Shallow Land Burial Research Articles

Activation calculation and radiation analysis for China Fusion Engineering Test Reactor

The activation calculation and analysis for the China Fusion Engineering Test Reactor (CFETR) will play an important role in its system design, maintenance, inspection and assessment of nuclear waste. Using the multi-particle transport code FLUKA and its associated data library, we calculated the radioactivity, specific activity, waste disposal rating from activation products, nuclides in the tritium breeding blanket, shielding layer, vacuum vessel and toroidal field coil (TFC) of CFETR. This paper presents the calculation results including neutron flux, activation products and waste disposal rating after one-year full operation of the CFETR. The findings show that, under the assumption of one-year operation at the 200MW fusion power, the total radioactivity inventory will be 1.05×1019Bq at shutdown and 1.03×1017Bq after ten years. The primary residual nuclide is found to be 55Fe in ten years after the shutdown. The waste disposal rating (WDR) values are very low (<<1), according to Class C limits, CFETR materials are qualified for shallow land burial. It is shown that CFETR has no serious activation safety issue.

Waste Management Assessment of Candidate Materials for HiPER Reaction Chamber

One of the critical decisions in the HiPER project is to select the most appropriate material for the reaction chamber. Within this framework, we investigate the performance of different steel alloys with respect to waste management. The capabilities of commercial steels, both austenitic and ferritic/martensitic, compared to reduced-activation ferritic/martensitic steels are evaluated as for different waste management strategies (near surface burial, clearance, hands-on and remote recycling). The examined materials are: SS304, SS316, mod.9Cr-1Mo and HT9 and EUROFER. Real impurities concentrations are taken into account, and their impact is analyzed. In the study, we have assumed the most exigent HiPER 4a irradiation scenario. Commercial steels revealed to be a suitable choice for the HiPER reaction chamber, as far as their waste management options do not differ significantly from those of the reduced activation ferritic steel case. We found that for mod.9Cr-1Mo and EUROFER hands-on recycling is already possible after a cooling time shorter than 50 years and that shallow-land burial is practicable for all the steel alloys studied. The impurities present in the real heats affects the cooling time for manual recycling but not significantly. Shallow-land burial feasibility is not perturbed by the presence of impurities in the real commercial heats. Moreover, the impact of activation cross section uncertainties on the waste management assessment of the irradiated steels has been analyzed, and it is found to be of no practical significance to determine eligibility of the considered steels for the HIPER 4a reaction chamber.

THE USE OF FLY ASH AND ZEOLITE MIXTURE AS BACKFILL MATERIAL IN THE RADIOACTIVE WASTE REPOSITORY

An investigation of the contribution of fly ash in the fly ash-zeolite mixture as the backfill material on the shallow land burial of radioactive waste has been done. The experiment objective is to know the effect of zeolite particle size and fly ash-zeolite weight ratio on physical properties such as permeability (K) and dispersion characteristic such as effective dispersion coefficient (De) in the fly ash-zeolite form as backfill material. The experiment was carried out by the fixed bed method in the column filled by the fly ash-zeolite mixture with a fly ash-to-zeolite weight percent ratio of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 in the water saturated condition flown by uranyl nitrate solution at concentration (Co) of 500 ppm. The concentration of uranium in the effluents in interval 15 minutes represented as Ct was analyzed by spectrophotometer, then using Co and Ct, data effective dispersion coefficient (De) in the backfill material were determined. The experiment data showed that -400 mesh fly ash and -70+80 mesh zeolite on fly ash-to-zeolite with weight percent ratio of 40/60 with K = 5.00x10-5cm/second and De = 1.11.10-5 cm2/second can be used as backfill material. Keywords: backfill material, fly ash, radioactive waste, zeolite

BENTONITE-QUARTZ SAND AS THE BACKFILL MATERIALS ON THE RADIOACTIVE WASTE REPOSITORY

An investigation of the contribution of quartz sand in the bentonite mixture as the backfill materials on the shallow land burial of radioactive waste has been done. The experiment objective is to determine the effect of quartz sand in a bentonite mixture with bentonite particle sizes of -20+40, -40+60, and -60+80 mesh on the retardation factor and the uranium dispersion in the simulation of uranium migration in the backfill materials. The experiment was carried out by the fixed bed method in the column filled by the bentonite mixture with a bentonite-to-quartz sand weight percent ratio of 0/100, 25/75, 50/50, 75/25, and 100/0 on the water saturated condition flown by uranyl nitrate solution at concentration (Co) of 500 ppm. The concentration of uranium in the effluents in interval 15 minutes represented as Ct was analyzed by spectrophotometer, then using Co and Ct, retardation factor (R) and dispersivity () were determined. The experiment data showed that the bentonite of -60+80 mesh and the quartz sand of -20+40 mesh on bentonite-to-quartz sand with weight percent ratio of 50/50 gave the highest retardation factor and dispersivity of 18.37 and 0.0363 cm, respectively. Keywords: bentonite, quartz sand, backfill materials, radioactive waste

Lack of international consensus on the disposition and storage of disused sealed sources

A lower-activity analogue of the trans-national problem of spent fuel management and disposal is the global problem of radioactive sealed source [source: The IAEA definition of a sealed source is “Radioactive material that is permanently sealed in a capsule or closely bonded and in a solid form.” Taken from glossary of Nuclear Waste Data Management found at http://www-ewmdb.iaea.org/showhelp.asp?Topic=8-1-1.] disposal. Sources are found in almost every country in the world because of their beneficial medical and commercial or industrial applications. Some of the isotopes used have short half-lives—iridium-192 (Ir-192), 73.8 days—while others have very long half-lives—americium-241 (Am-241), 432 years or plutonium-239 (Pu-239), 24,130 years. It is critically important, particularly for longer-lived isotopes, to find final disposition pathways. Lack of a permanent disposition pathway such as recycling or irretrievable disposal creates numerous problems, including the potential loss of regulatory control, which increases the risk of inadvertent or deliberate misuse of the material. The misuse of radioactive materials has the potential for substantial public health and economic damage. Disused sources also pose an inherent risk to the end-users from a liability, safety, and public health perspectives. This paper examines various disposition pathways employed by several key source manufacturing or possessing nation-states for disused sources. Examples of source disposition pathways include long-term storage, deep geological disposal, borehole disposal and shallow land burial. The Off-Site Source Recovery Project (OSRP), part of the office of Global Threat Reduction Initiative (GTRI), acts as an intermediary in the recovery and ultimate disposition of US origin sealed radiological materials. Several concepts that could help mitigate the challenge of a lack of long-term disposition options for sources are available, but these tools have not yet been applied by most nation-states. For example, regional consolidation and repatriation of sources to the country of manufacture would ease or eliminate the need for in situ disposal or storage in a number of developing nation-states.

Leaching of uranium, radium and thorium from vertisol soil by ground water

Shallow land burial is routinely used for the disposal of low-level radioactive waste. Natural processes causing leaching of radionuclides can lead to contamination of surrounding ground water and soil by the radionuclides. The comparative leachability of radionuclides U(nat), 226Ra, 228Ra and Th(nat) from the soil of a radioactive waste disposal site, by ground water was evaluated. The probability of leaching was obtained in the following order Ra (≈77%) > U (≈40%) > Th (≈20%). Observed ratios (OR) were calculated to correlate leachability of radionuclides to that of major cations Ca2+ and Mg2+. The leaching of the radionuclides was seen to be dependent on Ca2+ and SO42− leached from the soil. This study provides sitespecific leachability of radionuclides, that can be used as indicator of the tendency for migration or retention in soil. It can play an important role during an unforeseen accident like breach of containment at the waste disposal site leading to contamination of soil and ground water and causing hazard to public via drinking water route.

Safety assessment of Chinese ITER TBM design with helium-cooled solid breeder concept

Safety analyses have been performed for the Chinese ITER test blanket module (TBM) design with helium-cooled solid breeder (HCSB) concept for testing in a ITER device based on the design developed with the China blanket programme. Some safety issues associated with this TBM design are explored. In particular, radiological inventories, afterheat, waste disposal ratings (WDRs), electromagnetic analysis, LOCA accident analysis and tritium safety management for the CH HCSB TBM design are given. The analysis results show that the total radiological inventory in the TBM at shutdown is very low and drops rapidly within a minute due to the decay of the Mn-56 isotope, the total afterheat is as low as 8.77 × 10−3 MW, which is attributed mainly to the structure and the WDRs for Ni-59, Zr-93 and Nb-94 are 1.48 × 10−2, 4.72 × 10−9 and 1.034 × 10−3, respectively, and according to the US 10CFR61 regulation, these materials are qualified for shallow land burial. Results show that this test blanket concept is feasible within the existing domestic technologies and has good safety characteristics. In addition, future directions in the safety and environmental area are proposed.

Impact of N-isotope composition control of ferritic steel on classification of radioactive materials from fusion reactor

The possibility of reducing the concentration of 14C produced in irradiation of reduced activation ferritic steels (RAFS) intended as structural materials in first wall and blanket of fusion power plants was investigated. During RAFS irradiation 14C is mainly produced from the more abundant of the two isotopes of nitrogen present in the steel, namely 14N. The method proposed consists in increasing the enrichment of the other isotope ( 15N). One-dimension transport calculations show that for a typical DEMO blanket configuration an 15N enrichment from the natural value (0.37%) to 95% was sufficient to keep the end of life 14C concentration in the RAFS below the limit (3.7 × 10 7 Bq/kg) fixed by the Japanese Nuclear Safety Commission for qualifying it as a low level material (LLM) which can be disposed by shallow land burial.

Sensitivity of Shallow Land Burial to neutron environment and activation cross sections in IFE thick-liquid concepts

A comprehensive assessment on the eligibility of reduced activation (RA) steels as structural chamber material in Inertial Fusion Energy (IFE) thick-liquid concepts is performed. As far as alloying elements, it is shown that the activation of tungsten is a question to debate. Regarding impurity elements, it is analyzed if they could question the possibility of obtaining real RA steels for shallow land burial (SLB). The effect of the thickness of the liquid wall on the SLB response of alloying and impurity elements is computed. And above all, we have estimated the impact of cross section uncertainties when addressing the former questions, and we identify those to be improved. The necessary improvement of some tungsten and niobium cross sections is justified.

Effect of activation cross-section uncertainties in selecting steels for the HYLIFE-II chamber to successful waste management

We perform the waste management assessment of the different types of steels proposed as structural material for the inertial fusion energy (IFE) HYLIFE-II concept. Both recycling options, hands-on (HoR) and remote (RR), are unacceptable. Regarding shallow land burial (SLB), 304SS has a very good performance, and both Cr–W ferritic steels (FS) and oxide-dispersion-strengthened (ODS) FS are very likely to be acceptable. The only two impurity elements that question the possibility of obtaining reduced activation (RA) steels for SLB are niobium and molybdenum. The effect of activation cross-section uncertainties on SLB assessments is proved to be important. The necessary improvement of some tungsten and niobium cross-sections is justified.

Model of Analysis of the Groundwater-Protection Level in Shallow Land Burial of Radioactive Waste

A mathematical model of transfer of radionuclides in the aeration zone has been considered. The analytical solution for this model in a generalized form suitable for practical application has been obtained. A procedure enabling one to determine the optimum size of the protection zone and in which the maximum concentration of radionuclides does not exceed the maximum permissible levels for a specific kind of radionuclides has been proposed.

Activation Analysis for Two Molten Salt Dual-Coolant Blanket Concepts for the U.S. Demo Reactor

The US has considered, among other options, two blanket concepts for Demo reactor in which helium is primarily used to cool the first wall (FW) and structure whereas molten salt (MS) is used as both coolant and breeder. Conventional reduced activation ferritic steel (RAFS, F82H) is used as the structural material in both blanket concepts. The low melting point Flibe (~380°C) is used in the first option while the Flinabe (~305°C) is used in the second option. In this paper, we present the results for assessing the radioactivity and decay heat. This assessment is performed separately for the structural material, the Be multiplier and the breeder (Flibe/Flinabe). The Class C waste disposal rating (WDR) was estimated for each material. For Flibe, Flinabe and Be the WDR is much lower than unity. However, the WDR for F82H is ~0.6-1.3. They are attributed to reactions with Mo and Nb present in F82H with levels of 70 wppm and 4 wppm, respectively. To ensure that F82H qualifies for shallow land burial, it is suggested to reduce these two impurities to ~50 and ~3 wppm, respectively.

Safety and Environmental Aspects of Inertial Fusion Energy: An Overview of Recent Activities and Developments in the United States

During the past 2 yr, significant progress has been made in several areas related to the safety and environmental (S&E) aspects of inertial fusion energy (IFE). An updated methodology has been developed, and accident analyses have been performed for two IFE conceptual power plants and a target fabrication facility. Parallel to the consequence analyses of different accident scenarios, ongoing studies of accident initiating events are being used to support safety assessment and create a basic framework of types of events to consider in future risk characterization of new plant designs. Target designers/fabrication specialists have been provided with ranking information related to the S&E characteristics of candidate target materials. We have revisited waste management options for IFE, introducing the concept of clearance versus the traditional shallow land burial. A brief summary of results in each of these activities is given, and plans for future work are outlined.

Use of clearance indexes to assess waste disposal issues for the HYLIFE-II inertial fusion energy power plant design

Traditionally, waste management studies for fusion energy have used the waste disposal rating (WDR) to evaluate if radioactive material from irradiated structures could qualify for shallow land burial. However, given the space limitations and the negative public perception of large volumes of waste, there is a growing international motivation to develop a fusion waste management system that maximizes the amount of material that can be cleared or recycled. In this work, we present an updated assessment of the waste management options for the HYLIFE-II inertial fusion energy (IFE) power plant, using the concept of clearance index (CI) for radioactive waste disposal. With that purpose, we have performed a detailed neutronics analysis of the HYLIFE-II design, using the tart and acab computer codes for neutron transport and activation, respectively. Whereas the traditional version of acab only provided the user with the γ contact dose rate for recycling assessments and WDR as an index for waste disposal considerations, here we have modified the code to calculate CIs using the current international atomic energy agency (IAEA) clearance limits for radiological waste disposal. The results from the analysis are used to perform an assessment of the waste management options for the HYLIFE-II IFE design.

Overview of International Waste Management Activities in Fusion

The minimization of active waste from the operation and decommissioning of a fusion power plant is a common goal of fusion development programs in Europe and in the US. Approaches differ, according to national regulations, and include reuse and recycling, clearance for non-active disposal or free-release recycling, and shallow land burial. Adopting the minimization of active waste volume as a design requirement leads to a Low Activation Design if properly optimized for materials choices. Power plant studies, both ARIES in the US and the PPCS in Europe, are adopting strategies to meet such requirements. International collaboration, particularly in the frame of an IEA co-operative program, provides benefits in the development of waste management strategies.

Composition adjustment of low activation materials for shallow land burial

The three representative low activation materials for a fusion reactor are ferritic steel, V-alloy and SiC/SiC composite. The adjustment of the material composition of these materials to increase the fraction of shallow land burial in Japan was considered. In Japan, the fission waste having any single radionuclide exceeding the limiting concentration value, causing 100 μSv year −1 individual dose, determined by the Nuclear Safety Commission will not qualify as a low level waste (LLW), which could be disposed by shallow land burial. The limiting concentration values of radionuclides produced in fusion reactor were derived based on the methodology of the Nuclear Safety Commission. Radionuclide concentrations of the radwastes generated from the fusion power reactors using the three low activation materials based on the composition of existing materials were evaluated. Radwastes are classified into LLW and medium level waste (MLW), which is defined as the waste which does not qualify for LLW because one or more of the radionuclides exceeds the derived limiting concentration value. The weight fraction of MLW among the sum of LLW and MLW is found to be 10% for ferritic steel, 54% for V-alloy and 43% for SiC/SiC. The possibility of decreasing the MLW fraction by the material composition adjustment is considered. It is found that if Nb impurity content in V-alloy and N impurity content in SiC/SiC composite could be reduced, the MLW fraction can be significantly decreased. On the other hand, the content of the alloy component material (W), needs to be reduced to further decrease the MLW fraction in case of the ferritic steel F82H.

00/02985 Use of biofilters and suspended-growth reactors to treat VOCs

The three representative low activation materials for a fusion reactor are ferritic steel, V-alloy and SiC/SiC composite. The adjustment of the material composition of these materials to increase the fraction of shallow land burial in Japan was considered. In Japan, the fission waste having any single radionuclide exceeding the limiting concentration value, causing 100 μSv year−1 individual dose, determined by the Nuclear Safety Commission will not qualify as a low level waste (LLW), which could be disposed by shallow land burial. The limiting concentration values of radionuclides produced in fusion reactor were derived based on the methodology of the Nuclear Safety Commission. Radionuclide concentrations of the radwastes generated from the fusion power reactors using the three low activation materials based on the composition of existing materials were evaluated. Radwastes are classified into LLW and medium level waste (MLW), which is defined as the waste which does not qualify for LLW because one or more of the radionuclides exceeds the derived limiting concentration value. The weight fraction of MLW among the sum of LLW and MLW is found to be 10% for ferritic steel, 54% for V-alloy and 43% for SiC/SiC. The possibility of decreasing the MLW fraction by the material composition adjustment is considered. It is found that if Nb impurity content in V-alloy and N impurity content in SiC/SiC composite could be reduced, the MLW fraction can be significantly decreased. On the other hand, the content of the alloy component material (W), needs to be reduced to further decrease the MLW fraction in case of the ferritic steel F82H.

Impurity effects on reduced-activation ferritic steels developed for fusion applications

Reduced-activation steels are being developed for fusion applications by restricting alloying elements that produce long-lived radioactive isotopes when irradiated in the fusion neutron environment. Another source of long-lived isotopes is the impurities in the steel. To examine this, three heats of reduced-activation martensitic steel were analyzed by inductively coupled plasma mass spectrometry for low-level impurities that compromise the reduced-activation characteristics: a 5-ton heat of modified F82H (F82H-Mod) for which an effort was made during production to reduce detrimental impurities, a 1-ton heat of JLF-1, and an 18-kg heat of ORNL 9Cr–2WVTa. Specimens from commercial heats of modified 9Cr–1Mo and Sandvik HT9 were also analyzed. The objective was to determine the difference in the impurity levels in the F82H-Mod and steels for which less effort was used to ensure purity. Silver, molybdenum, and niobium were found to be the tramp impurities of most importance. The F82H-Mod had the lowest levels, but in some cases the levels were not much different from the other heats. The impurity levels in the F82H-Mod produced with present technology did not achieve the low-activation limits for either shallow land burial or recycling. The results indicate the progress that has been made and what still must be done before the reduced-activation criteria can be achieved.

Sorption Behavior of Radionuclides on Calcium-Leached Mortar

ABSTRACTLow-level radioactive waste (LLRW) is disposed of by shallow land burial. Reinforced concrete is used as the radioactive waste repository. However, the concrete structure is in contact with water and will gradually degrade over an extended period of time due to leaching.It is important to investigate the interaction between radionuclides and degraded concrete in the safety assessment of nuclear waste disposal. The authors measured the distribution coefficients (Kd) of various radionuclides for calcium-leached mortars. The calcium-leached mortars were prepared by an accelerated leaching test for mortar based on the electrical potential gradient. These degraded conditions are similar to that of degraded concrete in contact with water for a long period of time. The degradation degree of calcium-leached mortar is evaluated by the CaO/SiO2molarratio (Ca/Si ratio) of calcium silicate hydrate (C-S-H).As a result, the relationship between Kd and the Ca/Si ratio in C-S-H can be roughly grouped into the following three types:1.137Cs and85Sr – Kd decreases with an increase in the Ca/Si ratio.2.95mTc and110mAg – There is no correlation between Kd and the Ca/Si ratio.3.14C,241Am and125I – Kd increases with an increase in the Ca/Si ratio.

Fate and Transport of Hexavalent Chromium in Undisturbed Heterogeneous Soil

The disposal of toxic metals [e.g., Cr(VI)] generated by the Department of Energy during the cold war era has historically involved shallow land burial in unconfined pits and trenches. The objectives of this study were to investigate the impact of coupled hydrologic and geochemical processes on the fate and transport of Cr(VI) in undisturbed soil cores obtained from a fractured, acidic inceptisol that are commonly used in the disposal of waste at the Oak Ridge National Laboratory. The mobility of Cr(VI) was significantly retarded relative to a nonreactive Br- tracer, and the mobility decreased with increased loading of the solid phase with natural organic matter (NOM). A significant portion of added Cr(VI) did not elute from the columns, and X-ray absorption near-edge structure (XANES) revealed that both Cr(VI) and Cr(III) resided on the soil mineral surfaces. The reduction of Cr(VI) to Cr(III) was dramatically more significant on soils with higher levels of surface-bound NOM. This indicated that NOM was serving as a suitable reductant during Cr(VI) transport even in the presence of potentially competing geochemical oxidation reactions involving Cr. The redox reaction was catalyzed by the presence of soil mineral surfaces, and the reduced product Cr(III) was immobilized as a tightly bound moiety. The effectiveness of surface-bound NOM to reduce toxic Cr(VI) to Cr(III) under acidic conditions has important implications regarding the design and implementation of in situ remedial strategies.