Moderator Temperature 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.

The effect B impurity and Xe intrusion in the graphite moderator on void reactivity feedback of thermal MSR

Thermal Molten Salt Reactor (MSR) massively utilizes graphite as both moderator and reflector. The molten fuel circulates around the primary loop and directly contacts with the graphite. Gaseous fission products, such as neutron poison Xe-135, are normally separated out from the core. However, deposition of Xe-135 in graphite could take place through direct Xe-135 intrusion into graphite pores, or salt intrusion that eventually undergoes fission in the graphite and leaves the Xe-135 inside. Another possible way of strong neutron absorber deposition in the graphite is boron impurity. Beside reducing the core excess reactivity, the existence of strong neutron absorbers in graphite could alter the core reactivity feedbacks. Thus, this study is meant to analyse the effects of Xe-135 and natural boron deposition in the graphite to the core reactivity feedbacks. The evaluations were done by monte Carlo calculations on their effects on void coefficient, fuel, and moderator temperature reactivity feedbacks. The introduction of void into under-moderated thermal MSR core generally produces positive void reactivity feedback. The deposition of strong neutron absorbers in the graphite decreases the void reactivity coefficient, yet still positive. Despite of core excess reactivity loss, the existence of strong neutron absorber in the graphite is favourable in safety point of view.

A cross-section correction model for history effect treatment in pin-by-pin code NECP-Bamboo2.0

The consideration of the history effect is one of the foundations for ensuring the precision of the two-step calculation scheme. Since the actual condition could not be predicted in advance, the standard approach of the lattice calculation adopts the branch-based model in which the variations of nuclide number densities with burnup are only considered in averaged operating conditions. Consequently, the nuclide number densities associated with actual operating conditions are not considered properly in the 3D whole-core calculation, and the problem of the history effect arises then. The microscopic depletion method and serval correction methods based on different lattice calculation structures were developed to solve this problem and have been employed in pin-by-pin two-step calculation codes. This paper introduces a new spectral history correction method with the implementation of a new bidirectional branch model, a pin-cell burnup correction model, and a multiscale cross-section correction model. The work was carried out based on the pin-by-pin fuel management calculation code system NECP-Bamboo2.0 developed by Xi'an Jiaotong University. Encouraging conclusions have been obtained from the numerical results. For the heterogeneous history effect caused by CR movement, the bias of keff given by the classical corrections will exceed ± 1000pcm. In contrast, the improved method can reduce it to be smaller than ± 250pcm. For the histories triggered by the changes of other state parameters, including boron concentration, fuel, and/or moderator temperature(s), the new cross-section correction method is comparable in accuracy to the traditional microscopic cross-section correction method, both better than the microscopic depletion correction method.

Parametric neutronic analysis of different cladding options for ThO2 pellet of advanced dual-cooled annular PWR assembly

This study presents a parametric neutronic evaluation of different candidate claddings for a purposed candidate fuel (Th-233U-235U)O2 in an advanced dual-cooled annular PWR assembly. The investigated claddings include Zircaloy-2/4, zirconium carbide (ZrC), silicon carbide (SiC), ferritic Fe-20Cr-5Al (FeCrAl) and APMT alloys, and austenitic 310 (SS-310), compared to the by design cladding for the AP1000 assembly, Zirlo™. Neutronic calculations were performed using the deterministic code DRAGON, and the parameters included infinite multiplication factor, criticality periods, neutron flux spectra, radial power distributions, rim effects, and reactivity feedback coefficients.The results show that the candidate fuel with Zirlo™ achieves a higher criticality period than UO2 fuel in both solid and annular configurations. The candidate fuel with SS-310 cladding exhibits a higher criticality period penalty of about 12.60% compared to Zirlo™, APMT, and FeCrAl claddings show a criticality period penalty of about 10.94% and 8.79%, respectively. The candidate fuel with ceramic claddings, SiC and ZrC, can achieve slightly higher criticality periods compared to Zirlo™ of about 1.11% and 0.24%, respectively, due to their low thermal neutron absorption cross-sections. Regarding the power profile, the annular (Th-233U-235U)O2 fuel exhibits a flatter behavior compared to UO2 fuel. However, the claddings with lower capture cross-sections exhibit slightly increased fission power through the fuel pellet. The fuel temperature coefficient for candidate fuel exhibits more negative values as the capture cross-section of the cladding increases, while the molybdenum and zirconium-containing claddings exhibit more favorable moderator temperature coefficients.

Revolutionizing LWR SMR reactors: exploring the potential of (Th-233U-235U)O2 fuel through a parametric study

ABSTRACT Small modular reactors (SMRs) have garnered significant attention for their operational adaptability and ease of deployment. Thorium, with its well-documented advantages, shows promise as a viable fuel option for SMRs. However, the absence of intrinsic fissile components in thorium necessitates the exploration of different thorium-based fuel combinations. One such combination, (Th-233U)O2 fuel, has limitations due to the presence of pure U-233. To overcome this challenge, a new fuel mixture, (Th-233U-235U)O2, was investigated for SMRs. This study examined the reactor physics characteristics of the (Th-233U-235U)O2 fuel, including fuel burnup, neutron flux spectra, power distribution, the evolution of actinides, and reactivity coefficients. Results indicate that the (Th-233U-235U)O2 fuel allows for a longer criticality period compared to UO2 fuel, with up to a 14% improvement, while accumulating fewer plutonium and transuranic elements. Notably, it demonstrates significantly improved negative reactivity coefficients, particularly for moderator temperature, with an average improvement of 45% over (Th-233U)O2 fuel. The conceptual (Th-233U-235U)O2 fuel, therefore, exhibits promising neutronic properties, presenting possibilities for future studies. These findings contribute to the understanding and advancement of advanced fuel designs for SMRs.

Investigating the possibility of extending the BWR cycle length for 15 years of operation by mixing highly enriched UO2 fuel with burnable absorbers

This paper investigates the possibility to extend the cycle length of boiling water reactor bundles to 15 years of operation with three different burnable poisons; gadolinium, erbium, and boron carbide. This can be carried out by mixing highly enriched UO2 fuel (15–19.9% U-235) with high concentrations of Gadolinium oxide (3–14% Gd2O3) or Erbium oxide (2–4% Er2O3).The boron carbide B4C was modeled as (Al2O3-B4C) rods in the bundle guide tubes. MCNPX code 2.7 was used to evaluate infinite multiplication factor (K-inf), power distribution, peaking factor, void reactivity coefficient, fuel cycle length, depletion of U-235, and fissile inventory ratio for the three designs at 40% void. The MCNPX simulation showed that introducing gadolinium rods at the bundle periphery has the advantage of lowering reactivity swing throughout the exposure range. The uniform distribution of erbium in all fuel rods contributed to the flattening of peaking factor at all the burnup stages. For the B4C design, the author found that the assembly with B4C–Al performs best in terms of reactivity flattening when five of the B4C–AL2O3 rods are positioned in the central region of the assembly. Furthermore, the fuel temperature coefficient is more negative for gadolinium design at all burnup stages. On the other hand, the boron model delivers the lowest control rod worth. Finally, the moderator temperature coefficient is more negative for erbium and WABA designs due to the enhanced thermal neutrons capture by the effect of the strategic arrangement of WABA rods and the uniform distribution of erbium.

Neutronic investigation of enriched gadolinium, natural gadolinium and Zirconium-erbium as burnable poisons in the advanced PWR assemblies by MCNPX code

In Pressurized Water Reactors, burnable absorbers are implemented for controlling the excess reactivity and avoiding the high power peaking factor. A feasibility study was carried out for investigating the effect of enriched gadolinium (157Gd2O3), natural gadolinium (152 - 160Gd2O3) and Zirconium-Erbium (ZrEr2) as burnable absorbers on the safety parameters of advanced PWR assembly. Three poisoned assemblies are modeled and simulated by the well-known MCNPX code. To hold a fair comparison between these assemblies, the author adjusted the concentration of fissile materials and burnable absorbers in each assembly to provide approximately the same initial reactivity. The MCNPX simulation showed that the enriched gadolinium model achieved a number of advantages over Zirconium-Erbium as: 1) prolonging the fuel cycle length by 1.04%, 2) increasing the negativity of moderator temperature coefficient and fuel temperature coefficient by 10.3% and 9.3% at the BOL and this contributes to core safety, 3) minimizing residual reactivity penalty at the higher burnup steps and 4) more hardening of neutron spectrum at the BOL. On the other hand, the Zirconium – Erbium layer presented a number of benefits over enriched gadolinium as 1) reducing the peaking factor by 1.3%, 2.6% and 2% at 20, 30 and 50 GWd/ton, 2) increasing the reactivity worth by 1.6% at the BOL, 3) increasing the power flattening by 7.6% at the EOL, and this is considered an enormous economic benefit. Therefore, the modeling of enriched gadolinium and zirconium erbium in this study is suggested for improving the neutronic and safety parameters of PWR cores.

Investigation on Neutronic Behavior of Pebble Bed Reactor for TRU Transmutation

One alternative effort to eliminate the accumulation of transuranic (TRU) elements generated from LWR spent fuel is by utilizing TRU as nuclear fuel in pebble bed reactor. This work was aimed to investigate the neutronic behavior of a pebble bed reactor used for TRU transmutation. The reactor geometry is adopted from the HTR-Modul. MCNP6 multipurpose radiation transport code with ENDF/B-VII.1 neutron library was used to calculate the neutronic aspects. The results indicated that the initial effective multiplication factor (keff) values decrease as the amount of TRU fuel pebbles within the core increases, concurrent with the fuel burnup. Increasing the TRU fuel pebble ratio resulted in weakened Doppler temperature coefficients (DTCs) and improvement in moderator temperature coefficients (MTCs). The calculated total mass of plutonium and minor actinide transmutation demonstrated that a pebble bed reactor with 100% TRU pebble can reduce TRU mass up to 55.47% of its initial mass load.

Assembly-level analysis on temperature coefficient of reactivity in a graphite-moderated fuel salt reactor fueled with low-enriched uranium

To provide a reliable and comprehensive data reference for core geometry design of graphite-moderated and low-enriched uranium fueled molten salt reactors, the influences of geometric parameters on the temperature coefficient of reactivity (TCR) at an assembly level were characterized. A four-factor formula was introduced to explain how different reactivity coefficients behave in terms of the fuel salt volume fraction and assembly size. The results show that the fuel salt temperature coefficient (FSTC) is always negative owing to a more negative fuel salt density coefficient in the over-moderated region or a more negative Doppler coefficient in the under-moderated region. Depending on the fuel salt channel spacing, the graphite moderator temperature coefficient (MTC) can be negative or positive. Furthermore, an assembly with a smaller fuel salt channel spacing is more likely to exhibit a negative MTC. As the fuel salt volume fraction increases, the negative FSTC first weakens and then increases, owing to the fuel salt density effect gradually weakening from negative to positive feedback and then decreasing. Meanwhile, the MTC weakens as the thermal utilization coefficient caused by the graphite temperature effect deteriorates. Thus, the negative TCR first weakens and then strengthens, mainly because of the change in the fuel salt density coefficient. As the assembly size increases, the magnitude of the FSTC decreases monotonously owing to a monotonously weakened fuel salt Doppler coefficient, whereas the MTC changes from gradually weakened negative feedback to gradually enhanced positive feedback. Then, the negative TCR weakens. Therefore, to achieve a proper negative TCR, particularly a negative MTC, an assembly with a smaller fuel salt channel spacing in the under-moderated region is strongly recommended.

Modeling of irradiated dimensional change strain in MSR graphite moderator

When irradiated in the core of a thermal spectrum nuclear reactor, graphite undergoes dimensional changes which depend on the neutron spectrum. The graphite lifespan is a limiting parameter in MSR designs; it correlates strongly with the graphite’s dimensional changes. Coupled modeling of the thermal–hydraulic, neutronic, and thermal-mechanics conditions is necessary to evaluate the graphite thermal and mechanical responses to the MSR core conditions. To assess graphite dimensional changes associated with irradiation, a computational model was developed using GeN-Foam. In particular, the thermo-mechanical region solver was expanded to account for the moderator’s response to the core’s neutron kinetics. These capabilities enable modeling of the thermal deformation of graphite moderator structures as well as calculation of the moderator temperature at the thermo-mechanical mesh cell level. The GeN-Foam graphite moderator model was benchmarked using the experimental and computational data of graphite moderator thermal response in high temperature gas-cooled reactors.

Effect of Shade on Xylem Development of Moringa oleifera (Alrawag) Seedlings

This study investigated the effect of shade as a moderator oflight intensity and temperature on xylem development of seedlings ofMoringa oleifera. Three levels of shade were used: high shade, mediumshade and no shade. Four sequential harvests were carried out monthly forthe seedlings starting from four weeks after seed sowing. Cross sectionswere prepared from the lower and middle parts of the stems at eachharvest and vessel volume fraction, vessel diameters, number of vesselsand volume fraction of lignified cells were measured. The results showedthat the effect of shade was significant on the anatomical variables.Generally, the values of the anatomical variables of the stem wood of theseedlings decreased with increased shade level. The findings indicate theimportance of shade to M. oleifera seedling development and that itshould be considered in nursery conditions.

Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code

From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one.

Conceptual design of the double tube burnable poison for the next generation small modular reactor

In this study, a new application of an innovative burnable poison (BP), which is called Double Tube Burnable Poison (DTBP), is introduced. Two design cincepts of the DTBP were tested in a small modular reactor (SMR) to achieve a longer operating cycle with low soluble boron concentration. These two concepts can be loaded in different locations in the fuel assembly, which increases the degree of freedom of nuclear design. The flexibility of this design allows it to increase its poison effect with the ability to control the depletion speed of its absorber materials. The DTBP pins were combined with the Erbia pins to achieve a flat k-infinite letdown curve with a significant reduction of the excess reactivity. The assembly and core calculations were performed by DeCART2D and MASTER codes. The results showed that the combination of DTBP and Erbia core achieved a cycle length of more than 3.5 Effective Full Power Years. Also, the highest value of Critical Boron Concentration was reduced to 309 ppm within the design limitations of the moderator temperature coefficient and the maximum pin power peaking. The performance of this option is the best among the other options including the conventional BP designs.

Searching for optimal accident tolerant fuel for the VVER-1200 reactor from the neutronic point of view

ABSTRACT In an attempt to find an optimum fuel, which can withstand the harsh operation condition in the VVER-1200 reactor, different fuel types were investigated in this work. Uranium carbide and uranium nitride are proposed as alternative fuels to UO2. MCNPX 2.7 with cross-section library ENDF/B-VII was used to design 3D model of VVER-1200 to investigate the neutronic properties of the suggested fuel types. The output results of the burnup of the proposed fuels have been analyzed to determine which of them achieves the longest fuel cycle. It was found that UO2 and U12C achieved criticality till 35 GWD/MTHM, while U14N and U15N achieved criticality at 30 GWD/MTHM and 24 GWD/MTHM, respectively. The traditional method of reactivity management has been tested on the proposed fuel to determine to what extent it is valid for them. To ensure the operation safety of the proposed fuel, some safety factors such as moderator temperature coefficient (MTC‎), fuel temperature coefficient (FTC‎), and effective delayed neutron fraction (βeff) have been calculated. The distribution of the thermal neutron flux and normalized power has been displayed to give us a good insight view of the VVER-1200 assembly to determine which of the proposed fuels is optimum compared to UO2.

Neutronic Evaluation and Optimization of the Centrifugal Nuclear Thermal Rocket Concept

The centrifugal nuclear thermal rocket is a concept for a liquid-fueled nuclear system that would allow for a much higher specific impulse than the more traditional solid-fueled nuclear thermal propulsion designs. Although some preliminary neutronics analyses have been done on conceptual designs, this work seeks to perform a more systematic analysis and optimization of design parameters and to investigate additional neutronics properties such as power distributions and reactivity coefficients. This work used OpenMC for neutronics analysis and Dakota for the parametric study and optimization. Inter- and intra-fuel element power distributions were calculated, and a strong radial dependence was noted within fuel elements that may pose a challenge to thermal constraints. A positive moderator temperature coefficient of 3.78 0.16 pcm/K was calculated for the reference model, which may pose a challenge for system design and control. The optimization study of reflector size, fuel spacing, fuel mass, and fuel element radius indicated many trade-offs in the design considerations, and that the baseline model can be significantly improved in all respects. Positive reactivity feedback can be minimized by reducing moderation, and peaking factors can be reduced by limiting the amount of fuel per fuel element, which also minimizes the system mass.

Monte Carlo Study of an Electron-Based Neutron Source for Bragg Edge Imaging

Neutron imaging is a powerful and nondestructive tool for testing materials in industrial and research applications. Compact accelerator neutron sources are gaining interest in neutron application techniques, such as Bragg edge transmission imaging. Delivering a high neutron flux with a narrow pulse width and suppressed photons at the sample position are fundamental factors for designing a neutron source for Bragg edge imaging. In this study, Monte Carlo calculations were performed to simulate a 40-MeV electron beam impinging on a cylindrical tungsten target. Different target moderator and reflector (TMR) geometries were investigated to produce cold neutrons, and their results were compared. Polyethylene (PE) and graphite were used as the moderator and the reflector, respectively. The structures and dimensions of the moderator and reflector were optimized using a Monte Carlo simulation with the PHITS-3.28 code. The effect of the PE moderator temperature on the cold neutron flux was investigated. The results showed that the optimum size of the PE at 77 K inside the reflector was 3 × 15 × 15 cm3 to achieve the wavelength resolution of 1.05% and the neutron flux of 1.16 × 104 n/cm2/s at 1000 cm from the target station by assuming the electron beam current of 275 µA. In addition, the FLUKA 4-2.1 code was used to calculate the neutron spectrum from the designed neutron production target at room temperature, and the results were consistent with the PHITS calculations. The neutron spectrum together with its pulse width from the designed TMR were used to simulate the Bragg edges of an α -Fe sample, and it was concluded that the TMR is suitable for performing Bragg edge imaging.

Advanced fuel and burnable absorbers designed for long-cycle operation of BNPP

The present study is based on the facts that are happening for the Bushehr Nuclear Power Plant (BNPP) and coincides with the operational changes and step-by-step modifications of the BNPP in the near future. Bushehr reactor core will change from BNPP core to TVS-2M core. In other words, a new fuel, called the TVS-2M, is being used by the Russians to upgrade WWER-1000 reactors to reduce power generation costs, increase cycle length and fuel consumption. The neutronic analysis performed in this work is in line with this change. It provides a feasibility study of BA change and increase of cycle length in BNPP reactor without changing the initial structure of the reactor core. This research provides a new core configuration that can achieve the stated goals to a large extent without changing the BNPP core to the TVS-2M core. This research provides a feasibility study of Integrated burnable absorber (IBA) change by maintaining the initial neutronic conditions and a new high-performance configuration aiming for a long-cycle operation in Iran's nuclear power reactor. In Water-Water Energetic Reactors (WWER) reactors, two alternatives to CrB2Al (current burnable absorber in BNPP) are gadolinia (Gd2O3-UO2), and erbia (Er2O3-UO2). The benefits of the combinational IBA concept are expected by compensating weak points of each other. The main contribution of this research is the use of two cylindrical coaxial IBA layers with the same volume of gadolinium pins plus erbium (separated by a thin layer of zirconium alloy (Zr + 1% Nb)) and gadolinium pins. A neutronic evaluation has been conducted to investigate the variations of effective multiplication factor versus burnup, reactivity swing, power peaking distribution, moderator temperature coefficient (MTC), critical boron concentration (CBC), and cycle length. The result shows that the cycle length can be extended from 296 days in the BNPP core to 366 days in the proposed core, which is comparable to the cycle length of the TVS-2M core. The average BA concentration in the proposed core is 5%, which is lower than the average BA concentration of 5%–8% in the TVS-2M core. Regarding the gain of 73.90 days in cycle length, the revenue is estimated to be around 19,503,000 USD, making it worth the implementation considering the number of gadolinium and erbium pins in the proposed core.

Operation Cycle Length Extension of a Konvoi PWR: Requirements and Experience from Operator’s Viewpoint

The extension of the operating domain of PreussenElektra’s Konvoi-type pressurized water reactors (PWRs) beyond the natural end of cycle is known as stretch-out operation. A range of possibilities exists to increase nuclear fuel utilization to continue operation after the boron concentration reaches its dilution limit. The most basic option is to continue operation with constant average moderator temperature, which results in a relatively fast decrease in reactor power. From a fuel utilization point of view, this is the least optimal procedure. In PreussenElektra’s PWR fleet, an enhanced operation mode is adopted, leading to a comparatively modest decrease in reactor power and very high utilization of nuclear fuel. Initially, the stretch-out mode provided an option to gain flexibility regarding outage planning. More recently, the stretch-out method has served as a practical approach to optimizing electricity generation costs during the last cycles before the final shutdown as stipulated by law, as operators can extend the cycle length in a range of 30 to 60 days after the natural end of cycle. This paper describes the licensing rationale, the feasibility of this type of operation, and the operating requirements and experience. The system parameters affected by stretch-out operation are discussed. Adjustments of set points of thermal-hydraulic variables in the primary and secondary systems are explained. Licensing requirements for safe reactor operation in stretch-out mode are reviewed. Furthermore, aspects of neutronic and thermal-hydraulic core surveillance are included. After more than 35 years and counting, the methods of increasing fuel utilization are not new, and an evaluation of experience and effectiveness is in order.

A Novel Core Design with Movable Moderator for a Fluoride Salt–Cooled High-Temperature Reactor

The Fluoride salt–cooled High-temperature Reactor (FHR) is a Generation IV reactor concept that can operate under near atmospheric pressure circumstances and further enhance inherent safety. In this study, an FHR core design with 165 MW of thermal output [MW(thermal)] is proposed. The reactor core employs tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks as the basic fuel form, FLiBe [lithium-beryllium fluoride (2 7LiF-BeF2)] as the primary coolant, and a three-batch fuel cycle scheme. Sensitivity analyses on various parameters were performed to optimize the cycle length and neutronic parameters. The fuel cycle of this core design was evaluated in detail from four aspects: cycle length, power peaking factor (PPF), discharge burnup, and temperature coefficient. It was found that a larger fuel channel pitch would have a relatively harder neutron spectrum and yield a relatively longer cycle length, lower PPF, and better fuel temperature coefficient and moderator temperature coefficient (MTC). In addition, burnable poison (BP) (Er2O3) can effectively reduce PPF, hold down the multiplication factor, and more importantly it can improve the MTC. The preliminary design of control blades is also presented in this paper. Furthermore, on the basis of the proposed 165-MW(thermal) core, we propose a novel core design that incorporates “fuel inside radial moderator (FIRM)” assemblies, movable moderator, and movable BP. This new design can extend the fuel cycle length by approximately 45 days for an 18-month fuel cycle. In addition, improvements were also found in PPF, discharge burnup, and temperature coefficients.

Development of cross-section calculation module for high-temperature gas-cooled reactor engineering simulator system using an optimal model tree algorithm

To improve the accuracy of reactor physics module in the high-temperature gas-cooled reactor engineering simulator system (HTGR-ESS), a more accurate cross-section calculation model is required under the premise of real-time calculation in the simulation. Due to the nonlinear relationship between cross-section and reactor state parameters, such as burnup, moderator temperature and fuel temperature, etc., it is difficult to achieve full-range accuracy using the current multiple linear regression method (MLR). In this study, a machine-learning-based cross-section calculation model for the reactor physics model is proposed. We made improvements to the model tree by setting a smoothing function on the boundary and using ridge regression at the leaf node to conduct a continuous cross-section calculation model without overfitting in the subspace over the full range, trained from a scattered database generated by V.S.O.P. In the numerical tests, the proposed method produced far more accurate cross-section and reactor physics calculation results than the current method and could also meet the real-time calculation requirement. Meanwhile, the responses of the power and helium outlet temperatures of the 10 MW HTGR (HTR-10) simulator under different operating conditions conform to the reactor laws of physics, further validating the applicability of the proposed method.