bluuebunny/crossref_metadata_2025 huggingface dataset:
Fracture Resistance of Cast Austenitic Stainless Steels
Authors: Chen Y. , Chen W-Y. , Rao A. S. , Li Z. , Yang Y. , Alexandreanu B. , Natesan K.
Published: October 2016

Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is also vulnerable to embrittlement when exposed to reactor service temperatures and fast neutron irradiations. In this study, the combined effect of thermal aging and neutron irradiation on the degradation of CASS alloys was investigated. Neutron-irradiated CASS specimens with and without prior thermal aging were tested in simulated light water reactor environments for crack growth rate and fracture toughness. Miniature compact-tension specimens of CF-3 and CF-8 alloys were tested to evaluate the extent of embrittlement resulting from thermal aging and neutron irradiation. The materials used are static casts containing more than 23% delta ferrite. Some specimens were thermally aged at 400 °C for 10,000 hours prior to the neutron irradiation to simulate thermal aging embrittlement. Both the unaged and aged specimens were irradiated at ∼320°C to a low displacement damage dose of 0.08 dpa. Crack growth rate and fracture toughness J-integral resistance curve tests were carried out on the irradiated and unirradiated control samples in simulated light water reactor environments with low corrosion potentials. While no elevated crack propagation rates were detected in the test environments, significant reductions in fracture toughness were observed after either thermal aging or neutron irradiation. The loss of fracture toughness due to neutron irradiation seemed more evident in the samples without prior thermal aging. Transmission electron microscope (TEM) examination was carried out on the thermally aged and neutron irradiated specimens. The result showed that both neutron irradiation and thermal aging can induce significant changes in the delta ferrite. A high density of G-phase precipitates was observed with TEM in the thermally aged specimens, consistent with previous results. Similar precipitate microstructures were also observed in the neutron-irradiated specimens with or without prior thermal aging. A more extensive precipitate microstructure can be seen in the samples subjected to both thermal aging and neutron irradiation. The similar precipitate microstructures resulting from thermal aging and neutron irradiation are consistent with the fracture toughness results, suggesting a common microstructural origin of the observed embrittlement after thermal aging and neutron irradiation.
A Methodology for the Integration of a Mechanistic Source Term Analysis in a Probabilistic Framework for Advanced Reactors
Authors: Grabaskas Dave , Brunett Acacia J. , Bucknor Matthew
Published: October 2016

GE Hitachi Nuclear Energy (GEH) and Argonne National Laboratory are currently engaged in a joint effort to modernize and develop probabilistic risk assessment (PRA) techniques for advanced non-light water reactors. At a high level, the primary outcome of this project will be the development of next-generation PRA methodologies that will enable risk-informed prioritization of safety- and reliability-focused research and development, while also identifying gaps that may be resolved through additional research. A subset of this effort is the development of PRA methodologies to conduct a mechanistic source term (MST) analysis for event sequences that could result in the release of radionuclides. The MST analysis seeks to realistically model and assess the transport, retention, and release of radionuclides from the reactor to the environment. The MST methods developed during this project seek to satisfy the requirements of the Mechanistic Source Term element of the ASME/ANS Non-LWR PRA standard. The MST methodology consists of separate analysis approaches for risk-significant and non-risk significant event sequences that may result in the release of radionuclides from the reactor. For risk-significant event sequences, the methodology focuses on a detailed assessment, using mechanistic models, of radionuclide release from the fuel, transport through and release from the primary system, transport in the containment, and finally release to the environment. The analysis approach for non-risk significant event sequences examines the possibility of large radionuclide releases due to events such as re-criticality or the complete loss of radionuclide barriers. This paper provides details on the MST methodology, including the interface between the MST analysis and other elements of the PRA, and provides a simplified example MST calculation for a sodium fast reactor.
Study of the Tank Set of the PRHR on SG Secondary Side
Authors: Xu Ran
Published: October 2016

Passive systems are widely used in nuclear power plants, to enhance the inherent safety of the plant. Especially in the accident of station blackout, Passive Residual Heat Removal system (PRHR) which is installed on the secondary side of steam generator (SG), highly resolved the problem of the residual heat removal after reactor trip. The plant safety is improved largely by PRHR, this is the trend of the new plant design. A water tank is commonly mounted on the passive residual heat removal system (PRHR) of steam generator secondary side, which heavily impacts the performance and the startup procedure of PRHR. Moreover, there are many different types of tank layout in various plants design. Therefore it is necessary to study the thermal hydraulic feature of the different type the PRHR tank set. In this paper, the station blackout accident is chosen to verify the performance of the tank set. Then the assessment criteria of PRHR is established from a thermal hydraulic perspective. RELAP5/mod3.2 is adopted to simulate the accident. Based on the performance comparison of different tank sets, the best tank layout is selected, which is selected for the engineering practice. At last, the comparison of the predicted results and the test data is done to verify the performance of selected tank set design.
Transient Heat Transfer for Helium Gas Flowing Over a Horizontal Cylinder Under Flow Decay Conditions
Authors: Liu Qiusheng , Wang Li , Shibahara Makoto , Fukuda Katsuya
Published: October 2016

The Very High Temperature Reactor (VHTR) is a new reactor that uses helium gas as primary coolant for conventional graphite matrix, coated fuels. It enables the achievements of high thermal efficiency and can supply heat with a high temperature of about 900–1000 °C. During the loss of coolant process, the fuel hot spot temperature should not get over a criteria value due to the temperature limitation of the fuel assembly. Traditionally, the VHTRs are designed to deal with loss of forced circulation conditions by using a passive mode decay heat removal system for the cavity cooling. However, even passive systems may experience some failure due to multiple undesired conditions even though the possibility is extremely low. Therefore, the VHTRs are now expected to be designed as naturally safe reactors with inherent safety features. Which means the decay heat removal is fully dependent on natural convection and radiation. To accomplish the tough task, a clear understanding of the heat transfer process during flow decay transient condition is quite necessary.

This study was conducted to investigate the transient heat transfer process between the solid surface and coolant (helium gas) in VHTR under flow decay conditions. Forced convection transient heat transfer for a horizontal cylinder under flow decay transient condition was experimentally studied. The experiment was conducted by using the helium gas as coolant. A uniform heat generation rate was added to the heater. With a certain flow rate of the helium gas, the heater temperature was maintained at a designed value. Then, the flow rate of the helium gas starts to decrease according to designed linear functions with different decreasing speed. Platinum cylinder with 1 mm in diameter was used as the test heater. The heat transfer coefficient and surface temperature were measured during the flow decay transient process under wide experimental conditions such as initial flow rate, flow decay time. It was found that the temperature of the test heater increases in curve shape with different gradients during this process, with a shorter flow decay time the increasing rate of heater surface temperature would be higher. The heat transfer coefficient versus time during the flow rate decreasing process was also obtained.
Relationship Between Particles Volume Fraction Distribution and Electrical Properties in Rotational Liquid
Authors: Akimoto Yuya , Nagae Fumiya , Okawa Kazuya , Iso Yoshiyuki , Ichijo Noriaki , Ikeda Ryousuke , Takei Masahiro
Published: October 2016

A wireless electrical resistance detector has been developed to measure the real-time particle volume fraction in a centrifuge for controlling a rotational velocity and particle supply rate. The purpose of this study is development of real-time particle concentration measurement technique in centrifuges because a technique for measuring the particle concentration in the centrifuges has not existed yet. The wireless electrical resistance detector was attached around a centrifuge for real centrifuge measurement experiment. This experimental centrifuge had electrodes which were covered with dielectric poly methyl methacrylate to avoid conductivity around the electrodes. The wireless electrical resistance detector was lightweight because the centrifuge condition is very severe because the rotational velocity was 3790rpm and gravity was 2000G. At first, sodium chloride solution was put in the centrifuge for the reaction stabilization of the wireless electrical resistance detector. Next, slurry of sodium chloride solution and acrylic particles with 50μm was continuously supplied in centrifuge for 10 minutes. As a result, the measured voltage was increased because the slurry was piled up in the vicinity of centrifuge inner wall. The measured resistance was increased with the lapse of time. The measured resistance coincides with an anticipated tendency of resistance. A change of particle volume fractions was measured in the centrifuge by the wireless electrical resistance detector.
Sorption of Radioactive Methyl Iodide by Silver Doped Zeolite for Filtered Venting System
Authors: Ishii Tasuku , Akiyama Daisuke , Kirishima Akira , Sato Nobuaki , Narabayashi Tadashi
Published: October 2016

Filtered containment venting system (FCVS) is an emergency response system for severe accident. FCVS reduces the pressure in reactor vessel and removes radioactive iodine and cesium.

Recently, silver doped zeolite (AgX) has been attracting attentions since it is believed to remove over 99.99% of CH3I in the ventilation gas. However, the reliable data of AgX for CH3I sorption under expected severe accident conditions are still limited. Therefore, the sorption behavior of CH3I on AgX was investigated under several expected conditions.

In this study, the CH3I labeled by 125I tracer was synthesized by isotope exchange reaction, and it was devoted for sorption experiment on AgX. Sorption experiment using stable CH3I was also conducted by the same procedure. From results of the sorption experiment, over 99.9% of 2.28g of CH3I were sorbed by 10g of AgX. After the sorption experiment, AgX samples were analyzed by powder X-ray diffraction (XRD) and SEM–EDX. These results indicated the distribution of Ag agreed with that of I in micro area of AgX. For the evaluation of apparent sorption capacity of AgX, breaking through experiment was carried out. The apparent sorption capacity of AgX for CH3I was determined as 0.25g / g (AgX) at 24°C. The breaking through experiment was also carried out for the evaluation of CH3I flux and humidity effects in the vent gas. The effect of temperature on the sorption capacity was also studied. When the flux of CH3I supply was increased from 1.09 × 10−4 to 2.33 × 10−4g/cm2s, apparent sorption capacity was unchanged (= 0.21g / g (AgX)). Besides, the apparent sorption capacity was unchanged in the humidity range of ∼0 % H2O to 0.01 % H2O, while it decreased to 0.19 g / g (AgX) at 3% H2O. When sorption temperature increased to 170 °C, the apparent sorption capacity also increased to 0.69 g / g (AgX).
Development of “Proactive Safety Review Guideline for Continuous Improvement on Nuclear Power Plants”
Authors: Narumiya Yoshiyuki , Okamoto Koji , Ueno Shingo , Hayashi Kentaro , Kuramoto Takahiro
Published: October 2016

The Atomic Energy Society of Japan (AESJ) has established and issued the “Proactive Safety Review (PSR+) Guideline for Continuous Improvement on Nuclear Power Plants: 2015” 1) through the discussion at the Periodic Safety Review Subcommittee under the System Safety Technical Committee of the AESJ Standards Committee (SC), and then approved by the both committees. This guideline defines the review concept, perspectives, and methodologies, relating to identifying areas of safety improvement when the licensees implement periodic review for continuous improvement of safety of their nuclear power plants, so called “Proactive Safety Review (PSR+).

This paper discusses the details of this newly established standard, including the concept, perspectives, and methodologies, relating to identifying areas of safety improvement. In addition, specific examples of applying this standard and the policy of utilizing this standard in continuously improving the safety of all the nuclear power plants in Japan are also described.
Start-Up Simulation of the Improved TOPAZ-II Space Power System
Authors: Zhang Wenwen , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2016

An system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II space reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop components models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original coolant loop radiator, is also modeled, including the two-dimensional heat pipe analysis model, the fin model and the coolant transport duct model. The start-up procedure of the improved TOPAZ-II system are calculated. The results show that the code can be used to obtained the thermal characteristics of the system start-up process.
Reliability Analysis of the Automatic Control System of Reactor Power in Nuclear Power Plant Based on DFM
Authors: Hao Wang , Cong Tian , Shiliang Zhou , Yu Yuan Liu , Ahmad Shahroze
Published: October 2016

Instrumentation and control (I&C) system is central nervous system of nuclear power plant (NPP), so its reliability is very important for the safety of NPP. Now Digital I&C system (DICS) is widely used in NPP, its reliability should be analyzed more carefully due to the more complex interaction among its components than analog I&C system. The interaction among DICS can be divided into type I and type II. The former is the interaction among sensor failure, controller failure and actuator failure, the latter is the interaction among hardware\software components of DICS. Limited by static Boolean logic, traditional fault tree is hard to model the dynamic interaction among DICS, so dynamic modelling method is required. In this paper, Dynamic flowgraph methodology (DFM) is used for reliability analysis of a subsystem of NPP DICS, namely APC (Automatic Power Control System). The proposed dynamic flowgraph modelled type I and type II interaction among APC and reliability of APC is assessed via inductive and deductive analysis. In the inductive analysis, three basic events (user defined combination of states) are considered, which includes sensor failure, SRB failure and master\checker processor failure, and the sequence of events caused by these basic events are tracked. In the deductive analysis, top event (combinations of possible system parameters and/or component states) is defined as actual control rod position is lower than demanded, and four combinations of basic events are tracked according to occurrence probability of their corresponding event.
Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident
Authors: Zanchetti Alexandre , Hassanaly Mickael , Cordier Hervé , Sanna Antonio , Mechitoua Namane , Mimouni Stéphane
Published: October 2016

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations.

The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.
Corrosion in Steel T91 Caused by Flowing Lead–Bismuth Eutectic at 400°C and 10−7 Mass% Solved Oxygen
Authors: Schroer Carsten , Tsisar Valentyn , Wedemeyer Olaf , Skrypnik Aleksandr , Konys Jürgen
Published: October 2016

Specimens produced from two different heats of ferritic/martensitic steel T91 were exposed to oxygen-containing flowing lead–bismuth eutectic (LBE) at 400 °C, 10−7 mass% solved oxygen and flow velocity of 2 m/s, for exposure times between around 1000 and 13,000 h. The occurring phenomena were analyzed and quantified using metallographic cross sections prepared after exposure. Oxidation causes a material loss of <10 μm after 13,000 h, while corrosion initiated by the solution of the steel elements may generally proceed around 15 to 30 μm deep into the material in the same amount of time. Oxide scales formed on both heats of T91 tend to buckle and detach. In the case of one of the investigated heats, a singular event of exceptionally severe solution-based corrosion was observed, with associated local material loss around 1.2 mm after 13,000 h. The results are compared especially with findings at 450 and 550 °C and otherwise similar conditions as well as austenitic steels tested in the identical experimental run.
Loss of Core Cooling Test Without One Cooling Line in Vessel Cooling System (VCS) of High Temperature Engineering Test Reactor (HTTR)
Authors: Fujiwara Yusuke , Nemoto Takahiro , Tochio Daisuke , Shinohara Masanori , Ono Masato , Hamamoto Shimpei , Iigaki Kazuhiko , Takada Shoji
Published: October 2016

In the High Temperature engineering Test Reactor (HTTR), the Vessel Cooling System (VCS) which is arranged around the reactor pressure vessel, removes residual heat and decay heat from the reactor core passively when the forced core cooling is lost. The test was carried out at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant temperature control system was isolated, and three helium gas circulators in the primary cooling system were stopped to simulate the loss of forced cooling flow in the core. Moreover, one cooling line of the VCS was stopped to simulate the partial loss of cooling function from the surface of the reactor pressure vessel. The test results showed that the reactor power immediately decreased to almost zero and was stable as soon as the helium gas circulators were stopped. The power decrease is caused by negative feedback effect of reactivity. On the other hand, temperature changes of core internal structures, the reactor pressure vessel and the biological shielding concrete were slowly during the test. The measured temperature of the reactor pressure vessel decreased for several degrees during the test. The measured temperature increase of biological shielding made of concrete was small within 1 °C. The numerical analysis showed that the temperature increase of VCS cooling tube was about 15°C which is sufficiently small, which did not significantly affect the temperature of biological shielding. As the results, it was confirmed that the cooling ability of VCS can be kept sufficiently even in case that one of two water cooling lines of VCS is lost.
The Fast Three-Dimensional Space-Time Neutron Kinetic Model for Cartesian Geometry and Cylindrical Geometry
Authors: Li Zhifeng , Wu Hongchun , Wan Chenghui , Hu Tianliang
Published: October 2016

In order to raise computation speed on the premise of enough numerical accuracy, the Predictor-Corrector Improved Quasi-Static (PC-IQS) method and Nodal Green’s Function Method (NGFM) were combined to solve the three-dimensional space-time neutron diffusion kinetics problems for Cartesian geometry. In addition, the improved quasi-static method and the Krylov algorithm were applied to solve the three-dimensional space-time neutron diffusion kinetics problems for cylindrical geometry. Based on the proposed model, the program of three-dimensional neutron space-time kinetic code has been tested by the two-dimensional and three-dimensional transient numerical benchmarks. The numerical results obtained by this work were in good agreement with the reference solutions.
Analysis of Small-Break LOCA at ACME Test Facility Using RELAP5/MOD3
Authors: Caihong Xu , Guobao Shi , Pu Fan
Published: October 2016

The Advanced Core-cooling Mechanism Experiment (ACME) is conducted to investigate the performance of passive core-cooling system (PXS) for the advanced CAP1400 Pressurized Water Reactor (PWR). The small-break LOCA experiments conducted at ACME integrated test facility are simulated with a SNERDI modified version of RE-LAP5/MOD3 code. Several typical SBLOCA test cases are simulated and one case (2 inch cold leg break) is presented in this paper. And the predicted results are compared with the test data to assess the performance of the modified code. The calculated results agree reasonably well with the test data.
Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development
Authors: Yamada Katsumi , Amri Abdallah , Bevington Lyndon , Vincze Pal
Published: October 2016

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues.

After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations.

One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities.

Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM.

The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation.

This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.
Non-Invasive On-Line Monitoring for Nuclear Power Plants Using Guided Waves Propagating in Steel Pipes With Different Types of Structural Complexity
Authors: Bertoncini Francesco , Cappelli Mauro , Cordella Francesco , Raugi Marco
Published: October 2016

On-line monitoring for installed piping in Nuclear Power Plants (NPPs), as well as for Oil & Gas and other kind of plants, is crucial to early detect local ageing effects and locate single defects before they may result in critical failures. All the actions able to prevent failures are of great value especially if non-invasive and allowing an In-Service Inspection (ISI). In particular the Long Term Operation (LTO) and Plant Life Extension (PLEX) may be invalidated from radiation, thermal, mechanical stresses besides their own ageing. Hence on-line monitoring techniques are of much interest especially if they assure the required safety levels and at the same time are simple and cost-effective. Guided Waves (GW) satisfy these requirements since they are structure-borne ultrasonic waves that propagate themselves without interfering along the same pipe structure, which in turns through its geometric boundaries serves as a confining structure for the GW used to test its integrity. The frequencies used for GW testing extend up to 250 kHz, thus allowing a long-range inspection of pipes (tens of meters in favorable circumstances). The experimental conditions (e.g. temperature, complex piping structure, wall thickness, materials) have to be considered since they strongly affect the results but GW generated through magnetostrictive sensors are expected to overcome such issues due to their robustness and positioning ease. In this paper, new experimental tests conducted using the proposed methodology for steel pipes having different types of structural complexity are described.
Radiation-Induced Electrolytic Corrosion of LWRS: (Part 2) — Verification of In- and Out-Core Redox Potential Differences
Authors: Saji Genn
Published: October 2016

The author recently identified that there should exist a “differential radiation cell” mechanism in the reactor water, prompting “radiation-induced electrolytic (RIE)” phenomena. This mechanism was identified while trying to theoretically reconstruct the potential differences observed in two in-pile test loops; NRI-Rez in Czech Republic and INCA Loop in Sweden.

Part 2 of this series focuses on the theoretical reconstruction of the observed potential differences. Assuming a state of equilibrium, the author tried to develop a formalism by extending the Nernst equation to reproduce the observed redox potential differences. The radiological potential shift term is separated from the Nernst equation where the latter deals only with stable molecular and ionic species. The radiological effect is described as a perturbation term to the Nernst equation representing a potential shift due to radiation-chemical reactions which should diminish to zero without radiation.

The theory generally reproduced the experimental results after fitting the theoretical curve at a single point of the potential for both PWR and BWR-NWC water chemistry environments. This discrepancy is likely due to the “conductive-dielectric property” of the reactor water.
Approach for the Evaluation of the Impact of Potential Software Failures in Software-Based Instrumentation and Control (I&C) Equipment in Nuclear Power Plants
Authors: Mbonjo Hervé , Jopen Manuela , Ulrich Birte , Sommer Dagmar
Published: October 2016

In this paper we present an approach for the evaluation and assessment of the impact of software failures in software-based I&C systems of NPPs. The proposed two-step approach includes at the first step the identification of software failure modes on the basis of review of operating experience gained with software-based I&C systems and equipment. All probable software failures in software-based I&C systems should be identified and classified according to e. g. the concerned system, the observed software failure mode and to their actual and potential safety relevance. In a second step an evaluation of the potential impact of identified safety relevant software failure modes in a software-based I&C system shall be performed. The evaluation shall be done by means of a failure mode and effects analysis (FMEA) using a generic model of the software-based I&C system, i.e. software failure modes are postulated in the I&C system and their potential safety-relevant impact is analyzed.
New Models of Droplet Deposition and Entrainment for Prediction of Dryout in Uniform and Non-Uniform Heated Annuli
Authors: Zhang H. B. , Hewitt G. F.
Published: October 2016

In this paper, we present a phenomenological modeling of annular flow for dryout prediction in vertical annuli by adopting a new set of correlations of droplet deposition and entrainment. The performance of the new correlations is tested by using the experimental data of Becker and Letzter [1] measured in a 3000mm annulus under bilateral as well as unilateral heating conditions, and the experimental data of Becker et al. [2] obtained in a 3650mm long annulus in respect of eight different axial heat flux profiles. The applicability of two widely used correlations of droplet deposition and entrainment in annuli which were derived from flows in vertical tubes was also checked. It was shown that large discrepancies were observed for the tube-based correlations, while for our present model the predicted CHFs as well as the positions of dryout occurrence in the case of non-uniformly heated annuli agree well with the experimental data.
Isothermal Rectangular Roughness Elements in a Rectangular Cavity Heated at Bottom
Authors: Yousaf M. , Usman S.
Published: October 2016

The purpose of the present research was to explore the role of rectangular roughness elements during natural convection in a two-dimensional rectangular cavity. The computational algorithm was developed based on the single relaxation time Bhatnagr-Gross and Krook (BGK) model of lattice Boltzmann method (LBM). Rectangular roughness elements were located on the horizontal walls. The computational algorithm was validated against benchmark studies using different numerical techniques, and a good agreement was found to exist. The range of the Rayleigh (Ra) number was explored from 103 to 106 for a Newtonian fluid of Prandtl number equal to 1.0. The dimensionless amplitude (h/H) of roughness elements was fixed to 0.1, while the spacing between these elements was equal to twice their height. The maximum reduction in the average heat transfer was calculated to be 27 percent at Ra number 1×106.
iB1350: Innovative Passive Containment Cooling System for the iB1350
Authors: Sato Takashi , Matsumoto Keiji , Hara Nobuhiro
Published: October 2016

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. The iB1350 uses innovative passive containment cooling system (iPCCS). The iPCCS is a part of the in-containment filtered venting system (IFVS). The vent pipe is submerged in the IFVS tank in the outer well (OW) of the Mark W containment. The conventional PCCS has a suction pipe only from the dry well (DW). On the contrary, the iPCCS has two suction pipes. One is normally opened to the wet well (WW) and another normally closed to the DW. The suction pipe in the conventional design cannot be connected to the WW because the PCCS vent pipe is connected to the WW. A PCCS functions using differential pressure between two nodes to discharge noncondensable gases in a PCCS heat exchanger (Hx). A suction pipe and a vent pipe must be connected to different nodes to use differential pressure. Therefore, the conventional PCCS never can cool the S/P. Although the S/P is the in-containment heat sink, heat up of the S/P is the most unfavorable for the conventional PCCS. In order to use the PCCS the conventional design must discharge steam directly into the DW instead of the S/P. Therefore, the conventional PCCS must open depressurization valves (DPV) at a SBO if the isolation condenser (IC) fails. On the contrary, the iPCCS can cool the S/P directly using the suction pipe connected to the WW and without DPV. Instead of DPV the iB1350 has modulating valves (MV) of which discharge lines are submerged in the S/P. Even if the IC fails during a SBO, the iB1350 can cool the core using the severe accident feedwater system (SAFWS), the SRV or the MV, and the iPCCS. The SAFWS makes up the core. The decay heat is carried by steam to the S/P through the SRV or the MV. The S/P works as in-containment heat sink. Once the S/P starts boiling the iPCCS automatically initiates cooling of the steam from the S/P. In the case of a core melt accident, a certain amount of FP is released into the S/P and heats up the S/P. Once the S/P starts boiling, the noncondensable gases in the WW is purged by the steam into the DW and then into the PCCS Hx. In order to purge the stagnant gases, the conventional PCCS needs an active fan in the long term. On the contrary, the iPCCS can easily purge noncondensable gases in the heat exchanger using differential pressure to the OW and need not any active fan even in the long term.
Methodology for Fragility Evaluation of a Seawall Against Tsunami Effects: Part 1 — Overflow and Physical Damage Associated With Tsunami Wave Pressure
Authors: Kaida Hideki , Miyagawa Yoshinori , Kihara Naoto
Published: October 2016

Successive safety improvements for nuclear power plants (NPPs) have been required by society as well as by regulatory agencies because of the nuclear accident at the Fukushima Daiichi Nuclear Power Plant due to the Great East Japan Earthquake and Tsunami. The establishment of a methodology for the fragility evaluation of seawalls is essential for developing a probabilistic risk assessment (PRA) for tsunamis that is applicable to NPPs where the hazard level of tsunamis is high.

In the present study, fragility evaluation methods of reinforced concrete (RC) seawalls are documented. Two main damage modes of the seawall, namely overflow and physical damage caused by tsunami wave pressure, were the primary focus. Using the documented fragility evaluation methodology, a conceptual study for evaluating the fragility of a RC seawall against overflow and the impact of tsunami wave pressure is performed, and fragility curves are obtained by considering the following uncertainties: evaluation accuracy of the inundation level and tsunami wave pressure, density of the seawater, compressive strength of concrete, yield strength of reinforcement, and evaluation accuracy of the shear capacity.
Development of Rules on Concrete Containment Vessels for NPP
Authors: Nishizawa Tomohiro , Nishikawa Takao , Takiguchi Katsuki , Sugita Hiroyuki , Hijioka Yasuo , Tsuji Hirokazu
Published: October 2016

“Rules on Concrete Containment Vessels for Nuclear Power Plants” was first published in 2003. It was revised in 2011 and the latest Edition was published in 2015. Endorsement of this Code is expected in the very near future.

There are two types of reactor containments: steel containment vessels (hereinafter referred to as “SCV”) and concrete containment vessels (hereinafter referred to as “CCV”). The former is addressed by Rules on Design and Construction for Nuclear Power Plants, and the latter by Rules on Concrete Containment Vessels for Nuclear Power Plants, including Reinforced Concrete Containment Vessel (hereinafter referred to as “RCCV”), Pre-stressed Concrete Containment Vessel (hereinafter referred to as “PCCV”), and hybrid containment vessels which are required further safety and reliability.
Implementation of Reliability Evaluation Into JSME Fast Reactor Codes: 2 — Development of a Code Case for Seismic Buckling Evaluation Based on LRFD
Authors: Takaya Shigeru , Sasaki Naoto , Asayama Tai , Kamishima Yoshio
Published: October 2016

In a previous study, we proposed an evaluation method of seismic buckling probability of a reactor vessel considering seismic hazard and showed seismic load had the most significant impact on buckling probability among the random variables considered in the evaluation. It means that more rational design of vessels can be realized by taking account of seismic load variation. The load and resistant factor design (LRFD) method enables us to determine design factors corresponding to target reliability by considering variations of random variables and it is used widely in civil engineering. Therefore, in this study, we have developed a draft code case of the rule for preventing buckling of vessels in JSME fast reactor codes based on LRFD. The equation in the draft code case is almost the same as the original but every random variables, seismic load and yield stress, have their own design factors. In addition, mean or median values are used for evaluation instead of design values including conservativeness. This is the first attempt to develop a code case based on reliability evaluation and is expected to lead further implementation of reliability evaluation into JSME fast reactor codes.
Design of Heat-Pipe Heat Exchanger (HPHE) for a Coal-to-Liquid Fuel Process Using Nuclear Energy
Authors: Tshamala Mubenga Carl , Dobson Robert T.
Published: October 2016

Coal is burnt in actual Fischer-Tropsch process to provide heat for steam generation required for coal gasification in a coal conversion-to-liquid (CTL) fuel process. In light of the world energy demand growth and consumption rate of the world reserves, the use of alternative energy resource for heat supply to a CTL process seems to be more attractive. It is expected that the use of alternative energy on the current process will not only contribute to the extension of the coal reserves life-time but will also significantly contribute to reduce the carbon foot-print. In this study, a high-temperature gas cooled reactor is considered to supply heat to a 240 MWth CTL process and a HPHE is conceptually designed for physical dimensions and its dynamic behaviour is observed. The process heat superheated steam is generated in a once-through heat pipe heat exchanger steam generator using sodium as working fluid. A heat pipe heat exchanger once through steam generator is modelled and simulated with the aim of predicting the transient and dynamic behaviour of the heat transfer process occurring in the HPHE.
Development and Testing of Fault-Diagnosis Algorithms for Reactor Plant Systems
Authors: Grelle Austin L. , Park Young S. , Vilim Richard B.
Published: October 2016

Argonne National Laboratory is further developing fault diagnosis algorithms for use by the operator of a nuclear plant to aid in improved monitoring of overall plant condition and performance. The objective is better management of plant upsets through more timely, informed decisions on control actions with the ultimate goal of improved plant safety, production, and cost management. Integration of these algorithms with visual aids for operators is taking place through a collaboration under the concept of an operator advisory system. This is a software entity whose purpose is to manage and distill the enormous amount of information an operator must process to understand the plant state, particularly in off-normal situations, and how the state trajectory will unfold in time.

The fault diagnosis algorithms were exhaustively tested using computer simulations of twenty different faults introduced into the chemical and volume control system (CVCS) of a pressurized water reactor (PWR). The algorithms are unique in that each new application to a facility requires providing only the piping and instrumentation diagram (PID) and no other plant-specific information; a subject-matter expert is not needed to install and maintain each instance of an application. The testing approach followed accepted procedures for verifying and validating software. It was shown that the code satisfies its functional requirement which is to accept sensor information, identify process variable trends based on this sensor information, and then to return an accurate diagnosis based on chains of rules related to these trends.

The validation and verification exercise made use of GPASS, a one-dimensional systems code, for simulating CVCS operation. Plant components were failed and the code generated the resulting plant response. Parametric studies with respect to the severity of the fault, the richness of the plant sensor set, and the accuracy of sensors were performed as part of the validation exercise.

The background and overview of the software will be presented to give an overview of the approach. Following, the verification and validation effort using the GPASS code for simulation of plant transients including a sensitivity study on important parameters will be presented. Finally, the planned future path will be highlighted.
Analysis of RPV Downcomer for the SBLOCA in ACME Integral Test Facility
Authors: Zhong Jia , Wang Nan , Li Yuquan , Shi Yang
Published: October 2016

The Advanced Core cooling Mechanism Experiment (ACME) facility is a 1/3 height scale, high pressure, 1/54.3 power scale and 1/94 volume scale simulation of CAP1400 PWR. A series of twenty-one small break loss of coolant accident (SBLOCA) tests were performed at ACME to collect thermal-hydraulic data for computer code assessment and to support the license of CAP1400. The tests included a range of hot-leg or cold-leg break, double ended direct vessel injection (DEDVI) line break, inadvertent ADS actuation and double ended pressure balance line (PBL) break accidents. The purpose of this paper is to describe the test data of the typical SBLOCA, and to analyze the mixing of liquid in Reactor Pressure Vessel (RPV) downcomer. Furthermore, the trend of RPV downcomer temperature and the temperature distribution in RPV downcomer are investigated.
Overview of Security Plan for Offshore Floating Nuclear Plant
Authors: Kindfuller Vincent , Todreas Neil , Buongiorno Jacopo , Golay Michael , Birch Arthur , Isdanavich Thomas , Thomas Ron , Stevens Harvey
Published: October 2016

A new Offshore Floating Nuclear Plant (OFNP) concept with high potential for attractive economics and an unprecedented level of safety is presented, along with an overview of work done in the area of security. The OFNP creatively combines state-of-the-art Light Water Reactors (LWRs) with floating platforms such as those used in offshore oil/gas operations, both of which are well-established technologies which can allow implementation on a time scale consistent with combating climate change in the near future. OFNP is a plant that can be entirely built within a floating platform in a shipyard, transferred to the site. OFNP eliminates earthquakes and tsunamis as accident precursors; its ocean-based passive safety systems eliminate the loss of ultimate heat sink accident by design. The defense of an OFNP poses new security opportunities and challenges compared to land-based plants. Such a plant can be more easily defended by virtue of the clear 360 degree lines of sight and the relative ease of identifying surface threats. Conversely the offshore plant is potentially vulnerable to underwater approaches by mini-submarines and divers. We investigate security considerations of the OFNP applicable to two potential plant options, an OFNP-300 with a 300 MWe reactor, and an OFNP-1100 with an 1100 MWe reactor. Three innovative security system approaches could be combined for the offshore plant. The first is a comprehensive detection system which integrates radar, sonar and unmanned vehicles for a long distance overview of the vicinity of the plant. The second approach is the use of passive physical barriers about 100 meters from the plant, which will force a fast-moving power boat to lose speed or stop at the barrier allowing the plant security force more time to respond. The third approach takes advantage of the offshore plant siting and the monthly or biweekly rotation of crew to reduce the total on-plant and onshore security force by using the off-duty security force on the plant as a reserve force. Through the use of these approaches, the OFNP-300 should be able to achieve a similar security cost (on a per Megawatt basis) as land-based plants of similar or somewhat larger power rating. Due to non-linear scaling of cost, the security cost of the OFNP-1100 has the potential to be reduced significantly compared to its land-based equivalents.