bluuebunny/crossref_metadata_2025_split huggingface dataset:
Physics Design of Epi-Thermal Neutron Beam for BNCT Based on C-MNSR
Authors: Dan Peng , Xiaobo Wu , Jin Lu , Qian Hao , Jingyan Hong , Yiguo Li 
Published: October 2017

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) (235U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT.
The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 109n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 107n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10−11Gy · cm2 · n−1 and gamma dose contamination (Dγ/φepi) 2.18× 10−14 Gy·cm2·n−1.
Experimental Study of Performance of Cold Crucible Induction Melter in Melting Corium Simulant
Authors: Sun Caixin , Chen Ying , Chen Lian , Chang Huajian , Pei Jie , Tian Fang , Xu Xinchen 
Published: October 2017

Cold crucible induction melter (CCIM) technology has been widely used in melting of non-metallic refractory materials with a high melting point and a requirement of high purity. This paper discusses the application of the CCIM technology in melting of reactor corium (simulated as zirconium oxide in this experiment).
In this experiment, the power supply is set at constant power control mode, with a design frequency of ∼100 kHz and a total output power of ∼ 400 kW. The cold crucible consists of copper tubes and a stainless bottom plate which has a dimension of 220 mm in diameter and 420 mm in height. The crucible is charged with zirconium oxide (ZrO2) powder and surrounded by a set of water-cooling induction coils which is designed shorter than the crucible in order to reduce energy loss in the water-cooling bottom plate. The top of the crucible and the coils are at the same height. A Zirconium ring is used to initiate heating process. ∼ kg ZrO2 powder is used for each experiment. Electrical variables of the induction melter and temperature of the melt are detected and recorded in the entire heating process.
For the result, almost 3/4 of the ZrO2 powder is melt and a thin crust (∼ 2–3 mm) is generated between the melt and the crucible. To be specific, the ZrO2 powder above the bottom of the induction coils is melt, but others are not. In the melting process, the maximum surface temperature of the melt has reached 2230°C. Furthermore, for better understanding of the relationship between the melting process and the changing process of the electrical variables, an equivalent circuit of the heating system is established and analyzed. It can be concluded that some electrical variables such as the quality factor (Q) of the induction coils and the practical frequency can reflect melting status. The conclusion is confirmed by the records transmitted by sensors in the melting process. By comparison between the recorded data and the melting status, it is clear that the Q value can be accurately used in indicating the melting status due to its obvious change in the entire melting process. As the melting pool becoming larger, the Q value tends to be smaller. In contrast, the practical frequency becomes higher as the melt becoming larger. Furthermore, both of them are not influenced by the output power. In addition, the output current and the output voltage also can be used to describe the melting status. In specific, a larger melting pool corresponds to a lower output current and a higher output voltage.
Conceptual Design for CHP System for Class III Power
Authors: Sarvendran Vajran , Patel Heli , Harvel Glenn 
Published: October 2017

Nuclear power plants (NPPs) typically have four classes of electrical supply systems to provide redundant and resilient power to ensure safe operation, cooling, and shutdown of NPPs. Class III emergency power traditionally uses standby diesel generators to fulfill equipment electrical demands. This requires a reliable source of diesel energy and reliance on one particular kind of fuel source for backup power.
During the North American blackout of 2003, hospitals had to run on backup generators that were mostly diesel fired. A prolonged blackout could have negatively affected hospital infrastructure and patient health. With current diesel generators, capable of supporting critical systems for up to 72 hours (for most NPPs), a prolonged loss of power or transient could similarly also have adverse effects on NPPs.
A Combined Heat and Power (CHP) system is a power generation system that generates electricity, in addition to concurrently having heating and cooling capabilities. Often, gas or steam turbines are used to generate electricity. CHP heating and cooling capabilities can be met via absorption coolers and heat pumps. Low-grade exhaust heat from turbines could be used for the absorption processes. Absorption coolers/heaters produce cold/warm fluid using a heat source via the vapor compression cycle, taking advantage of high affinity fluids. The absorbent allows for the refrigerant to boil at lower operating conditions, which allows for heat transfer.
CHPs’ that are capable of accepting natural gas as a form of fuel will increase resiliency of NPP power supply. This ensures a reduced risk of running out of fuel during prolonged transients due to continual supply via pipelines already in place.
The objective of this work is to generate a conceptual design of a multi-fuel source CHP system that is capable of at least accepting natural gas as an alternative to Class III diesel generators for a NPP. The system will be capable of supplying Class III power to Small Modular Reactor (SMRs), as well as commercial NPPs, such as CANDU, PWR, and BWR types.
Current state of the art Class III backup power systems, CHP systems, CHP thermodynamic cycles, multiple compatible fuels, as well as absorber heaters and coolers have been investigated. Ranking systems were used to determine the top three designs. The parameters included turbine type, efficiency, flow rates, operating temperature/pressure, fuel type, fuel to energy ratio, fuel availability, absorber chiller/heater fluid efficiency, operating temperature and thermal conductivity.
Based on the ranking system parameters, a thermodynamic model including mass, energy and entropy balance of an 8 MWe CHP with heating and cooling capability between the ranges of 15°C to 27°C was conceptually designed.
HEPA Filter Clogging and Volatile Material Release Under Solvent Fire Accident in Fuel Reprocessing Facility
Authors: Ono Takuya , Watanabe Koji , Tashiro Shinsuke , Amano Yuki , Abe Hitoshi 
Published: October 2017

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures.
One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters.
We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.
Feasibility Study on the Natural Circulation Cooling Procedure at the Loop Unbalanced Condition
Authors: Sakuma Wataru , Miyata Shinya , Maruyama Manabu , Ogawa Junto 
Published: October 2017

In typical pressurized water reactor (PWR) plant, in case that one steam generator (SG) is dried out and cannot be credited for the primary cooldown, at least one reactor coolant pump (RCP) has to be operated in order to homogenize the primary coolant temperature distribution among loops when the plant is cooled down to the cold shutdown state. For example, an accident such as steam line break (SLB) and feedwater line break (FLB) leads to this situation. If the natural circulation condition is established due to unavailability of all the RCPs, the natural circulation in the primary loop connected to the affected SG would be interrupted in the plant cooldown phase. In this situation, the continuous cooldown disturbs the smooth depressurization because it leads to void generation at the top of the affected SG tube where the high temperature coolant is left.
In addition, there is a possibility that all RCPs cannot be operated in case of the earthquake or the fire if the RCPs are not earthquake-proof and fire-resistant. Therefore the establishment of the cooldown procedure without RCPs operation under the temperature unbalanced condition among the primary loops can contribute to the safety enhancement for typical PWR plants.
The several experiments have been already performed to observe the natural circulation phenomena under the temperature unbalanced condition. It has been reported that the plant can be continuously cooled down with smooth depressurization by stepwise cooling manner using MSRVs of the intact SGs.
In this study, Mitsubishi Heavy Industries, Ltd. (MHI) performed the transient analyses to simulate the natural circulation cooldown test under the temperature unbalanced condition among loops performed by Large Scale Test Facility (JAEA ROSA/LSTF) using M-RELAP5, which was a modified plant system transient code by MHI based on RELAP5-3D. Based on the analysis results, the thermal hydraulic phenomena of natural circulation cooldown under the temperature unbalanced condition were investigated. As a result, the mechanism of natural circulation interruption was clarified, and this paper shows the outline of the cooldown procedure under the temperature unbalanced condition which could be applied to the PWR plants.
Experiment and Numerical Analysis of Mixing Process of Two Component Gases in Vertical Fluid Layer
Authors: Takeda Tetsuaki 
Published: October 2017

A depressurization accident is the one of the design-basis accidents of a Very-High-Temperature Reactor (VHTR). When a depressurization accident occurs, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Therefore, it is important to understand the mixing processes of different kinds of gases in the stable and unstable stratified fluid layers. In particular, it is also important to examine the influence of localized natural convection and molecular diffusion on the mixing process from a safety viewpoint. Therefore, in order to predict or analyze the air ingress phenomena during a depressurization accident, it is important to develop a method for the prevention of air ingress during an accident. We carried out experiments and numerical analysis using three-dimensional (3D) CFD code to obtain the mixing process of two-component gases and the flow characteristics of localized natural convection. This study also investigated a control method for the natural circulation of air through the injection of helium gas.
The numerical model consists of a storage tank and a reverse U-shaped vertical rectangular passage. They are separated by a horizontal partition plate. One sidewall of the high-temperature side vertical passage is heated and the other sidewall is cooled. The low-temperature vertical passage is cooled by ambient air. The storage tank is filled with heavy gas and the reverse U-shaped vertical passage is filled with a light gas. In the vertical passage of the high-temperature side, localized natural convection is generated by the temperature difference between the vertical walls. The results obtained from the experiments were quantitatively simulated using 3D numerical analysis. The two component gases were mixed via molecular diffusion and natural convection. After some time elapsed, natural circulation occurred through the reverse U-shaped vertical passage. These flow characteristics are the same as those of phenomena generated in the passage between a permanent reflector and a pressure vessel wall of the VHTR.
Design Investigation on Flow Diffuser With Code_Saturne: CFD Simulation Analysis
Authors: Xu Tingting , Min Jiesheng , Bellet Serge , Howard Richard , Alvarez Dominique , Chen Guofei 
Published: October 2017

The flow rate distribution at the entrance of the core plays a key role for reactor design since it has important implications for the performance, and efficient safety of a nuclear reactor. When the coolant passes from the downcomer to the core, it changes direction due to the inertia force and the curvature of the bottom vessel head. The internal components inside lower plenum work to homogenize the flow distribution. Their purpose is to prevent the formation of instabilities and the creation of vortices due to the flow reversal. In the frame of EDF’s new reactor design there is a desire to identify an optimal flow diffuser. The future intention is to study five different types of flow diffuser including EPR, VVER, Konvoï, APR+ and Westinghouse to look at the pros and cons of each design. The authors underline that the geometries of each Reactor Pressure Vessel (RPV) and associated diffuser device are quite different therefore a generic form needs to be used to make an equivalent comparison. The goal of the present work is to find the optimal mesh refinement and associated numerical parameters for the simulation of the lower plenum flow. This work is a preliminary step for a future study to compare existing diffuser concepts. Thus in the future work only the section containing the flow diffuser structure will be changed.
The PIRT methodology is applied to better define the physical phenomena and key parameters that will influence the flow distribution at the entrance of the core. In order to better understand the fluid distribution and the function of the diffuser component, 3D computational fluid dynamics (CFD) simulations are launched to improve our knowledge on the flow pattern inside the lower plenum.
Both the geometry and mesh are generated by Salomé1. Simulations are carried out using Code_Saturne2, an EDF in-house open-source CFD code. The generic test case is a 1/5 scale EDF “BORA” 4 loop mock-up with a flow rate of 0.1 m3/s injected into each cold leg. The unsteady flow algorithm with standard k-epsilon turbulence model has been used with a full explicit meshing except for the reactor core where a porous approach is adopted. The physical time for each calculation case is 5s for a converged simulation. Mesh sensitivity tests have been carried out ranging from 8 million cells to 28 million cells. A mesh of 22 million cells is found to provide the most appropriate balance between simulation quality and feasibility. Due to the size of the simulations, high performance computers are necessary to provide timely results. The results indicate that CFD can provide extra capacity to engineers for reactor design to evaluate the pros and cons of different existing diffuser concepts.
Study of the Impact of Stellite Material on PWR Co-60 Contamination
Authors: Zhang Liying , Wang Yaxiao , Wang Xiaoxia , Gao Bo , Mi Aijun , Mao Yawei 
Published: October 2017

According with the requirements of HAF-102[1], the purpose of radiation protection optimization is to achieve occupational exposures as low as reasonably achievable (ALARA). More than 80% of the collective exposure doses are received during the outage for PWRs. Activated corrosion products, especially Co-60, deposited on the surface of reactor coolant system are the main causes of these doses. Co-60 is the principle contributor to out-of-core radiation fields in PWRs, which is a high energy gamma emitter with a 5.3 year half-life period. The contamination by Co-60 plates out in plant piping systems and produces long term high radiation areas in the plant. Stellite, a hard facing alloy trade, which has high Co-59 content (approximately 60%), is identified as one of the most important source of Co-60 in reactors, whereas the contribution of Stellite to cobalt are always not paid enough attention to. Cobalt is released through ware and corrosion of Stellite. When cobalt is released in systems with a flow path to the reactor, it can become activated by neutron to Co-60.
This paper analyzes the Stellite material release into the primary coolant. Based on the above considerations, the impact of Stellite replacement on plant contamination by Co-60 is calculated and analyzed with the use of activated corrosion product source term calculation program. The primary conclusions are as follows: (1) Reducing the quantity of Stellite material used during the design and building of PWRs can significantly reduce the Co-60 deposition on primary circuit. (2) The replacement of Stellite material has a positive impact on PWR contamination by Co-60, but the outcome is not so significant. Based on the above analysis, recommendations are made. Stellite material replacement project can be used for the operating reactors to reduce the Co-60 contamination. But for the design and building of new reactors, reduction of the use of Stellite material is the best choice.
US Department of Energy’s CoDCon Project: An Aqueous Safeguards R&D Program
Authors: Bresee James C. , Paviet Patricia D. , Todd Terry A. 
Published: October 2017

In 2015, Office of Nuclear Energy of the U.S. Department of Energy established a research and development project at Pacific Northwest National Laboratory to evaluate the process control capability to produce a specific uranium/plutonium product from used commercial nuclear fuel using an aqueous co-decontamination separations process. The process is controlled using on-line instrumentation supported by a dynamic process model. The new program is called the CoDCon project.
The result of the study will be a quantitative measure of the current capability to produce a specific U/Pu product, using U-IV to reduce plutonium to Pu-III and prepare a mixed product, the composition predicted by a dynamic model, measured with on-line instrumentation and controlled by the adjustment of process variables. This approach would be an alternative to the use of the PUREX aqueous separations process that produces separate plutonium and uranium products that are later blended to prepare the desired mixture. Since plutonium is always accompanied by uranium, the project will provide a safeguards-by-design tool for possible use in future commercial separations plant designs. The effectiveness of the design tool will be quantified using a methodology published by B. Cipiti, et al1.
The paper describes the process and the control system used by the project and provides details on the current status of the research and development program. Since this advanced process control system may have international applications, arrangements will be described for possible foreign participation in the project.
Development of Five-Equation Drift Flux Model for Thermal Non-Equilibrium Phenomena in LWRs
Authors: Shi Shanbin , Zhang Xiaoqin , Sun Xiaodong , Zhang Hongbin 
Published: October 2017

The classic four-equation drift-flux model treats the two-phase flow as a mixture to formulate mass, momentum, and energy balance equations. The dispersed phase is modeled in the mass balance equation to describe the mass transfer between the two phases. However, this model implies an assumption of the thermal equilibrium in the mixture energy balance equation. Therefore, a five-equation drift-flux model and its constitutive equations have been developed to relax the thermal equilibrium assumption in the classic four-equation drift-flux model when it is applied to thermal non-equilibrium phenomena in Light Water Reactors (LWRs). The additional energy balance equation of the dispersed phase has been introduced to address thermal non-equilibrium phenomena in LWRs in the five-equation drift-flux model. In addition, the newly added energy equation for the dispersed phase does not change the basic kinematic parameters and uses the mixture velocity instead of the dispersed phase velocity.
The hyperbolicity of the drift-flux model system equations is investigated by examining if the partial differential equations have only real eigenvalues. The results show that the five-equation drift-flux model is a well-posed hyperbolic model, which can be solved by both advanced solvers using the finite element method, such as MOOSE Framework, and the traditional solvers using finite difference method, such as RELAP5 code.
In developing the thermal non-equilibrium five-equation drift-flux model, it is necessary to select, improve, or develop constitutive models as closure relations for variables used in the field equations. In our current study, a literature survey has been conducted to review appropriate constitutive models and correlations for the dispersed phase mass generation rate, interfacial energy transfer, distribution parameter, and drift velocity in the five-equation drift-flux model.