Spent Fuel Assemblies Research Articles

Determination of the Plutonium Mass and Curium Ratio of Spent Fuel Assemblies for Input Nuclear Material Accountancy of Pyroprocessing, and Analysis of Their Errors

In pyroprocessing it is important to determine the amount of Pu in the various streams of materials involved. This paper presents two approaches to determine the Pu mass of spent fuel assemblies using nondestructive assay and burnup simulation code. Cm balance is adopted and the concept of “Cm ratio,” the mass ratio of Pu to Cm, is used for the nuclear material accountancy for the model pyroprocessing facility. The biggest error of the nuclear material accountancy is expected to arise from the determination of Pu mass and Cm ratio in input homogeneously mixed uranium oxide powder, which is assayed nondestructively. One approach to determine the Pu mass and Cm ratio is to apply the average burnup of spent fuel and determine the Pu mass and Cm ratio by using the ORIGEN code. The estimated error in Pu mass determined by this method ranges from 0.94% to 2.33% for a total of 225 spent fuel assemblies of various burnup, initial enrichment, and cooling time. The other approach is to use the functional relationship between the neutron emission rate and Pu mass of spent fuel. The error in Pu mass calculated using this method ranges from -1.68% to 3.86%.

Heuristic Methods for Assigning Spent Nuclear Fuel Assemblies to Canisters for Final Disposal

The disposal of spent fuel assemblies (SFAs) by companies currently producing nuclear power in Finland is the responsibility of a company named Posiva Oy. Posiva Oy has decided to use the KBS-3 (Swedish abbreviation for nuclear fuel safety; version 3) concept. In KBS-3, SFAs are placed in metal canisters, which are themselves deposited deep into crystalline rock. The disposal process in Finland will last many decades. To efficiently assign SFAs to canisters, in this paper we study the minimax canister formation problem. In this problem, we assume we are given two sets of data: (a) a schedule specifying the number of disposal canisters per year and (b) the decay heat of each SFA for every disposal year. The goal in the problem is to assign SFAs to canisters so that the largest canister heat load is minimized. The minimax canister formation problem is a variant of a well-known optimization problem: makespan minimization on unrelated parallel machines. We developed heuristic methods for solving the minimax canister formation problem. Using our methods and predicted SFA amounts and properties for Finland, we obtained high-quality solutions in numerous test cases. We also investigated how the uncertainty in SFA burnups affects the canister heat loads.

Partial Defect Tester: A Novel Approach to Detect Partial Defects in Pressurized Water Reactor Spent Fuel

A novel methodology to detect diversion of spent fuel from pressurized water reactors (PWRs) has been developed in order to address a long unsolved safeguards verification problem for an international safeguards organization such as the International Atomic Energy Agency (IAEA) or European Atomic Energy Community (EURATOM). The concept involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly (SFA) and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines, and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed SFA. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the SFA will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard laptop computers.Monte Carlo simulation studies and a set of controlled experiments with actual commercial PWR SFAs were performed, and they validated this novel methodology. Based on the simulation studies and benchmarking measurements, the methodology promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the IAEA goal that for SFAs that can be dismantled at the facility—which is essentially the case for most PWR fuel—the partial defect test used should assure that at least half the fuel pins are present in each SFA. The methodology does not rely on any operator-provided data like burnup or cooling time and does not require movement of the SFA from the storage rack in the spent-fuel pool.

Optimization of Spent Nuclear Fuel Filling in Canisters for Deep Repository

We demonstrated the use of combinatorial methods to optimize the filling of spent nuclear fuel (SNF) in metal canisters for final deep SNF repository, according to the maximal allowed thermal power per canister Pmax and the limit of n = 4 spent-fuel assemblies per canister. As a next step, the deposition time can be optimized by minimizing the required number of canisters M and the interim storage time. The method has been tested in detail for a typical pressurized water reactor (PWR), nuclear power plant (NPP) Krsko, SNF for different numbers of reactor cycles and different Pmax. The results show that the time interval between the last reactor cycle and the optimal deposition time varies between 3 and 30 yr for a typical PWR. The most significant contribution to the uncertainty of the calculated SNF decay heat (thermal power) is due to inaccurate cross sections taken from generic cross-section libraries. The quality of the results was verified by comparing the calculated M to the theoretical lower boundary Mmin. The idea behind the optimization method is universal and thus can be implemented for any SNF, canister, and repository design.

Analysis of spent fuel assemblies using a lead slowing down spectrometer

We analyze the potential of using Lead Slowing Down Spectrometer technology for assaying spent fuel. This initial study demonstrates that it may be feasible to design a system that will provide approximately 1% statistical precision in the determination of the 239Pu concentration in a pressurized water reactor spent-fuel assembly, for intermediate-to-high burnup levels, using commercial neutron sources, and an array of ultra-high-purity 238U threshold fission detectors. LSDS technology can also determine the concentration of 241Pu and 235U. There is indication that missing pins can be detected, as can asymmetry in the fuel bundle. The analytical model used to perform the viability assessment is described, as are the systematic effects that were not incorporated in this analysis, but could significantly degrade actual performance. These results provide the justification and impetus for the initiation of followup studies that will incorporate the complete suite of effects that impact the accuracy of LSDS measurements.

Methodology to make a robust estimation of the carbon steel overpack lifetime with respect to the Belgian Supercontainer design

The Supercontainer (SC) design is the preferred Belgian option for the final disposal of vitrified high-level waste (VHLW) and spent fuel (SF) in deep underground clay layers. The SC consists of a carbon steel overpack, containing VHLW canisters or SF assemblies, surrounded by a thick concrete buffer, which in turn, is entirely encased in a stainless steel envelope. An integrated R&D strategy is developed to demonstrate and defend that the integrity of the carbon steel overpack can be ensured at least during the thermal phase. This integrated approach, proposed to estimate the lifetime of the carbon steel overpack, consists of three steps: lifetime prediction, validation, and confidence building. Under the predicted conditions within the SC (highly alkaline concrete buffer), the carbon steel overpack is expected to undergo uniform corrosion (passive dissolution). The methodology exists in demonstrating that corrosion forms other than uniform corrosion (e.g. localised corrosion such as pitting corrosion, crevice corrosion and stress corrosion cracking) cannot occur (‘exclusion principle’). This paper elaborates on how this methodology is implemented.

SPECT imaging as a tool to prevent proliferation of nuclear weapons

International efforts are taken to avoid the proliferation of material and technologies that may lead to the development of nuclear weapons. These activities are called safeguards and involve inspections of spent nuclear fuel at nuclear power plants and storage facilities. At these inspections, various measuring techniques are employed for verifying the presence and identity of spent nuclear fuel assemblies. However, a fuel assembly contains about 100–300 fuel rods and techniques are also required for verifying that no individual fuel rods have been removed from the assembly. For this purpose, a non-destructive tomographic measurement technique for spent-fuel assemblies is being developed at Uppsala University, based on single photon emission computed tomography (SPECT). The technique utilizes the γ-ray emission from spent fuel. The first step of the methodology is the recording of the γ-ray flux distribution in a large number of positions around the fuel assembly, using γ-ray detectors attached to a collimator system. In the following step, a cross-sectional image of the source distribution in the fuel assembly is reconstructed. Because the fuel rods are highly activated during reactor operation, and because they are stored in water with practically no radioactive content, they appear very clearly in this type of image. The technique has earlier been used for determining the power distribution in fuel assemblies [S. Jacobsson Svärd, A. Håkansson, et al., Nucl. Technol. 151(1) (2005) 70. [1]]. The images obtained in those measurements show that the technique also has great potential for safeguards’ application. In the on-going development of the technique specifically for safeguards, image analysis plays an important role. Some crucial points in the analysis are the identification and positioning of the assembly in the image and also the definition of the background activity level. Finally, proper criteria have to be set for confidently stating if a fuel rod would be considered to be missing.

Gamma and fast neutron radiation monitoring inside spent reactor fuel assemblies

Gamma and neutron signatures of spent reactor fuel were monitored by small-size silicon diode and track etch detectors, respectively, in a nuclear power plant (NPP). These signatures, reflecting gross gamma intensity and the 242,244Cm content, contain information on the burn-up (BU) and cooling time (CT) of the fuel. The small size of the detectors allows close access to inside parts of the assemblies out of reach of other methods. A commercial Si diode was encapsulated in a cylindrical steel case and was used for gross γ monitoring. CR-39 detectors were used for neutron measurements. Irradiation exposures at the NPP were implemented in the central dosimetric channel of spent fuel assemblies (SFAs) stored in borated water. Gross γ and neutron axial profiles were taken up by scanning with the aid of a long steel guide tube, lowered down to the spent fuel pond by crane and fitted to the headpiece of the fuel assemblies. Gamma measurements were performed using a long cable introduced in this tube, with the Si diode at the end. A long steel wire was also led through the guide tube, to which a chain of 15 sample holder capsules was attached, each containing a track detector. Gamma dose rates of 0.1–10 kGy h −1, while neutron fluxes in a range of (0.25–26) 10 4 cm −2 s −1 were recorded. The results are in good correlation with those of a calculation for spent fuel neutron yield.

Cost-effectiveness analysis of seismically isolated pool structures for the storage of nuclear spent-fuel assemblies

This study introduces a method for evaluating cost-effectiveness of seismically isolated pool structures, taking account of fluid–structure interaction effects. As a measure of cost-effectiveness, the procedure estimates minimum life-cycle cost, sum of initial construction cost and excepted damage cist over life-cycle of the structure. The expected damage cost is the function of failure probabilities, which are computed by frequency domain analysis for convenience. Input ground motions are represented by a power spectral density function compatible with site-dependent response spectrums. The interaction effects between flexible walls and contained fluid are considered in the form of the added mass matrix. The thickness of wall and the stiffness of isolator are considered two key design variables in achieving a design with minimum life-cycle cost. The numerical examples show that seismically isolated pool tanks are highly cost-effective in low-to-moderate seismic regions.

Calculations of the Neutron Flux Outside BWR 8 × 8 Spent-Fuel Assemblies and the Sensitivity to Fuel Pin Diversion

The possibility of detecting replaced fuel rods in a spent-fuel assembly by means of measurement of the emitted neutron- and gamma-ray radiation has been investigated by computer simulations. The radiation field outside a boiling water reactor 8 × 8 fuel assembly with varying patterns of fuel rods replaced with lead dummies was calculated using a simple model for the source distribution and the Monte Carlo code MCNP-4C for the radiation field. In particular, the sensitivity of the thermal neutron field as measured in a Fork detector to various replacement patterns was investigated. The results suggest a detection limit of 5% of the fuel mass replaced, i.e., 3 out of 63 rods, independently of the pattern of the replaced rods.

Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States

Because of the unavailability of off-site storage for spent power-reactor fuel, the NRC has allowed high-density storage of spent fuel in pools originally designed to hold much smaller inventories. As a result, virtually all U.S. spent-fuel pools have been re-racked to hold spent-fuel assemblies at densities that approach those in reactor cores. In order to prevent the spent fuel from going critical, the fuel assemblies are partitioned off from each other in metal boxes whose walls contain neutron-absorbing boron. It has been known for more than two decades that, in case of a loss of water in the pool, convective air cooling would be relatively ineffective in such a "dense-packed" pool. Spent fuel recently discharged from a reactor could heat up relatively rapidly to temperatures at which the zircaloy fuel cladding could catch fire and the fuel's volatile fission products, including 30-year half-life 137 Cs, would be released. The fire could well spread to older spent fuel. The long-term land-contamination consequences of such an event could be significantly worse than those from Chernobyl. No such event has occurred thus far. However, the consequences would affect such a large area that alternatives to dense-pack storage must be examined--especially in the context of concerns that terrorists might find nuclear facilities attractive targets. To reduce both the consequences and probability of a spent-fuel-pool fire, it is proposed that all spent fuel be transferred from wet to dry storage within five years of discharge. The cost of on-site dry-cask storage for an additional 35,000 tons of older spent fuel is estimated at $3.5-7 billion dollars or 0.03-0.06 cents per kilowatt-hour generated from that fuel. Later cost savings could offset some of this cost when the fuel is shipped off site. The transfer to dry storage could be accomplished within a decade. The removal of the older fuel would reduce the average inventory of 137 Cs in the pools by about a factor of four, bringing it down to about twice that in a reactor core. It would also make possible a return to open-rack storage for the remaining more recently discharged fuel. If accompanied by the installation of large emergency doors or blowers to provide large-scale airflow through the buildings housing the pools, natural convection air cooling of this spent fuel should be possible if airflow has not been blocked by collapse of the building or other cause. Other possible risk-reduction measures are also discussed. Our purpose in writing this article is to make this problem accessible to a broader audience than has been considering it, with the goal of encouraging further public discussion and analysis. More detailed technical discussions of scenarios that could result in loss-of-coolant from spent-fuel pools and of the likelihood of spent-fuel fires resulting are available in published reports prepared for the NRC over the past two decades. Although it may be necessary to keep some specific vulnerabilities confidential, we believe that a generic discussion of the type presented here can and must be made available so that interested experts and the concerned public can hold the NRC, nuclear-power-plant operators, and independent policy analysts such as ourselves accountable.

Gamma-Ray Spectroscopy Measurements of Decay Heat in Spent Nuclear Fuel

A method for determining the residual thermal power in spent nuclear fuel using gamma-ray spectroscopy is suggested. It is based on the correlation between the residual power and the 137Cs activity, which is nearly linear for fuel with cooling times between 10 and 50 yr. Using available data of calorimetrically measured values of the decay heat in 69 boiling water reactor and pressurized water reactor spent-fuel assemblies resulted in agreement with a standard deviation of 3%.

The passive nondestructive assay of the plutonium content of spent-fuel assemblies from the BN-350 fast-breeder reactor in the city of Aqtau, Kazakhstan

The International Atomic Energy Agency is presently interested in developing equipment and techniques to measure the plutonium content of breeder reactor spent-fuel assemblies located in storage ponds before they are relocated to more secure facilities. We present the first quantitative nondestructive assay of the plutonium content of fast-breeder reactor spent-fuel assemblies while still underwater in their facility storage pond. We have calibrated and installed an underwater neutron coincidence counter (Spent Fuel Coincidence Counter (SFCC)) in the BN-350 reactor spent-fuel pond in Aqtau, Kazakhstan. A procedure has been developed to convert singles and doubles (coincidence) neutron rates observed by the SFCC into the total plutonium content of a given BN-350 spent-fuel assembly. The plutonium content has been successfully determined for spent-fuel assemblies with a contact radiation level as high as ∼10 5 Rads/h. Using limited facility information and multiple measurements along the length of spent-fuel assemblies, the combined measurement and facility declaration error is ∼8%. A simplified one-point measurement procedure leads to a combined measurement and facility declaration error of ∼13%.

Radioactive Nuclear Waste

Radioactive nuclear waste includes a large number of radioactive elements. Behavior of these elements is unstable and unpredictable because the elements can change form and properties during their decay process. In order to control this type of waste effectively, the interaction mechanism(s) between radioactivity and the environment must be understood, especially with the soil-rock-water system. In this paper three major types of radioactive elements—namely uranium, radium, and radon gas—are discussed because they are closely related to environmental geotechnology. It focuses on the spent-fuel assemblies produced from nuclear power plants and random gas seeping from ground soil. Several new interaction mechanisms between nuclear elements and the environment are proposed and explained. Remedial action on waste controlling techniques are suggested and discussed.

A Comparison of Natural Convection Heat Transfer for a Staggered Versus an Aligned Array of Horizontal Spent Nuclear Fuel Rods Within a Rectangular Enclosure

Natural convection heat transfer was experimentally investigated in a staggered array of heated cylinders, oriented horizontally within a rectangular isothermal enclosure. The test conditions were characteristic of a spent-fuel assembly during transport or horizontal dry storage. The assembly was configured with a pitch-to-diameter ratio of 1.33 and backfilled with pressurized helium or nitrogen. The backfill pressure was varied between 1 and 5 atm, while the assembly power was varied between 1 and 5 W per heater rod. The resulting data are presented in the form of Nusselt-Rayleigh number correlations, where the Nusselt number has been corrected for thermal radiation using a numerical technique. The staggered-array data are compared to previous data for a similar-pitch aligned rod array (a simulated boiling water reactor fuel assembly) to determine if convective heat transfer is enhanced or hindered in a staggered configuration. For the overall array, both the staggered and aligned configurations yield Nusselt-Rayleigh curves with a three-regime trend, which suggests distinct conduction and convection regimes separated by a transition regime. For lower Rayleigh numbers (<106), representative of the conduction regime, the aligned-array Nusselt number is 10 to 12% higher than the corresponding staggered-array value. However, in the convection regime at higher Rayleigh numbers, the staggered-array Nusselt number slightly exceeds the aligned- array Nusselt number. This is attributed to the fact that the staggered array begins to transition into the convection regime at lower Rayleigh number than the aligned array. For both configurations, the slope of the Nusselt-Rayleigh curve in the convection regime suggests turbulent flow conditions.

Design of a Spent-Fuel Assay Device Using a Lead Spectrometer

The neutron slowing-down-time method for nondestructive assay of light water reactor spent fuel has been under development for many years. Results for a newly optimized design of a lead slowing-down-time spectrometer for spent-nuclear-fuel assay are presented. Monte Carlo analyses were performed to optimize the design of the assay device, determine its main parameters, investigate the effects of the spent-fuel assembly and the detector impurities on its performance, determine the fission signatures of the fissile isotopes in spent-fuel elements, and simulate the assay signal as a function of the slowing-down time, assuming threshold fission chambers for the assay detectors. The assay signals from the threshold detectors were analyzed to predict the unknown masses of the fissile isotopes in a typical spent commercial light water reactor fuel element. The broadened resolution of the system caused by the presence of the spent fuel inside the spectrometer pile was found sufficient to separate the signatures of the U and Pu fissiles in spent fuel.

A Numerical Investigation of Natural Convection Heat Transfer Within Horizontal Spent-Fuel Assemblies

A numerical investigation of natural convection heat transfer is carried out for a single, horizontal, spent-fuel assembly in an environment typical of spent-fuel transportation systems as well as some dry storage/disposal scenarios. The objective is to predict computationally the convective heat transfer trends for horizontal spent fuel and to compare the results to data taken in a supporting experimental effort. The predicted data consist of thermal and flow fields throughout the assembly for a wide range of Rayleigh number, as well as numerically obtained Nusselt-number data that are correlated as a function of Rayleigh number. Both laminar and turbulent approaches are examined for a Boussinesq fluid with Pr = 0.7. The data predict the existence of a conduction-dominated regime, a transition regime, and a convection regime. Compared with the laminar approach, a significant improvement in the predicted Nusselt number is obtained for large Rayleigh numbers when a turbulence model is employed. This lends additional support to the experimental evidence that a transition to turbulent flow occurs for Rayleigh numbers greater than 107. Overall, the numerically predicted heat transfer trends compare well with previously obtained experimental data, and the computed assembly Nusselt numbers generally reside within the range of experimental uncertainty. The predicted thermal and flow fields further provide a numerical flow visualization capability that enhances the understanding of natural convection in horizontal spent fuel and allows improved physical interpretation of the experimental data.

Status of the Gorleben Pilot Conditioning Plant

In 1994, because of a change in atomic law, the concept of direct final disposal was developed as an equivalent alternative to the concept of waste management that included reprocessing.Since 1979, tests for direct final disposal have been conducted in Germany. In 1985, the State and the utilities came to an agreement to develop this concept of waste management to technical maturity. Gesellschaft für Nuklear-Service was commissioned by the utilities with the following tasks: to develop and test components with regard to conditioning technology, to construct and operate the pilot conditioning plant (PCP), and to develop casks suitable for final disposal.Since 1980, the construction of the PCP has taken place at the Brennelementlager Gorleben site. The PCP has been designed as a multipurpose facility and can thus fulfill various tasks within the framework of conditioning and managing spent-fuel assemblies and radioactive waste. The pilot character of the plant allows for the development and testing in the field of spent-fuel-assembly conditioning.The objectives of the PCP may be summarized as follows: to condition spent-fuel assemblies, to reload spent-fuel assemblies and waste packages, to condition radioactive waste, and to do maintenance work on transport and storage casks as well as on waste packages. Currently, the buildings of the PCP are constructed and the technical facilities are installed based on the Atomic Law Agreement. The plant will be ready for service in the middle of 1998. It is the first plant of its kind in the world.

Evaluation of Direct Disposal from the Point of View of One Utility and Perspectives for Waste Management

Reprocessing is not a method of spent-fuel disposal but merely an intermediate step that may precede final disposal to save resources. Based on physical and economic factors, it may prove reasonable to directly dispose of spent-fuel assemblies according to their individual burnups, i.e., according to the quality of the residual materials they contain, either through direct final disposal or via reprocessing and recycling. Currently, higher burnups are making reprocessing impractical.Even if spent fuel is to be disposed of directly, the German concept provides that it must be kept in interim storage for as long as 40 yr after discharge from the reactor before it is brought into a repository. During this period, economic aspects must be continuously considered to decide whether the residual materials in the fuel assemblies could be economically used under circumstances other than those prevailing at present.

International Nuclear Material Safeguards for the Direct Final Disposal of Spent-Fuel Assemblies

One of the two German spent-fuel management options envisages the direct (i.e., no reprocessing) final disposal of spent-fuel assemblies from nuclear reactors in a geological repository. The reference repository will be located in a salt dome. The spent fuel will become inaccessible immediately after its emplacement in drifts, and this rules out a retrieval of the nuclear material from the repository. The Nuclear Non-Proliferation Treaty requires compliance verification by inspection regimes. A safeguards approach has been developed for the German reference disposal concept, which would be applicable today and which would meet the requirements of the International Atomic Energy Agency as well as the recommendations of the 1995 Consultants Meeting on Safeguards for Geological Repositories. It foresees cask verification in the aboveground area (material balance area I) and design information verification in the aboveground and belowground (material balance area II) areas of the repository. Safeguards have to continue even after closure and decommissioning of the repository mine.