Fast Neutron Spectrum Research Articles

Analysis of zirconium hydride moderator effect on the micro lead-based reactor

As a crucial parameter for the design of micro reactors, the neutron spectrum directly impacts the keff, the critical size, burnup characteristics, reactivity temperature coefficients and so on. When considering the optimization of the reactor size, designs utilizing thermal neutron spectrum and fast neutron spectrum each present their unique advantages and challenges. To investigate the impact of energy spectrum on the reactor performance, a micro lead-based reactor with annular channel fuel elements is proposed and the study of the influence of varying the volume ratio of moderator and fuel (the M/F ratio) on the keff, the critical size, burnup characteristic and reactivity temperature coefficients by using the Reactor Monte Carlo code (RMC code) is conducted. The results show that for the micro lead-based reactor proposed, when the reactor energy output demand is greater than 350 MWt·years, the design with fast neutron spectrum (without moderator) exhibits the minimum size. While the reactor energy output demand is less than 350 MWt·years, the design with a thermal neutron spectrum (with moderator) achieves the minimum size. Moreover, to ensure the negative reactivity temperature effect, the M/F ratio should be maintained below 3.4. The insights presented in this paper will serve as valuable references for the design of the micro reactor.

MTIR-SCP nuclear safety concept

The MTIR-SCP reactor is a research reactor with a light-water coolant having a fast neutron spectrum in nominal operation. None of the existing or projected reactors has a set of these features, therefore, special attention should be paid to the choice of the design of control rods and their location. The article considers various designs of control rods characteristic of the VVER-SCP power reactor currently under development, as well as existing research reactors with a liquid metal coolant having a fast neutron spectrum, taking into account the features of the MTIR-SCP: a seven-element control rod assembly design, as well as a ring-type absorbing element. Efficiency calculations are carried out for each of the structures under consideration and it was determined that the design of a control rod with an annular type of absorbing element has the highest specific efficiency. This design is chosen as the main one for the MTIR- SCP reactor. Criteria are proposed and optimization of the placement of control rods in the reactor core is carried out in order to reduce the uncertainty of obtaining neutron-physical characteristics in the central autonomous loop channel and increase the flow density in it. When optimizing the placement of control rods in the research reactor, their influence on the main neutron-physical characteristics, and especially on their distribution during operation under nominal conditions, is taken into account. Emergency protection and compensating rods that compensate for temperature and density effects when the reactor reaches the nominal power level have been removed, therefore their effect on the neutron-physical characteristics of the reactor and the loop channel during operation at power is not significant, and the compensating rods are located in the core and are removed as the fuel burns out, therefore their effect on the spatial The distribution of neutron-physical characteristics will be significant. To reduce this effect, burnout compensators and automatic regulators are proposed to be moved outside the core into the reflector. For various reactor conditions, the efficiency of the control and protection system was evaluated. It is shown that in various states, the control rods transfer the reactor to the required level of subcriticality, in accordance with NP-009-17.

Multiphysics and multi-scale design and analysis of nuclear thermal propulsion pellet bed reactor based on spherical cermet fuel element

Due to the advantages of high thrust and high specific impulse, nuclear thermal propulsion (NTP) is a preferred near-term rocket engine technology for future Mars missions. This article introduces the reactor design based on spherical cermet fuel element, with the aim of studying the feasibility of applying this new fuel in NTP system. First, design requirements are determined based on the unique features of NTP. In terms of neutronics, the effect of rocket launch accidents and fuel loss caused by high-temperature hydrogen are considered. Thermal hydraulics refer to the requirements of NASA's Mars Design Reference Architecture 5.0 (DRA 5.0) for impulse, thrust, and thrust to weight ratio. Second, based on the requirements above, the optimal reactor configuration parameters are obtained with the goal of maximizing the thrust to weight ratio: reactor radius 51.1 cm, height 106 cm, aspect ratio 1.04, reflector 13 cm, and reactor mass 3940 kg. Finally, the performances of the optimal reactor design are analyzed. The neutron spectrum shows a characteristic of fast neutron spectrum, and fuel loss analysis shows that the core can withstand a fuel mass loss of about 4 wt%. In addition, the core axial and radial power distribution are calculated, with an overall power peak factor of 1.26. The thermal hydraulic calculation provides a range of core operating conditions that satisfy the design requirements of DRA 5.0 and temperature, with power of 620 ∼ 950 MW and flow rate of 16 ∼ 24 kg/s. Compared with the ANL200/2000 reactor using prismatic cermet, it has the advantage of higher specific impulse. The main disadvantage is the large pressure drop, which prevents it from being applied to the scenarios of high power and high thrust. Therefore, for NTP reactor using spherical cermet elements, it should be applied to low to medium power scenarios.

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

As the demand for nuclear energy keeps rising, Generation IV reactor designs continuously improve. One such potential reactor design is the Gas-cooled Reactor. Neutronic analysis is the foundation for ongoing design modifications and alternatives for the Helium Gas-cooled Reactor. An important aspect of neutronic analysis is the nuclear data library. This research compares how different nuclear data libraries impact a Helium Gas-cooled Reactor design by comparing the neutron effective multiplication factor (k-eff) value. This research was conducted by the Monte Carlo method using OpenMC code. ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 were used as nuclear data libraries to simulate the reactor design for ten years of depletion time with 1-year timesteps. The reactor is designed in two different spectrums: the fast neutron spectrum design and the thermal neutron spectrum design. The differences between the two designs are in the cladding and reflector materials where the thermal neutron spectrum uses graphite moderator. The results for thermal spectrum design show that while each nuclear data library’s average k-eff values vary, they all produce the same results at timestep zero, with k-eff values around 1.02500. The average k-eff value from simulation using the ENDF/B-VII.1 library is 1.01477, the average k-eff value from simulation using the ENDF/B-VIII.0 library is 1.01122, and the average k-eff value from simulation using the JENDL-5.0 library is 1.01047. Each simulation result has an average uncertainty of 0.00040. The difference is due to the different amounts of data in each library. JENDL-5.0 has the highest amount of data inside its nuclear libraries, followed by ENDF/B-VIII.0 and ENDF/B-VII.1. Meanwhile, the average difference between the different libraries is insignificant for the thermal gas-cooled reactor simulation results. ENDF/B-VIII.0 produces an effective k-average of 1.04437, and JENDL-5.0 produces an effective k-average of 1.04273. Mass change over time results also shows that the transmutation of Uranium-238 into Plutonium-239 occurred. The fast spectrum reactor design shows a greater increase in plutonium-239 than the thermal spectrum design. The three libraries used did not show large differences in results in calculating changes in the mass of Uranium-238 over time due to Uranium-238‘s large mass, but they still showed differences in Plutonium-239‘s mass change over time. The average value difference between libraries in the fast neutron spectrum reactor design is 50 kg, and in the thermal spectrum it is 30 kg.

Status of the 226Ra nuclear data library and its impact on the production amount of 225Ac via the 226Ra (n,2n) reaction

ABSTRACT 225Ac is a promising alpha emitter for targeted alpha therapy. However, the current capability of 225Ac supply is limited to approximately 63 GBq/y, primarily relying on the natural source of 229Th stocked at a few institutes. Therefore, alternative 225Ac production methods are highly desired. The research and development of 225Ac and its parent nuclide production methods using accelerators and reactors to target 226Ra and 232Th are actively pursued worldwide. Hence, the authors focused on the 226Ra(n,2n)225Ra method using fast neutron spectra as an application of the experimental fast reactor Joyo. This study investigated the status of nuclear data libraries for 226Ra, an essential target for 225Ac production, and evaluated the impact of different nuclear data libraries on the amount of 225Ac produced. Consequently, cross-sections with covariance data were stored in TENDL-2021, JEFF-3.3, and EAF-2010 but not in ENDF or JENDL, the major nuclear data libraries. Furthermore, no consistency occurred among the respective nuclear data, and the 225Ra production amount varied.

Molten salts for 4th generation nuclear reactors?

If we are to combat climate change significantly, massive use of nuclear energy is essential. This requires upstream management of the risks associated with nuclear power and its acceptability to society. Nuclear generation will also have to be able to keep up with the load of a fleet that includes a large proportion of renewable energies. This article presents and discusses a new approach to meeting these challenges: liquid fuel reactors, or molten salt reactors, which are attracting growing interest. It describes the developments made in France with the CNRS MSFR (molten salt reactor with fast neutron spectrum) concept, which offers prospects for the incineration of transuranics and the management of intermittent energy sources.

Fast and thermal neutron spectrum dosimetry measurements in the Advanced Test Reactor large-B and small-l positions following the sixth core internals change-out

The Advanced Test Reactor (ATR) has a wide variety of irradiation positions that have had experiments that were developed by users from around the world. Most experiment irradiations rely on thoroughly benchmarked numerical models. However, some irradiation positions in ATR are not as well- characterized and are complicated by spectral perturbations from control cylinder orientation. The “small I” and “large-B” irradiation positions are located nearby control cylinders and suffer from these flux perturbations from control cylinder orientations that change during an irradiation cycle to maintain the desired core power distribution. Models predicted that these types of position would exhibit both spectral shifts and amplitude changes in neutron flux, but few measurements have been conducted to benchmark these predictions. Recently, requalification testing was performed to confirm the operational readiness of the ATR following the completion of the Core Internals Change-out (CIC). These tests provided a unique opportunity to validate the analytical methods used to simulate the ATR because nearly all components in the reactor were in a clean as-built state, significantly reducing modelling uncertainties. One subset of the testing included characterization of the fast and thermal neutron flux in the “small-I” and “large-B” positions using silver, cobalt, and nickel neutron dosimetry. In contrast to typical irradiation cycles, the control cylinders were held in position during the post-CIC nuclear testing. The specific activity of these dosimeter wires was measured following two low-power tests with different control cylinder positions.

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

Peculiarities of Changes in the Isotopic Composition of Pilot Fuel Elements of a VVER-SKD Reactor under Successive Irradiation in a Fast and Thermal Neutron Spectrum

Burnable Absorber Design Study for a Passively-Cooled Molten Salt Fast Reactor

The Passively-Cooled Molten Salt Fast Reactor (PMFR) is one of the advanced design concepts of the Molten Salt Fast Reactor (MSFR) which utilizes a natural circulation for the primary loop and aims to attain a long-life operation without any means of fuel reprocessing. For an extended operation period, it is necessary to have enough fissile material, i.e., high excess reactivity, at the onset of operation. Since the PMFR is based on a fast neutron spectrum, direct implementation of a burnable absorber concept for the control of excess reactivity would be ineffective. Therefore, a localized moderator concept that encircles the active core has been envisioned for the PMFR which enables the effective utilization of a burnable absorber to achieve low reactivity swing and long-life operation. The modified PMFR design that incorporates a moderator and burnable absorber is presented, where depletion calculation is performed to estimate the reactor lifetime and reactivity swing to assess the feasibility of the proposed design. All the presented neutronic analysis has been conducted based on the Monte Carlo Serpent2 code with ENDF/B-VII.1 library.

Measurement of fast neutron spectrum in the radial channel of VR-1 research reactor

Having knowledge of the spectrum of fast neutrons in the irradiation position of a reactor plays an important role in many experimental settings. If well described, it can be used as a reference neutron field for validating and testing measuring devices or detectors. This paper compares model calculations with measurements of the neutron spectrum in the research reactor VR-1. The fast neutron spectrum was measured in the radial channel of VR-1 reactor using a stilbene scintillation detector and the NGA-01 measuring device. The NGA-01 is a two-parameter spectrometric system for neutron and gamma radiation fields. The calculations were performed in a critical mode of the MCNP6 code using ENDF/B-VII.1 nuclear data library. Generally, a good agreement was achieved. However, a slight discrepancy in the neutron spectrum between the calculated and measured values was found in the 2–3 MeV region. In other regions, the differences between calculated and experimental values are comparable with uncertainties. Since the detector is distanced from the core, the neutron beam can be assumed parallel and therefore possible issues with stilbene anisotropy are eliminated. Thus, this experiment can be used for validation of a reactor leakage spectrum and nuclear cross sections in the energy region above 6 MeV, which is of interest for the nuclear data section of International Atomic Energy Agency.

Simulation of unfolding fast neutron spectrum from CR-39 track morphology

The aim of this study is to investigate the impact of various factors on neutron spectrum unfolding, including track parameters (major axis, minor axis and depth), polyethylene (PE) thickness, and etching time. The Monte Carlo code Geant4 and proton track parameter calculation program TRACK_P were applied to simulate the response matrix and track parameters spectrum. Four test spectra—252Cf spectrum, Gaussian broadened deuterium–deuterium fusion neutron spectrum, 241Am–Be spectrum, and Gaussian broadened deuterium–tritium fusion neutron spectrum—were studied using the maximum-likelihood expectation maximization algorithm for neutron spectrum unfolding. The unfolded spectra were in good agreement with the reference spectra. Results showed that the major and minor axes were more suitable than the track depth for neutron spectrum unfolding. For high-energy neutrons, a thick PE layer is required. Moreover, etching time does not significantly affect the unfolded spectra. This study explores the factors influencing the unfolding of fast neutron spectrum based on simulation calculations and preliminary experiments that can be useful for applying CR-39 to neutron spectrum measurements.

Proliferation aspects of partitioning and transmutation (P&T) fuel cycles

Abstract. Nuclear fuel is usually irradiated once but sometimes twice (MOX fuel). This procedure leaves a huge amount of spent fuel. Spent fuel is highly radioactive for a long time. Deep geological repositories are considered the best option for final disposal of this high-level radioactive waste. However, no such repository has been put into operation yet. Several countries consider partitioning and transmutation (P&T) a possible alternative. In P&T nuclear fuel cycles, the spent nuclear fuel is chopped into pieces and separated into different material streams such as plutonium, uranium and minor actinides. This process is called partitioning. From plutonium and minor actinides, new fuel elements are fabricated and used in nuclear reactors. If the overall physical conditions there are favorable, most prominently the presence of a fast neutron spectrum, at least a part of the minor actinides is fissioned and thus transmuted into other, hopefully easier-to-handle, nuclides. Especially when looking at the ingestion-based radiotoxicity index, reduction in minor actinides in nuclear waste is beneficial and might ease the requirements for a final repository. There are several drawbacks when looking at the concept of P&T: among others, several irradiation cycles are needed because the transmutation efficiency is so low. The technology for the processing of the spent fuel, the fuel fabrication and the irradiation is far from commercially available, and it is not clear whether this level can be met at all. This presentation focuses on the proliferation risk posed by possible P&T fuel cycles. The nexus between civil and military use of nuclear technology is widely known. If new fuel cycles are implemented, they should be more proliferation-resistant and not less. This would not be true for P&T fuel cycles: they require a closed nuclear fuel cycle with far more advanced reprocessing than today. Fissile material contained in spent nuclear fuel is not sent to final disposal but is separated into its components, which must be made available separately before they can be recombined into new fuel elements. Besides plutonium, this also includes other transuranic elements of small critical mass. The radiation barrier, an essential protection against unauthorized access to fissile material that could be used for nuclear weapons, is no longer present. The handling of those materials requires special facilities to cope with the immense amount of heat and radiation of the fuel containing high fractions of minor actinides. In those facilities, fissile material fitting for military purposes could be handled as well. Even if thorough safeguards are in place, detailed accountancy of all materials is not possible. The quantities needed to build a nuclear weapon are small compared to the quantities that need to be handled in a commercial-sized P&T fuel cycle. P&T technology is a civil application which clearly strengthens military capabilities. It would be hard to argue why it should only be available to a trustworthy partner (whoever that might be). The implementation of P&T fuel cycles would thus increase the risk of unauthorized diversion of nuclear material, technology and knowledge.

TRU oxide sample reactivity worths measured in the FCA-IX assemblies with systematically changed neutron energy spectra

ABSTRACT A series of integral experiments was conducted to evaluate the fission and the capture cross-sections of transuranic (TRU) nuclides at the fast critical facility FCA of the Japan Atomic Energy Agency (JAEA). The experiments were carried out using seven uranium-fueled assemblies of the FCA. The neutron energy spectra of the core regions were adjusted so as to change from an intermediate neutron spectrum to a fast neutron spectrum on an assembly-by-assembly basis. The integral data measured with these experimental configurations provide some neutron energy characteristics: 1) fission rate ratios (FRRs) of 237Np, 238Pu, 242Pu, 241Am, 243Am, and 244Cm relative to 239Pu by using absolutely calibrated fission chambers, 2) small sample reactivity worths (SRWs) of 237Np, 238Pu, 240Pu, 241Am, and 243Am where oxide powders of around 15 to 20 grams were used, 3) criticalities, and 4) spectral indices such as fission rate ratios of 238U relative to 235U. In this paper, details of the SRW measurements are reported, and the latest Japanese Evaluated Nuclear Data Library JENDL-5 is tested by using the integral data obtained in systematically varied neutron energy spectra.

Chance coincidence analysis for capture-gated neutron spectrometry with a composite scintillator

Neutron spectrometry is important for radiation metrology, radiation protection, accelerator physics etc. The spectral response of a capture-gated detector to mono-energetic neutrons contains a significant peak, making it easy to be analyzed. But this system output not only capture-gated events (CGE) but also chance coincidence events (CCE). It is critical to identify and eliminate CCE. However, chance coincidence has not been analyzed in detail before. This paper aims to find a way to eliminate CCE and get the spectrum of CGE. Thus, categories of CCE were analyzed and their probability equations were established. Then recorded pulses from a composite scintillator measured in mixed radiation fields were used to verify CCE and explore the influence of timing gate. Results show that n/γ discrimination is effective to distinguish CCE caused by γ-rays and calculate the amount of CCE caused by fast neutrons. Timing gate should be determined according to the different distributions of CGE and CCE. Spectrum of CGE can be acquired by subtracting that of fast neutron CCE from the fast neutron spectrum of coincidence events.

Fast neutron response of 6Li enriched CLYC and CLLB scintillators within 0.9–5.2 MeV

The performance of two 6Li enriched elpasolite scintillators Cs2LiYCl6 : Ce3+ (CLYC) and Cs2LiLaBr6 : Ce3+ (CLLB) were investigated using mono-energetic fast neutrons generated by a 3 MV tandem accelerator. The Pulse Shape Discrimination (PSD) ability, response and energy resolution to fast neutrons and scattered slow neutrons were studied. The 6Li enriched CLYC had excellent discrimination ability between neutron and gamma rays with the Figure of Merit (FOM) value better than 2.4. An Artificial neural network (ANN) model could discriminate fast neutrons and scattered slow neutrons effectively with an accuracy of 97%, which render the Li-6 enriched CLYC capable to detect thermal neutrons with high efficiency and the energy spectrum of fast neutrons simultaneously. The quenching factors are calculated to be 0.92 and 0.48 for the 35Cl(n, p)35S and 35Cl (n, α)32P reaction, while the detection efficiency is about 0.1%–0.75% for fast neutrons within the energy range of 0.9–5.2 MeV. Unlike CLYC, CLLB was only sensitive to slow neutrons, and the discrimination ability between neutrons and gamma rays is worse than CLYC, while the ANN model exhibit better performance which has an accuracy of 97%.

Measurement of spectrum-averaged cross sections of the (n,p) and (n,n′) reactions on strontium by fast neutrons of a TRIGA reactor: comparison with integrated data from excitation functions of various data libraries

Abstract Integral cross sections of the reactions 84Sr(n,p)84Rb, 86Sr(n,p)86Rb and 87Sr(n,n′)87mSr, over the fast neutron spectrum of a TRIGA reactor above 0.5 MeV, were measured for the first time. The results were used for integral tests of the evaluated excitation functions in the data libraries, ENDF/B-VIII.0, ROSFOND-2010, JEFF-3.3, JENDL-5 and TENDL-2021. The integral measurements are generally consistent with the integrated values within 12 %, except for a few cases. The two data libraries ENDF/B-VIII.0 and ROSFOND-2010 appear to need some improvement with regard to the reactions investigated in this work.

Numerical Study of Flow and Heat Transfer Characteristics in a Simplified Dual Fluid Reactor

This study presents the design and computational fluid dynamics (CFD) analysis of a mini demonstrator for a dual fluid reactor (DFR). The DFR is a novel concept currently under investigation. The DFR is characterized by the implementation of two distinct liquid loops dedicated to fuel and coolant. It integrates the principles of molten salt reactors and liquid metal cooled reactors; thus, it operates in a high temperature and fast neutron spectrum, presenting a distinct approach in the field of advanced nuclear reactor design. The mini demonstrator serves as a scaled-down version of the actual reactor, primarily aimed at gaining insights into the CFD analysis intricacies of the reactor while minimizing computational costs. The CFD modeling of the MD intends to add valuable data for the purpose of modeling validation against experiments to be conducted on the MD. These experiments can be used for DFR licensing and design optimization. The coolant and fuel utilized in the mini demonstrator are of low Prandtl number (Pr = 0.01) liquid lead, operating at two distinct inlet temperatures, namely 873 K and 1473 K. The study showed a rapid increase in turbulence due to intense mixing and abrupt changes in flow areas and directions, despite the relatively low inlet velocities. Hot spots characterized by elevated temperatures were identified, analyzed, and justified based on their spatial distribution and flow conditions. Flow swirling within pipes was identified and a remedy approach was suggested. Inconsistent mass flow rates were observed among the fuel pipes, with higher rates observed in the lateral pipes. Although lower fuel temperatures were observed in the lateral pipes, they consistently exhibited higher heat exchange characteristics. The study concludes by giving physical insights into the heat transfer and flow behavior, and proposing design considerations for the dual fluid reactor to enhance structural safety and durability, based on the preliminary analysis conducted.

Uncertainties in estimating production costs of future nuclear technologies: A model-based analysis of small modular reactors

Predicting future costs of technologies not yet developed is a complex exercise that includes many uncertain parameters and functional forms. In that context, small modular reactor (SMR) concepts that are in a rather early development stage claim to have cost advantages through learning effects, standardized design, modularization, co-siting economies, and other factors, such as better time-to-market even though they exhibit negative economies of scale in their construction costs due to their lower power output compared to conventional nuclear reactors. In this paper, we compare two different approaches from production theory and show that they have a theoretically equal structure. In the second step, we apply these approaches to estimate a range of potential construction costs for 15 SMR projects for which sufficient data is available. These include water cooled, high temperature, and fast neutron spectrum reactors. We then apply the Monte Carlo method to benchmark the cost projections assumed by the manufacturers by varying the investment costs, the weighted average cost of capital, the capacity factor, and the wholesale electricity price in simulations of the net present value (NPV) and the levelized cost of electricity (LCOE). We also test whether the differences between the manufacturer estimates and ours differ between technology families of SMR concepts and apply a sensitivity analysis. Here we contribute to an intensifying debate in the literature on the economics and finance of SMR concepts. The Monte Carlo analysis suggests a broad range of NPVs and LCOEs: Surprisingly, the lowest LCOE is calculated for a helium-cooled high-temperature reactor, whereas all of the light water reactors feature higher LCOEs. None of the tested concepts is able to compete economically with existing renewable technologies, not even when taking their variability and necessary system integration costs into account. The numerical results also confirm the importance of the choice of production theory and parameters. We conclude that any technology foresight has to take as much of the case specifics into account, including technological and institutional specifics; this also holds for SMR concepts.

The acceleration of improved Tone's method in advanced reactor lattice neutron spectrum calculation

This paper focuses on the acceleration of solving fixed-source equations in improved Tone's method. Since each equation is independent of the other with respect to the energy variable, the parallel calculation based on energy group domain decomposition was used in the advanced reactor lattice neutron spectrum calculation code TULIP. In order to balance the load of each CPU, the equally-divided strategy and a weighting-based strategy, which were used for the division of energy groups, were proposed and discussed. In the weighting-based strategy, the cross sections were used to calculate the weighting so that the iteration behavior could be caught. Therefore, the CPU load was balanced, and the parallel efficiency was improved. Moreover, the selection of the initial value was also evaluated. To verify the improvement, several tests based on fast and thermal neutron spectrum were calculated. According to the analysis of calculation time and parallel efficiency, using the initial value that was calculated with the homogeneous problem could save more than half of the calculation time. Meanwhile, the weighting-based strategy for parallel calculation could balance the CPU load so that the parallel efficiency was improved about 10%–80%, which is based on different tests.

Multiphysics simulations of Self-Regulating performance of an optimized molten metal fuel microreactor design

Advanced microreactors are expected to play an indispensable role in reliable energy solutions for civilian/military applications in off-grid regions and space applications such as power supplies for Lunar and Mars bases. An innovative fast neutron spectrum heat pipe microreactor (HP-MR) concept was recently proposed with two unique features: molten U-Mn fuel and heat pipes as the heat removal mechanism. The microreactor was designed to be self-regulated, solely relying on reactivity feedbacks from fuel temperature. The reactor design has been further optimized to provide sufficient engineering safety margin and effective heat removal. In this study, the self-regulating performance of the optimized molten metal fuel microreactor was investigated by multiphysics simulations (neutronics, heat-transfer and heat pipe modeling) based on the MOOSE MultiApps system. The responses of the reactor to a series of transient scenarios were simulated to demonstrate the exceptional and inherent safety features of the novel microreactor concept.