THE LA CROSSE BOILING WATER REACTOR, GENOA, WISCONSIN Prepared by N. A. Altic, CHP June 2019 FINAL REPORT Prepared for the U.S. Nuclear Regulatory Commission This document was prepared for the U.S. Nuclear Regulatory Commission (NRC) by the Oak Ridge Institute for Science and Education (ORISE) through an interagency agreement between the NRC and the U.S. Department of Energy (DOE). ORISE is managed by Oak Ridge Associated Universities under DOE contract number DE-SC0014664.

LACBWR Rx Bowl 5299-SR-04-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE REACTOR BUILDING BOWL AT THE LA CROSSE BOILING WATER REACTOR, GENOA, WISCONSIN Prepared by:

Date:

N. A. Altic, CHP, Health Physicist/Project Manager ORISE Reviewed by:

Date:

P. H. Benton, Quality Manager ORISE Reviewed by:

Date:

W. F. Smith, Senior Chemist ORISE Reviewed and approved for release by:

Date:

E. N. Bailey, Survey Projects Group Manager ORISE FINAL REPORT JUNE 2019

LACBWR Rx Bowl i

5299-SR-04-0 CONTENTS FIGURES........................................................................................................................................................... ii TABLES.............................................................................................................................................................. ii ACRONYMS.................................................................................................................................................... iii

1. INTRODUCTION....................................................................................................................................... 1

2. SITE DESCRIPTION................................................................................................................................. 2

3. DATA QUALITY OBJECTIVES............................................................................................................. 3 3.1 STATE THE PROBLEM........................................................................................................................ 4 3.2 IDENTIFY THE DECISION................................................................................................................. 4 3.3 IDENTIFY INPUTS TO THE DECISION............................................................................................. 5 3.3.1 Radionuclides of Concern and Release Guidelines.............................................................. 5 3.4 DEFINE THE STUDY BOUNDARIES................................................................................................. 7 3.5 DEVELOP A DECISION RULE........................................................................................................... 7 3.6 SPECIFY LIMITS ON DECISION ERRORS......................................................................................... 8 3.7 OPTIMIZE THE DESIGN FOR OBTAINING DATA.......................................................................... 9

4. PROCEDURES............................................................................................................................................ 9 4.1 REFERENCE SYSTEM......................................................................................................................... 9 4.2 SURFACE SCANS................................................................................................................................. 9 4.3 MEASUREMENT/SAMPLING LOCATION DETERMINATION....................................................... 10 4.3.1 ISOCS Measurements............................................................................................................ 10 4.3.2 Volumetric Sampling.............................................................................................................. 11

5. SAMPLE ANALYSIS AND DATA INTERPRETATION............................................................... 11

6. FINDINGS AND RESULTS................................................................................................................... 12 6.1 SURFACE SCANS............................................................................................................................... 12 6.2 IN SITU GAMMA SPECTROMETRY MEASUREMENTS................................................................... 12 6.3 ROC CONCENTRATIONS IN CONCRETE SAMPLES..................................................................... 14

7.

SUMMARY

AND CONCLUSIONS...................................................................................................... 16

8. REFERENCES........................................................................................................................................... 17 APPENDIX A FIGURES APPENDIX B DATA TABLES APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES APPENDIX D: MAJOR INSTRUMENTATION

LACBWR Rx Bowl ii 5299-SR-04-0 FIGURES Figure 2.1. LACBWR Site Location................................................................................................................ 2 Figure 2.2. Reactor Building Bowl (Shaded in Red) (Adapted from LS 2019)......................................... 3 Figure 6.1. Q-Q for the Reactor Building Bowl Floor, Upper Walls, and Lower Walls....................... 12 Figure 6.2. Comparison of ORISE Confirmatory Mean Concentrations and FSS Data...................... 14 TABLES Table 3.1. LACBWR Confirmatory Survey Decision Process.................................................................... 5 Table 3.2. DCGLs for the Reactor Building Basementa.............................................................................. 6 Table 6.1. Summary of Reactor Building Bowl Confirmatory In Situ Gamma Spectrometry Measurements................................................................................................................................................... 13 Table 6.2. Summary of Analytical ROC Results from the Eight Random Reactor Bowl Volumetric Concrete Samples............................................................................................................................................. 15 Table 6.3. Summary of H-3 and Ni-63 Results from the Eight Random Reactor Bowl Volumetric Concrete Samples............................................................................................................................................. 16

LACBWR Rx Bowl iii 5299-SR-04-0 ACRONYMS AA alternate action BWR boiling water reactor cpm counts per minute DCGL derived concentration guideline level DCGLOp operational derived concentration guideline level DCGLBC base case derived concentration guideline level DPC Dairyland Power Cooperative DQO data quality objective FESW fuel element storage well FOV field-of-view FSS final status survey ISFSI Independent Spent Fuel Storage Installation ISOCS In-Situ Object Counting System HPGe high purity germanium LACBWR La Crosse Boiling Water Reactor LTP license termination plan MeV mega-electron volt mrem/yr millirem per year NaI sodium iodide NRC U.S. Nuclear Regulatory Commission ORAU Oak Ridge Associated Universities ORISE Oak Ridge Institute for Science and Education pCi/g picocuries per gram pCi/m2 picocuries per square meter PSQ principal study question Q-Q quantile-quantile ROC radionuclide of concern SOF sum-of-fractions SU survey unit TAP total absorption peak TEDE total effective dose equivalent UCL95 95% upper confidence level VSP Visual Sample Plan

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5299-SR-04-0 INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE REACTOR BUILDING BOWL AT THE LA CROSSE BOILING WATER REACTOR, GENOA, WISCONSIN

1. INTRODUCTION The La Crosse Boiling Water Reactor (LACBWR), a 50 Megawatt Electric BWR located in Genoa, Wisconsin, was originally a demonstration plant funded by the U.S. Atomic Energy Commission.

The plant was later sold to Dairyland Power Cooperative (DPC) with a provisional operating license.

The BWR achieved initial criticality on July 11, 1967, and operated for 19 years until being permanently shut down on April 30, 1987. Consequentially, DPCs authority to operate LACBWR under Provisional Operating License DPR-45 (issued by the U.S. Nuclear Regulatory Commission

[NRC] on August 28, 1973) was amended via License Amendment 56 (August 4, 1987) to possession only authority (LS 2018).

Dismantling unused and offline systems and waste disposal operations began in 1994, the Reactor Pressure Vessel (head, internals, and 29 control rods sealed with concrete), stored waste in the Fuel Element Storage Well (FESW), and other Class B/C wastes were shipped offsite for disposal in June 2007. Other systems and componentssuch as spent fuel storage racks, gaseous waste disposal systems (excluding the underground gas storage tanks), condensate and feedwater system (excluding condensate storage tank and condenser), the turbine and generator, and various components located in the Turbine Building (cooling water system pumps, heat exchangers, piping, etc.)have also been removed. In September 2012, 333 irradiated fuel assemblies from the FESW were packaged in five dry casks and transferred to the Independent Spent Fuel Storage Installation (ISFSI) for storage (LS 2018). In May 2016, the NRC consented to having the possession, maintenance, and decommissioning authorities of the LACBWR site transferred from DPC to LaCrosseSolutions, LLC.

LACBWR has now submitted a License Termination Plan (LTP) to the NRC requesting the removal of all remaining open-land and structures, except for the fenced area surrounding the ISFSI, from License DPR-45 (LS 2018). Per the LTP, the Waste Gas Tank Vault and the Reactor Building structures will be demolished and removed to a depth of three feet below gradecorresponding to the 636-foot elevation. The radiologically impacted above grade structures that will remain are the:

LACBWR Administration Building, G-3 Crib House, Transmission Sub-Station Switch House, G-1 Crib House, Barge Wash Break Room, Back-up Control Center, and Security Station (LS 2018).

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5299-SR-04-0 LACBWR will subject all remaining land areas and structures to a final status survey (FSS) to demonstrate compliance with federal radiological release criteria.

NRC requested that the Oak Ridge Institute for Science and Education (ORISE) perform confirmatory survey activities to independently assess the final radiological condition of the Reactor Building basement. This report summarizes the confirmatory survey activities associated with the Reactor Building basement (also known as the Reactor Building Bowl).

2. SITE DESCRIPTION The LACBWR site enclosure comprises an area of 0.61 hectare (1.5 acres) within the 66.2 hectare licensed site located approximately 1.6 kilometers (1 mile) south from the village of Genoa, Wisconsin on the eastern shore of the Mississippi River. The licensed site is shared with the non-nuclear Genoa-3 Fossil Station. The operational fossil plants buildings and structures were classified as non-impacted and are not subject to the release surveys specified in the LTP (LS 2018).

Figure 2.1 provides an aerial view site prior to demolition activities.

Figure 2.1. LACBWR Site Location

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5299-SR-04-0 The Reactor Building bowl was constructed of concrete with an internal steel liner. Additional concrete inside the liner provided structural support and biological shielding. The steel liner, including all internal concrete, was removed prior to the confirmatory survey activities. Structural concrete, external to the steel liner, above the 636-ft elevation had also been removed. The remaining reactor building structural concrete is in the form of an ellipsoidal bowl and is depicted in Figure 2.2.

Figure 2.2. Reactor Building Bowl (Shaded in Red) (Adapted from LS 2019)

3. DATA QUALITY OBJECTIVES The data quality objectives (DQOs) process was applied to the design of the confirmatory survey.

The DQOs applied, and described herein, were consistent with the Guidance on Systematic Planning Using the Data Quality Objectives Process (EPA 2006) and provided a formalized method for planning confirmatory radiation surveys, improving survey efficiency and effectiveness, and ensuring that the type, quality, and quantity of data collected are adequate for the intended decision applications.

The seven steps of the DQO process are as follows:

1. State the problem

2. Identify the decision/objective

3. Identify inputs to the decision/objective

LACBWR Rx Bowl 4

5299-SR-04-0

4. Define the study boundaries

5. Develop a decision rule

6. Specify limits on decision errors

7. Optimize the design for obtaining data Confirmatory survey DQOs were originally presented in ORISE 2019 and are represented here for completeness.

3.1 STATE THE PROBLEM The first step in the DQO process defined the problem that necessitated the study, identified the planning team, and examined the project budget and schedule. LACBWR is requesting approval from the NRC to remove the remaining open-land and structures not associated with the ISFSI facility from its 10 Code of Federal Regulations Part 50 license. NRC requested that ORISE perform confirmatory surveys at LACBWR to provide independent radiological data to assist the NRC in evaluating the adequacy and accuracy of LACBWRs FSS results. Therefore, the problem statement is as follows:

Confirmatory surveys are necessary to generate independent radiological data for NRCs consideration in the evaluation of the FSS design, implementation, and results for demonstrating compliance with the release criteria.

3.2 IDENTIFY THE DECISION The second step in the DQO process identified the principal study questions (PSQs) and alternate actions (AAs), developed a decision statement, and organized multiple decisions, as appropriate.

This was done by specifying AAs that could result from a yes response to the PSQ and combining the PSQ and AAs into a decision statement. The PSQ, AAs, and combined decision statement were organized based on the survey unit (SU) type (i.e., the associated FSS methodology) and are presented in Table 3.1.

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5299-SR-04-0 Table 3.1. LACBWR Confirmatory Survey Decision Process Principal Study Question Alternate Actions Do confirmatory survey results agree with the final radiological survey data for the Reactor Building Bowl structure?

Yes:

Compile confirmatory data and report results to the NRC for their decision making. Provide independent interpretation that confirmatory field surveys did not identify anomalous areas of residual radioactivity, quantitative field and laboratory data satisfied the NRC-approved decommissioning criteria, and/or that statistical sample population examination/assessment conditions were met.

No:

Compile confirmatory data and report results to the NRC for their decision making. Provide independent interpretation of confirmatory survey results identifying any anomalous field or laboratory data and/or when statistical sample population examination/assessment conditions were not satisfied for the NRCs determination of the adequacy of the FSS data.

Decision Statement Confirmatory survey results did/did not identify anomalous results or other conditions that preclude the FSS data from demonstrating compliance with the release criteria.

3.3 IDENTIFY INPUTS TO THE DECISION The third step in the DQO process identified both the information needed and the sources of this information, determined the basis for action levels, and identified sampling and analytical methods that will meet data requirements. For this effort, information inputs included the following:

• Concrete characterization and FSS data for Reactor Building bowl

• Derived concentration guideline levels (DCGLs), discussed in subsection 3.3.1

• ORISE confirmatory survey results including: surface radiation scans, direct surface activity measurements, and in situ gamma spectrometry measurements with an In-Situ Object Counting System (ISOCS)

• ORISE volumetric sample analysis results for concrete and samples 3.3.1 Radionuclides of Concern and Release Guidelines The primary radionuclides of concern (ROCs) identified for LACBWR are beta-gamma emitters fission and activation productsresulting from reactor operations. At LACBWR, there are five

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5299-SR-04-0 distinct source term media considered as part of the dose modeling: basement structures, soils, buried piping, above grade buildings, and existing groundwater (LS 2018). LACBWR has developed site-specific DCGLs for remaining basement structures that correspond to a residual radioactive contamination level, above background, which could result in a total effective dose equivalent (TEDE) of 25 millirem per year (mrem/yr) to an average member of the critical group. These DCGLsdefined in LACBWRs LTP as Base Case DCGLsare radionuclide-specific and independently correspond to a TEDE of 25 mrem/yr for each source term. In order to ensure that total dose from all source terms is less than the NRCs release criteria, the Base Case DCGLs (DCGLBC) are further reduced to Operational DCGLs (DCGLOp). The Operational DCGLs are scaled to an expected dose from prior investigations and are used for remediation and FSS design purposes. The initial suite of ROCs present at LACBWR has been reduced based on an insignificant dose contribution from a number of radionuclides. Base Case and Operational DCGLs, accounting for insignificant dose contributors, for the Reactor Building Basement [bowl] are provided in Table 3.2.

Table 3.2. DCGLs for the Reactor Building Basementa ROC Base Case DCGL (pCi/m2)

Operational DCGL (pCi/m2)

Co-60 5.16E+06 3.61E+05 Sr-90 1.45E+07 1.02E+06 Cs-137 2.17E+07 1.52E+06 Eu-152 1.19E+07 8.33E+05 Eu-154 1.10E+07 7.71E+05 aRecreated from LS 2019.

pCi/m2 = picocuries per square-meter Because each individual Base Case DCGL represents a separate radiological dose, the sum-of-fractions (SOF) approach must be used to evaluate the total dose from the SU and demonstrate compliance with the dose limit. Per Equation 5-5 in the LTP, the SOF calculations for the FSS data are performed as follows:

=

Cmean,j DCGLBC,j

=1

+

,x Eq. (3-1)

Where Cmean,j is the mean concentration of ROC j, CElv,j is an elevated area of ROC j, DCGLBC,j is the Base Case DCGL for ROC j, and AF is the area factor for ROC j. The Base Case DCGLs

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5299-SR-04-0 for basement structures are scaled by area to account for elevated radioactivity. For basement SUs, the AF in the equation above is equal to the SU surface area divided by the surface area of the elevated hot-spot (AF = SUSA/HSSA). Note that radioactivity concentrations will not be corrected for structural material background contributions for conservatism.

Lastly, the FSS data must also demonstrate total dose compliance for each media by summing the maximum SOF for each of the five media, plus additional source terms not accounted for in the five media. If the summed result satisfies the unity rule (less than or equal to one), then the site meets the criterion for unrestricted release. Maximum SOFs are calculated using the maximum dose result from each source term over the respective Base Case DCGL, plus the dose from areas of elevated activity.

3.4 DEFINE THE STUDY BOUNDARIES The fourth step in the DQO process defined target populations and spatial boundaries, determined the timeframe for collecting data and making decisions, addressed practical constraints, and determined the smallest subpopulations, area, volume, and time for which separate decisions must be made. Confirmatory surveys were conducted for the Reactor Building bowl during the period of April 8-10, 2019. Temporal limitations prevented ORISE from achieving 100% areal coverage with the in situ measurements, as done with the FSS measurements. The Reactor Building bowl was assigned SU identification number B1-010-001D by LACBWR.

3.5 DEVELOP A DECISION RULE The fifth step in the DQO process specified appropriate population parameters (e.g., mean, median), confirmed action levels are above detection limits, and developed an ifthen decision rule statement. Decision rules for this survey were based on independent scan surveys, in situ gamma spectrometry measurements, and concrete sample results to assess whether there is a statistical bias relative to the FSS data. Typically, decision rules are based on a statistical comparison of the ORISE survey data and the FSS data using an appropriate test. However, the difference in the number of ORISE and FSS sample size is significant. For example, 43 FSS in situ gamma spectrometry measurements were collected for the Reactor Building bowl SU, whereas ORISE collected only 17 random measurements. The approximately 2:1 sample size ratio significantly reduces the

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5299-SR-04-0 statistical power to detect a difference between the two sample groups. Therefore, alternative assessment methods will be employed.

The parameter of interest for each survey area was the mean ROC concentration and the associated confidence interval level. The mean and associated uncertainty for the confirmatory measurements was compared directly to that of the FSS data. Agreement between the two sample sets was evaluated by the overlapping of confidence intervals. The SOF of the confirmatory survey data was compared directly to the DCGLBC. The aforementioned information was combined to formulate the decision rule for the Reactor Building bowlwhich is stated as follows:

If the mean ROC concentrations of the confirmatory and FSS sample populations overlap at the 95% confidence level and results are below the applicable limit (SOF via Eq. 4-1), conclude that confirmatory survey data agree with the FSS dataotherwise, perform further evaluation(s) and provide technical comments to the NRC.

3.6 SPECIFY LIMITS ON DECISION ERRORS The sixth step in the DQO process examined the consequences of making an incorrect decision and established bounds of decision errors. Decision errors were controlled both during the confirmatory investigations and during data quality assessment and were based on two orders of control.

The first order of control was related to sample size, which impacts the degree to which the estimated sample mean is bound. Visual Sample Plan (VSP), version 7.9, was used to determine the confirmatory survey sample size using the FSS/characterization data as planning inputs. The constraint on the estimated mean was not larger than the difference between the DCGL and the reported FSS data mean ( 1.00 - FSS Mean SOF). The confirmatory survey mean was estimated at the 95% confidence level (refer to section 4.3 of this report for additional details).

As stated in Section 3.4, temporal limitations prevented ORISE from achieving 100% areal coverage of the Reactor Building bowl with the in situ measurements. Thus, the presence of all of locations not represented by the arithmetic mean of the confirmatory in situ measurement data set could not be identified and accounted for using Equation 3-1. Therefore, the 95-percent upper confidence level (UCL95) of the mean was used for the SOF calculations. The UCL accounts for the

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5299-SR-04-0 uncertainty in the estimate of the mean due to sampling error, the probability of a Type I decision error is minimized (i.e., incorrectly concluding that the ROC concentration is less than DCGLBC).

The second order of control was to optimize the analytical MDCs with respect to sample count times of analyses performed by ORISE, both for field and laboratory measurements. Measurement MDCs for the in situ gamma spectrometry measurements were less than 50% of the operational DCGLs presented in Section 3.3.1. Nominal MDCs for laboratory instrumentation were less than 10% of the operational DCGLs.

3.7 OPTIMIZE THE DESIGN FOR OBTAINING DATA The seventh step in the DQO process was used to review DQO outputs, develop data collection design alternatives, formulate mathematical expressions for each design, select the sample size to satisfy DQOs, decide on the most resource-effective design of agreed alternatives, and document requisite details. Specific survey procedures are presented in Section 4.

4. PROCEDURES The ORISE survey team performed visual inspections, measurements, and sampling activities within the accessible survey areas specifically requested by the NRC. Survey activities were conducted in accordance with the Oak Ridge Associated Universities (ORAU) Radiological and Environmental Survey Procedures Manual and the ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2016a and ORAU 2018).

4.1 REFERENCE SYSTEM ORISE referenced confirmatory measurement/sampling in the Reactor Building bowl to the FSS measurement locations. Measurement and sampling locations were documented on detailed survey maps.

4.2 SURFACE SCANS Thallium doped sodium iodide (NaI[Tl]) detectors were used to evaluate direct gamma radiation levels on structural surfaces. The upper walls of the Reactor Building bowl received medium-to high-density (50-100%) scan coverage with priority given to areas that LACBWR identified

LACBWR Rx Bowl 10 5299-SR-04-0 contamination. The lower walls and floor of the Reactor Building bowl received high-density scan coverage.

All detectors were coupled to Ludlum Model 2221 ratemeter-scalers with audible indicators and were also coupled to data-loggers to electronically record all scanning data. Locations of elevated response that were audibly distinguishable from localized background levels, suggesting the presence of residual contamination, were marked for further investigation with in situ gamma spectrometry measurements and/or volumetric sampling.

4.3 MEASUREMENT/SAMPLING LOCATION DETERMINATION Measurements/samples were collected from both randomly and judgmentally selected locations.

VSP was used to assess the sample size required for decision making and to randomly place locations throughout the survey area. FSS data for the Reactor Building bowl SU were not available at the time of survey planning. Based on the temporal constraints and previous experience, 16 random in situ measurements were planned; however, 17 random measurements were collected.

Eight concrete core samples were randomly collected from the 17 in situ measurement locations.

One judgmental in situ measurement and concrete core sample were also collected from a region of interest that was identified as the survey progressed.

4.3.1 ISOCS Measurements In situ gamma spectrometry measurements were performed using a portable high-purity germanium (HPGe) detector. Data acquisition was performed via Canberras Genie 2000 software. Efficiency curvesencompassing the applicable ROCsfor the measurement geometry were generated using ISOCS calibration software. A circular cylinder with uniform source term was used as the ISOCS geometry template. The detector was positioned and collimated such that the field-of-view (FOV) was approximately 28 m2 to coincide with the LACBWR measurement FOV. Due to physical boundaries, the FOV was adjusted to approximately 5 m2 for some measurements.

The judgmental measurement location was selected based on surface scan results. The detector standoff was adjusted for the judgmental measurement such that the FOV was adjusted to approximately 5 m2.

LACBWR Rx Bowl 11 5299-SR-04-0 4.3.2 Volumetric Sampling Concrete core samples were collected using a concrete coring rig. Samples were collected from a depth of up to 15 cm (6 inches), or until refusal. The sample depth was noted at each location.

Sampling equipment was decontaminated after collection of each sample to prevent cross contamination using deionized water and laboratory-grade detergent. One rinsate-blank was collected and analyzed to demonstrate effective decontamination.

The concrete cores were sectioned into 1.27 cm (0.5 inch) increments with a thickness corresponding to the modeled contamination depth, for a maximum of twelve increments. The increments were analyzed individually starting from the top portion (i.e., the 0 to 1.27 cm portion).

The project-specific plan stated that in the event ROCs were not detected above the analytical MDC in an increment, the subsequent increments would not be analyzed. Concrete cores that exhibited elevated activity at depth had low concentrations, therefore it was unnecessary to continue analyzing subsequent increments. The NRC approved of this approach and did not request analysis of additional increments.

5. SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data collected on site were transferred to the ORISE facility for analysis and interpretation. Sample custody was transferred to the Radiological and Environmental Analytical Laboratory in Oak Ridge, Tennessee. Sample analyses were performed in accordance with the ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2017). Concrete samples were crushed, homogenized, and analyzed by gamma spectrometry for gamma-emitting fission and activation products. Concrete samples, based on the gamma spectrometry results, were processed by wet chemistry and analyzed after separation for strontium-90, tritium, and nickel-63 by low-background gas proportional counting and/or liquid scintillation. Volumetric sample results in units of picocuries per gram (pCi/g) were converted to units of pCi/m2, as necessary, based on the concrete density and sample depth. Measurement results from the in situ gamma spectrometry measurements were reported in units of pCi/m2. ProUCL, version 5.1, was used to calculate the UCL95 for both the confirmatory and FSS data set. The mean ROC concentration and associated 95% confidence level were plotted for direct comparison.

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6. FINDINGS AND RESULTS The results of the confirmatory survey activities are discussed in the subsections below.

6.1 SURFACE SCANS Overall, NaI detector scan responses ranged from approximately 7,400 to 11,000 counts per minute (cpm) for the floor, 6,500 to 11,000 cpm for the lower wall, and 4,600 to 12,000 for the upper wall of the Reactor Building bowl. Figure 6.1 presents the quantile-quantile (Q-Q) plot for the scan data.

Due to a data-logger error, a portion of the scan data for the upper walls was not electronically captured; therefore, the data in Figure 6.1 has a range of 4,600 to 9,800 cpm for the upper walls.

However, the surveyors noted in the field that the maximum response for the upper walls was approximately 12,000 cpm.

Figure 6.1. Q-Q for the Reactor Building Bowl Floor, Lower Walls, and Upper Walls 6.2 IN SITU GAMMA SPECTROMETRY MEASUREMENTS Eighteen in situ gamma spectrometry measurementsincluding 17 random and 1 judgmental measurementswere collected from the Reactor Building bowl. Individual in situ gamma

LACBWR Rx Bowl 13 5299-SR-04-0 spectrometry measurements collected from this area are presented in Appendix B, Table B-1.

Measurement locations from this area are depicted in Figure A-1. Table 6.1 provides a summary of the confirmatory random/systematic in situ gamma spectrometry measurements. One judgmental in situ measurement location (5299G18-Rx) was collected from the eastern wall of the Reactor Building bowl. This judgmental location was selected based on surface scan results.

Table 6.1. Summary of Reactor Building Bowl Confirmatory In Situ Gamma Spectrometry Measurements ROC Parameter (pCi/m2)

Fractionb Mean Median SD Min Max UCL95a Op.

BC Co-60 2.70E+03 3.34E+03 7.82E+03

-2.05E+04 1.24E+04 1.10E+04

<0.01 0.03 Cs-137 1.12E+04 8.34E+03 1.69E+04

-2.15E+04 4.67E+04 2.91E+04 0.02

<0.01 Eu-152

-3.70E+04

-4.20E+04 3.59E+04

-9.37E+04 2.81E+04

-c 0.03

<0.01 Eu-154

-1.13E+04

-5.22E+02 6.54E+04

-1.29E+05 1.00E+05

-c 0.13

<0.01 Totald 0.19 0.04 aUCL based on the Chebyshev Inequality bOp. represents the UCL95 divided by the Operational DCGL; BC represents the UCL95 divided by the Base Case DCGL.

When no UCL95 is available, the maximum value is used instead of UCL95.

cUCL95 not calculated for Eu-152 or Eu-154 due to large negative data set.

dDiscrepency in summation due to rounding.

SD = standard deviation The only gamma-emitting radionuclide identified above its MDC was Cs-137. Both Eu-152 and Eu-154 had a large negative dataset indicated by the average and median being below zero. Negative values are reported by the gamma spectrometry software when few or no radiation interactionsin the energy range of interestare recorded by the HPGe detector system. The gamma spectrometry software will correct for background contributions (due to Compton scattering from other photopeaks) in the energy region of interest. The background correction is based on surrounding data in the spectra and the resulting net response will be negative if there are more counts in the surrounding channels relative to the region of interest. Due to the numerous negative values, the maximum measurement result for Eu-152 and Eu-154 is used in the subsequent SOF calculation, in lieu of a UCL95.

The maximum measurement for each ROC was less than the respective DCGLOp and, therefore, less than the respective DCGLBC. The average SOF in this SU for gamma-emitting radionuclides is 0.04 based on the DCGLBC (or 0.19 when based on the DCGLOp).

LACBWR Rx Bowl 14 5299-SR-04-0 Mean concentrations of Co-60 and Cs-137 and their associated uncertainties for both ORISE and FSS data are plotted in Figure 6.2. The error bars in Figure 6.2 represent the uncertainty in the mean concentration, where the upper end is simply the UCL95. As indicated in Figure 6.2, both ROC mean concentrations overlap at the 95% confidence level.

Figure 6.2. Comparison of ORISE Confirmatory Mean Concentrations and FSS Data and Uncertainties for Gamma-emitting Radionuclides in the Reactor Building Bowl Biases are present when comparing ORISE measurements to the FSS data. FSS data are biased low when no residual activity is present; however, they are biased high compared to ORISE results where elevated activity is detected (such as measurement location 5299G18-Rx). The ORISE measurement collected at location 5299G18-Rx resulted in a Cs-137 concentration of 2.97x105 pCi/m2. Four FSS measurements were collected in this area ranging from 1.64x105 to 8.58x105 pCi/m2, with an average concentration of 5.08x105 pCi/m2. The FSS data concentration is biased high when compared to ORISE results which is likely due to the ORISE measurement having a larger FOV and thus averaging out the residual activity over a larger area.

6.3 ROC CONCENTRATIONS IN CONCRETE SAMPLES Nine volumetric concrete samplesincluding eight random and one judgmental samplewere collected from the Reactor Building bowl. The judgmental sample (M0009) was collected from the

LACBWR Rx Bowl 15 5299-SR-04-0 judgmental in situ gamma spectrometry measurement location based on surface scan results.

Individual results for volumetric concrete samples are provided in Appendix B, Table B.2. The concrete samples were collected from the in situ gamma spectrometry measurement locations presented in Figure A.1.

The eight random concrete sample analytical results were converted to units of pCi/m2 based on a sample depth of 1.27 cmsame as input into the ISOCS modeland a concrete density of 2.35 g/cm3. Table 6.2 provides a summary of the random volumetric concrete samples. None of the individual ROC concentrations are above their respective DCGLOp and, therefore, less than the respective DCGLBC.

Table 6.2. Summary of Analytical ROC Results from the Eight Random Reactor Bowl Volumetric Concrete Samples ROC Parameter (pCi/m2)

Mean Median SD Min Max Co-60

-8E+01

-4E+02 1E+03

-1.0E+03 2.7E+03 Cs-137 3.0E+04 8.3E+02 8.2E+04 3.9E+02 2.32E+05 Eu-152

-5E+02

-2E+02 1E+03

-2.2E+03 1E+03 Eu-154

-5E+03

-3E+03 4E+03

-1.3E+04

-2E+03 Sr-90

-2E+03

-3E+03 4E+03

-6.3E+03 4.5E+03 The only gamma-emitting radionuclide identified above its analytical MDC is Cs-137 in samples M0002, M0007, and M0009. The analytical result for Cs-137 in sample M0009 is 1.33x106 pCi/m2 for the 0 to 2.54 cm depth. This result is calculated by converting the average concentration of the samples representing 0 to 1.27 cm and 1.27 to 2.54 cm to units of pCi/m2 based on a sample depth of 2.54 cm and a concrete density of 2.35 g/cm3. This result is biased high compared to the in situ measurements acquired by ORISE and the FSS data because the activity is not averaged over a large area. Sr-90 was identified above the MDC in sample M0009. The mean SOF for Sr-90 is less than 0.01 based on the DCGLBC and the DCGLOp. Therefore, Sr-90 has a negligible contribution to the total SOF for the SU.

At the request of NRC, ORISE analyzed the concrete core samples for Ni-63 and H-3. The site did not specify DCGL values in units of pCi/m2, therefore volumetric results are presented in picocuries per gram (pCi/g). Individual results are provided in Table B.3 and a summary of the results are

LACBWR Rx Bowl 16 5299-SR-04-0 provided in Table 6.3 below. Ni-63 was identified above the MDC in samples M0006, M0007, and M0009. Additional depth increments were analyzed for samples M0007 (1.3 to 2.5 cm depth) and M0009 (1.3 to 3.8 cm depth) based on preliminary gamma spectrometry data. Ni-63 was identified in the additional increments for both samples, and Cs-137 was identified in the additional increments for sample M0009. All sample results for H-3 are below the MDC.

Table 6.3. Summary of H-3 and Ni-63 Results from the Eight Random Reactor Bowl Volumetric Concrete Samples Analyte Parameter (pCi/g)

Mean Median SD Min Max H-3

-2.4

-2.3 0.45

-3.0

-1.6 Ni-63 1.2

-2.3 0.65 0.61 2.4

7.

SUMMARY

AND CONCLUSIONS At the NRCs request, ORISE conducted confirmatory survey activities at LACBWR during the period of April 8-10, 2019. The survey activities included gamma surface scans, in situ gamma spectrometry measurements at 17 random/systematic and 1 judgmental location, and volumetric sampling at 8 of the 17 random/systematic and at the 1 judgmental location.

All individual confirmatory measurements, by both in situ measurements and volumetric samples, are well below the DCGLOp and, therefore, are also below the DCGLBC. Based on the overlap of confidence intervals and relative mean SOF magnitudes between the confirmatory and FSS data for the in situ gamma spectrometry measurements, ORISE did not identify issues that would preclude the use of gamma-emitting ROC FSS data for demonstrating compliance with release criteria. Sr-90 was identified above its MDC in the judgmental volumetric concrete core at a depth of 0 to 1.27 cm, but the total contribution of Sr-90 to the total SOF for the SU is negligible. Additional analyses for Ni-63 and H-3 were performed, at the request of NRC, on the concrete cores and only Ni-63 was detected with a max concentration of 2.40 pCi/g.

LACBWR Rx Bowl 17 5299-SR-04-0

8. REFERENCES EPA 2006. Guidance on Systematic Planning Using the Data Quality Objectives Process. EPA QA/G-4.

U.S. Environmental Protection Agency. Washington, D.C. February.

LS 2018. La Crosse Boiling Water Reactor License Termination Plan, Revision 1. LaCrosseSolutions. Genoa, Wisconsin. May.

LS 2019. Reactor Building (Rx) Bowl Final Status Survey Plan. LaCrosseSolutions. Genoa, Wisconsin.

March.

ORAU 2014. ORAU Radiation Protection Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. October.

ORAU 2016a. ORAU Radiological and Environmental Survey Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. November 10.

ORAU 2016b. ORAU Health and Safety Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. January.

ORAU 2017. ORAU Radiological and Environmental Analytical Laboratory Procedures Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. August 24.

ORAU 2018. ORAU Environmental Services and Radiation Training Quality Program Manual. Oak Ridge Associated Universities. Oak Ridge, Tennessee. July 20.

ORISE 2019. Project-Specific Plan for Confirmatory Survey Activities of the Reactor Building Bowl at the La Crosse Boiling Water Reactor Genoa, Wisconsin. Oak Ridge Institute for Science and Education. Oak Ridge, Tennessee. April 5.

LACBWR Rx Bowl 5299-SR-04-0 APPENDIX A FIGURES

LACBWR Rx Bowl A-1 5299-SR-04-0 Figure A.1. Confirmatory In Situ Gamma Spectrometry Measurement Locations

LACBWR Rx Bowl 5299-SR-04-0 APPENDIX B DATA TABLES

LACBWR Rx Bowl B-1 5299-SR-04-0 Table B.1 In Situ Gamma Spectrometry Measurements from the Reactor Building Bowl (pCi/m2)

Measurement ID Co-60 Cs-137 Eu-152 Eu-154 Result MDC Result MDC Result MDC Result MDC 5299G01-Rx 9.13E+03 2.65E+04 1.22E+04 2.96E+04 -9.37E+04 7.62E+04 2.06E+04 1.28E+05 5299G02-Rx

-4.51E+03 2.51E+04 8.34E+03 2.92E+04 -4.45E+04 7.74E+04 9.30E+04 1.39E+05 5299G03-Rx 2.16E+03 2.56E+04 -2.15E+04 2.66E+04 -3.23E+04 7.35E+04 -1.17E+05 1.01E+05 5299G04-Rx 5.75E+03 2.51E+04 3.61E+04 3.15E+04 -6.77E+04 7.39E+04 -2.83E+04 1.39E+05 5299G05-Rx 5.99E+03 2.78E+04 3.99E+03 2.80E+04 -2.78E+03 7.48E+04 -1.29E+05 1.12E+05 5299G06-Rx 5.52E+03 2.95E+04 2.60E+04 2.85E+04 2.81E+04 7.99E+04 4.27E+04 1.26E+05 5299G07-Rx

-2.49E+03 2.74E+04 6.82E+03 2.74E+04 -7.79E+03 7.43E+04 1.00E+05 1.29E+05 5299G08-Rx

-2.05E+04 2.20E+04 1.79E+04 3.01E+04 -4.20E+04 6.97E+04 -2.23E+04 1.16E+05 5299G09-Rx 8.72E+03 2.91E+04 -1.73E+04 2.70E+04 -4.47E+04 7.41E+04 -1.30E+04 1.39E+05 5299G10-Rx 1.24E+04 2.43E+04 -4.74E+02 3.00E+04 -7.21E+04 7.22E+04 1.85E+04 1.23E+05 5299G11-Rx

-4.66E+03 2.11E+04 5.65E+03 2.62E+04 -9.90E+03 7.08E+04 6.08E+03 1.20E+05 5299G12-Rx 1.12E+03 2.78E+04 1.70E+03 2.62E+04 1.54E+03 7.64E+04 -3.50E+04 1.22E+05 5299G13-Rx 3.16E+03 2.38E+04 1.66E+04 3.50E+04 -2.01E+04 6.91E+04 2.06E+04 1.27E+05 5299G14-Rx 1.28E+03 2.92E+04 1.79E+04 1.48E+04 -8.86E+04 6.69E+04 2.27E+04 1.24E+05 5299G15-Rx 1.07E+04 2.71E+04 6.91E+03 2.93E+04 -3.10E+02 7.01E+04 -5.22E+02 1.21E+05 5299G16-Rx 8.81E+03 2.56E+04 4.67E+04 1.85E+04 -4.79E+04 7.07E+04 -7.03E+04 1.11E+05 5299G17-Rx 3.34E+03 2.53E+04 2.25E+04 3.08E+04 -8.35E+04 6.84E+04 -1.01E+05 1.21E+05 5299G18-Rxa

-1.47E+04 2.39E+04 2.97E+05 2.79E+04 6.01E+03 9.19E+04 1.25E+04 1.37E+05 aJudgmental measurement

LACBWR Rx Bowl B-2 5299-SR-04-0 Table B.2 ROC Concentrations in Reactor Building Bowl Concrete Samples Sample IDa Measurement ID Concentration (pCi/m2)b Co-60 Cs-137 Eu-152 Eu-154 Sr-90 5299M0001A 5299G01-Rx

-5.1E+02 6.6E+02

-2.2E+03

-3.0E+03

-3.3E+03 5299M0002A 5299G03-Rx

-3.9E+02 2.1E+03 1E+03

-3E+03 1E+03 5299M0003A 5299G06-Rx 2E+02 1.0E+03

-2E+03

-2E+03

-5.1E+03 5299M0004A 5299G09-Rx

-9.8E+02 4.8E+02 8.1E+02

-2E+03

-2E+03 5299M0005A 5299G11-Rx

-3.3E+02 4.8E+02 3E+02

-1.3E+04

-5.1E+03 5299M0006A 5299G12-Rx

-1.0E+03 3.9E+02

-2E+03

-4.5E+03

-6.3E+03 5299M0007A 5299G14-Rx 2.7E+03 2.32E+05

-3E+02

-3.9E+03 4.5E+03 5299M0007B 5299G14-Rx

-8.7E+02 1.1E+03 1.5E+03

-7.2E+03 2E+03 5299M0008A 5299G17-Rx

-3E+02 2.0E+03 0.0E+00

-1.2E+04

-3.6E+03 5299M0009Ac 5299G18-Rx

-1.3E+03 1.29E+06

-3.3E+03 2E+03 1.2E+04 5299M0009Bc 5299G18-Rx 1.1E+03 3.307E+04

-1E+03

-6.6E+03

-2E+03 5299M0009Cc 5299G18-Rx 2E+02 6.86E+03 2E+03

-3.0E+03 2E+03 a"A" values represents 0 to 1.27 cm increment; "B" represents 1.27 to 2.54 cm increment; "C" represents 2.54 to 3.81 cm increment.

bValues converted to units of pCi/m2 based on a concrete density of 2.35 g/cm3 and sample depth of 1.27 cm.

cJudgmental sample

LACBWR Rx Bowl B-3 5299-SR-04-0 Table B.3 H-3 and Ni-63 Concentrations in Reactor Building Bowl Concrete Samples Sample IDa In Situ Measurement ID Concentration (pCi/g)b H-3 Ni-63 5299M0001A 5299G01-Rx

-1.6

+/- 1.6 0.80 +/- 0.88 5299M0002A 5299G03-Rx

-3.0

+/- 2.0 1.10 +/- 0.88 5299M0003A 5299G06-Rx

-2.3

+/- 1.9 1.01 +/- 0.84 5299M0004A 5299G09-Rx

-2.6

+/- 1.8 0.83 +/- 0.81 5299M0005A 5299G11-Rx

-1.6

+/- 1.6 0.61 +/- 0.78 5299M0006A 5299G12-Rx

-2.3

+/- 2.0 2.04 +/- 0.83 5299M0007A 5299G14-Rx

-2.8

+/- 2.0 2.40 +/- 0.94 5299M0007B 5299G14-Rx

-2.0

+/- 2.1 1.59 +/- 0.93 5299M0008A 5299G17-Rx

-2.3

+/- 1.9 0.89 +/- 0.88 5299M0009Ab 5299G18-Rx

-2.2

+/- 1.9 1.82 +/- 0.87 5299M0009Bb 5299G18-Rx

-2.9

+/- 1.9 2.18 +/- 0.96 5299M0009Cc 5299G18-Rx

-2.2

+/- 1.9 1.46 +/- 0.89 a"A" values represents 0 to 1.27 cm increment; "B" represents 1.27 to 2.54 cm increment; "C" represents 2.54 to 3.81 cm increment.

bUncertainties represent the total propagated uncertainty reported at the 95% confidence level cJudgmental sample

LACBWR Rx Bowl 5299-SR-04-0 APPENDIX C: SURVEY AND ANALYTICAL PROCEDURES

LACBWR Rx Bowl C-1 5299-SR-04-0 C.1.

PROJECT HEALTH AND SAFETY ORISE performed all survey activities in accordance with the ORAU Radiation Protection Manual, the ORAU Health and Safety Manual, and the ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2014, ORAU 2016b, and ORAU 2016a). Prior to on-site activities, a work-specific hazard checklist was completed for the project and discussed with field personnel. The planned activities were thoroughly discussed with site personnel prior to implementation to identify hazards present.

Additionally, prior to performing work, a pre-job briefing and walkdown of the survey areas were completed with field personnel to identify hazards present and discuss safety concerns. Should ORISE have identified a hazard not covered in the ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2016a) or the projects work-specific hazard checklist for the planned survey and sampling procedures, work would not have been initiated or continued until the hazard was addressed by an appropriate job hazard analysis and hazard controls.

C.2.

CALIBRATION AND QUALITY ASSURANCE Calibration of all field instrumentation was based on standards/sources, traceable to National Institute of Standards and Technology (NIST).

Field survey activities were conducted in accordance with procedures from the following documents:

• ORAU Radiological and Environmental Survey Procedures Manual (ORAU 2016)

• ORAU Radiological and Environmental Analytical Laboratory Procedures Manual (ORAU 2017)

• ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2018)

The procedures contained in these manuals were developed to meet the requirements of U.S. Department of Energy (DOE) Order 414.1D and the NRC Quality Assurance Manual for the Office of Nuclear Material Safety and Safeguards and contain measures to assess processes during their performance.

Quality control procedures include:

• Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.

LACBWR Rx Bowl C-2 5299-SR-04-0

• Participation in Mixed-Analyte Performance Evaluation Program and Intercomparison Testing Program laboratory quality assurance programs.

• Training and certification of all individuals performing procedures.

• Periodic internal and external audits.

C.3 IN SITU GAMMA SPECTROMETRY MEASUREMENTS Canberras In Situ Object Counting System (ISOCS) software was used to model efficiency curves for each location measured with the high purity germanium (HPGe) detector. The geometry templates and specific ISOCS inputs for measurement locations in the Reactor Building bowl are discussed below.

A circular plane geometry was used to assess ROC concentrations in measurements collected from this area. An illustration of the circular plane geometry appears in Figure C.1. The geometry template was modeled in the same manner as presented in LACBWRs technical support document. The circular plane model consists of a concrete source layer of thickness of 0.0127 m (corresponding to a 0.5-inch depth) covering a non-radioactive concrete backing of 1 m. Based on the volumetric core samples the geometry depth was increased to 0.0254 m (1-inch depth) for the judgmental measurement location. Source to detector distance varied from one measurement location to another; this distance ranged from 1.3 m to 3.1 m.

Figure C.1. ISOCS Circular Plan Template (Canberra 2009)

LACBWR Rx Bowl C-3 5299-SR-04-0 C.4 RADIOLOGICAL SAMPLE ANALYSIS C.4.1 Gamma Spectroscopy Samples were mixed, crushed, and/or homogenized as necessary, and a portion sealed in a 0.06-liter container for analysis. The quantity placed in the beaker was chosen to reproduce the calibrated counting geometry. Net material weights were determined and the samples counted using intrinsic, high purity, germanium detectors coupled to a pulse height analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. All total absorption peaks (TAPs) associated with the radionuclides of concern were reviewed for consistency of activity. Spectra were also reviewed for other identifiable TAPs. TAPs used for determining the activities of radionuclides and the typical associated MDCs for a four-hour count time are presented in Table C.1.

Table C.1. Typical MDCs Total Absorption Peak Radionuclide TAP (MeV)a MDC (pCi/g)

Co-60 1.332 0.1 Cs-137 0.662 0.1 Eu-152 0.344 0.3 Eu-154 0.723 0.5 aMeV = mega electron volt C.4.2 Ni-63 Analysis Soil samples were spiked with a nickel and cobalt carrier and digested with a mixture of nitric and hydrochloric acids. Unwanted elements, such as iron and cobalt, are then removed by running the slurry via anion exchange chromatography. Nickel is then separated from the slurry using a nickel selective resin cartridge. The purified nickel is then eluted off of the column with a dilute nitric acid solution. Ni-63 activity is then determined via liquid scintillation counting. The typical MDC for a 60-minute count time using this procedure is 0.8 pCi/g.

C.4.3 Radioactive Strontium Analysis Sr-90 concentrations were quantified by total sample dissolution followed by radiochemical separation and counted on a low background proportional counter. Samples were homogenized and dissolved by a combination of potassium hydrogen fluoride and pyrosulfate fusions. The fusion cakes were dissolved, and strontium was co-precipitated on lead sulfate. The sulfate-salt complex was dissolved in EDTA at a pH of 8.0. The strontium was separated from residual calcium and lead by reprecipitating strontium sulfate from EDTA at a pH of 4.0. Strontium was separated from barium by complexing the strontium in DTPA while precipitating barium as barium chromate. The

LACBWR Rx Bowl C-4 5299-SR-04-0 strontium was ultimately converted to strontium carbonate and counted on a low-background gas proportional counter. The typical MDC for a 60-minute count time using this procedure is 0.4 pCi/g.

C.4.4 H-3 Analysis Tritium (H-3) analyses were performed using a material oxidizer and counted by liquid scintillation.

The Material Oxidizer combusts samples in a stream of oxygen gas and passes the products (including CO2 and H2O vapor) through a series of catalysts. The H-3 is carried by water and is captured in a trapping scintillation cocktail specific to water. The typical MDC for H-3 for a 60-minute count time using this procedure is 3-5 pCi/g.

LACBWR Rx Bowl 5299-SR-04-0 APPENDIX D: MAJOR INSTRUMENTATION

LACBWR Rx Bowl D-1 5299-SR-04-0 The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the author or his employer.

D.1 SCANNING AND MEASUREMENT INSTRUMENT/DETECTOR COMBINATIONS D.1.1 Gamma Ludlum NaI Scintillation Detector Model 44-10, Crystal: 5.1 cm x 5.1 cm coupled to: Ludlum Ratemeter-scaler Model 2221 coupled to: Trimble Geo 7X High Purity, Broad Energy Germanium Detector Canberra Model No. BE3825 Used in conjunction with:

Canberra Inspector 2000 multi-channel analyzer, Canberra In-Situ Object Counting System and Genie 2000 software, Canberra 50 mm, 90-degree FOV lead collimator, and Dell laptop (Canberra, Meriden, CT)

D.2 LABORATORY ANALYTICAL INSTRUMENTATION High-Purity, Extended Range Intrinsic Detector Canberra/Tennelec Model No: ERVDS30-25195 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, Connecticut)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Dell Workstation (Canberra, Meriden, CT)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. GMX-45200-5 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, CT)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Dell Workstation (Canberra, Meriden, CT)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. GMX-30P4 Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software

LACBWR Rx Bowl D-2 5299-SR-04-0 (Canberra, Meriden, CT)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Dell Workstation (Canberra, Meriden, CT)

High-Purity, Intrinsic Detector EG&G ORTEC Model No. CDG-SV-76/GEM-MX5970-S Canberra Lynx Multichannel Analyzer Canberra Gamma-Apex Software (Canberra, Meriden, CT)

Used in conjunction with:

Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Dell Workstation (Canberra, Meriden, CT)

Low-Background Gas Proportional Counter Series 5 XLB (Canberra, Meriden, CT)

Used in conjunction with:

Eclipse Software Dell Workstation (Canberra, Meriden, CT)

Liquid Scintillation Analyzer Perkin Elmer Model Tri-Carb 5100 TR (Perkin Elmer, Shelton, CT)

Used in conjunction with:

Quantamart Software Perkin Elmer, Shelton, CT)