bluuebunny/crossref_metadata_2025 huggingface dataset:
Bleed and Feed Analysis for Loss of Secondary Heat Sink to Support the Development of EOPs
Authors: Yang Changjiang , Huang Shuliang , Zhan Jingxiang
Published: October 2018

The loss of secondary heat sink accident caused by loss of all feed water for a PWR with medium pressure safety injection system (MPSI) design is analyzed in order to support the development of one of the emergency operating procedures (EOPs), “response to loss of secondary heat sink”. The bleed and feed (B-F) measure to mitigate the accident is analyzed in detail. For the bleed and feed measure, the safety injection flow is established only after the reactor coolant system (RCS) pressure is lower than the pump head of MPSI, which needs the open of the pressurizer (PRZ) safety valves to depressurize the RCS ahead. To prevent the core uncovery or minimize the fuel cladding temperature during the bleed and feed process, the time and the number of PRZ safety valves needed to be opened are investigated. A best estimated plant model with RELAP5 code is applied for the sequences calculation. The result shows that the core water level and fuel cladding temperature will be benefited from an earlier open of all safety valves.
Assessment of Time and Space High-Order Schemes for Two-Fluid Seven-Equation Two-Pressure Model Using the Reversed Water Faucet Problem
Authors: Chao Fei , Shan Jianqiang , Gou Junli , Wu Pan , Ge Li
Published: October 2018

The Ransom water faucet problem has become one of the most important benchmark tests to study two-fluid/phase flow because it exists an analytical discontinuous solution. The water faucet problem is a gravity-driven wave problem and its analytical solution is derived from the liquid free fall motion. In this paper, the opposite gravity-driven wave problem named the reversed water faucet problem where the liquid admits a rising motion with reduced speed driven by gravity is studied. With assumptions of decoupled phasic pressures, approximate incompressible flow, no liquid pressure gradient, no phase change, no wall and interfacial drag, the analytical solutions of the gas volume fraction and liquid velocity distribution are derived. From the gas volume fraction analytical solution, there is a moving discontinuity which is a very important point for testing accuracy of the numerical scheme and its stability near discontinuities. Two-fluid seven-equation two-pressure model is of particular interest due to the nature of inherent well-posed advantage in all situations. What’s more, high-order accuracy schemes have attracted great increasing attention to overcome the challenge of serious numerical diffusion from 1st-order scheme for accurately simulating many nuclear thermal-hydraulics applications such as long term transient natural circulation problems. In this paper, the solution algorithms with high-order accuracy in space and time are developed for this well-posed two-fluid model and its robustness and accuracy are verified and assessed against the derived analytical solutions. The numerical results show that high-order schemes could prevent excessive numerical diffusion and are more accurate than first-order time and space schemes; the space high-order scheme could give more accurate numerical results than the time high-order scheme for discontinuous solutions.
Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator
Authors: Huang Jun , Gou Junli , Ma Haifu , Fan Jie , Shan Jianqiang
Published: October 2018

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.
Upgrade of Accident Sequence Model of PRA for LOCA Inside PCV
Authors: Yamagishi Shohei , Tanno Shunsuke , Sato Teruyoshi , Saito Toshiteru , Hiraide Masayuki , Kita Toshinobu
Published: October 2018

In Japan, Risk-informed activity using PRA is becoming increasingly active and Reactor Oversight Process (ROP) will be introduced in 2020. Then, PRA has attracted more attention, and it is necessary to immediately develop its infrastructures.

Therefore, as the first step, TEPSYS/TEPCO conduct to upgrade the internal event level 1 PRA that is the basis of all events targeting unit 7 (ABWR) of TEPCO Kashiwazaki-Kariwa Nuclear Power Plant. In particular, LOCA inside PCV for all piping connecting RPV including instrumentation and control piping reclassified based on its break size, break location, success or failure of the vapor suppression function, the accident progression during mission time, and PCV control procedures.

Especially, the finding presented in this paper address the newly grouping that break location is classified into not 2 (steam phase/liquid phase) groups [1] but 3 groups as below.

✓ RPV nozzles are located above L8 reactor water level (steam phase)

✓ RPV nozzles are located between TAF and L8 reactor water level (steam phase / liquid phase)

✓ RPV nozzles are located below TAF reactor water level (liquid phase)

In the Results, completeness of the accident sequence for LOCA inside PCV was improved to be applied to risk-informed activity and the proactive safety review.
Development and Application of Multi-Physics Safety Analysis Code for Advanced Liquid Metal Cooled Reactor
Authors: Wang Chi , Zhang Xuebei , Feng Jingchao , Shehzad Khan Muhammad , Ye Minyou , Chen Hongli
Published: October 2018

The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.
Effects of Parameters on the Two-Phase Flow Instability in a Microchannel
Authors: Liu Yefei , Liu Yang , Yang Xingtuan , Pan Liqiang
Published: October 2018

Series of experiments are conducted in a single microchannel, where subcooled water flows upward inside a transparent and vertical microchannel. The cross section of the channel is rectangle with the hydraulic diameter of 2.8mm and the aspect ratio of 20. The working fluid is 3–15K subcooled and surface heat flux on the channel is between 0–3.64 kW/m2, among which two-phase instability at low vapor quantity may occur. By using a novel transparent heating technique and a high-speed camera, visualization results are obtained. The parameters are acquired with a National Instruments Data Acquisition card. In the experiments, long-period oscillation and short-period oscillation are observed as the primary types of instability in a microchannel. Instability characteristics represented from signals correspond well with the flow pattern. Moreover, effects of several parameters are investigated. The results indicate that the oscillating period generally increases with the heat flux density and decreases with inlet subcooling, while the effects of inlet resistance are more complex.
Qualification Test of APR1400’s RCP Seal Under Extended SBO Condition
Authors: Cho Seok , Kim Seok , Bae Byoung-Uhn , Cho Yun-Je , Kim Yeon-Sik , Jeon Woo-Jin , Youn Young-Jung , Chu Sung-Min , Bang Sang-Youn
Published: October 2018

The US NRC’s post-Fukushima taskforce recommend extending the coping time up to at least 72 hours. In PWRs, the reactor coolant pump (RCP), and more particularly the shaft seal system located between the motor and the impeller, are directly impacted by these recommendations [AREVA, 2012]. Considering the importance of station black out (SBO) condition, seal performance of RCP shaft seal assembly shall be verified especially under an extended SBO condition. As a final stage of the localization program of the RCP, KAERI performed a seal LOCA test to verify seal leakage characteristics and to produce qualified experimental data for a plant safety analysis during the extended SBO condition. The seal verification test was performed using the localized seal assembly, fabricated by DOOSAN Heavy Industries & Construction Company of Korea, along with the commercial APR1400’s RCP for the Shin-Kori unit-5 nuclear plant to simulate an actual condition of the commercial nuclear power plant. Test condition was derived from a series of safety analysis for the APR1400 plant which reflected the FLEX procedure, and it consisted of three major sub-conditions such as high-pressure and high-temperature condition, saturated condition, and low-pressure and low-temperature condition. The major concerning point of the test is to measure a seal leakage flow rate during the test period. Total leakage amount and maximum leakage flow rate through the localized shaft seal assembly during the whole period of the test are in comparatively smaller ranges than the conventional input values used for a safety analysis of APR1400.
Evaluation of Heat Removal During the Failure of the Core Cooling for New Critical Assembly
Authors: Eguchi Yuta , Sugawara Takanori , Nishihara Kenji , Tazawa Yujiro , Tsujimoto Kazufumi
Published: October 2018

The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important.

This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system).

The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient.

Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.
Analysis of the Condensation Model Used in Severe Accident Integral Codes
Authors: Zhu Yiming , Liu Zhuo , Yang Xiaoming
Published: October 2018

Steam condensation plays a key role in removing heat from the containment atmosphere, therefore the calculation of the steam condensation has a significant effect on the judgment of severe accident process. This paper discussed the calculation models of condensation used in severe accident integral codes.

Firstly two kinds of condensation calculation models introduced and analyzed for application in integral codes: models based on experiment correlations — Uchida model, Dehbi model, and the models based on analogy of heat-mass transfer — Kreith model and Collier model. And then two calculation models based on Kreith model were proposed: simplified calculation model and completely iterative calculation model.

After that, a heat structure calculation code was designed and the two model was applied. The results of the code with the two model along with Uchida and Dehbi models were compared with the experiment result of Wisconsin. It was found that the simplified model had a satisfied calculation speed and the relative errors were under 13% in total situations, which was acceptable for the integral code. Besides, it was shown that the completely iterative calculation model had smaller relative errors, which was under 11%. In addition, this paper also investigated the effects of some parameters on the heat transfer coefficient for these different models, the velocity of the bulk of gas, length of surface, pressure of the gas and mole fraction of air were selected and studied separately.
Free Release of Radioactive Waste Containing Very Low Level Waste and Short Lived Radionuclides at Nucleco
Authors: Andreozzi C. , Bianchilli B. , Dodaro A. , Gagliardi F. , Mauro E. , Sisti M.
Published: October 2018

Since August 2015 a new classification of radioactive waste was issued by Italian Ministry of Economic Development, in order to adapt Italian historical classification to European standards. This new classification provides 6 categories, from exempt to high level waste, and it is based on the waste final destination: from free release to final disposal or interim storage (high level waste and intermediate level waste with α-content higher than 400 Bq/g) [1].

Nucleco is a State owned Company acting as Waste Management Organization for radioactive waste coming from hospitals, industries and research and development activities not related to electricity production by nuclear plants. Nucleco collects, safely manages and temporarily stores waste that will be sent to the National Repository (site definition phase is still ongoing), while for the Short Lived Radionuclides and Very Low Activity waste Nucleco performs all necessary operations to be compliant with the conditions of release prescribed by the Italian Control Authority.

Short Lived Radionuclides are those whose half-life is shorter than 100 days or reach the condition of non-radiological relevance in 5 years: they are mainly produced by bio-medical applications of radioactive materials. Very Low Activity waste are characterized by activity concentrations lower than 100 Bq/g (of which less than 10 Bq/g of α-emitting radionuclides) and reach the condition of non-radiological relevance in 10 years: these waste usually came from research institutions and industrial activities.

This work presents the authorized operating procedures, the radiological measurements criteria and the technical know-how put in place by Nucleco to fulfil the provisions of Italian regulations for unconditioned release of radioactive waste.

A case study of ISO 20’ containers is discussed in the current paper.

Main emphasis will be addressed to:

• gathering of historical information about the state of the material to be released and definition of the reference radiological spectrum;

• sampling procedures to ensure representativeness of the samples from homogeneous waste batch to be released and then subjected to radiological characterization;

• characterization phase consisting of the integration of several state-of-art techniques aiming to collect the most complete set of radiological data;

• data processing protocols needed for the calculation of the activity concentrations for each radionuclide of the reference spectrum (or other radionuclides eventually detected);

• evaluation of the main sources of uncertainty affecting the results;

• comparison of the activity concentration (including the uncertainty) of each radionuclide with the corresponding authorized concentration limits.
Application of RELAP/SCDAPSIM/MOD4.1 to the Analysis of Advanced Reactor/Fluid Systems With Liquid Molten Salt in the Presence of Non-Condensable Gases
Authors: Jiang S. , Perez-Ferragut M. , Fu Z. , Hohorst J. K.
Published: October 2018

In recent years, organizations both at home and abroad are actively carrying out a research on the Molten Salt Reactor systems (MSRs). For example, the Shanghai Institute of Applied Physics (SINAP), Chinese Academy of Science (CAS), is currently involved in the design and development of a 10MWth Solid Fuel Thorium Molten Salt Reactor (TMSR-SF1). SINAP started their analysis of TMSR using an earlier version of RELAP/SCDAPSIM, MOD4.0. MOD4.0 included models and correlations for molten salts but was unable to treat molten salts in the presence of non-condensable gases. Since that time SINAP and ISS have worked in parallel to extend the models and correlations for such systems. The SINAP modified code, using SINAP proprietary models and correlations, is described in the “open literature” under the name RELAP5-MSR. More general, but comparable, models developed by ISS for liquid metals/salts in the presence of non-condensable have been incorporated into RELAP/SCDAPSIM/MOD4.1. This extended option is currently being implemented for Li-Pb, Pb-Bi, molten salts, and Na.
A Numerical Study on Graphite Dust Deposition on Steam Generator Tubes in the High-Temperature Gas-Cooled Reactor (HTGR)
Authors: Wei Mingzhe , Zhang Yiyang , Fang Zhu , Wu Xinxin , Sun Libin
Published: October 2018

Graphite dust is an important issue for the operation and maintenance of high-temperature gas-cooled reactor (HTGR), because the transport of fission product (FP) is coupled closely with graphite dusts. For instance, vapor phase FP could condense as flowing through the steam generator (SG) and deposit on the surface of graphite dusts that are either air-borne or already deposited on SG tubes. In water ingress or loss-of-coolant accidents, these dusts may re-suspend and contribute to the source term. Despite the importance of graphite dusts in HTGRs, the transport and deposition of dust particle are far from being fully understood, neither particle-fluid nor particle-wall interactions. In this work we present a numerical study on the particle transport through upper 5 layers of SG tubes. Particularly, the particle impaction process is simulated by Finite Element Method (FEM) with adhesion and dissipation specially accounted. The FEM simulation predicts the critical adhesion velocity and restitution coefficient when rebound occurs. Then we substitute the particle impaction model into Eulerian-Lagrangian simulation of flow field and extract the deposition rate statistically. The result shows that for small particles (< 5 μm), the deposition rate is controlled by the collision rate, which is mainly determined by the interaction between turbulence and thermophoresis. The particle-vortex interaction is essentially important for the distribution of particles near wall and thus influences the deposition rate. For large particles the deposition rate is more affected by the sticking efficiency, which is simultaneously controlled by both the critical adhesion velocity and normal impaction velocity. Therefore, the deposition rate first increases then decreases with particle size and reaches maximum at about 5 μm.
Study on Thermal-Hydraulics Characteristics of the Flat and High-Thermal-Conductivity Core-Catcher
Authors: Takeyama Daiki , Iwaki Chikako , Onitsuka Yoichi , Tahara Mika
Published: October 2018

For mitigation of severe accident with core melting, core-catcher has been developed to catch and cool molten core. The core-catcher developed by Toshiba for Advanced Boiling Water Reactor was designed to be installed under Reactor Pressure Vessel (RPV) and catch molten core in basin with thermal-resistant material. It also has structure including risers and downcomers to generate natural circulation flow of cooling water. On the other hand, there is not enough space to install it in the existing plants due to the height of the inclined cooling channels.

Then, we have been developing flat core-catcher with flat cooling channels for the existing plants to reduce total height of the structure. Finned channels will be adopted to increase heat transfer rate by increasing heat transfer area. However, the thermal-hydraulics characteristics of such core-catcher has not been clarified due to the specific configuration, that is, the horizontal rectangular finned channel with the heated surface from upper side.

This present study investigated the natural circulation characteristics and heat transfer behavior in the horizontal rectangular finned channel by experiment. Pressure drop, natural circulation flow and temperature were measured by changing heat fluxes. The flow was visualized to obtain flow pattern in the finned channel by a high-speed camera. The maximum value of test range of heat flux was 250 kW/m2, which is the value when the total amount of the molten core would be dropped to the core-catcher. The partial simulated test section of finned channel with 3.5 m length was heated from upper side by heaters to simulate the heat flux from the molten core. This length is the same as the inner diameter of RPV pedestal of the existing plants.

The natural circulation mass flux increased as the heat flux increased and then two-phase flow pressure loss also increased. Consequently, circulation flow turned to decrease. As a result of the test, when the heat flux was 50 kW/m2, the circulation mass flux got to the maximum value, 230 kg/m2s. Under all the conditions except for the maximum heat flux of 250 kW/m2, the fin surface temperature was around the saturated temperature. At maximum heat flux 250 kW/m2, the temperature got 540 K. However, the structural soundness was maintained because it is lower than melting temperature of the fin. It can be concluded that the flat and high-thermal-conductivity core-catcher has enough cooling performance to catch and stabilize the molten core.
FMEDA and FIT-Based Safety Assessment of NPP I&C Systems Considering Expert Uncertainty
Authors: Yasko Alexander , Babeshko Eugene , Kharchenko Vyacheslav
Published: October 2018

The complexity of modern safety critical systems is becoming higher with technology level growth. Nowadays the most important and vital systems of automotive, aerospace, nuclear industries count millions of lines of software code and tens of thousands of hardware components and sensors. All of these constituents operate in integrated environment interacting with each other — this leads to enormous calculation task when testing and safety assessment are performed.

There are several formal methods that are used to assess reliability and safety of NPP I&C (Nuclear Power Plant Instrumentation and Control) systems. Most of them require significant involvement of experts and confidence in their experience which vastly affects trustworthiness of assessment results.

The goal of our research is to improve the quality of safety and reliability assessment as result of experts involvement mitigation by process automation. We propose usage of automated FMEDA (Failure Modes, Effects and Diagnostic Analysis) and FIT (Fault Insertion Testing) combination extended whith multiple faults approach as well as special methods for quantitative assessment of experts involvement level and their decisions uncertainty. These methods allow to perform safety and reliability assessment without specifying the degree of confidence in experts.

Traditional FMEDA approach has several bottlenecks like the need of manual processing of huge number of technical documents (system specification, datasheets etc.), manual assignment of failure modes and effects based on personal experience. Human factor is another source of uncertainty. Such things like tiredness, emotional disorders, distraction or lack of experience could be the reasons of under- and over-estimation.

Basing on our research in field of expert-related errors we propose expert involvement degree (EID) metric that indicates the level of technique automation and expert uncertainty degree (EUD) metric which is complex measure of experts decisions uncertainty within assessment. We propose usage of total expert trustworthiness degree (ETD) indicator as function of EID and EUD.

Expert uncertainty assessment and Multi-FIT as FMEDA verification are implemented in AXMEA (Automated X-Modes and Effects Analysis) software tool. Proposed Multi-FIT technique in combination with FMEDA was used during internal activities of SIL3 certification of FPGA-based (Field Programmable Gate Array) RadICS platform for NPP I&C systems. The proposed expert trustworthiness degree calculation is going to be used during production activities of RPC Radiy (Research and Production Corporation).

Our future work is related to research in expert uncertainty field and extension of AXMEA tool with new failure data sources as well as software optimization and further automation.
Experimental Investigation and Flow Visualization of the Two-Phase Flow Instability at Low Vapor Quality in a Vertical Narrow Channel
Authors: Pan Liqiang , Liu Yang , Li Weihua , Liu Yefei
Published: October 2018

The two-phase flow instability of forced convection has been experimentally investigated in a vertical narrow channel with the hydraulic diameter of 2.857mm and aspect ratio of 20. Transparent, metallic and conductive films on external surfaces of the test section can provide visualization and uniform heating for deionized water. The heat flux is 6–18.2 kW · m−2. When the instability occurs at low vapor quality, a series of parameters are measured and visualized images are obtained by a high-speed camera. The results show that the large amplitude of pressure drop between the inlet and outlet in the test section is due to the elongated bubble, and the value of pressure drop is positively correlated with the volume of the bubble. The oscillation period of pressure drop decreases with the increase of heat flux, and the period can be determined by the method of the Fast Fourier Transform. The backflow phenomenon is analyzed, which has a greater effect on the oscillation of pressure drop than bubble nucleation, bubble growth, bubble coalescence and recoiling of bubble boundary.
The Dose Constraint Calculation of High Radioactivity Level Waste Canister Surface
Authors: Yang Bo , Wei Qianglin , Wu Hexi , Luo Xujia , Liu Yibao
Published: October 2018

Radiation dose and personnel protection are among the safety goals of geological disposal of high-level radioactive waste. The calculation of the dose field on the surface of the packaging container is of great significance for the research on the dose constraint value of the repository. This paper built model consulting the Sweden KBS-3 canister, the temporal and spatial distribution of the dose rate on canister surface was calculated by Monte Carlo method, the temporal and spatial distribution of radiation dose rate of the tunnel was obtained. The research results showed that the photon dose rate on canister surface was greater than the neutron dose rate by 4 to 6 orders of magnitude, and the dose value of repository tunnel within 100 thousand years was lower than the ICRP recommended dose limit value (0.3 mSv/a) by 5 orders of magnitude.
The Compatibility of Nuclear Graphite With Molten Salt in the Molten Salt Reactor
Authors: He Zhoutong , Tang Hui , Zhang Can , Gao Yantao , Xia Huihao , Zhou Xingtai
Published: October 2018

In thermal Molten Salt Reactors, the nuclear graphite core is in direct contact with the molten salt coolant. Due to the porous nature of nuclear graphite, the molten salt may infiltrate the nuclear graphite, which may affect the mechanical strength and irradiation behavior of the nuclear graphite. In order to evaluate the infiltration behavior of molten salt in nuclear graphite, both FLiNaK and FLiBe salts were used to infiltrate two typical nuclear graphite grades: IG110 and NBG18. The pressure dependence of the infiltration weight gain ratio was measured. The influence of molten salt infiltration on the thermal properties of these two graphite grades, such as their thermal expansion behavior and thermal conductivity, was also measured. The mechanical strength of the FLiNaK-infiltrated graphite was measured at room temperature and elevated temperature, and showed that the mechanical strength of the nuclear graphite was enhanced at room temperature and weakened at elevated temperature by molten salt infiltration. Finally, the thermal expansion coefficient and the fracture surface analysis measured after FLiNaK infiltration indicated that the stress induced by molten salt infiltration could be one of the reasons for the graphite property changes.
Study of Fission Product Behavior in Containment Vessel Using Modified ART Mod 2: Update of Cesium and Iodine Compound Models
Authors: Vechgama Wasin , Silva Kampanart
Published: October 2018

From the Fukushima accident, Thailand has studied fission product behavior in containment vessel using ART Mod 2 code. Cesium iodide (CsI), cesium hydroxide (CsOH) and iodine (I2) behaviors are studied using modified ART Mod 2 code. However, there are other compounds which are not included in the codes especially cesium and iodine compounds such as from Phébus FPT3 experiment including cesium molybdate (Cs2MoO4), cesium telluride (Cs2Te), methyl iodide (CH3I) and iodine pentoxide (I2O5). The paper objective is to add the four compounds in the codes in order to enlarge the coverage of the code in evaluation fission product behavior in the containment vessel. Physical parameters and models of the four compounds are updated in the codes. It is found that deposition phenomena of Cs2MoO4, Cs2Te CH3I and I2O5 are close to the experiment in case of no chemical reaction.
Study on the Turbulent Mixed Convection Phenomena Inside the Air-Cooled RCCS Riser
Authors: Shin Dong-Ho , Kim Sin-Yeob , Kim Chan Soo , Park Goon-Cherl , Cho Hyoung Kyu
Published: October 2018

High Temperature Gas cooled Reactor (HTGR) that is one of GEN-IV reactor types enhances its safety adopting a passive cooling system, Reactor Cavity Cooling System (RCCS). When the active cooling system for the reactor core in a HTGR doesn’t work, the decay heat from the reactor core is transferred to the reactor vessel and the concrete wall of the reactor cavity, which are cooled by the RCCS. The RCCS consists of the vertical rectangular ducts, called risers, surrounding reactor vessel at a certain distance and chimneys that are connected to risers. Risers receiving the decay heat from the reactor vessel, the air inside the riser is heated and flows up to the chimney which accelerates exhalation of the air to the external atmosphere. The RCCS performance depends on the heat transfer rate inside the riser ducts, but the turbulent mixed convection that may occur in the riser ducts can complicate the verification. In this study, an experimental facility had been constructed to investigate heat transfer phenomena inside a riser duct and experiments with various heat flux and flow rate conditions were carried out. The experimental results showed that the turbulent mixed convection occurred for certain experimental conditions in the riser duct, which leads to heat transfer deterioration. Therefore, a correlation was suggested to predict how much the heat transfer could decrease compared to the forced convection.

Large eddy simulation was carried out for CFD analysis. The heat transfer coefficients from LES showed consistent results with the correlation. The unique velocity profile and secondary flow was observed when the heat transfer was deteriorated. This study will contribute to the evaluation of the RCCS performance.
Sensitivity Analysis of RCS Depressurization Strategy Under a Postulated SGTR Accident in OPR1000
Authors: Choi Wonjun , Kim Taeseok , Jeon Joongoo , Kim Nam Kyung , Kim Sung Joong
Published: October 2018

Steam generator tube rupture (SGTR) accident is one of important accident that has high probability of resulting in severe accidents. As a bypass scenario, fission product can be directly released to the environment during the SGTR accident. Thus, the severe accident by SGTR should be carefully managed by severe accident management guidance (SAMG). In Korea, SAMG for optimized power reactor 1000 (OPR1000) has been developed in 1999 and used to mitigate the severe accident of OPR1000 with seven mitigation strategies. Among the mitigation strategies, ‘Depressurization of reactor coolant system (RCS)’ is one of the most powerful strategies to reduce direct release of the fission product. To reduce the RCS pressure, indirect depressurization using steam generator is generally recommended. However, depending on the RCS condition, the indirect depressurization can be ineffective to reduce the RCS pressure. In this case, direct depressurization using pilot operated relief valve (PORV) should be performed as a second plan. From this point of view, sensitivity study of RCS depressurization was performed to investigate priority of depressurization in this study. The severe accident scenario initiated by SGTR accident was selected from probabilistic safety assessment (PSA) level 1 report and simulated using MELCOR 2.1. For the mitigation strategy, various timing of depressurization, the number of opening valves and flow rate of feed water were applied to simulate the possible depressurization strategies during the severe accident. The MELCOR code simulation shows that if depressurization was performed at 30 minutes after SAMG entrance, the direct depressurization was more efficient to reduce the RCS pressure and the fission product release. Therefore, it was recommended to use direct depressurization rather than indirect depressurization in certain time. The sensitivity of flow rate of feed water and different number of opening valves were insignificant for progress of the accident and fission product release. In conclusion, operators should select the way of depressurization to reduce the RCS pressure and the fission product release during the SGTR accident, considering the condition of the plant such as accident progress and availability of safety features. To suggest more proper information for depressurization, more sensitivity analysis and detailed thermal-hydraulic analysis should be performed for the future work.
A Core Design of Innovative Breeder BWR
Authors: Rui Guo , Yamaji Akifumi , Cai Yun , Peng Xingjie
Published: October 2018

High breeding with light water cooling has been studied for decades, though is not easy to be achieved. The main obstacle is the moderating effect of light water, which softens the neutron spectrum. To harden the neutron spectrum and thereby to enhance the fuel utilization or even to achieve breeding with light water cooling, the tight-lattice assembly was proposed and applied to High Conversion LWRs. Nonetheless, none of them achieved high breeding. Until recently, the tightly packed fuel assembly (TPFA) is designed for the purpose of high breeding. The ratio of hydrogen atoms to heavy metal atoms (H/HM) in this assembly is significantly reduced to be less than 0.1. Super Fast Breeding Reactor (Super FBR) adopts TPFA and achieves breeding performance with compound system doubling time (CSDT) of 43 years. In this study, the breeder BWR core also applies TPFA and achieves CSDT of 50 years. BWR is one type of the most extensively built reactors in the world, with abundant operation experience and mature technologies. Breeder BWR is considered to be capable of being incorporated into the current BWR plants with a handful of modifications, thus obtaining optimal economy.
Study on Current Status and Future Developments in Nuclear-Power Industry of the World
Authors: Pioro R. , Zvorykin A. , Machrafi R. , Pioro I.
Published: October 2018

Currently, nuclear power plays a quite visible role in the world electricity generation (∼11%). However, before the Fukushima Nuclear Power Plant (NPP) severe accident in March of 2011, NPPs generated about 14% of the world’s electricity. Accounting that after, mainly, Chernobyl NPP severe accident a number of power reactors built and put into operation in the world decreased from 120 within 1985–1990 to about 22 per 5 years (within 1995–2015), we might face a significant shortage of operating power reactors within 2030–2040.

Therefore, it is important to evaluate current status of nuclear-power industry and to make projections on near (5–10 years) and far away (10–25 years and beyond) future trends in nuclear-power industry. In the current paper statistics on all current nuclear-power reactors were analyzed and based on that future trends were estimated in terms of types of reactors to be left after 10 years, new types of reactors to be put into operation, projections of how many reactors and of which types will be build.

To make any projections an average operating term of power reactors should be estimated. In the current paper a nuclear-power-reactor operating term of 45 years was considered. Also, rates of building and putting into operation power reactors worldwide were estimated, and several scenarios of future developments in nuclear-power industry in the world and in selected countries were considered.
Development of Numerical Simulation Method to Evaluate Molten Material Behaviors in Nuclear Reactors: Estimation of Fuel Debris Distribution in the Pedestal
Authors: Yamashita Susumu , Yoshida Hiroyuki
Published: October 2018

In the accident at the Fukushima Daiichi Nuclear Power Plant, fuel rods and reactor internals were overheated due to the radioactive decay heat and the oxidization of fuel cladding, and relocated to the lower region of the pedestal. To retrieve the fuel debris, location, distribution and characteristics of the fuel debris are quite important. However, this information still has not been clarified. To estimate the distribution of fuel debris, since experiments cannot provide sufficient information, numerical simulation will be a useful tool. There are some numerical codes for calculating conditions inside a reactor at severe accidents, however, those codes include assumptions and/or related to melt relocation, and cannot simulate relocation process directly.

In order to simulate the relocation and corium spreading behavior in PCVs without assumptions and scenario, a numerical simulation code that can phenomenologically evaluate the melting phenomena is required. Then, a phenomenologically-based numerical simulation code for predicting the melting core behavior, including solidification and relocation based on the computational fluid dynamics, JUPITER, has been developed in JAEA.

In this paper, by using JUPITER, we simulated the corium spreading process inside a pedestal. As a result, we confirmed that JUPITER has potential to estimate the debris distribution inside the pedestal region.
Technical Workforce Education and Training Program at Abu Dhabi Polytechnic: Integration of Academia and Industry Requirements
Authors: Hechanova Anthony
Published: October 2018

The United Arab Emirates (UAE) is a developing affluent nation. The leaders of the UAE announced the pursuit of peaceful nuclear power in 2008 and by the end of the following year established its Nuclear Energy Program Implementing Organization (the Emirates Nuclear Energy Corporation (ENEC)), Federal Authority for Nuclear Regulation (FANR), and ordered four APR-1400 pressurized water reactors from the Korean Electric Power Company (KEPCO). Nuclear Engineering programs were initiated soon afterwards at Khalifa University for graduate students and the University of Sharjah for undergraduate students. The technical workforce including nuclear power plant local operators and chemistry and radiation protection personnel was established by ENEC and the Institute of Applied Technology as an inaugural program of Abu Dhabi Polytechnic (AD Poly) in 2011. This paper describes the development of the dual education and training program at AD Poly, the experience of the initial cohorts who conducted their training at the APR-1400 units at the Shin Kori Nuclear Power Plant in Korea, and the current program between the AD Poly Abu Dhabi campus and the new Barakah Nuclear Power Plant based on lessons learned from the earlier years.
Study on Variation of HPGE Detector Dead Layer Thickness and its Effect on the Measurements of the Detector Response and Samples Characterization Using Monte Carlo Simulation
Authors: Abd El Gawad K. , Song Yushou
Published: October 2018

The main objective of this work is to produce an optimal modeling for our aged Planar-HPGe detector using Monte Carlo method (MC). That optimization included the analysis of the germanium dead (inactive) layer thickness for our old detection system (planar-HPGe detector). DL is one of the important parameters needed in order to obtain the smallest discrepancy between simulated and experimental measurements of detector efficiency. Also, precise determination of 235U enrichment for UO2 samples which is necessary for purposes of nuclear materials verification in the field of nuclear safeguards.

The thickness of Germanium dead layer (DL) can be vary by time as it is not well known due to the existence of a transition zone where photons are strongly attenuated and absorbed, that cannot contribute to the total photon energy absorption which causes a significant decrease in efficiency. Therefore, using data provided by manufacturers since long years (manufacture date) in the detector simulation model is not convenient. As a result, some strong discrepancies appear between measured and simulated efficiency, in addition to that non-accurate results for 235U enrichment determination. The Monte Carlo method applied to overcome this difficulty was to vary the thickness of dead layer step by step in simulation, a good agreement (minimum deviation) between estimated and experimental efficiency was reached when a suitable germanium dead layer thickness was chosen. Calculations and measurements were performed for radioactive nuclear material samples in the form of UO2 powder with different sizes and enrichments at different locations, under different gamma-lines emitted after a-decay of the 235U nuclei. Results indicated that a good agreement between simulated and measured efficiencies is obtained using a value for the germanium dead layer thickness approximately (2.45 mm) six in comparison with (0.389 mm) provided by the detector manufacturer.
Uranium Enrichment Measurement Using Enrichment-Meter Approach
Authors: Abd El Gawad K. , Song Yushou
Published: October 2018

The 235U enrichment is one of the most important characteristics of nuclear materials for nuclear safeguards purposes. The multi-group γ-ray analysis method for uranium (MGAU) is an important non-destructive gamma spectroscopy method for 235U enrichment determination. Using that method, the typical measurement bias is below 3% for uranium material with abundance from 0.3 to 93 %. However, it is not applicable for the samples with thick container or without isotopic decay equilibrium. In this work, the enrichment meter method was studied with two uranium dioxide samples (235U abundance 0.71 % and 3.167 %). The nuclear materials spectra were measured using a planar high-purity germanium detector. Based on the specific gamma peak (185.71 keV) of relative high intensity, this traditional enrichment meter approach gives measurement bias more than 10 %. Thus, this work represents two objects: (1) an alternative approach which was investigated, where the calibration is performed through Monte Carlo simulation (MCNP5) instead of experiment in advance, as the measurement bias was reduced to be around 5 %. Thus, to use this approach, one should have the sample details, such as dimensions, chemical composition and container. (2) The influence of the container wall thickness on the measurement accuracy by Monte Carlo simulation. So, if the container wall thickness is not modeled correctly the measurement accuracy is influenced, which is investigated by simulation.
Severe Accident Analysis With Spatial Discretized Model by MAAP: Part 1 — Parametric Study on Fukushima-Daiichi Unit-2
Authors: Kanda Kenichi , Nishi Yoshihisa , Abe Kazuma , Nishimura Satoshi , Nakamura Koichi , Furuya Masahiro , Ui Atsushi
Published: October 2018

Accident analyses of the Fukushima-Daiichi unit-2 nuclear power plant were performed with MAAP (Modular Accident Analysis Program) version 5.03. We assumed RCIC, SRV operation and alternative water injection in order to reproduce the measured pressure and temperature values in RPV and PCV. From parametric studies, it was found that the analysis results were in good agreement with the measured data. In this paper, the results of the parametric studies are reported. Furthermore, spatial discretization of compartments (such as rooms in the reactor building, etc.) into small parts successfully demonstrated the transient distribution and deposition of fission products (FPs) across the rooms. Such special discretization is particularly important for the forensic investigation of severe accidents and the deposited amount in the R/B might be estimated by using this detailed model.
Analyzing the Impact of Solutes on PKA Spectrum for Simulation of Neutron Induced-Radiation Damage in Zr-Based Metals
Authors: Cai Guang-bo , Zhang Ya-dong , Han Yu-xiang , Yu Jia-pei
Published: October 2018

Radiation-induced damage and degradation to shielding and structure materials in nuclear reactors is one of the key limiting factors that affect the safety considerations and the lifetime. Neutron radiation damages materials mainly by exciting a number of Primary Knock-on Atom (PKAs). PKAs induce displacement cascades, causing microstructure changes and mechanical degradations in materials. Computer simulations are used to model this complex process. Knowing the PKA spectrum accurately is important because PKA spectrum and the input of computer simulation are coupled. In this work, we aim to obtain the PKA spectrum in Zr-based alloys with relatively low Zr concentration by using the Geant4 software. Some new functions were added by reprogramming in Geant4. We found that the energy spectra of PKA in Zr2Cu and Zr2Ni are mainly caused by Zr atom, and the shape of the average energy spectra are similar with pure Zr. The number of PKA distributions and the energy deposition in these two Zr-based metals are similar with pure Zr but different than those in pure Ni. These finding indicate that the metallic elements of Cu and Ni have small impact on PKA spectrum in Zr-based alloys, which illustrate that simplified simulation models are feasible when using computer for simulating. Moreover, it has great significant for the calculation of irradiation damage and the computer simulation for the process of collision cascade after one PKA is formed.