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{{#Wiki_filter:The term Unimportant Quantities will be removed from the Decommissioning Plan. | {{#Wiki_filter:The term Unimportant Quantities will be removed from the Decommissioning Plan. | ||
The response is presented in the attached document Technical Basis for Demolition Debris Release Veri"cation. | |||
The response is | TECHNICAL BASIS FOR DEMOLITION DEBRIS RELEASE VERIFICATION University of Missouri Pickard Hall 405 S. Ninth Street Columbia, MO 65211-1420 Revision 1 May 19, 2024 Prepared by: | ||
Chase Environmental Group, Inc. | |||
200 Sam Rayburn Parkway Lenoir City, TN 37771 865-816-6015 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page i of i TABLE OF CONTENTS | |||
==1.0 INTRODUCTION== | |||
............................................................................................................. 1 2.0 RELEASE CRITERIA..................................................................................................... 2 3.0 DOSE MODELING METHODS..................................................................................... 3 3.1 Radionuclide Inputs......................................................................................................... 3 3.2 Geometry and Dose Points.............................................................................................. 5 4.0 DOSE MODELING RESULTS....................................................................................... 5 4.1 MicroRem Detector Response......................................................................................... 5 4.2 NaI Detector Response.................................................................................................... 5 5.0 SENSITIVITY ANALYSIS.............................................................................................. 6 5.1 Dose Point Geometry...................................................................................................... 6 5.2 Density............................................................................................................................. 7 5.1 Source Term Heterogeneity............................................................................................. 9 6.0 DOSE MODELING RESULTS | |||
==SUMMARY== | |||
................................................................. 9 7.0 SURVEY METHODS..................................................................................................... 10 TABLES Table 2-1: Radionuclide Concentrations at Unity........................................................................... 2 Table 3-1: Ra-226+C MicroShield Inputs....................................................................................... 3 Table 3-2: Th-232+C MicroShield Inputs....................................................................................... 4 Table 3-3: U Tailings MicroShield Inputs....................................................................................... 4 Table 4-1: Response at Unity (3 cm from Surface)......................................................................... 5 Table 4-2: Response at Unity (30 cm from Surface)....................................................................... 6 Table 4-3: Response at Unity (100 cm from Surface)..................................................................... 6 Table 5-1: Response at Unity (Center of Side of Dump Bed)......................................................... 6 Table 5-2: Response at Unity (1/4 Length Side of Dump Bed)...................................................... 7 Table 5-3: Response at Unity (Center Rear of Dump Bed)............................................................. 7 Table 5-4: Response at Unity (3 cm, 88 lb/ft3)................................................................................ 7 Table 5-5: Response at Unity (30 cm, 88 lb/ft3).............................................................................. 7 Table 5-6: Response at Unity (100 cm, 88 lb/ft3)............................................................................ 8 Table 5-7: Response at Unity (3 cm, 132 lb/ft3).............................................................................. 8 Table 5-8: Response at Unity (30 cm, 132 lb/ft3)............................................................................ 8 Table 5-9: Response at Unity (100 cm, 132 lb/ft3).......................................................................... 8 Table 6-1: 2 x 2 NaI Detector Response at Various Dose Points................................................ 9 Table 6-2: 2 x 2 NaI Detector Response at Various Densities..................................................... 9 Table 6-3: MicroRem Detector Response at Various Dose Points.................................................. 9 Table 6-4: MicroRem Detector Response at Various Densities...................................................... 9 APPENDICES Appendix A - MicroShieldTM Output Reports for Base Case Appendix B - Base Case Tables Appendix C - MicroShieldTM Output Reports for Geometry Analysis Appendix D - Geometry Analysis Tables Appendix E - MicroShieldTM Output Reports for Density Analysis Appendix F - Density Analysis Tables | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 1 of 10 | |||
==1.0 INTRODUCTION== | |||
Debris from demolition of Pickard Hall building structures above the basement meeting the release criteria will be released for unrestricted use and disposed as non-radioactive waste. The debris from the three brick ducts with elevated radioactivity will be encapsulated, marked for identification during demolition, and segregated to the extent practical. Building structures with significant surface contamination that have been encapsulated including the brick ducts discussed above and all structures in the basement will be disposed as radioactive waste consistent with the Decommissioning Plan. Debris with concentrations below the release criteria will be disposed at the Columbia City Landfill, while demolition debris with concentrations above the release criteria will be disposed at the WCS site in Andrews, TX. | |||
The release criteria for demolition debris are conservatively the same as the derived concentration guideline levels (DCGLs) for soils that will be left in place after post-demolition soil remediation. | |||
Calculations based on extensive characterization data collected with the building intact demonstrate that demolition debris from building structures above the basement will be a small fraction of the DCGLs. The estimated average activity concentration in demolition debris above the basement (prior to additional remediation of structures that will occur before demolition) is 0.055 pCi/g; this is less than 0.4% of the release criteria. However, there may be unidentified locations with residual radioactivity that hasnt been accounted for by conservatisms of the radioactivity estimates. The severity of any potential unidentified locations of activity are bounded to a low concentration by extensive 2 x 2 NaI measurements conducted during characterization. | |||
In addition to demonstrating the debris will meet the release criteria based on characterization data, MU will conservatively perform a two-step verification process of loaded debris consisting of 1) manual surveys and 2) passing loaded trucks through a truck radiation monitoring system prior to releasing material for transport to the City landfill. | |||
The gamma exposure rates associated with various debris configurations in a dump truck were modeled using MicroShieldTM software to determine the response for a 2 x 2 NaI detector and for a Microrem meter. For the base case model, a typical dump bed geometry and a typical density of demolition rubble were used. Sensitivity analysis was performed for the geometry of the dose points and for the density of debris. All models assume a uniform radioactivity concentration. | |||
This technical basis is intended to document the methods used to correlate debris concentrations to detector responses for verification surveys. Once a demolition contractor has been selected, the actual geometries of the dump trucks will be modeled in this same manner for project use if there are significant differences from the model assumptions. | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 2 of 10 2.0 RELEASE CRITERIA The soil DCGLs are conservatively used for the release of demolition debris for unrestricted release and disposal at the City landfill. Because multiple radionuclides are present, unity is required; therefore, a gross measurement DCGL is calculated using Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) Equation 4-4: | |||
= | |||
1 226 226 + | |||
232 232 + | |||
238 238 Where: | |||
fRa-226 = Ra-226 fraction fTh-232 = Th-232 fraction fU-238 = U-238 fraction DCGLRa-226 = Ra-226 DCGL (pCi/g) | |||
DCGLTh-232 = Th-232 DCGL (pCi/g) | |||
DCGLU-238 = U-238 DCGL (pCi/g) | |||
The calculation is provided below with appropriate substitutions: | |||
= | |||
1 0.888 15 | |||
+ 0.086 12 | |||
+ 0.026 371 | |||
= 15 / | |||
The concentration of each nuclide making up the gross DCGL is calculated below: | |||
Table 2-1: Radionuclide Concentrations at Unity Nuclide Gross DCGL (pCi/g Ra-226 + | |||
Th-232+U-238) | |||
Fraction of Total Activity Concentration at Unity (pCi/g) | |||
Ra-226 15 0.888 13.3 Th-232 0.086 1.3 U-238 0.026 0.4 | |||
University of Missouri | University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 3 of 10 3.0 DOSE MODELING METHODS The gamma emissions from demolition debris within a full dump truck load were modeled using MicroShieldTM at various dose points to estimate the detector response for verification surveys. This is accomplished in three steps: | ||
: 1. Model the exposure rate from a dump truck of debris at a unit concentration of 1 pCi/g of the parent nuclide (Ra-226, Th-232, and U-238) using MicroShieldTM at distances of 3 cm, 30 cm, and 100 cm. 1 pCi/g is converted to µCi/cc by multiplying by the density in g/cc and dividing by 1E6 pCi/µCi. | |||
: 2. Correlate the exposure rate to the instrument response for each nuclide decay chain. | |||
: 3. Determine the detector response to the sum of the nuclides using the nuclide ratios. | |||
3.1 Radionuclide Inputs The unit concentrations of radionuclides were converted to µCi/cc using the fractions of the parent nuclides and a density of 1.76 g/cc (110 lb/ft3) for input into MicroShieldTM as presented in the tables below. | |||
Table 3-1: Ra-226+C MicroShield Inputs Nuclide Fraction of Ra-226 Activity Input Value (pCi/g) | |||
Density 110 lb/ ft3 (g/cc) | |||
MicroShieldTM Input | |||
(µCi/cc) | |||
Ra-226 1 | |||
1 1.76 1.76E-06 Rn-222 1 | |||
1 1.76 1.76E-06 Po-218 1 | |||
1 1.76 1.76E-06 Pb-214 1 | |||
1 1.76 1.76E-06 Bi-214 1 | |||
1 1.76 1.76E-06 Po-214 1 | |||
1 1.76 1.76E-06 Pb-210 1 | |||
1 1.76 1.76E-06 Bi-210 1 | |||
1 1.76 1.76E-06 Po-210 1 | |||
1 1.76 1.76E-06 | |||
Revision 1 May 19, 2024 | University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 4 of 10 Table 3-2: Th-232+C MicroShield Inputs Nuclide Fraction of Th-232 Activity Input Value (pCi/g) | ||
Density 110 lb/ ft3 (g/cc) | |||
MicroShieldTM Input | |||
(µCi/cc) | |||
Th-232 1 | |||
1 1.76 1.76E-06 Ra-228 1 | |||
1 1.76 1.76E-06 Ac-228 1 | |||
1 1.76 1.76E-06 Th-228 1 | |||
1 1.76 1.76E-06 Ra-224 1 | |||
1 1.76 1.76E-06 Rn-220 1 | |||
1 1.76 1.76E-06 Po-216 1 | |||
1 1.76 1.76E-06 Pb-212 1 | |||
1 1.76 1.76E-06 Bi-212 0.64 0.64 1.76 1.13E-06 Bi-212 0.36 0.36 1.76 6.34E-07 Po-212 0.64 0.64 1.76 1.13E-06 Tl-208 0.36 0.36 1.76 6.34E-07 Table 3-3: U Tailings MicroShield Inputs Nuclide Fraction of U-238 Activity Input Value (pCi/g) | |||
Density 110 lb/ ft3 (g/cc) | |||
MicroShieldTM Input | |||
(µCi/cc) | |||
U-238 1 | |||
1 1.76 1.76E-06 Th-234 1 | |||
1 1.76 1.76E-06 Pa-234m 1 | |||
1 1.76 1.76E-06 U-234 1.02 1.02 1.76 1.80E-06 Th-230 1.02 1.02 1.76 1.80E-06 U-235 0.048 0.048 1.76 8.45E-08 Th-231 0.048 0.048 1.76 8.45E-08 Pa-231 0.048 0.048 1.76 8.45E-08 Ac-227 0.048 0.048 1.76 8.45E-08 Th-227 0.048 0.048 1.76 8.45E-08 Ra-223 0.048 0.048 1.76 8.45E-08 Rn-219 0.048 0.048 1.76 8.45E-08 Po-215 0.048 0.048 1.76 8.45E-08 Pb-211 0.048 0.048 1.76 8.45E-08 Bi-211 0.048 0.048 1.76 8.45E-08 Tl-207 0.048 0.048 1.76 8.45E-08 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 5 of 10 3.2 Geometry and Dose Points The geometry of a typical dump truck was input to MicroShieldTM: | |||
Geometry: Rectangular Volume (Dump Bed) | |||
Dimensions of debris: 8.2 ft wide x 27.9 ft long x 10.5 ft high Shielding (dump bed thickness): 0.1875 inches steel Dose points centered on the side of the dump bed Distances from bed surface: 3 cm, 30 cm and 100 cm Density of steel: 7.86 g/cm3 Density of debris: 1.76 g/cm3 (110 lb/ft3) 4.0 DOSE MODELING RESULTS MicroShieldTM output reports for each nuclide are presented in Appendix A. | |||
4.1 MicroRem Detector Response MicroShieldTM performed the calculations and determined a total exposure rate with buildup of in mrem/hr. Because the MicroRem has a tissue equivalent crystal, no energy correction is required to convert the MicroShieldTM output to a detector response. | |||
4.2 NaI Detector Response The NaI detector is energy dependent and therefore requires energy corrections to correlate debris concentrations to the detector response. MicroShieldTM provides the exposure rates for a number of gamma energies associated with the source term inputs. This data was used to convert the exposure rate at each energy to counts per minute (cpm) specific to the source term. These calculations are presented in the tables in Appendix B and are summarized in the tables below. Each table presents the detector responses to radionuclide concentrations at the gross DCGL. | |||
Table 4-1: Response at Unity (3 cm from Surface) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 12,139 18.2 Th-232 1.3 1,560 2.7 U-238 0.4 21 0.0 Total 15 13,720 20.8 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 6 of 10 Table 4-2: Response at Unity (30 cm from Surface) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 11,421 17.1 Th-232 1.3 1,463 2.5 U-238 0.4 19 0.0 Total 15 12,904 19.6 Table 4-3: Response at Unity (100 cm from Surface) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 8,769 12.8 Th-232 1.3 1,132 1.9 U-238 0.4 16 0.0 Total 15 9,916 14.7 5.0 SENSITIVITY ANALYSIS Sensitivity analysis is performed for dose point geometry and for debris density as described below. | |||
5.1 Dose Point Geometry Additional dose points were modeled at 1/4 of the length at center height on the side of the dump bed and at the center of the rear of the dump bed to determine the sensitivity of geometry differences. The MicroShieldTM output reports are presented in Appendices A and C, and calculations are presented in the tables in Appendix D. The results are summarized in the tables below. Each table presents the detector responses to radionuclide concentrations at the gross DCGL. | |||
Table 5-1: Response at Unity (Center of Side of Dump Bed) | |||
Distance (cm) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) 3 cm 13,720 20.8 30 cm 12,904 19.6 100 cm 9,916 14.7 | |||
University of Missouri | University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 7 of 10 Table 5-2: Response at Unity (1/4 Length Side of Dump Bed) | ||
Distance (cm) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) 3 cm 13,536 20.6 30 cm 12,871 19.5 100 cm 9,545 14.1 Table 5-3: Response at Unity (Center Rear of Dump Bed) | |||
Distance (cm) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) 3 cm 10,376 17.0 30 cm 10,198 16.6 100 cm 7,192 10.9 5.2 Density The debris density was modeled at +/- 20% of the base case to determine the sensitivity of density differences. The inputs described in Section 4.0 were modified to 1.41 g/cc (88 lb/ft3) and 2.11 g/cc (132 lb/ft3). The MicroShieldTM output reports are presented in Appendix E and calculations are presented in the tables in Appendix F. The results are summarized in the tables below. Each table presents the detector responses to radionuclide concentrations at the gross DCGL. | |||
Table 5-4: Response at Unity (3 cm, 88 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 12,124 18.1 Th-232 1.3 1,558 2.7 U-238 0.4 21 0.0 Total 15 13,702 20.8 Table 5-5: Response at Unity (30 cm, 88 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 11,440 17.1 Th-232 1.3 1,465 2.5 U-238 0.4 19 0.0 Total 15 12,925 19.6 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 8 of 10 Table 5-6: Response at Unity (100 cm, 88 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 8,692 12.7 Th-232 1.3 1,121 1.9 U-238 0.4 16 0.0 Total 15 9,829 14.6 Table 5-7: Response at Unity (3 cm, 132 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 12,133 18.1 Th-232 1.3 1,455 2.7 U-238 0.4 20 0.0 Total 15 13,609 20.8 Table 5-8: Response at Unity (30 cm, 132 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 11,373 17.0 Th-232 1.3 1,455 2.5 U-238 0.4 19 0.0 Total 15 12,848 19.6 Table 5-9: Response at Unity (100 cm, 132 lb/ft3) | |||
Nuclide Concentration at Unity (pCi/g) 2 x 2 NaI (cpm) | |||
Dose Rate | |||
(µR/hr) | |||
Ra-226 13.3 8,831 12.9 Th-232 1.3 1,138 1.9 U-238 0.4 16 0.0 Total 15 9,985 14.8 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 9 of 10 5.1 Source Term Heterogeneity Heterogeneity is not a major concern because debris will be inherently mixed during demolition as materials are removed from the building and staged for loading using heavy equipment, and then loaded into dump trucks with a different piece of heavy equipment. | |||
Heterogeneity effects are expected to be minor such that the effects are compensated by conservatisms of the release criteria and verification methods; however, verification survey procedures will include assessments of heterogeneity. | |||
6.0 DOSE MODELING RESULTS | |||
==SUMMARY== | ==SUMMARY== | ||
................................................................. 9 7.0 | The results of dose modeling of demolition debris are summarized in the tables below, including the results of sensitivity analysis. The results demonstrate that the detector responses are insensitive to density changes but are mildly sensitive to the location of the dose point, particularly at the 100 cm dose point. | ||
Table 6-1: 2 x 2 NaI Detector Response at Various Dose Points Distance 2 x 2 NaI Response at Unity (cpm) 1/4 Side Center Side Center Rear 3 | |||
13,536 13,720 10,376 30 12,871 12,904 10,198 100 9,545 9,916 7,192 Table 6-2: 2 x 2 NaI Detector Response at Various Densities Distance 2 x 2 NaI Response at Unity (cpm) 88 lb/ ft3 110 lb/ ft3 132 lb/ ft3 3 | |||
13,702 13,720 13,609 30 12,925 12,904 12,848 100 9,829 9,916 9,985 Table 6-3: MicroRem Detector Response at Various Dose Points Distance Dose Rate at Unity (µR/hr) 1/4 Side Center Side Center Rear 3 | |||
20.6 20.8 17.0 30 19.5 19.6 16.6 100 14.1 14.7 10.9 Table 6-4: MicroRem Detector Response at Various Densities Distance Dose Rate at Unity (µR/hr) 88 lb/ ft3 110 lb/ ft3 132 lb/ ft3 3 | |||
20.8 20.8 20.8 30 19.6 19.6 19.6 100 14.6 14.7 14.8 | |||
University of Missouri NRC License No. 24-00513-32 Revision 1, May 19, 2024 Technical Basis for Demolition Debris Release Verification Page 10 of 10 7.0 SURVEY METHODS Step one of the verification process is a manual survey of loaded dump trucks using a 2 x 2 sodium iodide detector at a distance of 30 cm from accessible surfaces of the dump bed and comparing the detector response (above the background level) to the response associated with the release criteria. | |||
* If the detector response is greater than the response correlating to the release criteria at any location, the dump truck fails. | |||
* If the detector response (above the background level) is less than 50% of the response correlating to the release criteria at all locations, the dump truck passes. | |||
* If the detector response is less than the response correlating to the release criteria at all locations, and any location is greater than 50%, and the results of the highest and lowest locations do not vary more than a factor of three, the dump truck passes. | |||
* If the detector response is less than the response correlating to the release criteria at all locations, and any result is above 50% of the release criteria, and the results vary more than a factor of three, the Radiation Control Supervisor may evaluate the results and make a decision whether to pass or fail the dump truck. | |||
If a dump truck passes Step 1, it will then move to step 2 of the verification process. Step 2 consists of passing the truck through a truck radiation monitoring system set to alarm at the release criteria. The monitor system is expected to include detectors located at a distance of 1 meter from each side of the bed centered vertically. | |||
If a dump truck fails, the Radiation Control Supervisor may decide to reject the entire load and dispose as radioactive waste, or to dump the load and segregate the materials as radioactive waste and non-radioactive waste. | |||
Appendix A Dump Truck 110 Ra-226+C, Pages A.1 to A.4 Natural Thorium, Pages A.5 to A.8 Natural Uranium, Pages A.9 to A.12 | |||