7.
ASSAY OF RADIOACTIVE WASTE

7.1 INTRODUCTION

The determination of levels of radioactivity for the disposal of low-level radwaste has historically been done more on a semi-quantitative basis than a quantitative one. The use of gamma spectrometry and/or direct radiation measurements on bulk containers of waste has been the general approach. General difficulty in quantitative sampling and analysis is a well-known fact, and the radiation exposure costs associated with such sampling and analysis is also a factor in determining the methodology of radioactivity assay of radwaste. Nevertheless, the methods which have been used in the management of these low-level wastes have produced satisfactory results and have not led to circumstances inimical to the public health and safety.

In December 1983, the U.S. Nuclear Regulatory Commission (NRC) put into effect new standards and criteria (known as Regulation 10CFR61) governing the land disposal of radioactive wastes. The new rules defined three classes of low-level waste (designated classes A, B and C) based on the half-lives and quantities of radioactivity of specified radionuclides and assumed pathway models. The radionuclide concentration limits from 10CFR61 are reproduced in Table 7–1.(1) Some major characteristics of the nuclides needed to be analyzed are listed in Table 7–2 Many of these nuclides do not emit measurable gamma rays in their decay and can be measured only through difficult and time-consuming radiochemical and nuclear counting processes, requiring techniques and equipment beyond the capability of most nuclear power plant on-site laboratories. Even in some reputable laboratories, significant differences in the results of some comparative analyses have been reported(2).

The NRC has accepted an indirect methodology for the difficult-to-measure nuclides (i.e., nuclides which do not emit easily measured gamma rays) listed in 10CFR61. One of these involves establishing ratios or scaling factors between the difficult-to-measure nuclides and those which are both easy to measure and possess similar chemical and physical properties.



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Radiochemistry in Nuclear Power Reactors 7. ASSAY OF RADIOACTIVE WASTE 7.1 INTRODUCTION The determination of levels of radioactivity for the disposal of low-level radwaste has historically been done more on a semi-quantitative basis than a quantitative one. The use of gamma spectrometry and/or direct radiation measurements on bulk containers of waste has been the general approach. General difficulty in quantitative sampling and analysis is a well-known fact, and the radiation exposure costs associated with such sampling and analysis is also a factor in determining the methodology of radioactivity assay of radwaste. Nevertheless, the methods which have been used in the management of these low-level wastes have produced satisfactory results and have not led to circumstances inimical to the public health and safety. In December 1983, the U.S. Nuclear Regulatory Commission (NRC) put into effect new standards and criteria (known as Regulation 10CFR61) governing the land disposal of radioactive wastes. The new rules defined three classes of low-level waste (designated classes A, B and C) based on the half-lives and quantities of radioactivity of specified radionuclides and assumed pathway models. The radionuclide concentration limits from 10CFR61 are reproduced in Table 7–1.(1) Some major characteristics of the nuclides needed to be analyzed are listed in Table 7–2 Many of these nuclides do not emit measurable gamma rays in their decay and can be measured only through difficult and time-consuming radiochemical and nuclear counting processes, requiring techniques and equipment beyond the capability of most nuclear power plant on-site laboratories. Even in some reputable laboratories, significant differences in the results of some comparative analyses have been reported(2). The NRC has accepted an indirect methodology for the difficult-to-measure nuclides (i.e., nuclides which do not emit easily measured gamma rays) listed in 10CFR61. One of these involves establishing ratios or scaling factors between the difficult-to-measure nuclides and those which are both easy to measure and possess similar chemical and physical properties.

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Radiochemistry in Nuclear Power Reactors Table 7–1 10 CFR 61 WASTE CLASSIFICATION ACTIVITY LIMITS (Ref. 1) (From 10 CFR Part 61.55)   Concentration, μCi/cm3 Nuclide Class A Class Ba Class C Long Lived   C-14 0.8 NA 8 C-14 in activated metal 8.0 NA 80 Ni-59 in activated metal 22.0 NA 220 Nb-94 in activated metal 0.02 NA 0.2 Tc-99 0.3 NA 3 I-129 0.008 NA 0.08 Alpha emitting transuranics with half-lives greater than 5 years 10b NA 100b Pu-241 350b NA 3,500b Cm-242 2,000b NA 20,000b Short Lived   Total of all nuclides with half-lives less than 5 years 700 (c) (c) H-3 40 (c) (c) Co-60 700 (c) (c) Ni-63 3.5 70 700 Ni-63 in activated metal 35 700 7,000 Sr-90 0.04 150 7,000 Cs-137 1 44 4,600 a. There is no Class B category for waste exceeding the Class A limit for long lived nuclides. Such wastes are automatically Class C or are unacceptable for shallow burial. For waste containing a mixture of nuclides, the sum of the fractions of limits must be less than one with each section (long or short lived) considered separately. The appropriate limits must all be taken from the same column in the table. b. Units are nanocuries per gram. c. There are no specific limits for these nuclides in these classes. Practical considerations such as heating effect, external package radiation on shipping, or maximum specific activity determine the maximum concentration for these nuclides.

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Radiochemistry in Nuclear Power Reactors Table 7–2 NUCLEAR DATA FOR DIFFICULT-TO-MEASURE RADIONUCLIDES Isotope Half Life (years) Radiation Emitted Principal Means of Production Nuclides Origin H-3 12.3 Beta Fission; Li-7 (n,nα); B-10 (n,2α); H-2 (n,γ) Fuel & Coolant C-14 5730 Beta N-14 (n,p), 0–17 (n,α) Fuel & Coolant Fe-55 2.76 x-ray Fe-54 (n,γ) In-core surfaces Ni-63 99.5 Beta Ni-62 (n,γ) In-core surfaces Ni-59 7.5×104 x-ray Ni-58 (n,γ) In-core surfaces Sr-90 28.1 Beta Fission Fuel Tc-99 2.12×105 Beta Fission; Mo-98 (n,γ) Mo-99 (beta) Fuel I-129 1.57×107 Beta, Gamma Fission Fuel TRU* Variable Mostly alpha Multiple n-capture Fuel *TRU=Transuranic isotopes, excluding beta emitter Pu-241 and shorter-lived Cm-242

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Radiochemistry in Nuclear Power Reactors Some typical BWR and PWR nuclide concentrations in various waste streams are shown in Tables 7–3 and 7–4, respectively. Some reports dealing with the methodologies for the radionuclide assay and correlations in radwaste have been published(2–8). A brief review and discussion of those methodologies are presented in the following sections. 7.2 SAMPLING AND SAMPLE PREPARATION The primary objective of sampling is to obtain a small quantity of the waste material which is representative of the whole quantity of the waste. The nuclide concentrations measured in the sample are then extrapolated to the whole waste quantity. In practice, true homogeneity of the waste is rarely achieved in a system to be sampled. It can only be approached by the application of mixing and sampling techniques. Thus, the accuracy of radionuclide determination depends largely on the sampling technique, more than the techniques of radiochemical analysis. An excellent discussion of radwaste sampling methodology and sample size requirements can be found elsewhere(6). The mixing and sampling techniques applied to a given system depend on the property of waste material being measured. Radwaste streams can best be characterized as complex mixtures containing various components, each with different properties and concentrations. Due to these variabilities, they should be considered as heterogeneous substances from a sampling standpoint, which may vary both in time and space depending on system conditions. In order to obtain a representative sample, one should: control process conditions to approximate a homogeneous solution, utilize sampling equipment/techniques which do not bias samples, and adopt sampling techniques which approximate random sampling. Controlling process conditions to approximate a homogenous solution cannot always be accomplished in a radwaste system. However, since the most practical application of automatic sampling equipment would be in pipe systems on holding tanks, the following guidelines should be observed(6).

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Radiochemistry in Nuclear Power Reactors Table 7–3 TYPICAL BWR NUCLIDE CONCENTRATION BY WASTE STREAM (μCi/cm3) (Reproduced with Permission, EPRI NP-5983, Ref. 2) Nuclide RWCU Resins Condensate Resins Radwaste Resins Evaporator Bottoms DAWc Co-60 25.0 2.6 11.2 3.4 4.0E-3 Cs-137 14.1 1.1 6.1 1.9 1.6E-3 Ni-63 3.6E-1 2.3E-1 1.1–1 1.2E-1 1.0E-4 Fe-55 10.6 2.5 1.3 2.2 4.6E-3 C-14 3.7E-3 7.7E-3 4.6E-3 4.6E-4 2.6E-6 I-129 3.6E-5 3.7E-5 3.0E-5a 3.0E-5a 1.6E-7 Tc-99 5.6E-4 2.3E-4 2.4E-4a 7.6E-5a 1.8E-7 Sr-90 4.5E-2 4.9E-2 1.8E-6 5.6E-2 3.2E-5 Pu-241b 8.5 3.0 7.8E-1 6.7 8.8E-6 Cm-242b 1.2E-1 4.1E-2 1.1E-2 9.3E-2 2.4E-7 TRUb 1.5E-1 5.2E-2 1.4E-2 1.2E-1 2.4E-7 a. Values derived from Cs-137 concentrations multiplied by industry-wide scaling factors. b. Concentration units for these nuclides are nCi/g. c. DAW=dry active waste; concentrations are based on contact dose rates of 10mR/h.

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Radiochemistry in Nuclear Power Reactors Table 7–4 TYPICAL PWR NUCLIDE CONCENTRATIONS BY WASTE STREAM (μCi/cm3) (Reproduced with Permission, EPRI NP-5983, Ref. 2) Nuclide Primary Resins Primary Filters Non-Primary Resins Non-Primary Filters Evaporator Bottoms DAWc Co-60 50 247 3.1 32.0 5.3E-2 2.6E-3 Cs-137 51.0 — 8.41 — 3.5E-2 2.0E-3 Ni-63 16.1 76.2 1.8 18.0 1.1E-1 1.14E-3 Fe-55 5.5 250 3.6 26.5 5.2E-2 7.25E-3 C-14 6.2E-2 1.4 3.1E-2 5.5E-1 1.0E-3 3.25E-5 I-129 1.7E-4 — 6.1E-5 — 4.0E-7a 1.3E-7 Tc-99 6.6E-4 — 1.8E-4 — 4.0E-6a 9.0E-8 Sr-90 1.9E-1 1.4E-1 1.4E-2 1.8E-2 7.7E-5 5.4E-6 Pu-241b 1400 3822 94 289 9.4E-1 6.2E-5 Cm-242b 19.4 53 1.3 4.0 1.3E-2 9.4E-7 TRUb 25 68 1.7 5.1 1.7E-2 1.6E-6 a. These values were derived by applying industry scaling factors to the cesium-137 concentrations. b. The unit of concentration for these nuclides is nCi/g. c. DAW=dry active waste; concentrations are based on contact dose rates of 10 mR/h.

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Radiochemistry in Nuclear Power Reactors Prior to transfer and/or sampling, holding tanks should be thoroughly mixed with mixing systems which promote homogeneity. The time interval between mixing and sampling should be a minimum to prevent settling of particles. The location of sampling points should be selected where stream stratification is not expected. Pump speed for transferring the waste stream should provide sufficient velocity to prevent settling and stratification of particles in the line or pipe being sampled. It is important that the sampling equipment produces the required size and quantity of individual grabs required for compositing a representative sample which can be handled for laboratory analytical evaluations. The sampling process should be performed on the transfer line to the waste container during the period of transfer. The initiating factor for each grab may be time, flow, radiation level, or other variable depending on the objective of sampling. Sampling variations have been identified(2) as one of the major uncertainties in the 10CFR61 compliance program. One way to reduce sampling error is to increase both the number and size of samples, but the sampling error is frequently compromised by personnel radiation exposure and operation costs associated with sampling. The composite sample is further properly mixed in a laboratory and prepared for radiochemical analysis. Figure 7–1 shows a flow diagram of the necessary sample preparation required to properly analyze the waste sample(5). 7.3 RADIOCHEMICAL ANALYSIS As pointed out earlier, due to difficulties of some analyses, most of the waste radioassay has been performed by outside contractor laboratories. However, the on-site laboratory can generally perform relatively easy-to-measure nuclides, including major gamma-emitting nuclides, which can be measured directly by gamma spectrometry, and some of

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Radiochemistry in Nuclear Power Reactors Figure 7–1. Flow Diagram of Sample Preparation (Reproduced with Permission, EPRI NP-4037, Ref. 5) Note (a) Ge(Li) analysis is performed on the appropriate counting form for the sample type, such as Marinelli beaker or other container for water, counting dish for evaporated water, petri dish for dried or ashed solids, etc. (b) Water with reducing preservative for 3H, 14C, 129I and 99Tc. Water with pulp for “prime”.

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Radiochemistry in Nuclear Power Reactors the beta-emitting nuclides H-3, Sr-90/Y-90, Fe-55 and Ni-63 which have been frequently measured in liquid waste stream samples. The radiochemical procedures for these nuclides are given in Appendix E. In many cases, the major gamma-emitting nuclides can be measured directly in transfer lines or in waste drums. Discussion of direct assay techniques will be presented in the next section. Description of the analytical procedures used by outside contractor laboratories should be available in their literature. A schematic of the typical analysis program used by Science Applications International Corporation (SAIC) is shown in Figure 7–2. A brief summary of measurement methods developed by EAL Corporation has been described by Wassman and Leventhal(5). There are two primary problems associated with the radiochemical analyses of the difficult-to-measure nuclides. These are analytical sensitivity and radiochemical purity. The former is mainly a function of sample size, length of counting times, and measurement equipment. These are variables which are well-defined and easily controlled. In the case of the long-lived, beta-emitting nuclides, the requirement of radiochemical purity is extremely severe and the degree of purity is difficult to measure or control, particularly as many of the potential impurities are also pure beta emitters. Thus it is not unusual to require chemical separation factors on the order of 106 to 107. Detailed evaluation of published radiochemical procedures for those difficult-to-measure nuclides is beyond the scope of this monograph. However, some of the major concerns in the analyses of those nuclides have been reported by Vance et al.(2). 7.4 DIRECT ASSAY TECHNIQUES Direct assay of waste radioactivity by gamma-ray spectrometry is a well-established technique. Generally, a high resolution germanium detector is installed in a heavily shielded collimator, which is positioned properly toward the target waste container or waste transfer line. The direct assay technique generally gives fast and accurate assays of bulk waste with no personnel radiation exposure from sampling.

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Radiochemistry in Nuclear Power Reactors Figure 7–2. Schematic of the Sample Chemical Analysis Program (Ref. 3) LSC: Liquid Scintillation Counter Nal(Tl): Thin window thallium activated sodium iodide scintillation counter Si(Au): Barrier type, silicon alpha energy counter

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Radiochemistry in Nuclear Power Reactors Direct assay is of particular importance for heterogeneous wastes, since representative sampling of those wastes is nearly impossible. Similar to all other radioactivity counting systems, the direct gamma-ray assay system requires a quantitative calibration. Calibration of the system makes it possible to convert the measured gamma-ray intensities into concentrations of gamma-emitting nuclides contained in the radwaste. The conversion factors should include the correction factors for various collimators, detector configurations and counting geometries, in addition to counting efficiencies as a function of gamma-ray energy. Unlike the laboratory counting system, it is rather difficult* to use a known “standard” for counting efficiency and geometry calibration in the counting system for the radwaste. Different calibration techniques have been reported by several investigators.(6–9) One of those calibration procedure has been developed by SAIC(6,7) and used with the direct assay gamma scanner. The procedure incorporates a semi-empirical model for detector efficiency that treats the detector as a point on the axis whose location is a function of gamma-ray energy and whose response to volume elements located off the axis is a function of both energy and the off-axis distance. The calibration procedure for each collimator and detector configuration is to determine the model parameters by mapping the response to a calibrated gamma-ray source (standard) located anywhere in front of the detector-collimator system. The model calculates detection efficiencies for a source-detector geometry through numerical integration of the detector response function over the volume of the source. Calculations include gamma-ray self-attenuation in the source material as well as the attenuation by the source container and any external absorber. There are at least three general rules of practice in application of the gamma-ray direct-assay technique to the assay of radwaste: (1) the background radiation fields in the area should be low, so that a clear gamma-ray spectrum can be obtained from the radwaste for analysis (the background spectrum is always needed for the net activity calculation); (2) establish the proper detection efficiency for the chosen geometry; and (3) choose *   It is difficult, but not impossible, to prepare a simulated waste material, spiked with known quantity of radioactivity, contained in a waste container for direct calibration.

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Radiochemistry in Nuclear Power Reactors the geometry of scan optimal to the waste container. Choosing the optimal scan geometry is often a practical as well as an accuracy consideration. Since the waste is frequently inhomogeneously mixed in the container, it is desirable to scan the entire portion of the container containing the waste. An ideal geometry is with the axis of the detector collimator aligned with the center of the wasteland there is a sufficient space between the detector and the waste container so that the entire container can be scanned.(10) A technique of direct assay of transuranics by passive neutron measurement has been developed by PNL (Battelle Pacific Northwest Laboratories)(6). Neutrons can be produced in the radwaste containing transuranic isotopes, either by spontaneous fission or by the (α, n) reaction. Although the technique has been demonstrated to be a viable technique for assaying measurable levels of transuranics, the efficiency of the counting system is unlikely to achieve the required LLD of 0.1 nCi/g. 7.5 RADIONUCLIDE CORRELATIONS AND SCALING FACTORS As mentioned earlier, in order to report the content of radionuclide in the radwaste, the “difficult-to-measure” nuclides may be inferred by taking a ratio of their activity to the activity of nuclides that can be readily measured. These ratios are referred to as scaling or correlation factors and are expected to be developed on a plant and waste stream specific basis. Ideally, the nuclides used for scaling factors should meet the following criteria: There should be a constant ratio in the rate of formation and release of both nuclides. Both nuclides should possess similar chemical and physical transport properties. The nuclide on which the correlation is based should be readily measurable by gamma-ray spectrometry.

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Radiochemistry in Nuclear Power Reactors Studies have been undertaken(3–5,11) to determine or review the scaling factors for all the difficult-to-measure nuclides. A careful examination of the correlation pairs used by many operating reactors reveals that most of the correlations are based on either Co-60 or Cs-137. There is a considerable disparity between the actual and ideal correlation pairs. As a result, the correlation ratios are considerably dispersive; differences of orders of magnitude have been reported. Variation in the correlation ratios is expected to arise from at least three major areas, which could be significantly different from plant to plant, and could change from time to time. The first area is a change in defective fuel conditions which would give rise to different release rates between nuclides. Detailed discussion on the variation of fission product release rate as a function of fuel failure condition has been presented previously in Section 3. The second area is a change in the primary system materials, from which most of the activated corrosion products originate. In recent years considerable effort has been made in many plants to remove the sources of cobalt (Co-60) as the major step taken to reduce personnel exposures. The Co-60 concentration in the BWR primary coolant has been seen decreasing from ≈ 0.5 μCi/kg in earlier years to <0.1 μCi/kg in some reactors. Reduction of iron crud input from feedwater is also an example of changing the source of activated corrosion products. The third area is a change in the source of inputs to the radwaste system. The compositions of nuclides in different waste streams are expected to be different. These areas and the difficulties of developing the reliable scaling factors for those difficult-to-measure nuclides are discussed below: Tritium (H-3) is produced in the coolant by neutron capture in the 2H(n, γ)3H reaction. It is also produced by high energy neutron reactions with lithium isotopes, 6Li(n, α)3H, and 7Li(n, nα)3H, and with boron isotopes 10B(n, 2α)3H, 10B(n, α)7Li (n, nα)3H. The boron and lithium reactions are predominant sources of tritium in PWR. Tritium is also produced by ternary fission of U-235, but only a small fraction (≈1%) of the total H-3 produced in the fuel is believed to diffuse through the cladding into the coolant. H-3 does not relate to any other nuclide in the waste but can be estimated from the coolant content of the radwaste. The concentration of tritium in the coolant is used to calculate the tritium content. The coolant content of the specific stream of the radwaste is, in fact, predetermined by measuring the H-3 concentration in a series of representative samples.

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Radiochemistry in Nuclear Power Reactors Carbon-14 is produced by neutron capture from the 17O(n,α)14C and 14N(n,p)14C reactions. Carbon-14 is released from PWRs in the gaseous radwaste system as organic species, predominantly methane. In BWRs, it is released in the condenser offgas system as CO2. A small fraction of C-14 as CO3= or HCO3− is retained in the reactor water cleanup system. No suitable correlation nuclide can be found in the radwaste. Iron-55 and Ni-63 are produced by neutron activation of Fe-54 and Ni-62, 54Fe(n,γ)55Fe and 62Ni(n,γ)63Ni, respectively. These two nuclides and Co-60 result from the activation of reactor material corrosion and/or wear products. All three species have similar chemical properties, although cobalt and nickel are probably more soluble than iron in the reactor coolant. The major activation process involves the deposition of corrosion products on the fuel surfaces followed by their activation and subsequent release and transport to the radwaste system. There are also some important sources of Co-60 and Ni-63 in the core construction materials. Because there are some differences in the material sources, and the primary coolant chemistry conditions are entirely different between BWRs and PWRs, the ratios of Fe-55/Co-60 and Ni-63/Co-60 are expected to be different between BWRs and PWRs. Nickel-59 is produced by the 58Ni(n,γ)59Ni reaction. Since the ratio of Ni-58 and Ni-62 in natural nickel is invariant, and the product of Ni-59 is quite similar to that of Ni-63, the ratio of the production rate of Ni-59 to Ni-63 should be a constant. This Ni-59/Ni-63 activity ratio has been calculated to be approximately 0.01 irrespective of the flux or core location. Since the Ni-63/Co-60 correlation factor can be measured and established for a specific waste stream, the Ni-59/Co-60 can also be estimated. Strontium-90 is produced by fission. It decays by beta emission to Y-90 which is also a beta emitter. The Y-90 (t1/2=64 hour) is practically in equilibrium with Sr-90, but unlike Sr-90, Y-90 is normally in insoluble forms. Since Sr-90 and Cs-137 are generally found in soluble cationic forms, the activity ratio may be easier to establish as long as the fuel conditions do not change significantly. Technitium-99 is mainly produced by fission. The valence states of technitium may vary from −1 to +7. In the BWR coolant, shorter-lived technitium activities (Tc-99m, −101, − 104) are normally found in the anionic forms, most likely TcO4−, but they would behave

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Radiochemistry in Nuclear Power Reactors totally differently in the PWR coolant. It is very difficult to find a good correlation partner. Although Cs-137 has been suggested, there are not only differences in chemical properties, but the release mechanisms from fuel are also significantly different. Tc-99 is also produced by activation of Mo-98. Iodine-129 is produced by fission as a decay product of Te-129. Iodine activities are always found in either the BWR or PWR coolant as anion species. In BWRs, iodine activities are also volatile and a large fraction of iodine activities released from the failed fuel can be transported with steam into the condensate (Section 3.4). Similar to Tc-99, a gamma-emitting correlation partner for I-129 may not exist. Transuranic isotopes are those produced by successive neutron activation of uranium and its products in a reactor. The chain of production and the radiation characteristic of each isotope are shown in Figure 2–9 in Section 2.6. As discussed earlier, the activity buildup varies quite rapidly with the fuel burnup (Figure 2–10); however, the predominant alpha activity is Cm-242 (90 to 95%) because of its shorter half-life. In the reactor coolant, most of the transuranic (TRU) alpha activities are found to be in insoluble forms, and can be correlated with insoluble fission products, Zr-95 and Ru-103 (Section 8.5) and Ce-144(6). The determination of the TRU content of low-level waste has long been a problem confronting utilities, and the determination of gamma-emitting nuclides Zr-95, Ru-103 or Ce-144 is also very difficult by direct spectrometry because of their low intensities and short half-lives (relative to Cs-137, Co-60 and other major activities in the radwaste). In summary, the correlation or scaling factors for those difficult-to-measure nuclides in the radwaste are very difficult to determine accurately and consistently. The plant and stream specific correlation factors can only be determined for a short operating duration because the correlation may vary from time to time depending on many factors discussed previously.

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Radiochemistry in Nuclear Power Reactors 7.6 REFERENCES (1) U.S. Nuclear Regulatory Commission, “Licensing Requirements for Land Disposal of Radioactive Waste,” Code of Federal Regulations, Title 10, Part 61, Federal Register, 47, p. 57446, December 1982. (2) J.N.Vance, et al., “Assessing the Impact of NRC Regulation 10CFR61 on the Nuclear Industry”, EPRI NP-5983 (August 1988). (3) J.E.Cline, et al., “Assay of Long-Lived Radionuclides in Low-Level Wastes from Power Reactors”, U.S. Nuclear Regulatory Commission NUREG/CR-4101 (April 1985). (4) J.A.Lieberman, et al., “Methodologies for Classification of Low-Level Radioactive Wastes from Nuclear Power Plants” , AIF/NESP-027 (December 1983). (5) W.T.Best, et al., “Radionuclide Correlations in Low-Level Radwaste”, EPRI NP-4037 (June 1985) (6) J.E.Cline, et al., “Advanced Radioactive Waste Assay Methods”, EPRI NP-5497 (November 1987). (7) J.E.Cline, et al., “Direct Assay of Drummed Evaporator Bottoms, Dry-Active Waste and Filter Cartridges at the GINNA Nuclear Station”, ANS Int. Topical Meeting on Waste Management and Decontamination and Decommissioning (1987). (8) M.Tsutsumi, T.Gato and H.Kato, “Assay System for Gamma Radionuclides in Miscellaneous Solid Waste Packed in Drums”, 1988 JAIF International Conf. Water Chemistry in Nuclear Power Plants, Tokyo, Japan, Vol. 2, p. 813, (1988). (9) W.Inder Schmitten, B.Sohnius and E.Wehner, Nucl. Tech., 92, 374 (1990). (10) R.L.Brodzinski, et al., “Design and Operation of a Passive Neutron Monitor for Assaying the TRU Content of Solid Wastes”, PNL-4966 (February 1984). (11) W.T.Best and A.D.Miller, “Updating Scaling Factors in Low-Level Radwaste”, EPRI NP-5077 (March 1987).