Moderator Temperature Coefficient Of Reactivity Research Articles

Neutronic analysis of the BYU molten salt micro reactor

Background: With the rising need for reliable and clean energy, researchers at Brigham Young University created a molten salt microreactor (MSMR) concept to help meet the world’s growing energy demands. The MSMR is rated at 45 MWth and is uniquely capable of passively dissipating all decay heat using conduction only. Methods: This work presents the results of the simulated neutronic behavior of the MSMR using the Monte Carlo neutronics code OpenMC with the ENDF/B-VII.1 cross-section library at steady state conditions. The specific characteristics discussed in this work are: keff, power profile, axial and radial peak power factors, Doppler coefficient of reactivity, and the moderator temperature feedback coefficient. Results: At steady state operating conditions, keff was calculated to be 1.07091 ± 0.00087 indicating that this configuration can achieve criticality. The axial peaking power factor was found to be 1.003 and the radial peaking power factor was found to be 1.45. Feedbacks were calculated by offsetting the temperature of the fuel and the moderator respectively. Doppler coefficient of reactivity to be -12.585 pcm/K and the moderator temperature coefficient of reactivity to be -0.34332 pcm/K. Conclusions: It is worth noting that the largest contributor of the negative reactivity comes from the fuel salt and not the moderator, further demonstrating the safety of the reactor. Future work will include a full neutronic optimization of the core, fuel burn up analysis, and exploration of a suitable burnable neutron poison.

Analysis for the use of thorium based fuel in LWRs

A detailed study has been taken up to assess the use of different fuel cycles based on thorium in current LWRs. Implications of use of thorium based oxide fuel in the existing LWRs, without requiring any structural and geometrical change, has been explored and compared with the prevailing uranium oxide fuel cycle. Owing to lack of intrinsic fissile component in thorium several possible fuel types have been considered which are mixture of thorium with a fissile component. The study has only been made with respect to the operational reactor physics aspects of use of such thorium based fuel. The fuel cycles considered are based on U235, Pu and U233 with thorium. The salient operational reactor physics characteristics, such as burnup of the fuel, reactivity coefficients, worth of the dissolved boron together with the kinetic parameters which has effect on the reactor operation and transients have been highlighted. The study shows that it is possible to use thorium fuel cycles in LWRs without any compromise on reactivity coefficients, except for the U233 + Th fuel where the moderator temperature coefficient of reactivity becomes positive even at modest concentration of dissolved boron.

Diagnostic methods applied to Esfahan light water subcritical reactor (ELWSCR)

In this work, Esfahan light water subcritical reactor (ELWSCR) is analysed using experimental and theoretical diagnostic methods. Important neutronic parameters of the system such as prompt neutron lifetime, delayed neutron fraction, prompt neutron decay constant, negative reactivity of the core, fuel and moderator temperature coefficient of reactivity, and overall and local void coefficient of reactivity are estimated. Also, neutron flux distribution, reflector saving, water level effect, and lattice pitch of the core including operating point of the facility are studied in details. Theoretical results are calculated by MCNPX and measurements are performed utilizing zero power reactor noise method. Detailed descriptions of the results are explained in the text.

Determination of the combined effect of boron and gadolinia on the reactivity and safety parameters of (U, Th)O2 fuel

This study investigated the combined effect of boron in light water moderator and gadolinia in fuel on the neutronic properties of (U, Th)O2 fuel in Westinghouse small modular reactor. The reactivity control of boron and gadolinia, and the potential of gadolinia in increasing thermal conductivity of its mixed fuel pellet motivated this study. Monte Carlo N-particle code integrated with CINDER90 was used for the calculations. Four fuel assemblies were used: one without boron and gadolinia, one with borated water only while the other two contained borated water plus 16 and 24 gadolinia fuel rods in each of the FAs. The BOC safety parameters was calculated at hot zero power (293.6 K) and hot full power (600 K). The results showed that the moderator temperature coefficient of reactivity was highest in FA without boron and gadolinia and smallest in FA containing boron plus 24 gadolinia fuel rods, while a reverse phenomenon was observed for the fuel temperature coefficient of reactivity. The effective delayed neutron fraction was slightly larger in FA without neutron absorber. The results show that boron and gadolinia decreased significantly the MTC but increased FTC slightly in addition to their initial neutron multiplication suppression role.

Comparison of homogeneous and heterogeneous thorium fuel blocks with four drivers in advanced high temperature reactors

Thorium is an attractive potential fuel owing to its abundance and unique thermal, neutronic, and chemical properties. One way to utilize thorium in block-type advanced high temperature reactors (AHTRs) is to homogenously mix the driver fuel (eg, U-233) with thorium in every fuel kernel of the tristructural-isotropic particles in a fuel block, which is the mixed oxide fuel (MOX) concept. Application of the seed and blanket (S&B) concept, that is, driver fuel in the seed region of a fuel block and thorium in the blanket region, is also another method to utilize thorium. To investigate the differences in the utilization of thorium using the MOX and S&B concepts in AHTRs, their multiplication factors (k∞), discharge burnups, conversion ratios, and reactivity temperature coefficients are compared. The investigated drivers include U-233 (U3), weapons-grade uranium (WU), weapons-grade plutonium (WPu), and reactor-grade plutonium (RPu). The results demonstrate that the MOX block with plutonium drivers has a higher discharge burnup in comparison with the S&B block. When the heavy metal mass is 6 kg and the initial mass of the fissile materials is 0.65 kg per block, the MOX block achieves 27% higher discharge burnup than the S&B block with the RPu driver. In contrast, the S&B block achieves higher discharge burnup in the case of uranium drivers (U3 and WU). The MOX block achieves a higher conversion ratio for all the drivers. Furthermore, the MOX block achieves a stronger negative moderator temperature coefficient of reactivity than the S&B block for all the driver fuels.

Influence of 235U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

To optimize the temperature coefficient of reactivity (TCR) for a graphite-moderated and liquid-fueled molten salt reactor, the effects of fuel salt composition on the fuel salt temperature coefficient of reactivity (FSTC) were investigated in our earlier work. In this study, we aim to provide a more comprehensive analysis of the TCR by considering the effects of the graphite-moderator temperature coefficient of reactivity (MTC). The effects of $$^{235}$$U enrichment and heavy metal (HM) proportion in the salt mixture on the MTC are investigated from the perspective of the six-factor formula based on a full-core model. For the MTC (labeled “$$\alpha _{{\mathrm{TM}}}$$”), the temperature coefficient of the fast fission factors ($$\alpha _{{\mathrm{TM}}}(\varepsilon )$$) is positive, while those of the resonance escape probability ($$\alpha _{{\mathrm{TM}}}(p )$$), the thermal reproduction factor ($$\alpha _{{\mathrm{TM}}}(\eta )$$), the thermal utilization factor ($$\alpha _{{\mathrm{TM}}}(f )$$), and the total non-leakage probability ($$\alpha _{{\mathrm{TM}}}(\varLambda )$$) are negative. The results reveal that the magnitudes of $$\alpha _{{\mathrm{TM}}}(\varepsilon )$$ and $$\alpha _{{\mathrm{TM}}}(p )$$ for the MTC are similar. Thus, variations in the MTC with $$^{235}$$U enrichment for different HM proportions are mainly dependent on $$\alpha _{{\mathrm{TM}}}(\eta )$$, $$\alpha _{{\mathrm{TM}}}(\varLambda )$$, and $$\alpha _{{\mathrm{TM}}}(f )$$, but especially on the former two. To obtain a more negative MTC, a lower HM proportion and/or a lower $$^{235}$$U enrichment is recommended. Together with our previous studies on the FSTC, a relatively soft neutron spectrum could strengthen the TCR with a sufficiently negative MTC.

Experimental Estimation of Moderator Temperature Coefficient of the IPEN-MB-01 Reactor

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28x26 rectangular array of UO2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4oC, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and (Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them.

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.

Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors

Abstract Our aim was to develop a method based on noise diagnostics for the estimation of the moderator temperature coefficient of reactivity (MTC) for the Paks VVER-440 units in normal operation. The method requires determining core average neutron flux and temperature fluctuations. The circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one-year-long period at Paks NPP.

An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors

Abstract In this paper, an investigation on the dependence of the effective multiplication factor, keff, on moderator temperature for various thicknesses of the upper beryllium reflector in reactor conditions with different fuel burnups for the Miniature Neutron Source Reactor is carried out. Based on the linear dependence of keff on moderator temperature, an approach to calculate the moderator temperature coefficient of reactivity, αT, at different temperatures and its average value, α, in a range of temperatures directly through the moderator temperature is developed. Calculations are performed to evaluate the effect of change in the upper reflector thickness on the moderator temperature coefficient of reactivity for the fresh core and reactor conditions with different fuel burnups. Calculated results indicate that α T increases with the increased beryllium thickness, but decreases with the increasing fuel burnup. Analysis of calculated results provides an additional insight into the relation of the upper reflector thickness, the neutron energy spectrum in the reactor core, and the moderator temperature coefficient of reactivity.

Direct reuse of spent nuclear fuel

In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO2 fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

Reducing axial offset and improving stability in PWRs by using uranium–thorium fuel

The axial offset (AO) is a parameter used to monitor the core axial power distribution, and is important for safe reactor operation. Previous works have showed that a new type uranium–thorium (U–Th) based fuel assembly could be used to achieve a more homogenous radial power distribution, and thereby improve the thermal margins in the core. In this paper, it is shown that a core containing U–Th fuel assemblies also reduces the AO and improves the core stability. For an equilibrium U–Th core, the maximum AO could be reduced by 33%, and the minimum AO could be reduced by 8% compared with a standard uranium core. In the U–Th core, the maximum AO was influenced by the more negative moderator temperature coefficient of reactivity, while the minimum AO is mainly affected by the more homogenous radial power distribution. These findings conclude that the U–Th core displays a more homogenous power distribution both axially and radially, thus offering better thermal margins for power uprates and flexible power operation. Moreover, xenon oscillations in the core due to perturbations induced by the control rods are 30% more effectively damped for the U–Th core compared with the standard core, improving the stability in AO.

CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.

CALCULATION OF THE MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY FOR MINIATURE NEUTRON SOURCE REACTORS

This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (MNSR) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study, the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and it’s average value in a range of temperatures directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment.

Development of a Method for Measuring the Moderator Temperature Coefficient by Noise Analysis and Its Experimental Verification in Ringhals-2

This paper deals with the estimation of the moderator temperature coefficient of reactivity (MTC) by noise analysis. The current noise analysis-based MTC measurement, relying on the cross correlation between the neutron noise measured by a single in-core neutron detector and the local temperature noise given by a single core-exit thermocouple located at the top of the same fuel assembly, or of a neighboring fuel assembly, is not accurate. The MTC is systematically underestimated by a factor of 2 to 5 compared to its design-predicted value. A theoretical study shows that, in case of nonhomogeneous moderator temperature noise, the core-averaged moderator temperature noise should be used for the MTC estimation. The new estimation method can reach up to 3% accuracy as compared with the results of core calculations for the Swedish Ringhals-2 pressurized water reactor (PWR). We show via noise measurements performed at the Ringhals-2 PWR that the moderator temperature noise is actually radially strongly heterogeneous and loosely coupled. The new MTC noise estimator is demonstrated to provide an accurate MTC evaluation, with the core-averaged moderator temperature noise estimated via the use of many radial in-core gamma-thermometers. More important, different forms of weighting functions are suggested to calculate the core-averaged moderator temperature noise. This new MTC noise estimator, which is nonintrusive and free of calibration, can therefore be applied to monitor the MTC throughout the cycle.

Calculation of the moderator temperature coefficient of reactivity for water moderated reactors

An American nuclear standard has been recently issued, providing guidance and qualitative recommendations on how the moderator temperature coefficient (MTC) of reactivity should be determined through reactor calculations in water moderated (and cooled) reactor cores (ANS, 1997. Calculation and measurement of the moderator temperature coefficient of reactivity for water moderated power reactors. American Nuclear Society, American National Standard ANSI/ANS-19.11-1997). The present work provides quantitative information on areas of concern and effects addressed in this standard, namely, the effect of the core reflector region and the effect of the magnitude of the moderator temperature change. The calculations consist of a sequence of static criticality analyses, and are performed with a purposely developed two-dimensional reactor model based on two neutron energy groups. The numerical results indicate that the thickness of the reflector has a measurable effect on the accuracy of the MTC value only for a small core, introducing uncertainties of the order of 10%. Instead, the effect of temperature change is found to be negligible within the recommended range of 3 to 5°C.

Estimation of the moderator temperature coefficient of reactivity via noise analysis using closed-loop transfer functions

The accuracy of the estimation, via noise analysis, of the moderator temperature coefficient (MTC) of reactivity is investigated, assuming that closed-loop data are used in the estimators. The noise field is generated through numerical simulations. The focus is on the approach using the proportionality between estimated and actual MTC values during the fuel cycle. The results show that the bias induced on the proportionality constant (calibration factor) depends, as expected, on the strength of the feedback. However, it remains within tolerable limits, considering that the introduced uncertainty is ⩽30%, which is comparable with the measurement uncertainty of the standard boron dilution method.

Moderator temperature effect on reactivity in light water moderated experimental reactors

The moderator temperature coefficient of reactivity has been measured in a U-233 fuelled plate type light water moderated reactor (KAMINI). Using neutron cross sections based on WIMS library, lattice homogenization code SMAXY and 3D core calculation code COMESH the moderator temperature coefficient of reactivity is predicted very well (within 0.6 pcm per °C). For U-235 and Pu-239 fuelled KRITZ experiments also, moderator temperature coefficients of reactivity have been predicted and compared with Studsvik results using lattice code MURLI and 2D core calculations. Our predictions of the moderator temperature coefficient of reactivity is fairly good for these cores. For U-233 fuelled KAMINI, the close agreement on moderator temperature coefficient of reactivity was not found using more recent U-233 cross-sections derived from ENDF/B-IV (WIMKAL). By comparing the α (capture to fission ratio) of U-233 at low energies between WIMS and ENDF/B-IV and VI, the better comparison with WIMS is attributed to the flatter α behavior with energy. There appears to be a need for measurement of α of U-233 below 0.1 eV.

Moderator Temperature Coefficient of Reactivity in Pressurized Water Reactors: Theoretical Investigation and Numerical Simulations

Estimation of the moderator temperature coefficient (MTC) of reactivity of pressurized water reactors by noise analysis is investigated, both theoretically and through numerical simulations. It is found that using local neutron signals, all estimators of MTC will underestimate the actual MTC value, as indeed is observed in all measurements reported in the literature. However, the relative deviation from the actual MTC value remains practically constant over a wide range of MTC values. This suggests that it is possible to calibrate the estimate at the beginning of the fuel cycle and then use this calibration until the end of the fuel cycle. A new estimator is proposed based on integral (point model) concepts. This estimator is found capable of providing the actual value of MTC, provided that coolant velocity fluctuations are at least as strong as coolant inlet temperature fluctuations in inducing neutron and coolant temperature fluctuations.

Evaluation of the moderator temperature coefficient of reactivity in a PWR by means of noise analysis

The neutron and temperature fluctuations of sensors belonging to one string in a 1300 Mw pressurized water reactor (PWR) were selected for evaluating the magnitude of the moderator temperature coefficient of reactivity (MTC). These fluctuations are less affected by subcooled boiling. A model describing the temperature fluctuations inside and outside the core was developed and the bias of the temperature fluctuations due to position corrected. The resulting MTC magnitude was compared with the design-predicted value.