bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Soil Structure Interaction on the Seismic Fragility of a Nuclear Reactor Building
Authors: El-Bahey Samer , Alzeni Yasser
Published: October 2016

In recent years, the nuclear industry and the Nuclear Regulatory Commission (NRC) have made a tremendous effort to assess the safety of nuclear power plants as advances in seismology have led to the perception that the potential earthquake hazard in the United States may be higher than originally assumed. The Seismic Probabilistic Risk Assessment (S-PRA) is a systematic approach used in the nuclear power plants in the U.S. to realistically quantify the seismic risk as by performing an S-PRA, the dominant contributors to seismic risk and core damage can be identified. The assessment of component fragility is a crucial task in the S-PRA and because of the conservatism in the design process imposed by stringent codes and regulations for safety related structures, structures and safety related items are capable of withstanding earthquakes larger than the Safe Shutdown Earthquake (SSE). One major aspect of conservatism in the design is neglecting the effect of Soil-Structure-Interaction (SSI), from which conservative estimates of In-Structure Response Spectra (ISRS) are calculated resulting in conservative seismic demands for plant equipment.

In this paper, a typical Reactor Building is chosen for a case study by discretizing the building into a lumped mass stick model (LMSM) taking into account model eccentricities and concrete cracking for higher demand. The model is first analyzed for a fixed base condition using the free field ground motion imposed at the foundation level from which ISRS are calculated at different elevations. Computations taking into account the SSI effects are then performed using the subtraction method accounting for inertial interactions by using frequency dependent foundation impedance functions depicting the flexibility of the foundation as well as the damping associated with foundation-soil interaction. Kinematic interactions are also taken into account in the SSI analysis by using frequency dependent transfer functions relating the free-field motion to the motion that would occur at the foundation level as the presence of foundation elements in soil causes foundation motions to deviate from free-field motions as a result of ground motion incoherence and foundation embedment.

Comparing the results of the seismic response analyses, the effects of the SSI is quantified on the overall seismic risk and the SSI margin is calculated. A family of realistic seismic fragility curves of the structure are then developed using common industry safety factors (capacity, ductility, response, and strength factors), and also variability estimates for randomness and uncertainty. Realistic fragility estimates for structures directly enhances the component fragilities from which enhanced values of Core Damage Frequency (CDF) and Large Energy Release Frequency (LERF) are quantified as a final S-PRA deliverable.
ESS Target Helium Experiments at Lund University
Authors: Nilsson Per , Schönborn Alessandro , Klingmann Jens
Published: October 2016

In ESS, a 5 MW proton beam will hit a helium cooled tungsten target to generate neutrons by spallation. For the purpose of investigating various aspects of cooling with helium, an experimental system has been constructed at Lund University, Department of Energy Sciences. A helium flow of 3 g/s is circulated in a closed loop with pre-heater, cooler, filters and a test vessel. The vessel has windows for measurements and is designed for 10 bar and above 400 °C. In the test vessel, a tube forms a helium jet of up to 150 m/s onto a coin shaped material sample.

One purpose of the system is to investigate erosion of sample surfaces. Drilling and attaching equipment to the surface had to be avoided, so the setup has been benchmarked and the comparison was used to tune a CFD model. Then the model is used to assess the conditions in the experiments.

The setup, as well as the tests and calculations, are described and it is concluded that the temperature and velocity estimates are sufficient for the experiments to be performed.
Optimization of Laser-Induced Breakdown Spectroscopy Parameters in a Novel Molten Salt Aerosol System
Authors: Williams Ammon , Phongikaroon Supathorn
Published: October 2016

Pyroprocessing of used nuclear fuel has shown potential as an alternative fuel reprocessing technology to the traditional aqueous technology. Having a clear picture of the actinide, fission product, and rare-earth elements within the salt in realtime is important from processes control, efficiency and material safeguards perspectives. However, measuring the molten salt electrolyte composition within the system is challenging due the high temperature and radiation involved. Laser induced breakdown spectroscopy (LIBS) has been proposed to measure the molten salt composition via a static liquid surface or solid surface approach. These approaches can yield compositional results near real-time; however, concerns with sample homogeneity, splashing, and poor repeatability present significant challenges. A novel molten salt aerosol-LIBS system has been developed to mitigate some of the aforementioned challenges. Here, modifications to the system using a 1-jet nebulizer and sampling chamber are being discussed. Preliminary results demonstrate the advantages and success of the modifications. Experiments were conducted to optimize the spectrometer gate delay and results indicate that the optimal gate delay is greater than 9 μs. In addition, the percent relative standard deviations (%RSD) for this system were found to be approximately 7%.
Evaluation of Thermodynamic Properties of Cerium in Liquid Cadmium Cathode Within LiCl-KCl Salt System
Authors: Yoon Dalsung , Phongikaroon Supathorn
Published: October 2016

Thermodynamic properties of Ce-Cd intermetallic compound were investigated in LiCl-KCl-CeCl3-CdCl2 molten salt system at various temperatures. Six Ce-Cd intermetallic compounds, CeCd, CeCd2, CeCd3, Ce13Cd58, CeCd6, and CeCd11 could be observed via cyclic voltammetry (CV) and the relative Gibbs free energies for the intermetallic formations were estimated by the analysis of the anodic peaks in the CV experiments. Furthermore, an open circuit chronopotentiometry which is known as a suitable method for measuring thermodynamic values of intermetallic compounds were performed. The linear trends of the relative Gibbs free energies were found by using potential difference from CV and CP methods. It can be noted that CV method can be easy and fast tool to estimate relative Gibbs energy for intermetallic compounds of Cd-Ms (metals). For the determination of standard Gibbs free energy, enthalpy, and entropy for the intermetallic formation, CP results were used. The linear trend of the Gibbs energy were obtained against temperature changing from 698 K to 823 K. From the linear relationship, the enthalpy and entropy of the formation were calculated.
Seismic Analysis and Design Upgrade of Condensate Storage Tanks
Authors: Moreno Daniel , O’Sullivan John , Tseng Tsiming
Published: October 2016

Following the accident at the Fukushima Daiichi Nuclear Power Plant resulting from the March 11, 2011 Great Tohoku Earthquake and subsequent tsunami, there was a general concern regarding the beyond design basis capability of existing nuclear power plants. The Condensate Storage Tanks (CST) in the nuclear power plant were originally designed to withstand an earthquake with a peak ground acceleration (PGA) of 0.3g.

The government regulatory commission increased the required PGA to 0.4g, therefore, an upgrade of the design basis for the CSTs was required. Due to the vintage of the existing nuclear power plant, the United States Nuclear Regulatory Commission (USNRC) Unresolved Safety Issue (USI) A-46 methodology may be used for seismic upgrade work of the mechanical and electrical equipment. The Condensate Storage Tanks (CST) belong to the mechanical and electric equipment and therefore were required to have seismic upgrade work because of significant deficiencies that were found in the anchorage of the tanks to the concrete foundation.

A seismic analysis and design upgrade of the Condensate Storage Tanks (CST) was performed to resolve exceedances for base shear, overturning moment and sloshing due to recently updated seismic loads. A detailed analysis of the CST showed that the as-built anchor chair detail did not provide sufficient margin for the Beyond Design Basis Earthquake Event (BDBEE). The as-built anchor chair detail did not provide the required strength to transfer the shear and pull-out loads from the tank shell to the concrete foundation, i.e. no clear load path was provided for the updated seismic loads. Therefore, as a result of the inadequate anchorage, the main issues to resolve in the CST were tank sliding, shell buckling and sloshing due to earthquake loading. The principal challenges encountered during the analysis, design and construction stages were (1) not allowing to loosen the double nut configuration attaching the anchor bolt to the existing anchor chair, and therefore allowing to (2) remove or replace only a few components of the as-built anchor chair, and (3) the retrofit design had to be implemented while tanks were operable, i.e. filled with fluid. An additional challenge faced during the design of the new anchor chair components was the limit in anchor chair height imposed by the numerous interferences in the CST such as nozzles, reinforcing plates and existing welds. A mitigation strategy is analyzed, designed and successfully implemented for retrofitting the ninety-six anchor chairs and allowing for full development of the anchor bolt shear and pull-out strength.
Effect of Cooling Rates on Mechanical Behavior of U-10Mo Monolithic Mini-Plate
Authors: Roh Hee Seok , Mohamed Walid
Published: October 2016

To investigate the effect of cooling on the thermo-mechanical behavior of U-10Mo fuel plate during shutdown step, Finite Element (FE) analysis was performed on the plate L1P756 from RERTR-12 experiments [1]. Changes in cooling rates were simulated by varying the coolant velocity on the two sides of the plate. Since coolant velocity was directly related to heat transfer coefficient (hc), different cooling velocities have been implemented by changing heat transfer coefficient corresponding to coolant velocity ranging from 10% to 200% of the baseline coolant velocity. Also, this study investigated the effect of strain rate on residual stresses of the mini-plates, which may be caused by the cooling rate.

From numerical analysis results, it was found that cooling time increases as the coolant velocity decreases. It was observed that the cooling time is seven times longer if the coolant velocity is reduced 90%. A plate with two times faster coolant than the baseline reduced the cooling time by half of the original cooling time. As the cooling proceeded, von Mises stress was being increased in the plate and the highest stress at a certain time during the shutdown period was observed in the plate with the fastest coolant flow. However, no difference in residual stress was found at all different cooling rates at the end of the shutdown step. For strain rate effect analysis, the maximum strain rate was calculated to be 3 s−1 as soon as the cooling was started and the strain rate drastically decreased close to zero. The change of strain rate in time was found the same in all cases with different cooling rates. Therefore, it turned out that the cooling rate did not affect the residual stress of the cladding considered in this study.
Tensioned Metastable Fluid Detectors for High Efficiency Thermal and Fast Neutron Sensitivity
Authors: Sansone Anthony , Webster Jeff A. , Taleyarkhan Rusi P. , Archambault Brian
Published: October 2016

Advancements in tension metastable fluid detector (TMFD) technology have led to an extension in detector sensitivity to now also detect and distinguish thermal energy (∼.02 eV) neutrons in addition to fast neutrons — spanning 109 orders of magnitude. The unique nature of detector operation and inherent detection mechanism in TMFDs offers a distinct advantage over conventional systems. TMFDs now posses the capabilities for simultaneous sensitivity to fast and thermal neutrons with high intrinsic efficiency, ascertaining directional and spectroscopic source information, all while remaining completely blind to background gamma and beta irradiation. The additional of thermal energy sensitivity was enabled via inclusion of boron in the detection fluid mixture; a compound composed of decaflouropentane (DFP), trimethyl borate (TMB) and methanol. Experimental benchmarking studies were conducted using the spontaneous fission based neutron source 252Cf, in conjunction with theoretical assessments using the nuclear particle transport package MCNP. Source neutron thermalization was accomplished through submersion of the source in a block of ice, such that the moderated spectrum contained an approximate 1:1 ratio between the fast and thermal flux magnitudes. Experimental results show that borated detection fluids yielded up to 6x improvements in the detection rate over their non-borated counterparts. Implications of the current results in regards to the applicability of TMFDs in the field of special nuclear material (SNM) interrogation and detection are discussed.
Calculation of Turbulent Kinetic Energy Budgets for Flow Through a Pebble Bed Using DNS
Authors: Fick Lambert H. , Merzari Elia , Hassan Yassin A.
Published: October 2016

Pebble bed high temperature reactors (PBR) are currently being investigated as a potential successor for light water reactor based designs in the future. Past analyses of flows through PBR cores using Reynolds averaged Navier-Stokes (RANS) approaches have had limited success. Due to a lack of available high fidelity experimental or computational data, optimization of RANS closure models for these geometries has not been extensively done. In the present study, direct numerical simulation (DNS) is employed to develop a high fidelity data set which may be used for the optimization of RANS closure models for pebble bed flows. Calculated parameters include turbulence statistics, as well as values for the turbulent kinetic energy (TKE) budget terms. A well documented, single face centered cubic domain with periodic boundaries was used. Flow was simulated at a Reynolds number of 9308. Tests were conducted to ensure sufficient spatial and temporal resolution for conforming to the requirements for DNS. A selection of the generated statistics and TKE budget terms is presented here.
Vortexing and Air Withdrawal Evaluations for Storage Tanks Using Physical Hydraulic Model Studies
Authors: Humphrey Facendola Amie , Haber Ludwig , Johansson Andrew
Published: October 2016

Knowing and minimizing the critical submergence of storage tanks for various operating conditions in power plants is very valuable to engineers. The goal is to maximize the usable emergency water volume and maintain tank operations such that no air (from vortexing or otherwise) is ever withdrawn into tank suction nozzle(s), that then could jeopardize continued operation of system pumps. While empirically derived curves for air withdrawal predictions can provide some general guidance, they cannot predict actual submergence requirements, especially when return flow is present and water level drops continuously. Physical hydraulic models can be used effectively to determine the critical submergence for plant specific geometries at all possible operating conditions. This paper presents a range of results of physical model studies conducted to determine critical submergence at various operating conditions and compares them with empirically derived curves to determine any possible trends or definitive rules. Critical submergence for all points on all studies was below the Hydraulic Institute boundary curve for pump intake design and below the Reddy and Pickford boundary curve for all cases without return flow to the tank and most with return flow to the tank. Both are much more conservative estimates at high Froude numbers as compared with lower Froude numbers. The Harleman curve, which was derived for the selective withdrawal of density stratified fluids, is neither predictive nor conservative.
Experimental Research on Non-Condensable Gases Effects in Passive Decay Heat Removal System
Authors: Liu Yang , Jia Haijun , Weihua Li
Published: October 2016

Passive decay heat removal (PDHR) system is important to the safety of integral pressurized water reactor (IPWR). In small break LOCA sequence, the depressurization of the reactor pressure vessel (RPV) is achieved by the PDHR that remove the decay heat by condensing steam directly through the SGs inside the RPV at high pressure. The non-condensable gases in the RPV significantly weaken the heat transfer capability of PDHR. This paper focus on the non-condensable gas effects in passive decay heat removal system at high pressure. A series of experiments are conducted in the Institute of Nuclear and New Energy Technology test facility with various heating power and non-condensable gas volume ratio. The results are significant to the optimizing design of the PDHR and the safety operation of the IPWR.
Research on Severe Accident Mitigation Strategy of Injecting Into the OTSG of ACP100
Authors: Zhiqiang Zou , Ming Zhang , Chunrui Deng , Jian Deng , Qingan Xiang , Zhiyi Yang
Published: October 2016

The China multipurpose module small reactor named ACP100 is developed by the China National Nuclear Corporation (CNNC). The severe accident management guidelines (SAMGs) of ACP100 reactor is necessary to be developed to mitigate the severe accident induced by in the multi-cause extreme conditions. Inject into the steam generators (SGs) is one of the most significant strategies, this strategy aims at protecting the SGs tubes from creep rupture, removing the heat from the reactor coolant system (RCS) and depressurizing the RCS. The entry condition of the severe accident guideline of injecting into the SGs, as usual, is low SG Water level in the most of the present pressurized water reactors that adopt the inverted U-tubes SGs. The integral reactor of ACP100 adopts the design of annular tube once-through SGs. In the normal conditions, the secondary side of the OTSGs is operating in a given steady pressure, any water level measure tool is impossibly set in the OTSGs. Therefore, on the basis of the design features of ACP100, the strategy of injecting into the OTSG at severe accident is studied, the potential entry condition of the corresponding guideline is discussed. And the validity and feasibility of the strategy is validated by the analysis of the integral systems analysis code.
Preliminary Probabilistic Risk Assessment of Accident-Tolerant Fuels on PWR Performance During Accident Conditions
Authors: Yanfei Qiu
Published: October 2016

The motivation for exploring the potential development of accident tolerant fuels in PWR to replace existing Zr alloy clad monolithic oxide fuel is outlined. The assessment includes a brief review of core degradation processes under transient conditions and introduces the enhancement of the accident tolerance of PWR by the development of fuels/cladding that can tolerate loss of active cooling in the core and loss of coolant water for a considerably longer time period while maintaining or improving the fuel performance during accident conditions. probabilistic risk assessment are presented that illustrate the impact of these new candidate fuel/cladding materials on the reduction of core damage frequence.
Intelligent Plant Configuration Management Using Wireless Sensors
Authors: Agarwal Vivek , Buttles John W. , Beaty Lawrence , Naser Joseph
Published: October 2016

Plant configuration management is an essential element of nuclear power plant (NPP) design, construction, and operation. In the current operating model of NPPs, plant configuration management is highly dependent on large technical staffs. This dependency is because NPPs have a large number of systems and most operations are manually performed. Work processes tend to be fairly complex due to nuclear quality and documentation requirements. NPPs conduct a substantial number of ongoing surveillance activities to verify that plant components are in their required positions (open/close, on/off, etc.) for current and upcoming plant configuration. This puts nuclear energy at somewhat of a long-term economic disadvantage compared to non-nuclear energy generation sources with rising labor costs. Also, it presents human error opportunities, regulatory compliance impacts, and personnel safety hazards. Furthermore, some of these components are located in radiation control zones and result in dose to the surveillance personnel, thereby creating potential nuclear safety hazards. Technology can play a key role in NPP configuration management in offsetting labor costs by automating manually performed plant activities, such as determining the current state of equipment and process parameters. Alternatively, current NPP instrumentation and control systems are approaching their end-of-life and are facing age-related issues, which presents opportunity to upgrade the systems to reduce dependence on manual activities. This paper presents a proof-of-concept prototype intelligent plant configuration management system using available wireless component position sensors to help reduce operating costs for field-based component position verification activities as well as reduce operational challenges due to component position errors. The work focuses on position sensors for selected manual valve types. The wireless network implemented in this work is based on The Internet of Things network since it enables many different devices to communicate between each other across the same network. The proof-of-concept prototype presented in this paper would benefit the nuclear industry in several ways including reduced labor costs, reduced radiation dose, reduced nuclear and personnel safety challenges, and improved plant and regulatory performance.
Thermal Mixing Behavior in the Annulus of Co-Axial Double-Walled Piping in HTGR
Authors: Tochio Daisuke , Fujiwara Yusuke , Ono Masato , Shinohara Masanori , Hamamoto Shimpei , Iigaki Kazuhiko , Takada Shoji
Published: October 2016

The future HTGR has been designed in JAEA. The HTGR uses the helium gas as the coolant in primary cooling system. The reactor has many merging points of the higher-temperature helium gas and the lower-temperature helium gas in the cooling system.

Previously, the reactor inlet coolant temperature was controlled lower than the specific one in the HTTR constructed in JAEA. It was confirmed that this event was caused by the lack of mixing of helium gas thermally at the measurement point of the reactor inlet coolant temperature. From this operational experience, it is needed to clear the thermal mixing characteristics of the helium gas at the annulus of the co-axial double-walled piping in HTGR from the viewpoint of the appropriate temperature control in future HTGR.

In this paper, thermal-hydraulic analysis on the helium gas at the annular flow path of the co-axial double pipe with T-junction was conducted to clarify the thermal mixing behavior of the helium gas.

It is shown that the thermal mixing behavior is not so much affected by the flow rate helium gases. Moreover it is difficult to mix the helium gas with the smaller height of the annular flow path. This is caused by smaller contact area of higher- and lower-temperature helium gas and the lack of helium gas flow in radial direction.

It is confirmed that it is difficult to mix the higher- and the lower-temperature helium gas in the annular flow path of the co-axial double-walled piping by using the hydraulic behavior, and it is necessary to arrange the mixing promotor in the annular flow path in order to mix the higher- and the lower-temperature helium gas.
Numerical Simulation of the Head of the Direct Action Solenoid Valve Under the High Temperature Condition
Authors: Liu Qianfeng , Zhao FuLong , Bo Hanliang
Published: October 2016

This solenoid valve is used in the Hydraulic Control Rod Drive Technology (HCRDT). The Technology is a newly invented patent and the Institute of Nuclear and New Energy Technology Tsinghua University owns HCRDT’s independent intellectual property rights. The Integrated valve which is made up of three direct action solenoid valves is the key part of this technology, so the performance of the solenoid valve directly affects the function of the integrated valve and the HCRDT. On one hand, based on the conditions occurring in the operation of the Control Rod Hydraulic Drive System, the thermal field of the head is analyzed by ANSYS. The result shows that the highest temperature is below the decomposition temperature of the coil. Second, the performance of the head is obtained in the high temperature. Third, a method is presented to monitor the performance of the valve while the reactor is working. On the other hand, the head of the direct action solenoid valve in high temperature is studied by the experiment. In a word, the solenoid valve has a good performance under high temperature condition.
The Application of an Integrated Head Assembly for Advanced Power Reactor 1400
Authors: Kang Tae Kyo , Jo Won Ho , Cho Yeon Ho , Chang Sang Gyoon , Lee Dae Hee
Published: October 2016

The reactor vessel head region consists of a number of components and systems including reactor vessel head, CEDMs with their cables, cooling air system with ducts and fans, missile shield, seismic supports, head lift rig and cable supports. Prior to refueling operation, those components must be dismantled separately, and moved to the designated storage area. It was a very complicated and time consuming process. As a result, the integrated head assembly (IHA) was introduced to simplify those disassembling procedures, reduce refueling outage period, and improve safety in the containment building as those components are combined into a single system.

To reduce refueling outage duration and radiation exposures to the workers by integrating the complicated reactor head region structures, KEPCO E&C has developed the IHA concept in the Korean Next Generation Reactor (KNGR) project [1]. The first application was implemented for the Optimized Power Reactor 1000 (OPR1000) at Shin-Kori units 1&2 and Shin-Wolsong units 1&2. With the past experience, the IHA was upgraded to be applied to the Advanced Power Reactor 1400 (APR1400). The design was patented in Korea [2], China, EU and the USA as modular reactor head area assembly. The IHA was applied for APR1400 nuclear power plants at Shin-Kori and Shin-Hanul, Korea. The design was also supplied to Barakah Nuclear Power Plants in the United Arab Emirates. This paper presents the design features and a variety of analysis which have been used for the APR1400 IHA.
Development and Application of the Neutronics/Thermal-Hydraulics Coupling Code for Safety Analysis of EBR-II Loss of Heat Sink Tests Without Scram
Authors: Lu Daogang , Guo Chao , Sui Danting
Published: October 2016

In the GEN IV technology evaluations, the LMFBR (Liquid Metal Fast Breeder Reactor) system which includes SFR (Sodium-cooled Fast Reactor) and LFR (Lead-cooled Fast Reactor) was top-ranked in sustainability due to its closed fuel cycle and it is top-ranked in proliferation resistance and physical protection because it employs a long-life core. It is necessary to conduct the coupled neutronics and thermal-hydraulics simulation when the feedback effects are significant in the safety analysis of Anticipated Transients Without Scram (ATWS) in LMFBR. Thus, a neutronics-thermalhydraulics coupling code for safety analysis of LMFBR was developed and used to analyze whole-plant transient behavior of the Experimental Breeder Reactor II (EBR-II) under Loss of Heat Sink Without Scram (LOHSWS) tests in this paper. The two mixing zone method for cold pool coupled with SAC-CFR was used and the predicted results agree well with measurements which are taken from EBR-II LOHSWS test data.
Estimation About Amounts of Radioactive Wastes From Fukushima Daiichi Nuclear Power Station
Authors: Sakai Taichi , Suzuki Shunichi , Okamoto Koji
Published: October 2016

In this research, the amounts of the radioactive wastes generated by the secondary wastewater treatment was focused on. In this paper, at first, cooling water circulating in the plant and the facilities which are removing radionuclides were modeled and how the radionuclides migrate with elapsed time was formulated. Then, the analytic value was fitted to the measured value to compute undetermined coefficients of the formula by setting a migration rate of these nuclides leaching from fuel debris as the parameter for this calculation. Secondly, based on the above calculated inventory, the concentration of those nuclides per vessel was estimated and classified, supposing that the current standard of waste classification for disposal in the normal decommissioning can be applied even for this accident. In this case, the stabilization processing of wastes inside the vessels was not taken into consideration, and these were presumed to be restored directly. By this fitting, the results show that the migration rates of Cs-137 and Cs-134 are 24.4 TBq / day and 10.5 TBq / day. And the classification results indicate that the concentrations of almost all vessels are classified as the equivalent of pit disposal or trench disposal group.
New IAEA Coordinated Research Project on Thermal-Hydraulics of Supercritical Water Cooled Reactors
Authors: Yamada Katsumi , Leung Laurence , Ambrosini Walter
Published: October 2016

In view of the high interest among a number of Member States in the Supercritical Water Cooled Reactor (SCWR) concept, the IAEA launched the second Coordinated Research Project (CRP) on thermal-hydraulics of SCWRs, entitled “Understanding and Prediction of Thermal-Hydraulics Phenomena Relevant to SCWRs” in 2014 to foster international collaboration. The key objectives of this new CRP are to (i) improve the understanding and prediction accuracy of thermal-hydraulics phenomena relevant to SCWRs and (ii) benchmark numerical toolsets for their analyses. At present, 12 institutes participate in the CRP from 10 IAEA Member States, and the OECD/NEA is in cooperation, based on a special agreement with the IAEA, to host a database housing experimental and analytical results contributed from the CRP participants. The expected outcomes from this CRP include (i) enhancement of the understanding of thermal-hydraulics phenomena, (ii) sharing of experimental and analytical results, and the prediction methods for key thermal-hydraulics parameters, and (iii) cross-training of personnel between participating institutes through their close interactions and collaboration.

This paper describes the plan of the new CRP: overall and specific research objectives; tasks and sub-tasks; schedule; and expected outcomes and outputs. It also introduces briefly other IAEA activities to facilitate and support R&D for SCWR technology in Member States, which include technical meetings and training courses.
Overview on the Standard Requirements for the Separation Measures and Practical Implementation in the I&C Design: Return of Experience
Authors: Ciriello Antonio , Kümmerling Stefan , Pickelmann Johannes
Published: October 2016

After giving a short overview about the published instrumentation and control standards worldwide with regard to the separation measures, as released by the different safety authorities, this paper provides a first assessment of the different needs and requirements to be considered in the frame of separation measures which are to be implemented in the instrumentation and control design. According to this analysis the corresponding design aspects and features are derived and analyzed. The experience gained both in nuclear power plant modernization projects and new builds show that there are margins both for interpretation possibilities and for possible technical solutions, which can be proposed to the customer in order to fulfill the concerned standards. A conclusion about the design approach and solutions as well as possible future scenarios is given.
Positron Emission Particle Tracking (PEPT) Validation for Jet Flow
Authors: Langford Seth , Wiggins Cody , Patel Nitant , Ruggles Arthur
Published: October 2016

Positron Emission Particle Tracking (PEPT) generates 4D Lagrangian particle trajectories and is increasingly used to evaluate flow in granular media and complex geometries where optical interrogation methods are not possible. A Multi-Particle PEPT (Multi-PEPT) approach was developed capable of finding and tracking many particles simultaneously to extend the utility of the PEPT method. This paper compares 4,014 trajectories generated using the Multi-PEPT method with 3,055 trajectories generated from High Speed Video (HSV) data. All trajectories are taken in an acrylic test section with water flow using resin beads. The flow geometry includes a flow restriction producing a jet of Reynolds number 23,500 ±800, with peak velocity near 2 m/s, and two recirculation zones. Variation between measurement outcomes is generally less than 0.1 m/s, and measured variations fall within measurement uncertainties. Data co-location uncertainty contributes to larger variations between Multi-PEPT and HSV velocities in regions of steep gradients.
Preliminary Analysis of the Post-Disassembly Expansion Phase and Structural Response Under Unprotected Loss of Flow Accident in Prototype Sodium Cooled Fast Reactor
Authors: Onoda Yuichi , Matsuba Ken-ichi , Tobita Yoshiharu , Suzuki Tohru
Published: October 2016

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident (ULOF) were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.
Fuel Assembly Flow Measurement and Calculation in an Oversized Test Flow Housing
Authors: Wood Carrie A. , Wang Jiwei , Lu Roger Y. , Conner Michael E. , Evans Paul M.
Published: October 2016

To determine fuel assembly mechanical damping under PWR conditions during a postulated seismic event or LOCA, fuel assembly flowing water damping tests were needed. Since flowing water damping values are high, pluck tests were used in which the fuel assembly was deflected, pluck mechanism retracted, and fuel assembly vibration decay observed and measured. To allow for the pluck vibration to occur, an oversized flow housing was utilized, which allowed additional flow to bypass the fuel assembly in the gaps between the assembly and the flow housing walls that would otherwise flow through the bundle rodded region. Since a direct measurement of the average bundle flow velocity could not be made, a method was developed for determining the flowrate through the bundle using fuel assembly lift force measurements.

This paper will present the method for determining the actual flowrate that was used in recent testing. The use of custom load cells that were designed for this testing to measure the fuel assembly lift force will be described in this paper. Additionally, the paper will show how the correlation between total test flow and bundle flow was developed from a series of flow calibration tests at different hydraulic conditions. This correlation was then used to prescribe the loop flow setpoints necessary to achieve targeted bundle flow velocities during the pluck test series.
SAFEST Roadmap for Corium Experimental Research in Europe
Authors: Journeau Christophe , Bouyer Viviane , Cassiaut-Louis Nathalie , Fouquart Pascal , Piluso Pascal , Ducros Gérard , Gossé Stéphane , Guéneau Christine , Quaini Andrea , Fluhrer Beatrix , Miassoedov Alexei , Stuckert Juri , Steinbrück Martin , Bechta Sevostian , Kudinov Pavel , Hozer Zoltan , Guba Attila , Manara Dario , Bottomley David , Fischer Manfred , Langrock Gert , Schmidt Holger , Kiselova Monika , Ždarek Jiri
Published: October 2016

SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on corium experimental research has been written to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. It is based on the research priorities determined by SARNET SARP group as well as those from the recently formulated in the NUGENIA Roadmap for severe accidents and the recently published NUGENIA Global Vision report. It also takes into account issues identified in the analysis of the European stress tests and from the interpretation of the Fukushima accident. 19 relevant issues related to corium have been selected during these prioritization efforts. These issues have been compared to a survey of the European corium experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. It shows a few lacks in EU corium infrastructures, especially in the domains of core late reflooding impact on source term, Reactor Pressure Vessel failure and corium release, Spent Fuel Pool accidents, as well as the need for a large mass (100–500 kg) prototypic corium facility.
Rotor Dynamics of Multistage Centrifugal Pump
Authors: He Hua , Zhang Bo , Pan Guo-pei , Zhao Gang
Published: October 2016

In a nuclear reactor, pumps are employed as key components for various applications, such as in the cooling system. The steady running of the pump relies greatly on the stability of the rotor. This paper analyses the non-damping natural vibration, critical speed and unbalance response of rotor in the multistage centrifugal pump. The calculation results indicate: (1) The supporting stiffness of either electromagnetic bearing or water film has a large influence on critical speed of first and second order rigid mode, but shows a minor influence on the critical speed of first and second order bending mode. In the case that the electromagnetic bearing has a supporting stiffness of 0.5 × 106 N/m, and the supporting stiffness of water film bearing takes twice of the electromagnetic bearing stiffness, the simulated frequency of first order stiffness will be larger than 50 Hz, which is the rated frequency of the rotor. The stiffness of electromagnetic bearing in our experiment was in the range of 1.3–1.5 × 107 N/m. Thus, the working speed of this rotor will be lower than the first-order rigid critical speed. (2) The unbalance responses of the rotor under four unbalance conditions show that the displacement amplitude of major rotor components (including electromagnetic bearing, water film bearing, first and final impeller) are all in the allowed range for engineering design.
Development of an MCNP6-ANSYS FLUENT Multiphysics Coupling Capability
Authors: Gurecky William , Schneider Erich
Published: October 2016

This work presents a novel core multiphysics coupling method and its application to geometries and thermal hydraulic operating conditions typical of U.S. PWRs. Monte Carlo based radiation transport from the MCNP v6.1.0 package and finite volume thermal hydraulic (TH) packages provided by ANSYS-FLUENT v14.0 are combined to produce results with intra-pin resolved spatial resolution equivalent to state-of-the-art reactor physics and multi-physics suites. The Virtual Environment for Reactor Applications (VERA) whose development is spearheaded at Oak Ridge National Laboratory is one such example package. Results from the MCNP-FLUENT coupling framework are compared to a deterministic solution provided by the MPACT-COBRA-TF (MPACT-CTF) package available in VERA. Comparisons between the MCNP-FLUENT methodology and the MPACT-CTF solutions are provided for a single pin case. Good power and eigenvalue agreement (+/−4%, 352[pcm] respectively) is achieved at hot full power conditions.
Optimization of the Radiation Shielding Program QAD
Authors: Wang Wenyi , Zhang Liguo , Cao Jianzhu , Xie Feng
Published: October 2016

The QAD program, based on the point kernel integration method, is widely used in the radiation shielding design of nuclear power plants and related fields. However, QAD-CGA, as the latest version of QAD program, still has some problems, which may affect calculation results and limit the application range. In this paper, the features, principles, and algorithms of QAD-CGA program will be described and several optimization will be introduced. The quantity of γ rays considered in each calculation has been expanded, which can supply more accurate results than those from the original program. Furthermore, the number of dose receivers has been increased, which can provide detailed distribution of the dose field. In addition, a method has been put forward to realize the discretization of source intensity automatically which can simplify the input of the program. Meanwhile, the compartmentalization of the discrete source in the program has been improved. If the size of the discrete source can be minimized small enough to be served as an ideological core, the accuracy of calculations of QAD-CGA program would be guaranteed. However, with the increase of the radius of a sphere or cylinder, the volume of the discrete source will be enlarged and the precondition “small enough” will be lost gradually which can result in the increase of the inaccuracy of calculations. A superior algorithm to solve the coordinate distribution of point kernel which is nonuniform has been proposed. It can reduce the inaccuracy from the discretization of the source intensity in spherical and cylindrical geometry effectively. The optimization of QAD-CGA program has been implemented, analyzed and compared to the original edition with a numerical example.
Design of Severe Accident Management Systems for Beyond Design Basis External Hazards at Paks NPP
Authors: Katona Tamás János , Vilimi András
Published: October 2016

Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some mew systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented and design the new post-Fukushima measures that is outlined in the paper. The concept and practice is presented on several examples.
Outline of Specification for Composite SC Walls in Nuclear Facilities
Authors: Bhardwaj Saahastaranshu R. , Varma Amit H. , Al-Shawaf Taha
Published: October 2016

Appendix N9 to AISC N690s1 presents the design provisions for steel-plate composite (SC) walls in safety related nuclear facilities. AISC N690s1 is Supplement No. 1 to AISC N690-12 specification for safety related steel structures in nuclear facilities and was published in October 2015. This paper discusses the outline of Appendix N9 as well as how the appendix can be used for the design of SC wall structures.

Appendix N9 establishes the minimum requirements that SC walls need to meet in order for the specification to be applicable. The requirements include minimum and maximum wall thickness and steel reinforcement ratio. Detailing requirements for SC wall panel sections are also discussed. The faceplate slenderness requirement to prevent the limit state of buckling before yielding is provided. Steel anchor requirements are based on developing adequate composite action, and preventing interfacial shear failure. Requirements for tie bars connecting the steel plates (faceplates) are provided to prevent splitting failure and out-of-plane shear failure. The detailing and design provisions for regions around openings in SC walls are also included. Appendix N9 provides a method of checking the design of SC walls for impactive and impulsive loads.

A discussion of the analysis requirements for SC walls is presented. The provisions include effective stiffnesses, accident thermal loading and model parameters for analysis.

The design strength equations for axial tension, axial compression, out-of-plane shear, out-of-plane flexure, in-plane shear, and for combined in-plane forces and out-of-plane moment demands are parts of the provisions of the appendix. The provisions also include interaction equations for evaluating tie bars resisting demands due to combination of out-of-plane and interfacial shear forces. Performance requirements for the anchorage of SC walls to concrete basemat, SC wall-to-wall connections and SC walls to floor slab connections are given in the appendix.

The provisions also include requirements for fabrication, inspection, and quality control of SC walls constructed for safety-related nuclear facilities.