End Of Cycle Research Articles

Modeling and load follow simulation of CPR1000 Nuclear Power Plant implementing Mechanical Shim control strategy

Load follow operation without changing soluble boron concentration is desirable for reasons of economic efficiency, automatic control capability and effluent reduction. Many researchers have investigated the implementations of Mechanical Shim (MSHIM) strategy to different types of Pressurized Water Reactors (PWRs) to achieve this goal. However, their studies mainly focused on reactor core calculations. This paper presents the modeling and load follow simulation of the CPR1000 Nuclear Power Plant (NPP) implementing MSHIM operation and control strategy in MATLAB/Simulink environment. The whole plant model consists of the Nuclear Steam Supply System (NSSS) and the Balance of Plant (BOP) system. The NSSS model consisting of the one-dimensional nodal reactor core model, nonequilibrium three-region pressurizer model and nonlinear, time-dependent lumped parameters U-Tube Steam Generator (UTSG) model was first developed. The BOP system model including dynamic models of key components, piping lines and valves was also developed. These models were then integrated in MATLAB/Simulink according to their coupled relationships with relevant control systems built to obtain the whole plant simulation model. Based on this model, the typical 12-3-6-3/100-50-100 daily load follow maneuvers with constant soluble boron concentrations at the Beginning of Cycle (BOC) and End of Cycle (EOC) were simulated under the MSHIM control strategy. The simulation results have demonstrated the feasibility and safety of the CPR1000 NPP implementing MSHIM control strategy to achieve daily load follow operations at both BOC and EOC. This study could offer a reference to the engineering application of MSHIM control strategy as well as whole plant simulations of PWRs.

PF205 RNA BASED IMMUNE SIGNATURES MAY PREDICT RESPONSES TO PD‐1 INHIBITION AND AZACITIDINE TREATMENT IN ACUTE MYELOID LEUKEMIA (AML): A SUBSET ANALYSIS OF A PHASE 2 STUDY

Background:Combining PD‐1/PD‐L1 blockade with hypomethylating agents (HMA) showed encouraging efficacy in relapsed/refractory (R/R) AML, but immune biomarkers of response are needed to optimize patient selection.Aims:Deciphering immune predictors of response to PD‐1 blockade in AML.Methods:We treated 70 R/R AML patients with azacitidine (AZA) (75 mg/m2 on Days 1 through 7) and nivolumab (Nivo) (3 mg/kg on Days 1 and 14 of a 28 to 35 day cycle) in a phase 2 clinical trial (Daver N et al Cancer Discovery 2018). Herein, we discuss results of 11 patients (5 responders and 6 non‐responders) who had pretreatment bone marrow (BM) NanoString RNA expression profiling of 1468 immune‐relevant genes. 17‐color established immune panel flow cytometric analysis were performed at baseline and end of cycle (EOC)1, 2 and/or 4 treatments on BMs. Results were correlated with clinical, pathological and molecular data.Results:11 patients with R/R AML, median age 65 years (range 47–73 years), 63% with adverse cytogenetics, 27% with TP53 mutation, 6/11 (55%) with prior hypomethylating agent‐based therapy had indepth flow‐cytometry and NanoString immune profiling.The median administered courses of AZA+Nivo was 3 (range 1–17). The CR/CRi rate was 45% (including 2 CR, 1 CRi, 1 CRn and 1 CRp), with a median time to response of 1.8 months (range 0.8–4.9 months). The median overall survival was 13 months with 27% patients alive at 1 year. We found significant positive correlations between proportions of T‐effector cells in baseline BM, and CD3+, CD8+, and T‐regulatory cells in EOC 1 BM (r>0.75, p < 0.01 for all). At EOC2, these correlations were no longer significant. However, there was a significant positive correlation between proportion of T‐effector cells in baseline BM and T‐regulatory cells (r = 1, p < 0.001) in EOC4 BM.Using NanoString analysis on pretreatment BM samples, we found 133 differentially expressed genes (fold change = 2, q<0.1) between responders (5/11) and non‐responders (NRs) (6/11). Pretreatment pathway analysis of differentially expressed BM genes between responders and non‐responders demonstrated enrichment of innate immune response, cytokine activity, cell adhesion, and chemotaxis pathways (q<0.05) in responders. Hallmark signaling pathway analysis also demonstrated significant enrichment of IL2‐STAT5, IL6‐JAK‐STAT3, IFN‐aand TNF signaling pathways (q<0.05) in pretherapy BM in responders compared with NRs. We then utilized z‐score distribution analysis to quantify the degree of activation of known immunologic pathways. We found signatures highly specific to neutrophils and macrophages were significantly (p < 0.05) upregulated in the pretreatment BMs of responders compared with NRs.Summary/Conclusion:Immune regulatory signatures related to IL2, IL6, IFNa, chemotaxis, neutrophils and macrophages in pretreatment BMs may be associated with augmented clinical response to PD‐1 based therapies in AML. Similar underlying pathways (primarily IL2‐STAT5 and IFNa), have consistently predicted for responses to PD‐1 inhibition in solid cancers. Further, responses to PD‐1 based therapy was higher in patients with more active innate immune system as demonstrated by macrophage and neutrophil up‐regulated signatures, suggesting a possible prerequisite of active innate immunity for optimal PD‐1 blockade responses. Identifying such predictive signatures may help select AML patients most likely to benefit from PD‐1 based approaches, further enhancing the benefit‐risk ratio with such therapies.image

IL2-STAT5 immune signatures to predict responses to PD-1 inhibition and azacitidine treatment in acute myeloid leukemia (AML): A subset analysis of a phase 2 study.

7041 Background: Combining PD-1/PD-L1 blockade with hypomethylating agents (HMA) shows encouraging preliminary efficacy in AML, but immune features predictive of response are lacking. Methods: We treated 11 relapsed/refractory (R/R) AML patients with azacitidine (AZA) and nivolumab (Nivo) in a phase 2 clinical trial ( Daver N et al Cancer Discovery 2018). Patient characteristics: median (med) age 65 years (47-73), 63% adverse cytogenetics, 27% TP53 mutated. Pretreatment bone marrow aspirates had immune-phenotypic 17-color flow analysis and NanoString RNA quantification of 1469 immune-relevant genes. Results were correlated with clinical, pathological and molecular data. Results: The median courses of AZA+Nivo administered was 3 (range 1-17). The CR/CRi rate was 45% (including 2 CR, 1 CRi, 1 CRN and 1 CRP), with a median time to response of 1.8 months (range 0.8-4.9 months). The median overall survival was 13 months with 27% patients alive at 1 year. We found significant positive correlations between proportions of T-effector cells at baseline, and CD3+, CD8+, and T-regulatory cells at end of cycle (EOC) 1 (r > 0.75, p < 0.01 for all). At EOC2, these correlations were no longer significant. However, there was a significant positive correlation between T-effector cells at baseline and T-regulatory cells (r = 1, p < 0.001) at EOC4. Using NanoString analysis, we found 105 differentially expressed genes (fold change = 1.5, p < 0.05) between responders (5/11) and non-responders (6/11) at pretreatment. IL2-STAT5, TP53 and TNF Hallmark pathways and immune response from GO gene sets were highly enriched (q < 5x10-4) in responders. We then utilized z-score distribution analysis to quantify the degree of activation of known immunologic pathways. We found that signatures highly specific to neutrophils, NK cells, T-cells and eosinophils were significantly (p < 0.05) upregulated in patients with CR compared to non-responders at pretreatment. Conclusions: Our data demonstrates that a signature suggestive of lymphocyte activation in the pretreatment BM may be associated with augmented clinical response to PD-1 based therapies. Similar underlying pathways that have consistently predicted for responses to PD-1 inhibition in solid cancers, primarily IL2-STAT5 genes, may have predictive relevance in AML. Such pretreatment flow and NanoString signatures may help select AML patients most likely to benefit from PD-1 blockade plus HMA, further enhancing the benefit-risk ratio with such therapies.

Fuel Burn-up and Radioactivity Inventory Analysis for New In-core Fuel Management of the RSG-GAS Research Reactor

A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new fuel management pattern is done so that the replacement of fuel in every operating cycle is more efficient. The new fuel replacement pattern required various safety analyzes, among others related to the fuel fraction and radioactivity inventory of the RSG-GAS. Both of those are the main elements in dose acceptance analysis for workers and communities around the reactor when the reactor operates both in normal or in abnormal conditions. Fuel burn-up and inventory analysis was performed using ORIGEN2 computer program. Inputs for ORIGEN2 are the fuel mass, the time required for one operating cycle as well as the peak power factor in each fuel in the specified fuel management pattern. The result for the 97th Core configuration (T97) is that the average burn-up fraction of each cycle is 6.79%, and the maximus fuel fraction is 52.36% for Fuel Elemet and 56.52% for Contro Elemet. It is also obtained that the largest and dangerous human inventory activity at the end of cycle (EOC) for Iodine radionuclide group is I-131 for 5.18E+04 Ci, and for the alkali metal radionuclide group Cs-137 of 7.65E+03 Ci.

Fully ceramic microencapsulated fuel in prismatic high-temperature gas-cooled reactors: Design basis accidents and fuel cycle cost

Fully ceramic microencapsulated (FCM) fuel may enhance the already strong inherent reactor safety characteristics of a conventional High-Temperature Gas-Cooled Reactor (HTGR). FCM fuel uses an SiC matrix that exhibits higher stability under irradiation with limited swelling relative to a conventional graphite matrix. Additionally, the SiC matrix is expected to exhibit improved mechanical performance relative to graphite. The feasibility of FCM-fueled HTGRs has been explored in a previous study (Lu et al., 2018), and three FCM fuel concepts were identified to be able to maintain the cycle length of the reference conventional General Atomics 350 MWt prismatic modular HTGR.In this work, the previous investigation of the reactor core safety characteristics of HTGRs fueled by these three FCM fuels was significantly extended to detailed thermal hydraulics and neutronics analysis of normal operation and design basis accident scenarios. We found that these FCM-fueled cores would have a maximum fuel temperature ∼20–40 K higher than that of the reference HTGR core at the beginning of cycle (BOC) and ∼30–60 K higher at the end of cycle (EOC) during normal operating conditions. Pressurized (P-) and depressurized (D-) loss of forced cooling (LOFC) accidents as well as the control rod withdrawal accident were explored. The FCM-fueled cores would at most have a maximum fuel temperature ∼30–50 K higher than that of the reference core in a P-LOFC accident, ∼10–30 K in a D-LOFC accident and ∼30–50 K in a rod withdrawal accident. The conclusion of the study is that FCM fuel has a small impact on the reactor performance and safety characteristics during normal operation and design basis accident conditions within the existing and generic 350 MWt prismatic modular HTGR design we considered.Additionally, a fuel cycle cost comparison was also performed in the current study. We found that the FCM fuels may be up to 40%–74% more expensive than the conventional fuel. Further analysis of the economics of FCM fuel are needed, because the uncertainties in the assumptions for this study are high.

Thorium application in the medium-sized sodium-cooled fast reactor

The report presents an analysis of the medium-sized Sodium-Cooled Fast Reactor (SFR) core with Thorium-based Mixed-Oxide fuel. The introduction of Transuranics (TRU) to the fuel was to allow long-lived nuclear waste incineration. The studied core is based on the modified Advanced Burner Reactor (ABR) 1000MWth core design, which was analysed in the OECD/NEA “Benchmark for Neutronic Analysis of Sodium-Cooled Fast Reactor Cores with Various Fuel Types and Core Sizes”. The full-core simulations with SERPENT 2.1.31 Monte Carlo computer code and ENDF library were performed, including static criticality and fuel burnup calculations for five fuel cycles. The core inventories at the Beginning of Cycle (BOC) and End of Cycle (EOC) were studied, and the impact of thorium fuel was assessed. The proposed core design is a burner reactor which uses thorium fuel. The excess core reactivity stays positive for long time despite large net consumption of transuranic elements as new fissile Uranium 233 is constantly breed from Thorium 232. Breeding of uranium allows longer fuel cycles.

Minor Actinides transmutation in candidate accident tolerant fuel-claddings U3Si2-FeCrAl and U3Si2-SiC

An advanced transmutation method is suggested that the long-lived Minor Actinides (MAs) in the spent fuel can be efficiently transmuted in the candidate Accident Tolerant Fuel (ATF). The transmutation of MAs is investigated through the Monte Carlo simulations in two potential fuel-claddings of ATF, U3Si2-FeCrAl and U3Si2-SiC. The critical loadings of MAs are determined through the Linear Reactivity Model (LRM) in order to keep the same reactivity as the current UO2-zircaloy system at the End of Cycle (EOC). In all cases, excellent transmutation efficiencies are found for the most important three MAs, 237Np, 241Am, and 243Am, of which the total transmutation rates are around 60%, 90%, and 60%, respectively. If only the longest-lived isotope 237Np is considered, one U3Si2-SiC assembly can transmute 237Np from six normal assemblies. The loading of MAs has little influences on the neutronic properties, such as the power distributions inside an assembly and inside a fuel rod. The transmutation of MAs in the ATF assembly is shown to be more efficient and safe comparing with the normal assembly, while other important properties are kept, such as the cycle length and the power distribution.

Neutronic Study of Burnable Poison Materials and Their Alternative Configurations in Fully Ceramic Microencapsulated Loaded Pressurized Water Reactor Fuel Assembly

Use of FCM fuel in light water reactors is an attractive option for existing and future generations of these reactors to make them accident tolerant in nature. This work focuses on the neutronic study of the use of burnable material in various configurations to control the excess reactivity and to keep the moderator temperature coefficient of reactivity (MTC) feedback negative for entire cycle length. Erbia and gadolinia, two conventional materials are used in three different configurations including quadruple isotropic (QUADRISO), bi-isotropic (BISO), and Matrix Mix forms. The results obtained from the implicit random treatment of the double heterogeneity of tri-structural isotropic (TRISO), QUADRISO, and BISO particles show that the erbia is the best material to be used in QUADRISO and Matrix Mix configurations with lowest reactivity swing for the life cycle and residual poison well below 0.5%. Gadolinia is usable in FCM environment only in the BISO form where enhanced self-shielding controls the depletion performance of the material. The gadolinia has almost zero residual poison at end of cycle (EOC); however, it has relatively large reactivity swing, which will need more micromanagement of the control rods during the plant operations. At the beginning of cycle (BOC), erbia-loaded assemblies have shown an increase in negative value of MTC compared with reference due to presence of resonance peak in erbium near 1 eV. The finally recommended material-configuration combinations have shown the excess reactivity containment in desired manner with good depletion performance and negative feedback of the MTC for life cycle.

Transient analysis of MOX-3600 and MET-1000 sodium-cooled fast reactor using SARAX code system

The capability of transient analysis has been implemented in the SARAX code system recently. Alternatively, two calculation schemes using the point-kinetics model and 3-D transport-based spatially-dependent model have been developed. Based on these, two transient benchmarks of sodium fast reactor (SFR) suggested by the Working Party on Reactor and System (WPRS) were calculated, which include the large size core loaded with oxide fuel (MOX-3600) and the medium size core loaded with metallic fuel (MET-1000). Two typical states including the state of begin-of-cycle (BOC) and state of end-of-cycle (EOC) were considered. The unprotected loss of flow (ULOF) transient, unprotected transient over power (UTOP) transient, and unprotected control rod withdrawal (UCRW) were calculated. For the large size MOX-fueled SFR, the ULOF transient is the most severe accident, which leads to the core damage in a very short time. At the EOC state, the thermal margin is smaller because of the change of fuel compositions due to the depletion. Besides, the comparisons show that it is very important to use the spatially-dependent model in predicting the power level in the UCRW transient, where the point-kinetics model significantly underestimates the power level by 7–17%.

Neutronics analysis of CNPP-II loaded with 1/3rd MOX fuel

This paper performs the neutronics analysis of Chashma Nuclear Power Plant, Unit-II (CNPP-II) loaded with 1/3rd Mixed Oxide (MOX) fuel employing a burnup code Linkage of ORIGEN2.2 and OpenMC (LOOP). This is a PWR type reactor employed with UO2 fuel and is selected as a benchmark. The appropriate MOX composition which can give the same cycle length as that of pure UO2 core is found using linear reactivity model. The best suitable placement of MOX fuel is found to be at the periphery of the core. The important neutronics safety parameters for 1/3rd MOX fueled core at the Beginning of Cycle (BOC), Mid of Cycle (MOC) and End of Cycle (EOC) are compared with pure UO2 core of CNPP-II. The parameters include reactivity coefficients, control rods worth, shut down margin, peaking factors and effective delayed neutron fraction. It is observed that with slight decrease in the effective delayed neutron fraction, the MOX fuel can safely be used in the PWR type SMR reactor.

Burn-up characteristics and criticality effect of impurities in the graphite structure of a commercial-scale prismatic HTGR

This study investigates the burn-up characteristics and the criticality effect of impurities in the graphite structure of commercial-scale prismatic High Temperature Gas-cooled Reactor (HTGR), and thereby reconsiders the necessity of high-grade graphite material. In an HTGR, the core is filled with the graphite, and the impurities in the graphite have a non-negligible poison effect on the criticality. To account for the effect of the reflector blocks deployed adjacent to the fuel blocks, GTHTR300, commercial-scale HTGR, employed fine purified grade graphite material IG-110. Ideally, the fuel blocks should also employ IG-110; however, for economic purposes they are constructed from an un-purified grade graphite material IG-11. The poisoning effect of the impurity (which behaves like 10B burn-up and is expressed in boron equivalents) decreases exponentially and eventually saturates at 1% of the initial boron equivalent. However, the reactivity worth of the fuel and reflector blocks with 0.03 ppm boron equivalents (equivalent to 1% of IG-11) is negligible (i.e., <0.01%Δ k/kk′). Because the poisoning effect of the impurity mimics that of naturally occurring boron, it was evaluated in whole-core burn-up calculations with the impurities represented by naturally occurring boron.According to the results, the criticality of the commercial-scale HTGR is unaffected by the impurity levels (even in the un-purified grade IG-11) because the impurities burn cleanly until the End of Cycle (EOC). Therefore, the economy of electricity generation by HTGRs can be improved by using the un-purified grade IG-11 instead of the fine purified grade graphite IG-110.

Validation of LOOP through PWR measured data

This paper serves two purposes. First it demonstrates the validation of newly developed and verified burnup code, Linkage of ORIGEN and OpenMC (LOOP) against measured data of Post Irradiation Experiments (PIE) performed at Takaham-3 PWR. Secondly, it assesses the qualification of OpenMC model of Chashma Nuclear Power Plant unit-II (CNPP-II) PWR to predict different operating conditions of the reactor core. The LOOP code is then employed to perform the burnup simulations of CNPP-II first cycle with fresh UO2 fuel. The LOOP calculated results of power distribution of the first cycle at Beginning of Cycle (BOC), Middle of Cycle (MOC) and End of Cycle (EOC) are then compared with the operational data of the reactor. The maximum difference between the LOOP calculated radial and axial power distribution results and the measured data is 9.1% and 2.3% at BOC. At EOC, this difference further decreases to 7.2% and 1.5%. The depletion of burnable poison rods in the reactor was also compared with the FSAR predicted data which are in exact agreement with each other. The production of Cs-137 and Cs-134 was also investigated for the validity of LOOP burnup simulations.

Simplified load follow schemes to simulate long-term daily load follow operation

Simulation of Daily Load Follow Operation (DLFO) for a whole cycle has long computation time as well as tedious lengthy input files. The purpose of this study is to reduce the computation burden associated with long-term DLFO simulation by developing a Simplified Load Follow Scheme (SLFS) simulation that can reproduce the core characteristics of long-term DLFO simulation (reference scheme) at End of Cycle (EOC). To achieve this aim, the SLFS employs two constraints: energy production and cycle length must equal those of the reference scheme. The solution variables in a SLFS then becomes operation time and power level satisfying the two constraints. In this paper, low power levels of 50%, 63%, 75%, and a maximum power level of 100% are employed for the SLFSs. Operation times are determined to satisfy the two constraints for the given power level. The reference scheme in this paper employs OPR1000 Mode-K daily load follow scheme for the Small Modular Reactor (SMR).The results show the SLFSs can predict Core Radial Plane Burnup (CRPB) and Core Radial Plane Power (CRPP) distributions of a long-term DLFO with errors lower than 6% across the core height. Axial Offset (AO) and Control Element Assembly (CEA) critical height of the SLFSs are also comparable with the reference scheme with low errors of 1.5% and 4.5% respectivelyat EOC. The computational time of the simplified schemes is reduced by more than 29 times than that of the reference scheme. Single Control Element Assembly (SCEA) Ejection is simulated to validate the applicability of the simplified schemes. The SCEA ejection simulations also show CRPP distribution errors less than 6%. In conclusion, the SLFSs can be used to simulate long-term DLFO with marginal errors of less than 6%.

A semi-analytical treatment of xenon oscillations

A novel approach is developed to investigate xenon oscillations within a two-group two-region coupled core reactor model incorporating thermal feedback and poison effects. Group-wise neutronic coupling coefficients between the core regions are calculated applying the associated fundamental and first mode eigenvalue separation values. The resultant nonlinear state space representation of the core behavior is quite suitable for evaluation of reactivity induced power transients such as load following operation. The model however comprises a multi-physics coupling of sub-systems with extremely distant relaxation times whose stiffness treatment inquire costly multistep implicit numerical methods. An adiabatic treatment of the sluggish poison dynamics is therefore proposed as a way out. The approach helps further investigate the nonlinear system within a linear time varying (LTV) framework whereby a semi-analytical framework is established. This scheme incorporates a matrix exponential analytical solution of the perturbed system as a quite efficient tool to study load following operation and control purposes. Poison dynamics are updated within larger intervals which exclude the need for specific numerical schemes of stiff systems. Simulation results of the axial offset conducted on a VVER-1000 reactor at the beginning (BOC) and the end of cycle (EOC) display quite acceptable results compared with available benchmarks. The LTV reactor model is further investigated within a stability analysis of the associated time varying systems at these two stages employing the concept of Lyapunov exponent.

Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

New techniques for designing the initial and reload cores with constant long cycle lengths

Several utilities have increased the output power of their nuclear power plant to increase their income and profit. Thus, the utility increases the power density of the reactor, which has other consequences. One consequence is to increase the depletion of the fuel assemblies (FAs) and reduce the end-of-cycle (EOC) sum of fissionable nuclides in each FA, ∑EOC. The power density and the ∑EOC remaining in the FAs at EOC must be sufficiently large in many FAs when designing the loading pattern, LP, for the first and reload cycles to maintain constant cycle lengths at minimum fuel cost. Also of importance is the cycle length as well as several other factors. In fact, the most important result of this study is to understand that the ∑EOCs in the FAs must be such that in the next cycle they can sustain the energy during depletion to prevent too much power shifting to the fresh FAs and, thus, sending the maximum peak pin power, PPPmax, above its constraint. This paper presents new methods for designing the LPs for the initial and follow on cycles to minimize the fuel costs. Studsvik’s CMS code system provides a 1000MWe LP design in their sample inputs, which is applied in this study. The first 3 cycles of this core are analyzed to minimize fuel costs, and all three cycles have the same cycle length of ∼650days. Cycle 1 is designed to allow many used FAs to be loaded into cycles 2 and 3 to reduce their fuel costs. This could not be achieved if cycle 1 was a low leakage LP (Shi et al., 2015). Significant fuel cost savings are achieved when the new designs are applied to the higher leakage LP designs. There are many factors, such as the core power density, cycle length, fuel cost, time between core shutdown and return to power, cost of replacement power during shutdown, loss of income during shutdown, cost of storing used FAs, and the income accrued over the same period of operation, which the utilities must consider when trying to increase their profit. The purpose of this paper is to provide the information to give guidance in making these decisions. Relative cost calculations are presented in establishing this guidance by comparing the utility’s profit accrued over a total cycle for the different core designs and cycle lengths.

Impact of the thermal scattering law of H in H2O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

The contribution of the thermal scattering law of hydrogen in light water to isothermal temperature reactivity coefficients for UOX and MOX lattices was studied in the frame of the MISTRAL critical experiments carried out in the zero power reactor EOLE of CEA Cadarache (France). The interpretation of the core residual reactivity measured between 6 °C to 80 °C (by step of 5 °C) was performed with the Monte-Carlo code TRIPOLI4® . The nuclear data from the JEFF-3.1.1 library were used in the calculations. Three different thermal scattering laws of hydrogen in light water were tested in order to evaluate their impact on the MISTRAL calculations. The thermal scattering laws of interest were firstly those recommended in JEFF-3.1.1 and ENDF/B-VII.1 and also that recently produced at the atomic center of Bariloche (CAB, Argentina) with molecular dynamic simulations. The present work indicates that the calculation-to-experimpental bias is −0.4 ± 0.3 pcm/°C in the UOX core and −1.0 ± 0.3 pcm/°C in the MOX cores, when the JEFF-3.1.1 library is used. An improvement is observed over the whole temperature range with the CAB model. The calculation-to-experimpental bias vanishes for the UOX core (−0.02 pcm/°C) and becomes close to −0.7 pcm/°C for the MOX cores. The magnitude of these bias have to be connected to the typical value of the temperature reactivity coefficient that ranges from −5 pcm/°C at Begining Of Cycle (BOC) up to −50 pcm/°C at End Of Cycle (EOC), in PWR conditions.

Burn up Effects on Kinetic Parameter’s Variations in an Electron Accelerator Driven Subcritical Reactor

One of the key components in nuclear reactors is calculation of the neutronic parameters and the burnup effect on the variation of these parameters. In this study, neutronic parameters such as neutron flux and power distributions, effective multiplication factor (Keff), neutron mean generation time (Λ), effective delayed neutron fraction (βeff) and fuel burn up are calculated in an Accelerator Driven Subcritical ALMR reactor to transmutation of TRUs by MCNPX code for the parabolic spatial distribution of electron beam. The results this study shows that with increasing in the fuel burn up the values of Keff and TRU decrease while the fission products rate increase. In addition, it observes that in comparison with the beginning-of-cycle (BOC) values, at end-of-cycle (EOC), the Λ parameter increases but the βeff parameter more or less constant as a function of burn up in the multiplication factor of 0.93. Also, according to the results, the average relative difference between the results of calculations with reference data is within 3%.

New burnable absorber for long-cycle low boron operation of PWRs

This paper presents a new high performance burnable absorber (BA) design for advanced Pressurized Water Reactors (PWRs) aiming for a long-cycle operation with a low soluble boron concentration. The new BA consists of a UO2–157Gd2O3 rod covered with a thin layer of Zr167Er2. A key feature of this new BA is that enriched isotopes, 157Gd and 167Er, are used as absorber materials. Since the high absorption cross section of 157Gd can reduce the mass fraction of Gd2O3 in UO2–Gd2O3, the thermal margin of fuel rods will increase with higher heat conductivity. Also, the 157Gd transmutes into 158Gd by neutron absorption and therefore the residual penalty at the end of cycle (EOC) will decrease. Since 167Er has a resonance near the thermal neutron energy region, the moderator temperature coefficient (MTC) will become more negative and the control rod worth will increase. These advantages of the new BA are demonstrated with three verification cases: a 17×17 Westinghouse (WH) type fuel assembly, a 16×16 Combustion Engineering (CE) type fuel assembly, and an OPR-1000 equilibrium core.

Study on erbium loading method to improve reactivity coefficients for low radiotoxic spent fuel HTGR

Erbium loading methods are investigated to improve reactivity coefficients of Low Radiotoxic Spent Fuel High Temperature Gas-cooled Reactor (LRSF-HTGR). Highly enriched uranium is used for fuel to reduce the generation of toxicity from uranium-238. The power coefficients are positive without the use of any additive. Then, the erbium is loaded into the core to obtain negative reactivity coefficients owing to the large resonance the peak of neutron capture reaction of erbium-167. The loading methods are attempted to find the suitable method for LRSF-HTGR. The erbium is mixed in a CPF fuel kernel, loaded by binary packing with fuel particles and erbium particles, and embedded into the graphite shaft deployed in the center of the fuel compact.It is found that erbium loading causes negative reactivity as moderator temperature reactivity, and from the viewpoint of heat transfer, it should be loaded into fuel pin elements for pin-in-block type fuel. Moreover, the erbium should be incinerated slowly to obtain negative reactivity coefficients even at the End Of Cycle (EOC). A loading method that effectively causes self-shielding should be selected to avoid incineration with burn-up. The incineration mechanism is elucidated using the Bondarenko approach.As a result, it is concluded that erbium embedded into graphite shaft is preferable for LRSF-HTGR to ensure that the reactivity coefficients remain negative at EOC.