When fuel rods have defects, the fission products in the fuel rods will come out and enter the reactor coolant through the defects. The release of the fission products will result in the increase of radiation dose, the indeterminacy of the fuel management and will influence the economics and safety of the nuclear power station. Based on the analysis of the typical nuclides activities in reactor coolant, the evaluation of the defect fuel rods can be realized. This paper studied the related analysis around the world and determined the methods to evaluate the number, the defect type (open or tight) and the burn-up of the defect fuel rods. The evaluation method of this paper can be used to evaluate the defects of fuel rods, and can provide valuable information for the fuel management and dose analysis, and also can be a useful technical support to the operation of nuclear power station.

The High Flux Engineering Test Reactor (HFETR) is loaded with complex geometry assemblies, taking on strong heterogeneity. These features call for more advanced core analysis and fuel management methodologies compared with traditional PWR. Based on the two-step method, a fuel management simulation platform HEFT was developed, which is a 3-D core fuel management transport calculation program of HFETR. In HEFT, the lattice code HEFT-lat has the abilitiesto describe complex assembly geometriesof HFETR, whereby mitigate errors brought by geometrical approximations.DNTR, a triangular mesh nodal SN method code, is employed to solve the transport equation. This paper introduces the development backgrounds of HEFT, introduces the principle and function of the three main modules of HEFT, that is HEFT-lat?HEFT-core and HEFT-int, the calculation method of lattice parameter and reactor loop calculation method are described. By follow-up calculation of 81-I to 89-II of HFETR core and comparison of intial critical keff, the control rod position during shutting down and the neutron flux, thus obtains the calculation deviation. The result shows the validity of the HEFT code, the lattice and reactor core calculation model. Thus, HEFT has already successfully applied to the fuel management calculation of HFETR.

One of the safety requirements in dry storage of spent fuel is to ensure the cladding integrity. In this regard, the understanding of the cladding mechanical performance along the storage period is indispensable, both to analyse the failure probability and to characterize the state of the cladding so that fuel management is conducted with accurate knowledge of the material conditions. The main interest is focused on cladding degrading mechanisms as creep and hydrogen related (e.g. hydrides embrittlement), which are strongly influenced by temperature. Therefore, cladding thermal characterization along dry storage is an important element to predict fuel rod mechanical performance. Cladding temperature decay models found in the literature are fuel burnup independent and they cannot be applied to storage periods longer than some decades. The goal of this work is to develop a simplified model of cladding temperature as a function of burnup that spans up to 300 years of cask storage. To do so, a methodology is established based on FLUENT steady state calculations fed by heat decay data found in the literature for different burnups (33–63 MWd/kgU). From the results, a temperature correlation as a function of burnup and out-of-reactor time has been derived. It shows an average relative error less than 2% with respect FLUENT calculations. Finally, significance of having an accurate thermal characterization of the fuel rod has been highlighted by comparing fuel rod thermo-mechanics based on the derived correlation and the one resulting from using a correlation developed by EPRI.

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal S N method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT. In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

Safety margins are particularly tight in natural uranium-fuelled CANDU reactors which are refueled on-power. During on-power refueling, the insertion of xenon-free fresh fuel bundles into the reactor core affects the reactor’s excess reactivity in such a way that this could lead to temporary power derating. It is desirable from a fuel management perspective, and to maintain safety margins to eliminate this xenon-free effect and any other power ripples such as the subsequent plutonium reactivity peak. A redesign of the CANDU NU fuel bundle with an appropriate combination of elements, with some including neutron-absorbers, could well address the issue of the xenon-free initial portion of the bundle’s irradiation and also lower the plutonium-peak that occurs shortly thereafter. This may improve the fuel utilization (by further optimizing the fuelling strategy) and provide improved safety margins (by lowering the maximum channel and bundle powers). The use of neutron-absorbers in fuel design and manufacturing has been a regular practice in Light Water Reactor fuels for more than three decades. In CANDU applications, neutron absorbers have also been considered for the conceptual Advanced CANDU Reactor and the Low Void Reactivity fuel designs, for which the fissile content is made of low enriched uranium (LEU) or MOX fuels. The application to CANDU natural uranium (NU) fuel, however, especially as burnable poisons, is a relative novel approach. The reason for this is that the neutron economy in natural uranium-fuelled CANDU reactors is a prime concern, thus the addition of extra neutron absorbers is generally shunned. In our proposed application of burnable poisons to existing CANDU NU fuel design, because of low excess reactivity for NU fuel, the amount of neutron-absorber is expected to be restricted to small quantities and in a manner whereby the poison effect is restricted to the initial period of excess reactivity of a newly inserted fuel bundle. This implies that the impact on neutron economy would be relatively minimal, but the fuel performance would be significantly improved. Small amounts and appropriate mixtures of neutron absorbers were selected (approximately 500 mg of absorbers in a CANDU fuel bundle having a nominal weight of 24 kg). Preliminary results indicate that the fuelling transient and the subsequent reactivity peak can be lowered to improve the reactor’s operating margins. A parametric study using the Los Alamos National Laboratories’ MCNP 5 and Atomic Energy of Canada Limited’s WIMS-AECL 3.1 codes is presented in this paper. Details of this project and future work are also to be discussed.

Plutonium is a significant proliferation concern as well as a major contributor to the long-term toxicity of nuclear waste. Partial incineration in PWRs with uranium-MOX fuel is often considered to mitigate these concerns. Thorium-MOX is an alternative fuel with superior material properties and higher plutonium destruction rates, as shown in multiple feasibility studies. However, the core performance and operational characteristics (e.g. discharge burn-up, feasibility of controlling the core) are ultimately dependent on the core loading pattern (LP) and burnable poison (BP) design. In this paper, the LP for Th-Pu fuel of various compositions is optimized for (1) discharge burn-up, (2) radial form factor (RFF), (3) cycle length, (4) moderator temperature coefficient (MTC), and (5) reactivity swing over cycle. Maximizing the cycle length makes the discharge burn-up and reactivity swing worse due to placement of once- and twice-burnt fuel near the core periphery. It also makes the MTC less negative. The harder neutron spectrum of Th-Pu fuel compared to conventional U fuel favours the use of distributed integral burnable poisons to control the reactivity swing over the cycle. This leads to a significant amount of dissimilarity between LPs with relatively similar performance measures, and between optimal LPs for different Pu loadings in the fuel. The RFF can vary throughout the cycle but a careful placement of the assemblies can mitigate this. The cycle reactivity swing is controlled using enriched soluble boron, which makes the MTC worse, and this constrains feasibility for high Pu loading in the fuel.

A neutronic analysis of the Massachusetts Institute of Technology Research Reactor (MITR) is performed using state-of-the-art computational tools: the continuous-energy Monte Carlo code MCNP5 and the point-depletion code ORIGEN2.2. These codes are externally coupled by the in-house code package, MCODE (MCNP-ORIGEN Coupled Depletion Program), more recently, it being extended to MCODE-FM (Fuel Management). The latter features automated input file generation, data manipulation, and post-processing of the output data for the fuel cycle analysis, so that it is used to simulate the fuel management of the MITR. MCODE-FM also has an optional criticality search algorithm to simulate control blade movement. The code validation is carried out by comparing the calculated results to experimental data. Two sets of the comparisons are made in the present paper: 1) the Xe-135 reactivity effect during the reactor start-up and shutdown and 2) the thermal and fast neutron flux in an irradiation capsule in the reactor core. Good agreements have been found. The validated MCODE-FM is therefore useful for neutronic analysis and the fuel cycle simulation of the MITR. The time dependent variation of the key parameters, viz. the control blades’ axial position (maintaining criticality) and the fissile inventory in the fuel, is presented.

A type of SCWR core design with Double-row fuel rods assembly will be introduced in this paper. The Double-row fuel rods SCWR (DL-SCWR) core is designed for reaching a thermal power of 2280 MW with an equivalent diameter of 3.16 m and length of 4.20 m. There are 249 fuel assemblies in the core, and 180 fuel rods in each assembly. The inlet coolant temperature is 280 °C and the operation pressure is 25 MPa. There are 80 fresh assemblies in the equilibrium cycle. The numerical results show that the average discharged burnup of the equilibrium is 37 GWD/tU, and the average temperature of outlet coolant is 500°C. In the equilibrium cycle, the maximum cladding surface temperature is less than 650 °C.

An in-core transmutation analysis and evaluation code, named CATE, considering in-core fine flux calculation and fine depletion process, is verified and validated in the present paper. Verification and Validation of implementations for the OECD/NEA PWR cell benchmark for actinides transmutation, IAEA PWR benchmark and infinite homogenized plate problem to confirm reliability and numerical accuracy for the code have been performed in presented paper. The numerical performance of the code system is demonstrated in the analyses of the in-core fuel management calculation. It is found that the present code system gives stability in prediction of critical concentration of boric solution and radial power distribution. Based on the verifications and validations, a preliminary LLFP transmutation pattern is calculated. Numerical results indicate that CATE can be used not only for the fuel management calculation, but also for in-core transmutation evaluation of PWR.

Currently, the advanced nuclear irradiation facility MYRRHA is under development at the Belgian Nuclear Research Centre SCK•CEN in Mol, Belgium. MYRRHA is intended as a multi-purpose machine for material and fuel irradiations, radioisotopes production, silicon doping, transmutation research, education, etc. A special effort is being made within the project to develop an in-core fuel management tool for MYRRHA. Various requirements have been imposed on the core management tool; e.g., the tool should allow a user to specify a variety of optimization objectives and constraints. Based on that, a modular design of the tool was developed that makes it independent of the particular metaheuristic method used for the optimization as well as of the reactor analysis code used for an evaluation of candidate solutions during the search. In the paper, the current version of the core management tool for MYRRHA is presented together with some results on an illustrative in-core fuel management optimization problem.

Verification and validation of AEGIS/SCOPE2, a state-of-the-art in-core fuel management system for PWRs, was conducted. Verification of implementations for processes of tabulation and reconstruction of cross sections data in the system was confirmed through benchmark problems to confirm consistency between AEGIS and SCOPE2 in terms of reactivity and pin power distribution with identical computational condition in single- and multi-assembly geometry. In validation study numerical performance of the system was demonstrated in analyses of the B&W’s critical experiment and tracking calculations for commercial PWRs in combination of ENDF/B-VII.0 library. It is found the present system gives stability in prediction of critical boron concentration and radial power distribution.

The EPR™ reactor has been designed by AREVA to support economical fuel cycles. The progress in the reactor and systems design improves the reactor safety, and allows the EPRTM reactor to support the large range of high performance fuel management strategies covering cycle length from 12 to 24 months. Different fuel management strategies with 12, 18 and 24 month cycles are described. Economic analyses are performed to illustrate the low uranium consumption and the high fuel cycle performance compared with the fuel managements implemented in most current traditional PWR reactors.

Core loading pattern design has great influence on nuclear power plant operation. An excellent core loading pattern can not only enhance operation factor, reduce operation cost, but also increase operation safety. Under the premise of nuclear safety, AP1000 first core loading pattern achieves the goal of low leakage loading by simulating the reactivity distribution of the 18-month Equilibrium Cycle design. The fuel management presented in this paper illustrates the economic performance and technical feasibility of the advanced 18-month cycle first core fuel. The advanced feature of this first core design include: the development of loading pattern that simulates the reactivity distribution of a typical low leakage reload core in order to reduced leakage, the use of radial enrichment zoning in higher enriched assemblies to lower peaking factors, and the use of axial burnable absorber zoning to improve axial power shape control and reduce axial peaking. The discussion provided in this paper demonstrates the ability of the advanced first core design operate safety and efficiently, and the core is designed with adequate peaking factor margin in both base-load and load-follow operation. Finally, this paper analyses the impact brought about by multi-enrichment on the nuclear power plant operation capacity and operation cost, and arrive at a conclusion that precise fuel cycle evaluation such that the enrich uranium and operation costs can be accurately quantified and control in a detailed economic evaluation.

The HTR-10 is a pebble bed High-temperature gas-cooled reactor (HTGR) with a nominal thermal power of 10 MW. The on-line nondestructive burn-up (BU) measurement for the spherical fuel elements with TRISO coated fuel particles is an important issue in the fuel management of the HTR-10. The HTR-10 employs an HPGe gamma spectroscopy system to determine the BU values of the discharged fuel elements, and the accuracy of the BU measurement system is a crucial issue. The calibration requirement and the absolute calibration method are discussed and analyzed in this paper. In this work, the long-lived fission product 137 Cs is demonstrated as the best BU indicator in the HTR-10, whereas the inner-calibrating source method using the ratio of the amounts of 134 Cs to 137 Cs is not suitable. The BU value of a fuel element is proportional to the intensity of 137 Cs gamma-ray and independent on the irradiation history of the fuel element, except for a small correction term related to the radioactive decay of 137 Cs. The efficiency of the measurement system is calibrated by using a 137 Cs standard source embedded into the center of a graphite sphere with the same size and the same graphite material as the fuel element, which was located at the measurement position of the fuel sphere, to eliminate the possible systematic errors from the measurement system. The difference in geometry and self-absorption between the BU measurement and the calibration measurement is corrected by applying the MCNP simulation. The error sources of the calibration results are analyzed, in which the uncertainty of the source activity (∼5%) is found to dominate. The other major error sources include the fluctuation of the uranium loading in the fuel elements, the errors of the nuclear data, the counting errors of the detector, and the correction procedure for the detection efficiency. More than 1,900 fuel elements discharged from the HTR-10 have been measured by far, and the BU values of these elements spread in the range from 8.4±0.5 GWD/t to 20±1.0 GWD/t, where the relative errors of the burn-ups consist of the calibration errors about 5.2% and the counting errors around the order of 1%. The behavior of the measured burn-ups has been discussed qualitatively.

The objective of this paper is to develop a new hybrid mutation in genetic algorithm (GA) for designing the loading pattern (LP) in pressurized water reactors. Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new hybrid mutation is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKNOGA, based on well established MAKGA code is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.

The objective of this paper is to develop a new genetic algorithm (GA) for designing the loading pattern (LP) for pressurized water reactors (PWR). Because of huge number of possible combinations for the fuel assemblies (FA’s) loading in a core, finding the optimum solution is truly a complex problem. In common genetic algorithm the mutation and crossover techniques are used to optimize an objective function but in this paper a new modified crossover along a unique technique is presented. In this study flattening of power inside a reactor core is chosen as an objective function. To obtain optimal FA arrangement, a core reload package code, MAKGA, is developed. This code is applicable for all types of PWR core having different geometries and designs with an unlimited number of FA types. The result is well improved in comparison with pattern proposed by designer.

Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a “conventional” assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical “blanket”) and the “driving” part of the core (a supercritical “seed”). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER’s/PWR’s. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30% of standard reactor. In addition, the isotopic compositions of the RTF’s and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.