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6 Measurement Issues The quantitative determination of the identity and activity of radionuclides present in a sample is a process that ranges from straightforward to complex, depending on the radionuclides, their distribution on or within the sample, the instrumentation available, the material matrix, and the pattern of radionuclide distribution within the matrix. Many radionuclides that emit gamma photons are relatively easy to identify and quantify. Most radionuclides that decay only by particle emission can be detected if they are on the surface of a solid material, but identification of the specific radionuclides present is often difficult or complex. Further, when particle-emitting radionuclides are distributed through the volume of a solid material, determining the amount of a radionuclides present can require sophisticated technology beyond simple survey instruments. Dose cannot be measured directly. Instead, the dose received is estimated by first determining activities for the radionuclides to be released (identity and quan- tity of each radionuclides then using a factor to convert from activity to dose. Specifically, a screening level of activity is set by two quantities, the primary dose standard and the dose factor that relates the secondary activity standard (or screening level) to the primary dose standard, as discussed in Box 5-1. The dose factors, which are derived by modeling, vary by radionuclide and by the expert group that computed them. The relationship between source concentration and dose is affected by many factors, including but not limited to the following: . The magnitudes of the dose factors chosen to derive screening levels from the primary dose standards; The specific instrumentation used in measuring radioactive material con- centrations in a source: 115

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116 THE DISPOSITION DILEMMA The counting conditions, including background radiation levels; The sample's physical and chemical characteristics; The inventory (identity and quantity) of the radionuclides present; and The nonradioactive material present. NUREG-15()7, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions (USNRC, 1997) discusses each of these factors in detail, including the factor's impact on the minimum detectable concentration (MDC). The MDC is defined in NUREG- 1507 as "the minimum activity concentration on a surface or within a material volume, that an instrument is expected to detect (e.g., activity expected to be detected with 95% confidence)" (USNRC, 1997, p. 3-1~. This discussion assumes that (1) the concentration of any radionuclides in samples to be measured is low relative to licensed levels and (2) the dose received by individuals from contact with these materials after their release is a small fraction of the natural background doses. As the activity in the sample increases, detecting, identifying, and quantifying the radiation source or sources become easier.) When clearance for materials is considered, the process starts with an assay of a sample having an unknown inventory of radionuclides. The instrument se- lected to perform the assay will depend on the type of radiation that may be present. The Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (EPA et al., 2000) specifies a methodology, which is discussed later in this chapter, for accomplishing a statistically valid assay of radioactivity in potentially clearable material. It also provides guidance on instrument selection. NUREG-1507 provides detailed information on instrument capabilities (USNRC, 1997~. Instrument selection is straightforward when it is known which radionu- clides could be present. An example would be a medical licensee that uses only three radionuclides. However, if the licensee operates a reactor where a large number of radionuclides are present and neutron activation of materials is a possibility, instrument selection may be more complex. A series of measure- ments may be required, using different instruments, each of which can detect a different radiation type. Each measurement will yield a number of counts ob- tained in a counting period. The counts per unit time are converted to units of radioactivity, using the known properties of the detector and the geometry of the configuration for counting (see Appendix E). An important issue is whether one or more radionuclides may be present. Each radionuclide has its own activity, which in most circumstances will differ from the activities of other radionuclides present in the sample. However, as the "Appendix E of this report provides tutorial-level information on radiation, radioactivity, and radiation detection.

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MEASUREMENT ISSUES 117 number of radionuclides present increases, it becomes increasingly likely that the radiation from one will mask (be sufficiently close in energy to) the radiation from another, complicating the process of identifying and quantifying them. Detection limits for both field survey instruments and laboratory instruments play a critical role in selecting the instrumentation and measurement procedures used in the analysis. Background radiation from naturally occurring radionu- clides and cosmic radiation influence the sensitivity of the measurement process. As discussed in Appendix E and NUREG-1507, a detection limit in effect repre- sents a practical trade-off between the acceptable statistical chances of obtaining a false positive or a false negative indication of the presence of radioactive material. LEVELS OF DETECTABILITY A reasonable question to ask is whether a radionuclide can be measured at the concentrations corresponding to (i.e., derived from) proposed primary stan- dards. The Environmental Protection Agency's (EPA's) Technical Support Docu- ment 97 ("TSD 97") presents MDC data derived from 24 laboratories (EPA, 1997a). The authors of TSD 97 recognized that increasing the count time or sample size could lower the reported MDC, but they concluded that the values reported represented the state of the art at the time (1995) for practical measure- ments. For most radionuclides, the background count rates were less than one count per minute and the lower limits of the detectors were less than 0.037 Bq/g (1 pCi/g). A variety of instruments were used, depending on the radionuclide. Count times ranged from 20 to 1,000 minutes. Sample masses ranged from 0.1 to 750 grams. A review of the dose factor data illustrates the wide range of screening levels for volume contamination (picocuries per gram) obtained from different reports for the same radionuclide. Table 6-1 presents the screening levels for selected radionuclides from three reports, based on a 1 mrem/yr primary dose standard. In the two right-hand columns are the volumetric MDCs from TSD 97. Despite the variations, these derived (secondary) screening levels2 are all greater than the lower MDC from TSD 97, except for the i29I dose factor for NUREG- 1640. Even this screening level could probably be detected if longer counting times were used to lower the MDC. Thus, under practical measurement conditions, existing measurement capabilities are sufficiently sensitive to meet almost all of the de- 2Derived (secondary) screening levels (i.e., secondary dose standards) can be derived by dividing the primary standard (in units of microsieverts per year) by the highest dose, from the most critical scenario, per year per becquerel per gram for volume sources, or by the highest dose per year per becquerel per square centimeter for surface-contaminated sources (see Box 5-1).

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118 THE DISPOSITION DILEMMA TABLE 6-1 Comparison of Derived Screening Levels and Laboratory Minimum Detectable Concentrations (MDCs) for Selected Radionuclides ~pci/g~a Derived Screening Level MDC USNRC IAEA TECDOC 855 EPA TSD 97 Values Table 1.6b Table ~ gb ANSI/HPS NUREG- 1640 Radionuclide 13.12-1999 Table 2.6 Low High Low High 137Cs 30 1 5.4 2,432 0.007 0.3 60Co 30 1 13.5 2,432 0.01 0.3 63Ni 3,000 27,000 2.15 x 105 2.7 x 107 1 100 1291 300 0.1 270 21,000 0.4 2 14C 3,000 17 2,700 1.9 x 105 0.2 37 239Pu 3 1.2 2.16 18,000 0.02 0.4 99Tc 3,000 2.3 1,100 1.6 x 106 0.3 15 230Th 3 1.2 2.7 216 0.05 0.5 NOTE: ANSI/HPS = American National Standards Institute and Health Physics Society; IAEA = International Atomic Energy Agency; USNRC = U.S. Nuclear Regulatory Commission. a Based on 1 mrem/yr. bLow and high indicate the extremes of the screening level range presented in the reference. rived (secondary) screening levels for volume contamination derived in the tech- nical analyses reviewed by the committee. TSD 97 also evaluated the detectability of surface contamination and reached a similar conclusion. Namely, existing measurement capabilities for surface con- tamination are sufficiently sensitive to reach the screening levels for surface contamination derived in these same technical analyses. Although the Health Physics Society (HPS) Standards Working Group evaluated a different set of instruments and measurement procedures for the American National Standards Institute (ANSI)-Health Physics Society Standard N13.12-1999, the conclusion about detectability at the derived activity levels was the same (ANSI/HPS, 1999, Sections B.4 and B.5~: . . . in most cases the minimum detectable activities were significantly lower than the derived screening levels. These results indicate that, with a careful selection of alpha and gamma spectroscopy instruments and methods, it should be possible to attain a minimum detectable activity lower than the screening levels for most groups of radionuclides identified in this standard. The ANSI/HPS report uses the term "minimum detectable activity" instead of MDC.

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MEASUREMENT ISSUES TABLE 6-2 Detectability of Selected Radionuclides by Laboratory Analysis Relative to Derived Screening Level (DSL) from TSD 97 (pCi/g~a 119 MDC DSL Detectable at All Radionuclide Low High 15 mrem/yr 1 mrem/yr 0.1 mrem/yr Levels? 137Cs 0.007 0~3 170 11 1.1 Yes 60Co 0.01 0.3 17 1.1 0.11 Yes 63Ni 1 100 1.4 x 106 93,000 9,300 Yes 91 0.4 2 19 1.3 0.13 No 14C 0.2 37 17,000 1,200 12 Yes 239pu 0.02 0.4 21 1.4 0.14 Yes 99Tc 0.3 15 700,000 46,000 4,600 Yes 230Th 0.05 0~5 23 1.6 0.16 Yes aLow and high represent the extremes of the derived screen levels in this reference. SOURCE: EPA (1997a, Table 8.9). Table 6-2 compares the MDCs from TSD 97 with the derived screening levels from TSD97 for volumetric contamination corresponding to primary dose standards of 15 mrem/yr, 1 mrem/yr, and 0.1 mrem/yr. The scenario used to derive the screening levels was the normalized dose to individuals exposed to radiation as the result of recycling scrap metal from nuclear facilities. Again, the MDCs are lower than the screening levels in all cases except for i29I at the 0.1 mrem/yr primary dose limit. TSD 97 reports similar results for surface-contaminated materials, when large-area detectors are used for surface scans (EPA, 1997a, Table 8-6~. For large-area detectors used in the scan mode with a distributed source, the TSD 97 analysis concludes that in the laboratory, all 40 radionuclides considered would be detectable at the surface contamination screening levels (in units of disintegra- tions per minute per 100 cm2) derived from a primary dose limit of 1 mrem/yr. These results assume a scanning rate of one-third of the detector width per second for beta and alpha detection and 15 cm/s for gamma detection. For small-area detectors, which TSD 97 assumes would be used in field conditions, detectability becomes more difficult when factors such as human error, small nonhomogeneous contamination areas, realistic distances from source to detector, the condition of the material's surface, and surface coating are in- cluded. The fraction of radionuclides detectable under field conditions at the derived screening levels decreases from 39 of 40 for a primary dose limit of 15 mrem/yr to 31 of 40 for 1 mrem/yr and only 11 of 40 for 0.1 mrem/yr. Whenever the potential exists for the presence of radionuclides that are not detectable with the detection method being used for the survey, it is necessary to

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120 THE DISPOSITION DILEMMA change or modify the method to increase the sensitivity of the measurement by lowering the scan rate, changing to a larger area detector, or changing from a field measurement to a laboratory measurement. The conclusion of TSD 97 is that at levels corresponding to the screening levels utilized in that study of 15 mrem/yr and 1 mrem/yr, "100% of the radionuclides evaluated can be detected." Even at screening levels corresponding to 0.1 mrem/yr, "85% of the radionuclides are detectable" (EPA, 1997a, p. ES-17. Thus, for both volume-contaminated and surface-contaminated solid materi- als, measurement of radionuclide activity concentrations at levels being consid- ered for dose-based standards is not the limiting factor if the primary dose stan- dard is at or above 1 mrem/yr in both laboratory and field measurements. MEASUREMENT COST The cost of measuring activities at these levels depends on the difficulty of analysis. The instrumentation to perform alpha, beta, and gamma spectroscopy is similar in cost to the most sophisticated systems for chemical analysis. Alpha and beta spectrometers cost approximately $50,000 each, but many systems can be adapted to analyze either particle by changing the detector. Gamma spectroscopy systems range from $50,000 to $200,000. A reasonable cost to set up a state-of- the-practice radionuclide analysis laboratory would be less than half a million dollars. The major operating expense is for the trained personnel needed to per- form the sample preparation analyses correctly, especially on difficult samples. The TSD 97 authors referenced an article by Cox and Guenther (1995) that presented a range of MDCs as reported by 24 commercial and governmental laboratories. Table 8-5 of TSD 97 presents detection costs in 1995 dollars per sample that range from $40 to $375, depending on the radionuclide. There is some increase in per-sample cost as the required sensitivity increases: activities in the 10 pCi/g range, cost $40 to $250 per sample; in the 1 pCi/g range, $75 to $300 per sample; and in the 0.1 pCi/g range, $100 to $375 per sample. However, the increase is not as large as would be expected if most laboratories offering detection services were not already working with instruments and measurement procedures adequate to detect activities at the 0.1 pCi/g level. If clearance is an option, the tradeoff between the cost of clearance and the cost of disposal as low-level radioactive waste (LLRW) will ultimately determine which option a licensee chooses. Chapter 4 estimates that costs for LLRW dis- posal will range from $3,120 to $16,800 per cubic meter. LLRW densities in the United States are usually between 50 and 120 pounds per cubic foot (0.8 to 1.92 metric tons/m3~. If a nominal density of 75 pounds per cubic foot, disposal costs of $30 per metric ton and $110 per metric ton at Subtitle D and C landfills, respectively, and a fixed sampling cost of $20 per sample (collection and prepa- ration) are assumed, one can estimate the number of samples that can be taken at the break-even cost relative to LLRW disposal. Table 6-3 presents the results of

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MEASUREMENT ISSUES 12 TABLE 6-3 Estimated Number of Analyzed Samples per Metric Ton of Waste at Breakeven Between Clearance and Low-Level Radioactive Waste Disposal LLRW Disposal ($2,590 per metric ton) Number of Samples Analyzed at LLRW Disposal ($13,950 per metric ton) Number of Samples Analyzed at $40 $375 $40 $375 Alternative Disposal site per sample per sample per sample per sample Subtitle D at $30 42 6 232 35 per metric ton Subtitle c at $110 41 6 232 35 per metric ton NOTE: Calculation assumes no difference in transportation costs and constant sampling costs of $20 per sample. this estimation.3 The number of samples required to characterize the waste stream adequately will depend on the degree of certainty that the waste is homogeneous. However, at the higher LLRW cost, the number of samples that could be taken for the same cost ranges from 35 to 232, which is large enough to characterize a homogenous ton of waste. If the lower cost of LLRW disposal and the high sample analysis cost are used in the estimation, the six samples at equivalent cost are probably too small for adequate sampling, unless the waste is known to be homogeneous. Depending on the waste stream and the sampling protocol, it may be possible to aggregate samples and resample. This approach would reduce the typical number of samples to be analyzed per ton of waste. Thus, the cost of sampling and analysis by itself does not appear to be a limiting factor when selecting a primary dose standard at or above 0.1 mrem/yr. (However, as noted above, at screening levels corresponding to a primary dose standard of 0.1 mrem/yr, the detection capability of field instruments is such that only 11 of 40 key radioncuclides can be detected.) This conclusion on costs is confirmed by the operation of a commercial waste management service, Duratek, Inc., which uses the derived screening levels from ANSI/HPS Standard N13.12 (see Table 6-1) to make decisions on waste disposition. Duratek, Inc. provided 3For example, if you had one ton of waste and access to LLRW disposal at $2,590 per ton, one option is to send that ton of waste to such an LLRW disposal facility. On the other hand, to send the same ton of waste to a Subtitle D facility, you would have to sample sufficiently to show it meets clearance levels and do so within a budget of $2,590-$30 = $2,560, the amount left after $30 tipping fee per ton. At $40 per sample characterization plus $20 per sample for sampling, this allows 42 samples to be taken within the break-even budget. If more samples are needed to show the waste meets clearance criteria, it is cheaper to send it to LLRW disposal.

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22 THE DISPOSITION DILEMMA the committee with information on its process and procedures, as discussed in the next section. CURRENT MEASUREMENT PRACTICES OF A WASTE BROKER Radioactive waste is generated daily from hospitals, research laboratories, and nuclear power plants. Licensees that generate controlled materials during operations currently survey all potentially contaminated waste materials prior to shipment. Those that are determined to have no licensee-generated radioactive material present are treated as nonradioactive waste. Materials that have surface contamination are either treated as LLRW or cleared using the criteria in Regula- tory Guide 1.86 (AEC, 1974), license conditions, or approval obtained on a case- by-case basis from either the USNRC or the agreement state regulator. During decommissioning, potentially radioactive materials are typically cleared on a case-by-case basis or sent to a waste processor for clearance. Known radioactive materials are disposed as appropriate for their radioactive waste classification. In 2000, about 30,000 tons of LLRW were processed in the United States. Waste brokers and processors handle a significant fraction of this waste. Waste brokers provide services to direct the disposition of LLRW and to prevent the release of contaminated materials into general commerce. A broker may trans- port, collect, or consolidate shipments or process radioactive waste. The survey of the incoming waste stream is an essential step in a waste processor's manage- ment of customer materials. The incoming shipment is scanned with handheld counters as an initial screen. (The licensee shipping the material has already certified that the waste has a low activity level and can be evaluated for clear- ance.) The material is then examined in either a box or a drum assay system. At the facilities of the waste broker Duratek, Inc., high-purity germanium (HPGe) detectors are employed for gamma spectroscopy, sodium iodide detectors are used for micro-dose rate determinations, and the records for each assay are stored digitally. If the material is clean (no activity at or above detectable limits), it is shipped to a Subtitle D landfill. As a further check, portal monitors at the facility exits are used to ensure that "clean" material shipped to the local Subtitle D landfill will not trigger portal monitors upon arrival there. If the material is contaminated at levels above those that would allow landfill disposal, it is either returned to the generator or, at the direction of the generator, disposed of as LLRW. Prior to disposal as LLRW, material is processed by melting, compac- tion, incineration, or a combination of these processes, to reduce its volume (which reduces disposal costs). THE MARSSIM METHODOLOGY Determination of an appropriate sampling program is a major consideration in the measurement process. MARSSIM methodology could be a valuable tool

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MEASUREMENT ISSUES 123 for licensees in demonstrating compliance with the type of dose-based standards under consideration for releasing SRSM. The MARSSIM includes a statistical sampling methodology suitable for release of land and buildings potentially con- taining residual radioactive material in surface soil or on building surfaces. At some licensed facilities, potentially clearable building materials may contain volume-distributed sources of radioactivity, in addition to surface sources. The MARS SIM methodology could also be expanded to be used as a decision tool in evaluating these solid materials. The number of measurements or samples needed in each survey unit for statistical testing of residual radioactive material against a release level depends on the expected variability in concentration of the radioactive material and the level of acceptable error. If a licensee is in doubt, MARSSIM encourages assum- ing a larger, rather than smaller, variability in the material. This conservative approach (presumption of less homogeneity) drives a MARSSIM-guided assess- ment toward taking a larger number of measurements or samples. A plethora of radiation detection instruments is available to scan surfaces and make direct measurements of residual radioactivity. The radionuclide~s) present and the magnitude of the release level are key factors in determining the appropriate instrument for a particular slightly radioactive solid material to be assessed. Several references, including MARSSIM (EPA et al., 2000), and NUREG 1507 (USNRC, 1997), provide MDCs for various types of radiation detection instruments under different sets of circumstances. The characteristics of the detector (probe area, detection efficiency, background response, etc.) en- able the licensee to relate the release level to a corresponding instrument re- sponse, which MARSSIM calls the Derived Concentration Guideline Level (DCGL). The instrument selected should have sensitivity as far below the DCGL as possible. MARSSIM recommends that the MDC should be less than 10 per- cent of the DCGL, although it is acceptable for the MDC to be as much as 50 percent of the DCGL. Having selected appropriate instrumentation, the licensee must next develop an integrated survey design combining some degree of scanning surveys with static measurements or sample collection. MARSSIM strongly recommends that the effort expended be weighted toward those survey units4 more likely to contain elevated levels of residual radioactive material. 4A geographical area consisting of structures or land areas of specified size and shape at a remediated site for which a separate decision will be made whether the unit attains the site- specific reference-based cleanup standard for the designated pollution parameter. Survey units are generally formed by grouping contiguous site areas with a similar use history and the same classification of contamination potential. Survey units are established to facilitate the survey process and the statistical analysis of survey data (EPA et al., 2000).

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124 THE DISPOSITION DILEMMA The assessment phase, which follows collection of the survey data, includes data validation as well as a reassessment of the quantity of data. For example, the number of measurements taken was based, in part, on an assumption about the variability of radionuclide concentration in the material. This assumption should be verified. If the variability was underestimated, more data should be collected to ensure that the desired statistical significance is attained. For survey units that are likely to contain elevated levels of radioactivity, MARSSIM also requires that an elevated measurement comparison (EMC) test be performed to demonstrate compliance for small areas with elevated activity concentrations. FINDINGS Finding 6.1. The concentration of radioactive material in released solids directly affects radiation detection requirements and costs. Measurement of the amount of radioactive material in a solid matrix is a complex task that involves a combina- tion of instrument characteristics, background radiation levels, and source char- acteristics. No single measurement method would be appropriate or adequate for all radionuclides. Finding 6.2. The overall measurement costs, including sampling (collection and preparation) and analysis and material disposition choices, affect clearance deci- sions. If the measurement costs are too high, it may be more cost-effective to dispose of the material as low-level radioactive waste. Finding 6.3. For a 1 mrem/yr or higher standard (and the corresponding derived secondary screening levels), the majority of radionuclides can be detected at reasonable costs in a laboratory setting, under most practical conditions. For a 0.1 mrem/yr standard, the measurement capability falls below the upper bound of minimum detectable concentrations for some radionuclides in some laboratories, although 85 percent of radionuclides are still detectable. Using field measure- ments, a more rapid fall-off of detectability is observed at more stringent radia- tion protection levels, with 31 of 40 key radionuclides detectable at 1 mrem/yr and 11 of 40 detectable at 0.1 mrem/yr.