Temperature Coefficient Of Reactivity Research Articles

Neutronic benchmark of the molten salt fast reactor in the frame of the EVOL and MARS collaborative projects

This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with 233U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.

Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

Study on Temperature Coefficient of Reactivity for Pebble Bed Reactor with Thorium Fuel

Study on Temperature Coefficient of Reactivity for Pebble Bed Reactor with Thorium Fuel

Development and verification of resonance elastic scattering kernel processing module in nuclear data processing code NECP-Atlas

Some efforts have been made to exactly consider the effect of neutron up-scattering caused by thermal motion of target nuclei and resonance elastic scattering on the multi-group cross sections and scattering matrices. Firstly, the resonance elastic scattering kernel (RESK) formulations for anisotropic scattering up to any Legendre order has been adopted to represent the exact Doppler broadened energy transfer kernels. A semi-analytical integration method is applied to perform the RESK calculations. Combining with the RESK calculations, a linearization algorithm is proposed to generate the RESK interpolation tables. The RESK data can be interpolated precisely based on the interpolation tables to reduce the calculation burden. Secondly, a neutron slowing-down equation solver is developed based on the RESK instead of the conventional asymptotic scattering kernel, where the effect of neutron up-scattering on the neutron energy spectrum can be exactly considered. The accuracy of the multi-group cross sections and scattering matrices are significantly improved when the exact energy spectrum is applied for the group collapsing calculations. All the methods mentioned above have been implemented into a newly developed nuclear data processing code called NECP-Atlas. Numerical results show that the proposed methods are capable of producing accurate multi-group cross sections for downstream calculations; the fuel Doppler coefficients and eigenvalues of the real problems will be improved if the up-scattering effect is incorporated into the multi-group cross sections and matrices.

Estimation of the Reactivity Temperature Coefficients of the AP-1000 Reactor by MCNPX Code

The estimation of temperature coefficients of reactivity is a very important task because of its relationship to the safety of reactor operation. The temperature coefficient of reactivity includes the Moderator Temperature Coefficient (MTC), Doppler temperature coefficient (DTC) and the isothermal temperature coefficient (ITC). The present work investigated the temperature coefficients of reactivity for the advanced first core of AP1000 PWR. The calculations were carried out using Monte Carlo code MCNPX and the Evaluated Neutron Data File library, ENDF/B-VII.1 (ENDF71x), cross section data. The calculated results have been compared with the same results of KENO and VERA-CS codes. The study was investigated at a boron concentration of 1300 ppm. The MCNPX values of the MTC, DTC, and ITC were observed to be: (-1.27), (-1.48), (-2.75) pcm/F respectively. The results showed the compatibility between the MCNPX and KENO and VERA codes. Also the results demonstrate that the MCNPX code showed good estimates of Temperature Coefficients of Reactivity of AP-1000 PWR. The results can be used as reference for reactor physics studies.

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.

Coupled Experimental and Computational Approach for CABRI Power Transients Analysis

CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute and operated by Commissariat a l’Energie Atomique et aux Energies Alternatives at the Cadarache Research Center. It is designed to study fuel behavior under reactivity-initiated accident conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in the so-called transient rods inside the reactor core. The CABRI reactivity injection system allows us to generate structured transients based on specific sequences of depressurization. For such transients, the time difference between the openings of two valves of the reactivity injection system has an important impact on the power pulse shape. A kinetic point code, SPARTE, was developed in order to replace the older DULCINEE code dedicated to the modeling and prediction of CABRI power transients. The SPARTE code includes new models of 3He depressurization based on CFD calculations, variable Doppler coefficient based on Monte Carlo calculations, and variable axial neutron flux profile. The density and Doppler models have a large impact on power transient prediction. For low initial pressure transients, the major uncertainty comes from the reactivity injected by the 3He depressurization. For high initial pressure transients, the 3He heating during the power pulse (“TOP effect”) is responsible of an additional injection of reactivity that needs to be modeled precisely.

Neutronic analysis of fuel assembly design in Small-PWR using uranium mononitride fully ceramic micro-encapsulated fuel using SCALE and Serpent codes

Abstract One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than UO2 pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of 14N. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to 235U enrichment ratio compared to the UO2/Zr fuel.

A temperature effect analysis of the KRITZ-1 benchmark based on keff decomposition and using the JENDL-4.0 and ENDF/B-VII.1 libraries

This paper provides a comparison between multiplication factors within the UO2 fuelled cores that were measured during the original KRITZ-1 experiments performed in the early 1970s. Results were calculated using the MCNP6.1 code and the most recent cross section libraries, ENDF/B-VII.1 and JENDL-4.0. The results of this study reveal a good level of agreement between calculated and measured outcomes; indeed, the maximum relative error of calculated keff from experimental data does not exceed 0.5% (absolute value) for all cores and for both evaluations. The reactivity temperature coefficient, expressed as the temperature dependence of the reactivity, was examined into details using the seven factors that incorporate the keff parameter rather than just the five that exist in the literature. In order to better investigate the influence of differences in neutron cross sections we also quantified the temperature effect on the five factors of the infinite multiplication factor on each one separately by admitting a pin cell model.Calculated thermal and fast non-leakage probabilities using JENDL-4.0 are found to overestimate the ones calculated using ENDF/B-VII.1 for all temperatures, possibly because the JENDL-4.0 library overestimates some absorptions, especially in the resonance region of fissionable nuclides. This discrepancy between the libraries decreases with increasing temperature, however, while the inverse tends to occur in the case of standard deviations. Our assessment has shown that the temperature coefficient in the thermal temperature range is linked to thermal spectrum and water density effects, while within the epithermal temperature range this is strongly dependent on the thermal shapes of the cross sections of the uranium isotopes; the resonance escape probability highlights that the error is mainly due to the Doppler broadening effect.

Effects of fuel salt composition on fuel salt temperature coefficient (FSTC) for an under-moderated molten salt reactor (MSR)

With respect to a liquid-fueled molten salt reactor (MSR), the temperature coefficient of reactivity mainly includes the moderator temperature coefficient (MTC) and the fuel salt temperature coefficient (FSTC). The FSTC is typically divided into the Doppler coefficient and the density coefficient. In order to compensate for the potentially positive MTC, the FSTC should be sufficiently negative, and this is mostly optimized in terms of the geometry aspect in pioneering studies. However, the properties of fuel salt also directly influence the FSTC. Thus, the effects of different fuel salt compositions including the $$^{235}$$ U enrichment, heavy metal proportion in salt phase (HM proportion), and the $$^{7}$$ Li enrichment on FSTC are investigated from the viewpoint of the essential six-factor formula. The analysis is based on an under-moderated MSR. With respect to the Doppler coefficient, the temperature coefficient of the fast fission factors ( $$\alpha _{\text {T}}(\varepsilon )$$ ) is positive and those of the resonance escape probability ( $$\alpha _{\text {T}}(p )$$ ), thermal reproduction factor ( $$\alpha _{\text {T}}(\eta )$$ ), thermal utilization factor ( $$\alpha _{\text {T}}(f )$$ ), and total non-leakage probability ( $$\alpha _{\text {T}}(\varLambda )$$ ) are negative. With respect to the density coefficient, $$\alpha _{\text {T}}(p )$$ and $$\alpha _{\text {T}}(\eta )$$ are positive, while the others are negative. The results indicate that the effects of the $$^{235}$$ U enrichment and HM on FSTC are mainly reflected in $$\alpha _{\text {T}}(\varepsilon )$$ and $$\alpha _{\text {T}}(p )$$ , which are the dominant factors when the neutron spectrum is relatively hard. Furthermore, the $$^{7}$$ Li enrichment influences FSTC by $$\alpha _{\text {T}}(f )$$ and $$\alpha _{\text {T}}(\varLambda )$$ , which are the key factors in a relative soft spectrum. In order to obtain a more negative FSTC for an under-moderated MSR, the possible positive density coefficient, especially its $$\alpha _{\text {T}}(p )$$ , should be suppressed. Thus, a lower $$^{235}$$ U enrichment (albeit higher than a certain value, 5 wt% in this article) along with a lower HM proportion and/or a higher $$^{7}$$ Li enrichment are recommended. The analyses provide an approach to achieve a more suitable fuel salt composition with a sufficiently negative FSTC.

Studies on Reactivity Coefficients of Thorium-Based Fuel (Th-233U)O2 with Molten Salt (Flibe) Cooled Pebble

A combination of the neutronics features of gas-cooled high-temperature reactors by using the fuel in the form of ceramic-coated particles, called tristructural-isotropic, and the heat removal feature of molten salt reactors by using molten salt as a coolant is an attractive option in designing a reactor with a high-power density operation without compromising the safety aspects. Neutronics feasibility of such a combination of the molten salt (LiF-BeF2) as a coolant and thorium-based fuel, in particular (Th-233U)O2, in a graphite-moderated system is investigated. This technical note presents the influence of the heavy metal (HM) loading on neutronics features of a pebble lattice cell, that is, infinite multiplication factor (K-inf), temperature coefficients of reactivity (TCR), the void reactivity coefficient, etc. In addition, enriched uranium fuel has also been studied just to make a comparison with thorium-based fuel. Furthermore, the minimum HM loading of fuel per pebble that is needed to achieve negative coolant-temperature reactivity coefficients and void reactivity coefficients has been estimated for molten salt coolant.The analyses show that Th2/U3 fuel gives a less negative fuel temperature reactivity coefficient as compared with that of uranium-based fuel. This study also shows that all the TCR of both fuel types improve, becoming less positive or more negative, by increasing HM loading per pebble. Further, the burnup dependence of K-inf and the reactivity coefficients are studied for limiting HM loadings, e.g., 30 g per pebble. The change in the spectrum and the four-factor formula are used to explain the behavior of the reactivity coefficients as a function of HM loading and burnup.

Influence of Coolant 6Li Concentration on the Neutronic Parameters of Lithium-Cooled Reactor

Lithium is an attractive coolant for space nuclear reactors due to its good thermal properties and low density. The main objective of this study is to investigate the effect of coolant 6Li concentration on the neutronic parameters of the lithium-cooled space reactor, with the aim to provide an appropriate reference for the purification of lithium coolant and the safety of the lithium-cooled space reactor. The neutronic calculations based on the lithium-cooled reactor are performed using the SuperMC code with the ENDF/B-VII cross-section database. The effects of coolant 6Li concentration were studied over the range of 0 to 8.5 at. % as well as the corresponding effects on the depletion, helium production, neutron spectrum, and temperature reactivity coefficient. The results show that the 6Li concentration of 0.01 at. % in the lithium coolant is appropriate in the lithium-cooled reactor. In this case, the neutronic parameters including the depletion, helium production, and neutron spectrum showed no obvious change compared to that of the pure 7Li, and the coolant reactivity coefficient has a negative effect on reactivity.

Fuel pebble optimization for the thorium-fueled Pebble Bed Fluoride salt-cooled high-temperature reactor (PB-TFHR)

Thorium-fueled Pebble Bed Fluoride salt-cooled High-temperature Reactor (PB-TFHR) is a newly developed reactor concept featuring the technology of High Temperature gas-cooled Reactor (HTR) and the molten salt cooling for thorium-based fuels. To attain a deep burnup with enough negative temperature coefficient of reactivity (TCR), a systematic study for the graphite-to-heavy metal ratio (C/HM) from 66 to 800 and the 233U-to-HM ratio (233U/HM) from 5.0% to 20.0% are carried out. Neutronics characteristics, including the effective multiplication factor (keff), the TCR and the conversion ratio (CR) are analyzed in terms of neutron usage, core safety and burnup. The results show that the fuel pebble with a higher 233U/HM owns a wider range of under-moderated area, which is favorable to improve TCR. Moreover, a strongly negative TCR is obtained by optimizing the C/HM and 233U/HM ratios. In order to evaluate the 233U saving for thorium based fuel, the burnup per fissile mass (Bu) is introduced. Bu can be improved through reducing C/HM, and it increases firstly and then achieves saturation as 233U/HM increases. A burnup level of 122 GWd/tHM can be achieved for the recommended pebble design with 233U/HM of about 12.5% and C/HM of about 124. With the optimized design, an improved TCR can be obtained with a value of about −2.77 pcm/K, which is much more superior to our previous work.

Haar wavelet for solving the inverse point kinetics equations and estimation of feedback reactivity coefficient under background noise

A new formalism is presented in this paper for solving the inverse point kinetics equations with six groups of delayed neutron precursors using Haar wavelet and estimation of feedback reactivity coefficient from the observed power transient under background noise. The Haar wavelet transforms the inverse point kinetics equations into a set of linear equations and these equations can be solved easily. Using this method the reactivity required for a desired power transient is obtained and also the feedback reactivity and the temperature coefficient of reactivity involved in the observed power transient are estimated for various background noise levels. This method is tested in two ways, i.e. (i) to estimate the reactivity required for different types of power transients in thermal reactor as well as in Indian Prototype Fast Breeder Reactor (PFBR) and (ii) to estimate the feedback reactivity and the temperature coefficient of reactivity involved in the power transients of thermal as well as Indian PFBR. In the case of Indian PFBR, the temperature coefficient of reactivity is estimated for various background noise levels in power transients. It is observed that as the noise level is reduced, the accuracy in the estimation of temperature coefficient of reactivity is increased. It is also shown that using Haar wavelet with beamforming, the temperature coefficient of reactivity can be estimated to a good accuracy even under high background noise. In this method the estimated feedback reactivity and the temperature coefficient of reactivity are found to be in good agreement with reference values. From the comparison of results it is observed that this method is efficient in estimating the reactivity required for different types of desired power transients in thermal as well as in fast reactors and this method is also efficient in estimating the temperature coefficient of reactivity under high background noise. This method is effective and simple to use.

Optimization of Th-U fuel breeding based on a single-fluid double-zone thorium molten salt reactor

Molten salt reactor (MSR) shows great promise of high thermal-electric conversion efficiency, inherent safety, and on-line reprocessing. Furthermore, fuel breeding can be realized in both thermal and fast spectrum reactors with Th-U fuel cycle, which can make use of the abundant resource of thorium.In order to combine the advantages of high breeding ratio (BR) in fast spectrum and low fissile inventory for criticality in thermal spectrum, a single-fluid double-zone thorium-based molten salt reactor (SD-TMSR) core configuration is proposed. First, assemblies in both the inner and the outer zones are optimized by adjusting the ratios of molten salt and graphite with consideration of BR, 233U inventory, double time (DT), and temperature coefficient of reactivity (TCR) at the startup time, aiming to have a minimal DT for fuel breeding and a negative TCR for security. The results show that DT has the optimum value when the ratios of molten salt and graphite in the inner zone and the outer zone are 0.357 and 1.162, respectively. And the TCR can be improved to −2 pcm/K when the side length of the graphite hexagonal prism is 7.5 cm. Then, based on the optimized geometry, the burn-up calculations with a series of on-line reprocessing rates are carried out with a self-developed MSR reprocessing sequence (MSR-RS). The results show that on-line reprocessing is beneficial to the Th-U fuel breeding. When the reprocessing rate is 200 l/d, iso-breeding can be achieved. When the reprocessing rate is 5 m3/d, the performance of Th-U breeding is improved significantly, and only 16 years is needed for the 233U doubling. The results also show that TCR remains negative and inherent safety is satisfied during all the operation time.

Neutronic design study of a small modular IPWR loaded with accident tolerant-fully ceramic micro-encapsulated (AT-FCM) fuel

Small modular reactors (SMRs) offer a green energy option that will be suitable to serve smaller energy markets with less developed infrastructure. In the present work, a neutronic model of a near term deployable SMR of a natural circulation integral pressurized water reactor (IPWR) design is developed using the TRISO type fuel embedded in SiC matrix. The model development is done using the Monte Carlo Code capable of treating the double heterogeneity in implicit random manner. The Low Enriched Uranium Carbide (LEUC) fuel in TRISO form embedded in SiC matrix and cladded with FeCrAl forms a perfect combination for an accident tolerant fuel for light water reactors due to enhanced capability of fission-product retention of TRISO fuel and lower hydrogen production by FeCrAl at elevated temperature. Length of assembly is altered for reduced active height suitable for SMR application. The depletion analysis is performed for the optimization of fuel enrichment and core loading pattern to achieve a uniform core power distribution. The optimized core is capable of producing 160 MW(th) power with a core life time of more than four EFPYs without refueling or shuffling. The Erbia (Er2O3), 80 w/o enriched in 167Er as an integral fuel burnable absorber (IFBA) in QUADRISO form, shown excellent depletion performance and control of power peaking throughout the life cycle. This study showed that Doppler, moderator temperature and coolant void reactivity coefficients are all negative over the core lifetime. Transuranic production is low, with relatively higher 240Pu content in the spent fuel. This leads to a highly proliferation resistant discharge fuel. A new design of RCCA loaded with HfB2 absorber is capable of providing the required shutdown margin in all core conditions. Therefore, the SM-IPWR concept satisfies completely the design criteria and the requirements of an ATF loaded PWR.

Advanced micro-reactor concepts

Nuclear power systems capable of outputting low powers (<100 MWth) are increasingly receiving interest internationally for deployment not only as electricity production systems, capable of operating off-grid, but also as systems able to provide industrial process heat. These ‘micro-reactor’ concepts must demonstrate economic competitiveness with other potential solutions capable of providing similar power outputs. With this in mind, reactor technologies that offer inherent advantages associated with improved power density and simplified operation, both of which are important attributes that determine economic competitiveness, are reviewed in the context of the fundamental safety functions provided by the IAEA.The reactor technology chosen based on the results of the review were: low vapour pressure coolants like molten salt or liquid metal; solid moderator material; and conventional solid UO2 fuel. Initial infinite lattice neutronic studies indicated a series of positive reactivity coefficients. A finite system was also modelled using a molten salt as the coolant. When modelling the finite system the coolant temperature reactivity coefficient became negative, the void coefficient strongly negative and moderator temperature coefficient negative to weakly positive. Given that a number of reactivity coefficients were negative to strongly negative in the finite system, the weakly positive moderator temperature coefficient is not thought to be prohibitive. Thus the design should exhibit acceptable safety performance.Whilst the importance of leakage in fast reactor cores is well known, a key outcome from this study is the strong influence of leakage on all safety related parameters for the thermal reactor designs considered here with solid moderator material. Thus it seems that safety studies for such small cores should be based on full core calculations instead of the traditional infinite lattice studies for fuel assemblies.

Calculation of core safety parameters and uncertainty analyses during unprotected transients for the ALLEGRO and a sodium-cooled fast reactor

The calculation of the essential core safety parameters in fast spectrum reactors are loaded with large uncertainties due to the particularly large uncertainty of nuclear data, modelling uncertainties and possible mistakes by users. These core safety parameters such as coefficients of reactivity, power peaking factors and point-kinetics parameters play an important role during unprotected transients. The contributions to the uncertainties of the target parameters (e.g. peak cladding temperature) from the core safety parameters can be determined by uncertainty and sensitivity analyses.Three-dimensional models of the ALLEGRO (demonstrator gas-cooled fast reactor) and a sodium-cooled fast reactor (SFR) concept were created in the Serpent Monte Carlo code, and several coefficients of reactivity were calculated. Furthermore, thermal hydraulic analyses with point kinetics were made for the selected transients by the ATHLET code system. An unprotected loss of flow and an unprotected overpower transient were studied at first using the best estimate approach to understand the behaviour of the reactors in these scenarios. In addition, uncertainty analyses were performed applying the GRS method, considering the uncertainties of the coefficients of reactivity, which were taken from benchmark results. Finally, the importance of the uncertain parameters was determined from sensitivity measurement.The results show that the thermal expansion of the core structural elements has a significant effect on the reactivity of these fast spectrum reactors. Moreover, it was found that during unprotected overpower transients the most important uncertain parameters are the Doppler coefficient and the fuel thermal expansion coefficient of reactivity. In addition, the large uncertainty of the positive coolant and cladding temperature coefficients of reactivity have a significant influence on the uncertainty of the calculated peak cladding temperature, especially in the case of unprotected loss of flow transients. The results imply that the uncertainty of some core safety parameters should be reduced. Lastly, the results adumbrate that additional safety shutdown systems are to be applied to mitigate the consequences of unprotected transients.

Isotopic and spectral effects of Pu quality in Th-Pu fueled PWRs

UK plutonium (Pu) management is expected to focus on the use of uranium-plutonium (U-Pu) mixed oxide (MOX) fuel. However, research has shown that thorium-plutonium (Th-Pu) may be a viable alternative, offering favourable performance characteristics. A scoping study was carried out to determine the effect of isotopic composition and spectral hardening in standard and reduced moderation Pressurised Water Reactors (PWRs and RMPWRs). Lattice calculations were performed using WIMS to investigate safety parameters (Doppler Coefficient (DC), Moderator Temperature Coefficient (MTC), Void Coefficient (VC) – in this case Fully Voided Reactivity (FVR) – and Boron Worth (BW)), maximum theoretically achievable discharge burnup, Pu consumption and transuranic (TRU) composition of spent nuclear fuel (SNF) for the two reactor types. Standard grades of Pu were compared to a predicted UK Pu vector.MTC and FVR were found to be strongly influenced by the isotopic composition of the fuel. MTC was determined to be particularly sensitive to positive ‘peak’ contributions from fissile isotopes in the energy range 0.1–1 eV which diminish as the Pu content increases. The more extreme nature of the perturbation in FVR cases results in key differences in the contributions from fissile isotopes in the thermal energy range when compared with MTC, with no positive contributions from any isotope <500 eV.Where the requirement for MTC to remain negative was the limiting factor, a higher maximum fissile loading, discharge burnup and Pu consumption rate were possible in the PWR than the RMPWR, although the two reactors types typically produced similar levels of U233. However, for the majority of Pu grades the total minor actinide (MA) content in SNF was shown to be significantly lower in the RMPWR. Where FVR is the limiting factor, the maximum fissile loading and discharge burnup are similar in both reactor types, while increased Pu consumption rates were possible in the PWR. In this case, lower concentrations of U233 and MAs were found to be present in the PWR. These results are for a single pass of fuel through a reactor and, while the response of fissile isotopes at given energies to temperature perturbations will not vary significantly, the maximum achievable discharge burnup, Pu consumption rate and TRU build-up would be very different in a multi-recycle scenario.

Farklı sıcaklıklar altında BSR yakıt demetinin efektif çoğalma faktörü ve yakıt sıcaklığı katsayısı hesaplamaları

This paper presents effective multiplication factor (keff)&#x0D; with different burnable absorbers and weight percentages at different&#x0D; temperatures as well as doppler coefficient results and number density&#x0D; calculations for Westinghouse type pressurized water reactor (PWR) Assembly.&#x0D; Integral fuel burnable absorber rods coated with ZrB2 and&#x0D; Gadolinia-Uranium (UO2-Gd2O3) integral&#x0D; burnable absorbers were considered to calculate reactor parameters (keff&#x0D; and doppler coefficient). The results compared with base fuel which&#x0D; does not contain burnable absorber at different temperatures. The results show&#x0D; that reactivity was decreased with increased temperature and doppler&#x0D; coefficients increased with temperatures but remained negative at all&#x0D; temperatures. At 1500 K, the effective multiplication factor for base fuel was&#x0D; found to be 1.46985 while the effective multiplication factors for 2% with Gd2O3,&#x0D; 8% with Gd2O3, and IFBA rods were 1.38976,&amp;nbsp; 1.37574, and 1.30337 respectively.