bluuebunny/crossref_metadata_2025_split huggingface dataset:
Neutronic Methodological Benchmarks With Simplified Geometries for the Gas Cooled Reactor Group Constant Generating Tools
Authors: Temesvari E. , Batki B. , Gren M. 
Published: October 2018

In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark.
The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.
Experimental Study on the Safety Injection Pump (SIP) Failure Accompanied by the Steam Generator Tube Rupture (SGTR)
Authors: Park Yusun , Bae Byoung Uhn , Kim Jongrok , Lee Jae Bong , Park Hae Min , Choi Nam Hyun , Kang Kyoung Ho 
Published: October 2018

The integral effect test to simulate the safety injection pump (SIP) failure accompanied by a steam generator tube rupture (SGTR), named a SGTR-SIP-01 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident.
In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomenon which can be occurred in the nuclear power plant, as realistically as possible.
In this SGTR-SIP-01 test, a rupture of five steam generator u-tubes on the steam generator hot-side was simulated. Due to the initiation of SGTR, a reactor was tripped by high steam generator level (HSGL) signal.
During the transient simulation for SGTR-SIP-01 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops, and the fluid temperatures, were measured and analyzed. Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as a SGTR accident with a total failure of SIPs. In addition to that, for improvement of the system code which are now on developing such SPACE or MARS-KS code, this test data can be utilized for validation and verification work.
Strength Limit of Thimble Tube With Material Plasticity Under Bending Moment and Axial Compression Force
Authors: Koike Hisashi , Mori Masaji , Fujiwara Daisuke , Shimomura Takashi 
Published: October 2018

The thimble tube, which is made of Zircaly-4, is one of the main components of a PWR fuel assembly. The thimble tube has an important role as a structural member of the skeleton. Another role of the thimble tube is to guide a rod cluster control assembly (RCCA) for insertion during the reactor operation, and the function has to be assured not only in normal operation but in a seismic event.
In a horizontal seismic event, the fuel assembly vibrates laterally, which gives bending moment to the thimble tube. In addition, axial compressive force acts on the thimble tube in a vertical seismic event.
The integrity of the thimble tube has to be maintained while this force and moment act. Mitsubishi has confirmed by the elastic stress analysis that the stress of the thimble tube is lower than the limit value requested for the seismic event. The stress evaluation method is based on the ASME code.
The ASME code also describes the limit analysis which is available when the predicted stress is beyond elastic region of the material. In the analysis, the material is assumed to be elastic-perfectly plastic, and the maximum load that the structure can carry is calculated. For the reason mentioned above, the allowable limit of the thimble tube should be determined as a function between the force and the moment. We are planning to examine the allowable limit experimentally.
As a step before testing, an analytical approach for the limit is discussed in this paper. Firstly, the allowable limit is calculated by a beam model assuming elastic-perfectly plastic material, based on the ASME code.
Secondly, a 3D model analysis with elastic-plastic material is performed to predict the practical strength. Based on the comparison with the analysis using elastic-perfectly plastic material, ASME based limit is considerably conservative compared with the one with the actual stress-strain curve. Conversely, this means there is enough room to rationalize the allowable limit.
As the future work, the experiment will be conducted to obtain the practical limit of the thimble tube and to verify the analysis results.
The EUR Assessment Process and Highlights of the Compliance Analysis for the EU-APR Standard Design
Authors: Declercq Cedric , Ballard Andrew , Ferraro Giovanni , Touré Anicet 
Published: October 2018

The EU-APR standard design is a two-loop pressurised water reactor of the range of 1400MWe, developed by Korean Hydro & Nuclear Power (KHNP) from the Korean APR1400 reference plant. It is the objective of the EU-APR to comply with the current main European regulatory and utility requirements for safe and competitive LWR NPPs. A full assessment of the EU-APR against the European Utility Requirements (EUR) can therefore be regarded as an important step in its introduction to the European market.
KHNP’s request to assess its EU-APR standard design was granted by the EUR Organisation and in autumn of 2015, a two-year assessment process started, during which the EU-APR was assessed against the twenty chapters of the EUR Document Volume 2, encompassing over four thousand individual requirements having in view principally NPP safety and performance.
The EU-APR standard design was assessed against each single requirement by experts of the appropriate discipline from within the EUR member companies. The resultant assessment reports were reviewed by the Coordination Group (working level team) and then validated by the EUR Administration Group and EUR Steering Committee. The syntheses of the requirement-by-requirement assessments per chapter were collected in the principal project output: the EU-APR dedicated subset of EUR Document Volume 3. This Volume 3 subset (Subset I) also contains a technical description of the plant, issued by KHNP and reviewed by the EUR organisation. Furthermore, it explains the assessment process and highlights the main assessment results. Volume 3 Subset I was approved by the EUR Steering Committee in September 2017.
The EU-APR standard design was found to be fully compliant for a majority (almost 90%) of all requirements applicable in this assessment. Furthermore, the EU-APR was found to be compliant with the objectives of an additional 6% of the investigated requirements. The EU-APR standard design can be considered to be in the late basic design phase. Nonetheless, the portion of requirements which were not assessable is very low (below 3%), explained by the abundant availability of reference plant (detailed design) documentation, which often helped demonstrate by means of extrapolatable examples KHNP’s adherence to design principles established for the EU-APR. In total, less than 1% of assessed requirements were found to be non-compliant. The main non-compliances relate to choices of design principles and analysis methods that are explicitly different from those stipulated by EUR, and to the relatively low level of development of reactor core and fuel studies to demonstrate operational nuclear plant manoeuvrability and fuel cycle optimisation in view of safety margins and economics.
Preliminary Research on the Oxidation Effect of the Carbon Steel Plate of Downward Facing Pool Boiling by Two-Dimensional Image
Authors: Wang Kai , Erkan Nejdet , Okamoto Koji 
Published: October 2018

When considering the safety of the reactor after the core melted, the external reactor vessel cooling (IVR-ERVC) is regarded as one of the most prominent method and is now widely being studied. And in order to apply this method more efficiently, CHF is the utmost part because it limits the upper threshold of the cooling effect. There has already been a large number of investigation on the CHF both by experiment and simulation. But for the experiment, most of them used the copper for its high thermal conductivity. However, the lower plenum is actually composed of the carbon steel instead of copper since the reactor pressure vessel and some core catchers in nuclear power plants are made of this material.
This CHF experiment here used carbon steel plate on a downward face. The carbon steel plate was attached to the copper base by solder. The results of the carbon steel plate experiment were then analyzed. After polishing the surface by sandpapers, the heat flux is being increased until CHF was reached. We found that the surface is somehow changed during the heating process. This process was repeated several times, and the growing tendency of the CHF was found. Detailed images of the heating surface acquired by high speed camera under different heat fluxes were also obtained and analyzed. It was found that the more the oxidization was, the fewer bubbles were generated and the bigger the CHF was. Finally some theories of the CHF mechanism were also analyzed. The authors hope that study could shed some light on surface effect on causing different CHF.
Evaluation of the Radiation Effects of Residents Living Around the NSRI Under the External Hazards
Authors: Motome Yuiko , Akiyama Yoshiya , Hiroyuki Murao 
Published: October 2018

The NSRR (Nuclear Safety Research Reactor) is a research reactor of TRIGA-ACPR (Annular Core Pulse Reactor) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity initiated accident (RIA) conditions.
Under the new regulation standards after the Fukushima Daiichi accident, the research reactors are being regulated according to the risk of the facility. Graded approach is introduced in the regulation. In order to apply the graded approach, the radiation effects of residents living around the NSRI under the external hazards were evaluated and the level of the risk of the NSRR facility was investigating. This report is summarized for the result of the evaluation in case the safety functions were lost by the tornado, earthquake and following tsunami.
As the result, the risk is confirmed to be low, since the effective dose of the residents has been below 5 mSv per event due to the loss of the safety functions by the tornado, earthquake and following tsunami.
Large Eddy Simulation of 5-Tube Bundle Helical Coil Steam Generator Test Section
Authors: Yildiz Mustafa Alper , Merzari Elia , Hassan Yassin A. 
Published: October 2018

Helical Coil Steam Generator (HCSG) is a specific type of shell and tube heat exchanger known for having a higher heat transfer coefficient than many other designs. For this reason, they are considered in Small Modular Reactor (SMR) and High Temperature Reactor (HTR) designs. Investigation of flow behavior in HTRs is important for better design. To investigate the flow on the shell side, five rod HCSG test section was built in Texas A&M University. Experiments were conducted at Reynolds number of approximately 9000.
The spectral element code NEK5000 was used to do high fidelity large-eddy simulation (LES) of the test section to validate against experimental results. In this paper preliminary simulation results were presented. Pressure drop in preliminary results with seventh and ninth order polynomial expansion was consistent with experimental data.
Modelling of X-Ray Radioscopy for Phase Topology Estimation During Corium Sodium Interaction
Authors: Singh Shifali , Cassiaut-Louis Nathalie , Journeau Christophe , Zabiégo Magali , Estre Nicolas , Tamagno Léonie 
Published: October 2018

In case of a severe accident scenario in a Sodium cooled Fast Reactor (SFR) such as the ASTRID demonstrator, the fuel might melt and interact with the coolant i.e. liquid sodium. This molten Fuel Coolant Interaction (FCI) can generate an energetic vapor explosion that can jeopardize the reactor structures. The yield of the vapor explosion is strongly dependent on the local distribution of the fragmented melt with respect to the local vapor fractions. The medium is composed of three phases, i.e. corium, liquid sodium and vapor sodium. Thus, a study of the three phase distribution within the system is a key to understand the extent of the explosion.
PLINIUS-2, the future large mass experimental platform of CEA Cadarache will be dedicated to conducting experiments to understand the behavior of prototypic corium in case of severe accidents. In order to study these interactions, a high energy X-ray imaging system is being developed. This system consists of a 15 MeV Linear accelerator producing high energy X-rays with significantly high flux, which are attenuated as it passes through the highly dense test section. The transmitted radiation is detected and re-emitted as visible light by the GADOX screen coupled to the CMOS camera. Using this system to study the interaction between corium and sodium is particularly challenging due to the small corium particulates of the size of the order of 1 mm. The qualification of the foreseen radioscopy system on the visualization of such an interaction requires the development of physical phantoms. This paper presents the preliminary simulations of expected images of corium fragments in sodium, vapor bubbles and vapor film around the fragments. The simulations are carried out using a CEA Cadarache in-house tool MODHERATO, which produces radiographic images in satisfactory agreement with the real time imaging. The simulation of particles is based on the knowledge of interaction phenomenology gained from past experiments and on the statistical analysis of the size of corium particles formed. The models which, according to MODHERATO results, qualify to be detected and resolved, help manufacturing physical phantoms to conduct the experiments.
Evaluation of the Mixing Vanes Effect on the Hydrogen Diffusion and Hydride Formation in the Fuel Cladding
Authors: Aly Ahmed , Petrov Victor , Avramova Maria , Manera Annalisa , Ivanov Kostadin 
Published: October 2018

The fuel cladding is an important barrier to the release of fission products to the environment. Its integrity must be conserved during the in-reactor lifetime and during the spent fuel pool and dry cask storage. The corrosive interaction between the cladding and the water coolant in light water reactors leads to the oxidation of the zirconium-based cladding. A fraction of the hydrogen released due to those corrosive interactions or the radiolysis of the water coolant is picked-up by the fuel cladding. It diffuses inside the cladding driven by the concentration and temperature gradients. Eventually, its concentration can increase beyond a certain limit above which hydrogen precipitates as hydrides. The formation of hydrides can embrittle the cladding and leads to micro-cracks that can compromise the cladding integrity.
At the spacer grids locations, the mixing vanes will create swirl flow and mixing of the coolant leading to a high temperature gradient on the fuel rod cladding. This temperature gradient is a strong driving force for hydrogen to diffuse from high to low temperature locations. Therefore, the hydrogen behavior around the spacer grids with mixing vanes is important to model. In this work, the computational fluid dynamics code START-CCM+ is used to model the effect of the mixing vanes on the temperature profile on the cladding outer surface. It ws coupled with the transport code MPACT and the fuel performance code BISON. The computational model consisted of a 5 × 5 fuel rods subassembly with a guide tube in the central location. The obtained cladding temperature profile on a fuel rod of interest was applied as a boundary condition to BISON to model the hydrogen behavior around the spacer grids in a three-dimensional manner. Three spacer grids were modeled at elevations of 217.9 cm, 270.14 cm and 322.35 cm. The hydrogen behavior at each of those locations is evaluated and compared to assess the importance order of those locations.
Measuring Effects of Radiation on Precipitates in Aluminum 7075-T6 Using Differential Scanning Calorimetry
Authors: Connick Rachel C. , Hirst Charles A. , Cao Penghui , So Kangpyo , Kemp R. Scott , Short Michael P. 
Published: October 2018

Radiation damage in structural materials for nuclear applications is not well-understood, especially when linking the atomic scale damage mechanisms to the macroscopic effects. On a microscopic level, particle radiation creates defects that can accumulate in the material. Defects can also interact with existing features in the material. Since both defects and features have different energies associated with them, investigation of the resulting energy spectrum in a macroscopic sample may offer insight into the connection between microscopic damage and macroscopic properties.
In alloys, changes in the size and number of precipitates will be reflected in the amount of energy required to dissolve the precipitates during thermal analysis. This can then be studied using differential scanning calorimetry (DSC). This work explores the sensitivity of the DSC measurement to detect irradiation-induced instability in metastable and secondary phase precipitates in the high-strength aluminum alloy 7075-T6 for extremely low doses of helium-ion and neutron irradiation. The precipitates in aluminum 7075-T6 are expected to grow or shrink, changing the energy spectrum measured by DSC. The magnitude of the change can then be compared to a model of irradiation-induced phase instability. This will demonstrate the ability of this thermal analysis technique to help bridge the gap between microscopic radiation effects and macroscopic properties.
Overview of PSA for the UK ABWR Generic Design Assessment
Authors: Hirokawa Naoki , Ishiwatari Yuki , Taniguchi Daisuke , Hisamochi Kohei 
Published: October 2018

A Probabilistic Safety Assessment (PSA) was provided as an integral part of the safety case for the United Kingdom Advanced Boiling Water Reactor (UK ABWR) Generic Design Assessment (GDA). The main objectives of PSA for GDA is to provide a demonstration of the compliance, for a single unit UK ABWR, with numerical risk targets defined in the UK Safety Assessment Principles (SAP) and to support the “As Low As Reasonably Practicable” (ALARP) assessment. This content of this paper includes:
• An overview of the UK ABWR PSA
• Identification of the PSA scope
• PSA model development to compare with numerical risk targets and inform design and operational features
• Illustration of PSA results, e.g., Core Damage Frequency (CDF), Large Release Frequency (LRF) and Large Early Release Frequency (LERF)
• Peer reviews
• Use of PSA in ALARP assessment
Simulation of JAERI Downcomer Effective Water Head Experiments With WCOBRA/TRAC-TF2
Authors: Kobelak Jeffrey R. , Liao Jun , Ohkawa Katsuhiro 
Published: October 2018

During the reflood phase of a postulated large break loss-of-coolant accident (LBLOCA), the liquid head in the reactor vessel downcomer provides the driving force to reflood the core. Since the reflood rate is a function of the downcomer inventory, the calculation of the downcomer liquid inventory is critical in simulating the reflood phase of a postulated LBLOCA accident in a pressurized water reactor.
Since the reactor coolant system pressure decreases rapidly after the onset of a LBLOCA transient, the walls surrounding the downcomer become superheated for the duration of the transient. The Japan Atomic Energy Research Institute (JAERI) downcomer effective water head test facility was designed to study boiling and steam-water interaction in the reactor vessel downcomer under prototypical reflood conditions. A number of tests were conducted at this facility with varying degrees of wall superheating (among other things) that cover the expected degree of superheating in a pressurized water reactor. The wall superheating achieved at the JAERI facility is greater than that of other large-scale facilities that are typically simulated to validate thermal-hydraulic system codes.
WCOBRA/TRAC-TF2 is the thermal-hydraulic system code utilized in the FULL SPECTRUM™ LOCA (FSLOCA™) evaluation model (EM). The ability of the WCOBRA/TRAC-TF2 code to predict phenomena occurring in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA has been previously validated. However, only limited wall superheating was present in the existing validation basis. As such, two experiments conducted at the JAERI downcomer effective water head test facility are simulated to provide additional information on the capability of WCOBRA/TRAC-TF2 to predict the liquid inventory in the reactor vessel downcomer during the reflood phase of a postulated LBLOCA. The code captured all the trends observed in the experimental data for both Run 115 and Run 121. The various collapsed liquid levels tended to be well-predicted or under-predicted by the code after the initial simulated accumulator injection period.