Treatments of Thermal Neutron Scattering Data and Their Effect on Neutronics Calculations

In the conventional method to generate thermal scattering cross section of moderator materials, only one of the coherent elastic scattering and incoherent elastic scattering is considered in neutronics calculations. For the inelastic scattering, fixed incident energy grid is used in the nuclear data processing codes. The multipoint linearization method is used to refine the incident energy grid for inelastic scattering. We select ZrHx (zirconium hydride) as an example to analyze the effects of the above described treatments on the reactivity of several critical benchmarks. The numerical results show that the incident energy grid has an obvious effect on the effective multiplication factor (keff) of the analyzed reactors; simultaneously considering the coherent and incoherent elastic scattering also affects keff by tens of pcm.

A study on the core loading pattern of the VVER-1200/V491

The VVER-1200/V491 was a selected candidate for the Ninh Thuan I Nuclear Power Plant.However, in the Feasibility Study Safety Analysis Report (FS-SAR) of the VVER-1200/V491, the core loading pattern of this reactor was not provided. To assess the safety features of the VVER- 1200/V491, finding the core loading patterns and verifying their safety characteristics are necessary. In this study, two core loading patterns of the VVER-1200/V491 were suggested. The first loading pattern was applied from the VVER-1000/V446 and the second was searched by core loading optimization program LPO-V. The calculations for power distribution, the effective multiplication factor (k-eff), and fuel burn-up were then calculated by SRAC code. To verify several safety parameters of loading patterns of the VVER-1200/V491, the neutron delayed fraction (DNF), fuel andmoderator temperature feedbacks (FTC and MTC) were investigated and compared with the safety standards in the VVER-1200/V491 FS-SAR or the VVER-1000/V392 ISAR.

Calculation of critical core configurations of a research reactor using MTR, IRT-4M, VVR-KN fuel assemblies

This paper presents calculation results to determine critical core configurations and aminimum number of fuel assemblies (FAs) or uranium mass of a research reactor loaded with three types of FAs such as MTR, IRT-4M and VVR-KN. The MCNP5 code and ENDF/B7.1 library were applied to estimate characteristics parameters of the fuel types and the whole core. Infinitive multiplication factor kinf, neutron flux distribution and neutron spectra of the fuels were calculated. The reactor core configurations with three fuel types were modeled in 3-dimensions, and then the effective multiplication factors keff, relative radial power distribution of each configuration were also evaluated. From calculation results, twelve fuel loading schemes were chosen based on lowest uranium mass or smallest number of FAs loaded into the core. In addition, two full core configurations using VVR-KN and MTR FAs and consisting of beryllium reflectors, vertical irradiation facilities, horizontal neutron beam ports, etc. have been proposed for further consideration in thermal hydraulic calculations and safety analysis.

Two Putative Cypovirus-Encoded miRNAs Co-regulate the Host Gene of GTP-Binding Nuclear Protein Ran and Facilitate Virus Replication

microRNA (miRNA) plays important roles in regulating various biological processes, including host-pathogen interaction. Recent studies have demonstrated that virus-encoded miRNAs can manipulate host gene expression to ensure viral effective multiplication. Bombyx mori cypovirus (BmCPV), a double-stranded RNA virus with a segmented genome, is one of the important pathogens for the economically important insect silkworm. Our present study indicated that two putative miRNAs encoded by BmCPV could promote viral replication by inhibiting the gene expression of B. mori GTP-binding nuclear protein Ran (BmRan), an essential component of the exportin-5-mediated nucleocytoplasmic transport of small RNAs. BmCPV-miR-1 and BmCPV-miR-3 are two of the BmCPV-encoded miRNAs identified in our previous studies. BmRan is a common target gene of them with binding sites all located in the 3′-untranslated region (3′-UTR) of its mRNA. The expression levels of the two miRNAs in the midgut of larvae infected with BmCPV gradually increased with the advance of infection, while the expression of the target gene BmRan decreased gradually. The miRNAs and the recombinant target gene consisting of reporter gene mCherry and 3′-UTR of BmRan mRNA were expressed in HEK293T cells for validating the interaction between the miRNAs and the target gene. qRT-PCR results revealed that BmCPV-miR-1 and BmCPV-miR-3 negatively regulate target gene expression not only separately but also cooperatively by binding to the 3′-UTR of BmRan mRNA. By transfecting miRNA mimics into BmN cells and injecting the mimics into the body of silkworm larvae, it was indicated that both BmCPV-miR-1 and BmCPV-miR-3 could repress the expression of BmRan in BmN cells and in the silkworm, and the cooperative action of the two miRNAs could enhance the repression of BmRan expression. Furthermore, the repression of BmRan could facilitate the replication of BmCPV genomic RNAs. It is speculated that BmCPV-miR-1 and BmCPV-miR-3 might reduce the generation of host miRNAs by inhibiting expression of BmRan, thus creating a favorable intracellular environment for virus replication. Our results are helpful to better understand the pathogenic mechanism of BmCPV to the silkworm, and provide insights into one of the evasion strategies used by viruses to counter the host defense for their effective multiplication.

A Monte Carlo study on burnup treatment in sodium-cooled reactor with Th fuel

Sodium Cooled Reactors is one of the Generation-IV plants selected to manage the long-lived minor actinides and to transmute the long-life radioactive elements. This study presents the comparison between two-designed SFR cores with 600 and 800 MWth total heating power. We have analyzed a conceptual core design and nuclear characteristic of SFR. Monte Carlo depletion calculations have been performed to investigate essential characteristics of the SFR core. The core calculations were performed by using the Serpent Monte Carlo code for determining the burnup behavior of the SFR, the power distribution and the effective multiplication factor. The neutronic and burn-up calculations were done by means of Serpent-2 Code with the ENDF-7 cross-sections library. Sodium Cooled Fast Reactor core was taken as the reference core for Th-232 burnup calculations. The results showed that SFR is an important option to deplete the minor actinides as well as for transmutation from Th-232 to U-233.

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

The Effects of the Number of Coated Fuel Particles on the Neutronic Aspects of 25 MWt Pebble Bed Reactor with Thorium Fuel

High-Temperature Gas Reactor (HTGR) is a type of reactor that continues to be developed because of its advantages in terms of economic aspects, proliferation resistance, and safety aspects. One of the safety aspect improvements is due to the use of the Coated Fuel Particle (CFP). A coated fuel particle is a fuel with a diameter smaller than 1 mm and is protected by several carbon layers. In the Pebble Bed Reactor (PBR) type of HTGR design, the CFP is placed in a 6 cm fuel ball. How much CFP is put into the fuel ball will determine the neutronic characteristics of the reactor. In this study, the effect of the amount of CFP in the fuel ball on the 25 MWt PBR design using Thorium fuel and its impact on several important neutronic aspects, such as the effective multiplication factor, the amount of fuel enrichment, the utilization of fissile material, and the density of the fissile material formed. The calculation was performed by the Monte Carlo MVP / MVP-BURN code. This study found that the coated fuel particle fraction of 15% was the optimum value for the studied neutronic parameters.

EFFECT OF β ON EFFECTIVE MULTIPLICATION FACTOR IN 1/fβ SPECTRUM RANDOM SYSTEM

We investigated the effect of β on effective multiplication factor(keff) in the 1/fβ spectrum random system. The random system was generated by the 1/fβ noise model. The model is a continuous space model based on the Randomized Weierstrass function and describes the component spatial distribution with a power spectrum of 1/fβ, where f and β are the frequency domain variable and the characteristic parameter related to randomness, respectively. In this work, the two-group Monte Carlo calculations were performed to obtain the keff for a simple cubic geometry that consisted of two materials (fuel burned to 12 GWd/t and concrete). A large number of replicas having different spatial distribution and characterized by the representative β values were generated using the model, and the distribution on keff was analyzed. We found the dependency on β of standard deviation, skewness, and kurtosis of keff distribution. This result is expected to help to predict the keff distribution due to the randomizing model.

3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems. Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data. Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created. These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations. This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed.