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ASME Conference Presenter Attendance Policy and Archival Proceedings

2016;():V001T00A001. doi:10.1115/ICONE24-NS1.
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This online compilation of papers from the 2016 24th International Conference on Nuclear Engineering (ICONE24) represents the archival version of the Conference Proceedings. According to ASME’s conference presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.

Commentary by Dr. Valentin Fuster

Operations and Maintenance, Aging Management and Plant Upgrades

2016;():V001T01A001. doi:10.1115/ICONE24-60031.

The presented experimental investigations of static characteristics of check valves are a contribution to work leading to improve check valve models in 1D network programs. Two designs of check valves were tested: a swing disc and a tilting disc check valve. For each type of these valves static characteristics for the two directions of water flow were performed.

Following physical quantities were measured during the experiments: the angular position of the valve closing element (disc), water discharge, pressure losses caused by the check valve, moment of the hydrodynamic forces acting on the valve disc and pressure at selected locations of the flow system of the test rig.

The paper describes the test rig designed for developing the static characteristics of the check valves and presents the results of tests for the characteristics of the swing and tilting disc check valve under steady flow conditions. These characteristics are presented as time-averaged parameters measured during the tests. Following dimensionless quantities are calculated basing on these parameters:

- the flow rate coefficient,

- the coefficient of the moment of hydrodynamic forces acting on the valve disc,

- the local loss coefficient of the valve.

The above-mentioned coefficients were determined for disc angular position of selected values (opening rate) and interpolated within the full range of openings for both directions of flow through the valve.

The static behaviors of the tested valves are discussed and compared with each other.

Topics: Valves , Disks
Commentary by Dr. Valentin Fuster
2016;():V001T01A002. doi:10.1115/ICONE24-60086.

The purpose of this paper is to provide a general overview of the organization and content of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code. This will involve a brief description of the regulatory requirements associated with Inservice Testing (IST) as well as a brief overview of the OM Code scope and requirements. This paper will discuss, in general, the regulations requiring IST as well as a brief discussion on when Preservice Testing (PST) and IST become required. A general organization of the ASME OM Code will be provided as well as general topics associated with how to determine when testing and examination intervals are established; what documentation is required; and general discussion regarding the various subsections of the OM Code and the components associated with the OM Code. Alternatives to the OM Code requirements and how to obtain these alternatives will also be provided as well as how the edition applicability of the ASME OM Code is determined. There is also discussion regarding a few general issues associated with the OM Code regarding existing reactor power plants as well as the “new builds” and advanced reactor plants and designs.

Commentary by Dr. Valentin Fuster
2016;():V001T01A003. doi:10.1115/ICONE24-60102.

The ability to extend calibration intervals for nuclear plant instrumentation has multiple benefits for improving productivity and reducing operating costs at nuclear plants. Benefits include fewer calibrations inside containment during an outage and associated reduced critical path time and ALARA exposure, reduced risk of calibration error or instrument damage during removal and replacement, and reduced operations and maintenance cost for instrument removal, calibration and replacement.

A good instrument calibration program ensures instruments are checked frequently enough to provide a high level of confidence that they are performing within acceptable limits, but no more frequently. Over-testing of plant instruments and equipment should be avoided for two reasons: valuable resources are expended on maintenance that might not measurably improve plant safety, reliability, or efficiency; and the potential exists for adjustment errors or equipment damage each time an instrument is removed from service for testing. Over-testing increases the risk of errors or damage being introduced without a justifiable improvement in reliability.

This paper discusses the regulatory framework for extending calibration intervals of safety related instruments for U.S. based nuclear power plants. Necessary changes to licensing, plant processes and procedures, training, and configuration management are summarized. An example application of pattern recognition modeling is provided to highlight the analytical support for the processes provided by active monitoring to confirm on-going instrument heath. The paper concludes with a listing of recommended steps to implement a practical program for extending calibration intervals of safety related instruments within the U.S. nuclear regulatory environment.

Commentary by Dr. Valentin Fuster
2016;():V001T01A004. doi:10.1115/ICONE24-60214.

This paper focuses on testing results of the impact of water contamination in diesel fuel on the ability of an emergency diesel generator (EDG) to successfully start and operate during an emergency. This testing program resulted from the discovery of degraded vent pipes on diesel fuel feed tanks that could have allowed rain water to enter and collect at the bottom of the diesel fuel system and potentially prevent satisfactory engine start-up and operation. The nuclear regulator notified the nuclear plant of a potential yellow finding.

The initial analysis effort focused on the use of diesel engine combustion software (Ricardo’s WAVE© 1D engine performance simulation software). Two medium-speed diesel engine models were analyzed with added water content ranging from 10% to 40% water in the diesel fuel. The analyses demonstrated that the engines could start and operate with those percentages of water in the fuel, but that the engine output would experience a power loss or derate proportional to the water content. The regulator was not convinced that the analysis was sufficient.

The validity of the analytical findings above was demonstrated by full-scale tests conducted by MPR Associates on a large diesel engine. To accomplish this, Entergy, the nuclear power plant owner constructed a simulation of the diesel fuel supply system at a facility having the same make and model EDG. Water was introduced into the diesel fuel day tank by two different approaches; slow trickle flow and large slugs of water. These conditions simulated either a steady rainfall while the EDG was operating or a large volume of water collected in the system while the EDG was in standby.

Under both water contamination scenarios, which consisted of more than 50 hours of testing, the diesel engine and generator set responded with negligible loss of frequency or voltage. Further, the engine was confirmed to have undergone no increased wear or degradation as a result of the high levels of water in the combustion chambers. Following a detailed review of the test program and the successful results, the regulator concluded that neither a non-cited violation nor a penalty was warranted.

Commentary by Dr. Valentin Fuster
2016;():V001T01A005. doi:10.1115/ICONE24-60256.

For safety operation of nuclear power plants, soundness assurance of structures has been strongly required. In order to evaluate properties of inner defects at plant structures quantitatively, non-destructive inspection using ultrasonic testing (UT) has performed an important role for plant maintenances. At nuclear power plants, there are many structures made of cast austenitic stainless steel (e.g. casings, valve gages, pipes and so on). However, UT has not achieved enough accuracy measurement at cast stainless steels due to the noise from large grains. In order to overcome the problem, we have developed comprehensively analyzable phased array ultrasonic testing (PAUT) system.

We have been noticing that dependency of echo intensity from defect is different from grain noises when PAUT conditions (for example, ultrasonic incident angles and focal depths) were continuously changed. Analyzing the tendency of echoes from comprehensive PAUT conditions, defect echoes could be distinguished from the noises. Meanwhile, in order to minimize the inspection time on-site, we have developed the algorithms and the full matrix capture (FMC) data acquisition system. In this paper, the authors confirmed the detectability of the PAUT system applying cast austenitic stainless steel (316 stainless steel) specimens which have sand-blasted surface and 3 slits which made by electric discharge machining (EDM).

Commentary by Dr. Valentin Fuster
2016;():V001T01A006. doi:10.1115/ICONE24-60293.

Based on the information provided by the operators, IRSN experts select and analyze different deviations presenting a possible generic nature which could affect the safety of power plants. Some of these deviations result in non-compliance (NC) with the safety requirements.

To maintain an acceptable level of safety, an operator has to implement corrective measures for any situation of non-compliance with safety requirements. IRSN, the Technical Support Organization of the French Nuclear Authority, analyzes the different deviations to assess the impacts on the concerned NPPs safety. Based on the impact on safety, measures should be applied immediately or during the next outages, on a reactor or on several of them. The permanent corrective measures schedule is defined taking into account the “NC” safety impact. However, for some of the “NCs”, it can be difficult to define and implement swift permanent corrective measures, especially when the lack of compliance affects several similar units and requires a design change.

This paper explains the French approach of deviations treatment and specifically the relationship between the Nuclear Safety Authority, the Technical Support Organization, IRSN and the Licensee, EDF during an outage.

Commentary by Dr. Valentin Fuster
2016;():V001T01A007. doi:10.1115/ICONE24-60311.

As part of the Light Water Reactor and Sustainability (LWRS) program in the U.S. Department of Energy (DOE) Office of Nuclear Energy, material aging and degradation research is currently geared to support the long-term operation of existing nuclear power plants (NPPs) as they move beyond their initial 40 year licenses. The goal of this research is to provide information so that NPPs can develop aging management programs (AMPs) to address replacement and monitoring needs as they look to operate for 20 years, and in some cases 40 years, beyond their initial, licensed operating lifetimes. For cable insulation and jacket materials that support instrument, control, and safety systems, accelerated aging data are needed to determine priorities in cable aging management programs.

Before accelerated thermal and radiation aging of harvested, representative cable insulation and jacket materials, the benchmark performance of a new test capability at Oak Ridge National Laboratory (ORNL) was evaluated for temperatures between 70 and 135°C, dose rates between 100 and 500 Gy/h, and accumulated doses up to 200 kGy. Samples that were characterized and are representative of current materials in use were harvested from the Callaway NPP near Fulton, Missouri, and the San Onofre NPP north of San Diego, California. From the Callaway NPP, a multiconductor control rod cable manufactured by Boston Insulated Wire (BIW), with a Hypalon/ chlorosulfonated polyethylene (CSPE) jacket and ethylene-propylene rubber (EPR) insulation, was harvested from the auxiliary space during a planned outage in 2013. This cable was placed into service when the plant was started in 1984. From the San Onofre NPP, a Rockbestos Firewall III (FRIII) cable with a Hypalon/ CSPE jacket with cross-linked polyethylene (XLPE) insulation was harvested from an on-site, climate-controlled storage area. This conductor, which was never placed into service, was procured around 2007 in anticipation of future operation that did not occur.

Benchmark aging for both jacket and insulation material was carried out in air at a temperature of 125°C or in a uniform 140 Gy/h gamma field over a period of 60 days. Their mechanical properties over the course of their exposures were compared with reference data from comparable cable jacket/insulation compositions and aging conditions. For both accelerated thermal and radiation aging, it was observed that the mechanical properties for the Callaway BIW control rod cable were consistent with those previously measured. However, for the San Onofre Rockbestos FRIII, there was an observable functional difference for accelerated thermal aging at 125°C. Details on possible sources for this difference and plans for resolving each source are given in this paper.

Commentary by Dr. Valentin Fuster
2016;():V001T01A008. doi:10.1115/ICONE24-60551.

Toshiba Corporation, a member of International Research Institute of Nuclear Decommissioning (IRID), has contributed to decontamination works throughout Fukushima Daiichi Nuclear Power Plant from outside ground to inside of the buildings. Speedy decontamination works allow workers to access and stay inside of the contaminated buildings for many hours, and as a result, all decommission works can be accelerated. Some remote decontamination machines to decontaminate the inside of the reactor buildings from a remote safe building have been developed for workers not to be radiated by high-level radiation. Conventionally, operators have remotely controlled the decontamination machine through multiple views of some remote surveillance cameras mounted on it. Because the position data such as GPS data is not available in the buildings, it was hard for operators to detect its absolute position and orientation in the building, and it took much time to recognize targets to be decontaminated. In order to reduce positioning time and make operation works easier, we constructed 3D positioning system to automatically detect the absolute 3D position of the decontamination machine in the reactor buildings from a remote safe building. Moreover, we can also keep records of decontamination works easily by tracking 3D position of the decontamination machine.

This paper shows the overview of our approach of 3D positioning and a result of examinations in the mock-up facility that simulates a part of the inside of the reactor building at Fukushima Daiichi NPP.

Commentary by Dr. Valentin Fuster
2016;():V001T01A009. doi:10.1115/ICONE24-60860.

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.

Topics: Manufacturing
Commentary by Dr. Valentin Fuster
2016;():V001T01A010. doi:10.1115/ICONE24-60885.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T01A011. doi:10.1115/ICONE24-60890.

Reactor recirculation motor generator lube oil twin screw pumps are commonly found in nuclear power plants and throughout industry. In a vertical mounting configuration in which the electric motor is bolted atop the twin screw pump in an unsupported manner the natural frequency of the pump/motor structure can be quite low, resulting in damaging vibration. When a structure’s natural frequency coincides at or near the operating speed, or multiple thereof, a phenomena known as resonance can occur. Resonance can occur when a driving force, in this case minor imbalances in either the motor or pump, begins to vibrate and excite the structure resulting in greatly amplified levels of vibration.

In this paper, finite element analysis software is utilized to first calculate the natural frequency of the pump/motor structure, and then potential modifications are modeled to determine their impact on eliminating harmful resonance.

Commentary by Dr. Valentin Fuster
2016;():V001T01A012. doi:10.1115/ICONE24-61114.

Degradation modeling and condition assessment of critical components are important issues in the maintenance of nuclear power plant, but modeling uncertainties must be taken into account seriously by considering the stochastic nature of degradation and observation process. Based on support vector regression algorithm, this article proposes a wall thinning model for carbon steel pipes in a nuclear power plant using in-service inspections data and further performs the uncertainty quantitive assessment for the proposed model. In the beginning, Latin hypercube sampling method is used to create new sample sets from the original observation with certain distribution of the mean values which are assumed from the observed data. Furthermore, part of the reconstructed sample sets are chosen as training sets to develop a wall thinning model and the remaining samples are used as test sets to verify the model. By comparing model predicted wall thickness values of the test sets and the observed values, a quantitative assessment of the degradation model uncertainty is obtained. The obtained results demonstrate that the deviations between observed thickness values and average model predicted values fluctuate around 1%, while model predicting variances are much smaller than the observed variances. This report concludes that the proposed support vector regression model for component degradation can provide accurate condition assessments with rather small variance.

Commentary by Dr. Valentin Fuster

Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory

2016;():V001T02A001. doi:10.1115/ICONE24-60011.

It would take a long time to solve the point kinetics equations by using full implicit Runge-Kutta (FIRK) for the strong stiffness. Diagonally implicit Runge-Kutta (DIRK) is a useful tool like FIRK to solve the stiff differential equations, while it could greatly reduce the computation compared to FIRK. By embedded low-order Runge-Kutta, DIRK is implemented with the time step adaptation technique, which improves the computation efficiency of DIRK. Through four typical cases with step, ramp sinusoidal and zig-zag reactivity insertions, it shows that the results obtained by DIRK are in perfect agreement with other available results and DIRK with adaptive time step technique has more efficiency than DIRK with the fixed time step.

Commentary by Dr. Valentin Fuster
2016;():V001T02A002. doi:10.1115/ICONE24-60041.

A procedure for Electron Beam Welding (EBW) was developed for the manufacturing of a follower fuel assembly made of an AA 6061-T6 aluminum straps for a U-Mo plate-type fuel proposed to be used in the future in Korea’s Kijang Research Reactor (KJRR) project. The initial welding trials of the weld samples were carried out with a high vacuum chamber using the EBW process. After investigating the welds, EB welding parameters for the full-sized samples were optimized for the required depth of penetration and weld quality. Subsequently, the weld samples made by the filler specimens showed higher shearing strengths than those of the non-filler specimens. This procedure made by EBW process was also confirmed based on the results of the shearing strength test, an examination of the macro-cross sections, and the fracture surfaces of the welded specimens.

Commentary by Dr. Valentin Fuster
2016;():V001T02A003. doi:10.1115/ICONE24-60046.

Tristructural-isotropic coated fuel particle is an important fuel design for high-temperature gas-cooled reactor. Irradiation-induced pyrocarbon (PyC) shrinkage and creep behavior will affect greatly the stresses of a TRISO-coated particle. In this study, 5 cases under different conditions by analytical solution were studied to calculate the particle stresses with different fuel behavior. These cases varied in particle geometries, the mount of gas pressure or fuel behavior. A comparison between the results and other benchmarking studies among different codes was made. The results indicated that the calculated results in this study were in good agreement with other codes.

Commentary by Dr. Valentin Fuster
2016;():V001T02A004. doi:10.1115/ICONE24-60052.

Power Ramp test (PRT) of a fuel element is generally conducted with a PRT irradiation rig within a research reactor, in order to study the fuel’s behavior and validate its safety under power transient. Neutronics characteristics of a new PRT irradiation rig within a typical HFETR (High Flux Engineering Test Reactor) core and its components’ heat generation rates are calculated with MCNP code in this paper. The range of the test fuel rod power is obtained with a coupled Neutronic-Thermal-Hydraulic calculation method which combines MCNP and CFX code. The results show that changing the density of 3He gas can vary the test fuel rod power effectively, and the 3He gas layer influences the neutron field intensely by reducing the thermal neutron current into the layer and decreasing the neutron flux in and near the irradiation rig. The test fuel rod power varies from 5.80kW to 15.3kW while decreasing the 3He gas pressure from 4.5MPa to 0.13MPa, along with 0.231$ reactivity addition. Power of the fuel pellet in the test rod increases monotonically along with the 3He gas pressure reducing, and its calculation results have good agreement with the curve fitting by a natural logarithm function.

Commentary by Dr. Valentin Fuster
2016;():V001T02A005. doi:10.1115/ICONE24-60058.

The recently developed lattice Boltzmann equation (LBE) framework [1] for radiation transport is extended to solve time-dependent nonequilibrium neutron transport problems. Dynamics of radiation and material energy exchange is modeled by coupling the radiation transport equation with the material energy equation in a one-dimensional isotropically scattering homogenous medium. The LBE equations are obtained for corresponding radiative or neutron transport in constant source and reactor criticality search problems. Furthermore, a two-dimensional D2Q8 & D2Q16 LBEs are proposed for solving the time-dependent neutron transport equation in a heterogenous media (e.g., a checkerboard lattice with pure scattering and absorbing cells). The results obtained with LBE are in good agreement with the existing discrete ordinate method results for the benchmark problem.

Commentary by Dr. Valentin Fuster
2016;():V001T02A006. doi:10.1115/ICONE24-60061.

The fundamental investigation reported here was motivated by the controversial and incomplete data for zirconium δ-hydride orientation relationships. For the studies, electron back scattered diffraction analysis (EBSD) was used. Formation of zirconium δ-hydrides and their bulk distribution in zirconium and Zircaloy-4 were investigated. On the basis of the grain boundary spectra and the analysis of microtexture, the orientation relationships between δ-hydrides and the α-Zr matrix were refined. It was found that almost all authors, who criticised each other for years, have established the correct relationships. Seven variants of δ-hydride accommodation in the zirconium matrix were summarised in the table of possible orientation relationships.

Topics: Zirconium
Commentary by Dr. Valentin Fuster
2016;():V001T02A007. doi:10.1115/ICONE24-60089.

Recently, the ratio of the UO2 powder made from dry process has increased obviously comparing those produced by wet process since the wet process has been gradually eliminated. Furthermore, the yield of the pellet is less than 85% during the UO2 pellet production process. This causes the increasing tendency of scrap stocked and production cost. It is because the wet process method cannot adapt to the dry process powder. In this article, improvement on the manufacture of UO2 pellet is researched. Firstly, To achieve the minimum defection of the pellet and residual volume of the grinding scrap, several variable processes such as the improving density of the green pellet, the redesigning cavity block size and the modified pressure maintain method were investigated. The optimum process is shown to be a much superior approach to improving the yield of the pellet, decreasing the amount of the scrap, the cost of production and contributing to further developing of pellet manufacturing technology.

Commentary by Dr. Valentin Fuster
2016;():V001T02A008. doi:10.1115/ICONE24-60116.

Nuclear fuel rods is mainly composed of uranium dioxide pellets and zirconium alloy cladding, there is a gap between pellets and cladding, which is filled with helium. Under the reactor operation conditions, pellets produce a lot of heat by nuclear fission reactions and at the same time also produce lots of radioactive fission products. Cladding serve as the first barrier to accommodate radioactive fission product, needs to maintain its structural integrity under the reactor operation conditions. Cladding stresses can be effectively limited by controlling power increase rates. However, pellet manufacturing defects such as missing pellet surface (MPS), may lead to cladding local stress significantly high to cause cladding failure. Simulating the impact of these defects correctly can help prevent these types of failure. MPS defects are 3D phenomenon, needs 3D modeling method to study the influence of these defects on the cladding .In this paper, stress update algorithm is derived, with the help of ABAQUS (a commercial finite element software), simulated the thermal-mechanical behaviors of the MPS defects fuel rod with a 3D FEM and completed the sensitivity analysis of MPS defects size for the fuel performance. The models included in this simulation, including pellet irradiation swelling (fission gas products induced swelling and fission solid products induced swelling), pellet densification, pellet relocation, pellet thermal expansion, pellet irradiation creep, pellet irradiation hardening, cladding irradiation growth, cladding thermal expansion, cladding thermal creep, cladding irradiation creep, cladding irradiation hardening and gap heat transfer (gas heat conduction, radiation heat transfer and contact heat conduction) etc. Furthermore, considering the effects of irradiation and temperature on the material parameters such as thermal conductivity, specific heat and young’s modulus etc. According to the simulation result, showing that MPS defects have a large impact on the performance of fuel rods, this impact will be more obvious with the size of MPS defects increase. The MPS defects cause larger gap distance between pellet and cladding, higher gap distance causes smaller gap conductance, and then causes elevated temperature at the center of the pellet and in the region of the pellet adjacent to the defect. The cladding temperature is reduced in the area immediately across from the defect, and is elevated in neighboring areas. Meanwhile, MPS defects clearly have a significant effect on stress distribution and maximum stress of the cladding, cause high tensile stresses in the inner surface of the cladding and high compressive stresses on the outer surface of the cladding at the center of the defect. Around the boundaries of the defect, the stresses are reversed, with high compressive stresses on the cladding interior and high tensile stresses on the cladding exterior.

Commentary by Dr. Valentin Fuster
2016;():V001T02A009. doi:10.1115/ICONE24-60120.

Monolithic, plate-type fuels are the proposed fuel form for the conversion of the research and test reactors to achieve higher uranium densities within the reactor core. This fuel type is comprised of a low enrichment, a high density U-10Mo alloy fuel-foil, which is sandwiched between diffusion barriers and encapsulated in a cladding material. To understand the irradiation performance, fuel-plates are being benchmarked for large number of parameters. In this work, effects of the cladding material were studied. In particular, a monolithic fuel-plate with U7Mo foil and Zry-4 cladding was simulated to explore feasibility of using Zircaloy as a surrogate cladding material. For this, a selected mini-plate from RERTR-7 tests was simulated first with as-run irradiation history. By using same irradiation parameters, a second case, a plate with U10Mo fuel and Al6061 cladding was simulated to make a comparative assessment. The results indicated that the plate with Zircaloy cladding would operate roughly 50 °C hotter compared with the plate with Aluminum cladding. Larger displacement profiles along the thickness for the plate with Zircaloy cladding were observed. Higher plastic strains occur for the plate with Aluminum cladding. The results have revealed that any pre-irradiation stresses would be relieved relatively fast in reactor and the fuel-foil would be essentially stress-free during irradiation. The fuel stresses however, develop at reactor shutdown. The plate with Zircaloy cladding would have higher residual stresses due to higher pre-shutdown temperatures. Similarly, the stresses magnitudes are higher in the foil core for the plates with Zircaloy cladding. Finally, pressure on the fuel is significantly higher for the plates with Zircaloy cladding. Overall, employing a Zircaloy as surrogate cladding material did not provide a better thermo-mechanical performance compared with the Aluminum cladding.

Commentary by Dr. Valentin Fuster
2016;():V001T02A010. doi:10.1115/ICONE24-60126.

A typical sandstone uranium deposit, located in the Tuhar basin, was selected to compare the effect of oxygen as the oxidizer with that of hydrogen peroxide. Based on the feasibility study of oxygenation of ferrous and uranium minerals, batch leaching, pressure column leaching and field testing were carried through. The results of feasibility study and laboratory leaching indicate that ferrous ion is inaccessible to being oxidized by pressure oxygen in acidic solutions with pH 2–2.5, and oxygen can oxidize the uranium minerals. Recovery of uranium is proportional to the oxygen pressure. Additionally, the low concentrations of aluminium and ferric ion alleviate the potential precipitation of aluminum and iron significantly. The further field test confirmed the feasibility of oxygen in acid leach. Oxygen has some extent effects of increasing uranium level and considerable effects of anti-precipitation and clogging. In general, oxygen has better applicability in this deposit.

Topics: Oxygen , Uranium
Commentary by Dr. Valentin Fuster
2016;():V001T02A011. doi:10.1115/ICONE24-60150.

Metal matrix microencapsulated (M3) fuel is one of the research directions on Accident Tolerant Fuel (ATF). In this article, it provides one of ATF design which consists of BISO (Bistructural ISOtropic) particles embedded in a zirconium alloy matrix, and the cladding coating with silicon carbon (SiC). The temperature distribution of the ATF element has been built, and then the center temperature has also been calculated based on the operation parameters of the large-scale pressurized-water reactor. Simultaneity, the several factors of fuel failure is preliminary analyzed and calculated, especially the pressure shell failure mechanism.

Commentary by Dr. Valentin Fuster
2016;():V001T02A012. doi:10.1115/ICONE24-60175.

Online burnup measurement is a unique feature for pebble bed gas-cooled reactor and the fuel balls undergo a multi-circulation on the basis of the online burnup assay. It is ascertained that the accuracy of the online burnup assay is related with the economy and safety of pebble bed reactor. In the economical perspective, the burnup assay accuracy allow some part of pebbles that are below the burnup limit in the orifice to be discharged out of the core. In the safety view, the burnup assay allow some part of pebbles in the reactor core to exceed the burnup limit. In this paper, a mathematical model is proposed to establish the relationship. The model is implemented based on some reasonable theoretical hypothesis, and the influence of assay accuracy on the reactor safety and fuel cost issues are discussed based on the simulated results given by different assay accuracy. It is ascertained that improvements on burnup assay accuracy could save the fuel cost and improve the PBR economical efficiency as well as reduce the probability of radioactive release due to over-irradiation and enhance the safe reliability of PBR. Further research on the burnup distribution of pebbles in and out of the core and the burnup assay model are expected to provide some implications on proposing reasonable requirements for accuracy of online burnup assay.

Topics: Safety , Economics
Commentary by Dr. Valentin Fuster
2016;():V001T02A013. doi:10.1115/ICONE24-60313.

Aiming at generating a 361-group library, this paper investigated neutron up-scattering effect in the 361-group Santamarina-Hfaiedh Energy Mesh (SHEM). Firstly, the Doppler Broadening Rejection Correction (DBRC) method is implemented to consider the neutron up-scattering effect in Monte Carlo (MC) method. Then the MC method is employed to prepare resonance integral table and scattering matrix for afterward calculation. Numerical results show that the neutron up-scattering affects kinf by ∼200 pcm at most for UO2 pin cell problems in the 361-group SHEM, while the fuel temperature coefficient (FTC) is also influenced by 12∼13%. It has also been found that both of the above two influences acts through scattering matrix rather than self-shielded absorption cross sections. In addition, the self-shielding effect of cladding is studied and it’s been found that it affects kinf by 30∼70 pcm.

Commentary by Dr. Valentin Fuster
2016;():V001T02A014. doi:10.1115/ICONE24-60330.

Plenum temperature is an important parameter in the calculation of LWR fuel rod internal pressure, since plenum contributes over 40% of free volume in typical LWR fuel design. In many fuel performance codes the plenum temperature are determined in a simple and empirical method. There is a concern about whether such a simple method can properly reflect the temperature variation during fast transients and accident conditions. Therefore, a detailed simulation of heat transfer process in plenum based on COMSOL Multiphysics software was performed to give a fine resolution of temperature and to investigate the heat transfer mechanisms. Since the COMSOL application is a stand-alone model, a new plenum temperature model which can be coupled with fuel performance code was also developed. Finally the validation of the new plenum temperature model is performed by comparing its results with COMSOL model and FUPAC code.

Commentary by Dr. Valentin Fuster
2016;():V001T02A015. doi:10.1115/ICONE24-60331.

A way to generate the few-group cross sections for fast reactor calculation is presented in this paper. It is based on the three steps computational scheme. In the first step, the ultrafine method is used to solve the slowing down equation based on the ultrafine group cross section generated by NJOY. Optional 0D or 1D calculation is used to collapse energy group into broad energy groups. In the second step, the 2D RZ calculation using SN method is performed to obtain the space dependent neutron spectra to collapse broad energy groups into few groups. The anisotropic scattering is well handled by the direct SN calculation. Finally, the full core calculation is performed by using the 3D SN nodal method. The results are compared with continuous energy Monte-Carlo calculation. Both the cross section generated in the first step and the final keff in the last step are compared. The results match well between the three steps calculation and Monte-Carlo calculation.

Commentary by Dr. Valentin Fuster
2016;():V001T02A016. doi:10.1115/ICONE24-60338.

The implementation of the Generation IV nuclear reactors and a spallation target of the accelerated driven system (ADS) concerning to the use of liquid lead-bismuth eutectic (LBE) alloy. The liquid LBE alloy should be fully characterized and especially its physical properties should be completely known to make sure the nuclear safety. Differential scanning calorimetry (DSC) experiments were employed on LBE alloy at a temperature range of room temperature (RT) to 500°C to detect the structure phase transition and obtain the thermal effect of LBE alloy.

The results of DSC curves showed that there existed two remarkable thermal signal events at the melting points zones of Pb and Bi during the melting process tested with a scan rate of 2°C/min. Even though the scan rate was increased to 5°C/min, the DSC signal still exhibited the changes of curve slop at the element melting point zone. Just this interesting DSC thermal signal change phenomenon around 300°C contributed to the relationship with the explanation on the severe embrittlement of T91 steel induced by liquid LBE alloy. The results suggest that there existed effective critical size and form of chemical clusters in liquid LBE alloy worked actively on the embrittlement of T91 Steel.

Commentary by Dr. Valentin Fuster
2016;():V001T02A017. doi:10.1115/ICONE24-60365.

The High Flux Engineering Test Reactor (HFETR) is loaded with complex geometry assemblies, taking on strong heterogeneity. These features call for more advanced core analysis and fuel management methodologies compared with traditional PWR. Based on the two-step method, a fuel management simulation platform HEFT was developed, which is a 3-D core fuel management transport calculation program of HFETR. In HEFT, the lattice code HEFT-lat has the abilitiesto describe complex assembly geometriesof HFETR, whereby mitigate errors brought by geometrical approximations.DNTR, a triangular mesh nodal SN method code, is employed to solve the transport equation. This paper introduces the development backgrounds of HEFT, introduces the principle and function of the three main modules of HEFT, that is HEFT-lat?HEFT-core and HEFT-int, the calculation method of lattice parameter and reactor loop calculation method are described. By follow-up calculation of 81-I to 89-II of HFETR core and comparison of intial critical keff, the control rod position during shutting down and the neutron flux, thus obtains the calculation deviation. The result shows the validity of the HEFT code, the lattice and reactor core calculation model. Thus, HEFT has already successfully applied to the fuel management calculation of HFETR.

Topics: Fuel management
Commentary by Dr. Valentin Fuster
2016;():V001T02A018. doi:10.1115/ICONE24-60372.

Aqueous Homogeneous Reactor demonstrated in this analysis uses Uranyl Nitrate water solution as fuel; for the use of medical isotope producing, AHR has many advantages like small size, low power, passive safety, low output of nuclear waste, etc. Liquid fuel and radiolytic gas bubbles lead to big negative feedback features and fuel fluctuation; these may result in transient excursion, which is extremely important and could affect the operation stability and design of AHRs. For the analysis of transient characteristic of AHR, we developed this transient physics procedure of AHR, it uses improved quasi-static approximation and is based on Multi-group Monte Carlo method, which could consider spatial effects onto transient behavior and has powerful geometry processing ability; both physical model and preliminary validation results will be presented in this article. This procedure applies as an analyzing tool for AHR’s transient behavior; characteristics of AHR in typical transient process are also presented in the article.

Commentary by Dr. Valentin Fuster
2016;():V001T02A019. doi:10.1115/ICONE24-60401.

Accurate nuclear cross-section sensitivity-coefficient evaluation is important for sensitivity and uncertainty analysis, similarity analysis, cross-section adjustment et al. A cross-section perturbation will affect the lattice-physics calculation results through the transport calculation directly and through the resonance calculation indirectly. The indirect effect was found to be important in some cases in the previous studies. To quantify the indirect effect on the lattice-physics calculation results for subgroup resonance calculation method, a sensitivity and uncertainty analysis code COLEUS was developed based the GPT-based method. The eigenvalue sensitivity to non-resonance nuclide cross sections was investigated. Numerical results show that in the traditional LWR, the sensitivity coefficients will be overestimated if implicit sensitivity is neglected. And in the BWR, the implicit sensitivity will become more important along with the temperature rise. But if resonance fission and resonance capture play a coequal role or the background cross section is big, the implicit sensitivity can be small.

Topics: Physics , Eigenvalues
Commentary by Dr. Valentin Fuster
2016;():V001T02A020. doi:10.1115/ICONE24-60443.

This work uses the 2-D C5G7 benchmark to verify the accuracy of the MOCUM code, a parallel neutronics program based on the method of characteristics (MOC) for solving arbitrary core geometry. Compared to the MCNP results, MOCUM k-eff, maximum assembly and pin power percentage errors are 0.02%, −0.06%, and 0.64%, respectively. The results demonstrate the high accuracy of the MOCUM code. The calculation uses a total of 56 threads, and the runtime on dual Intel Xeon E5-2699 v3 CPUs is 26 minutes, with speed up higher than 50 times. The sensitivity study of various MOC parameters using the calculation of the C5G7 benchmark problem is also performed. The study reveals that C5G7 requires the usage of 48 or more azimuthal angles. The strong flux gradient and the heterogeneous effects need fine unstructured meshes to resolve. The simulation uses 258 million zones with an average mesh size of 0.016 cm2. The investigation of the polar angle quadrature indicates that Leonard polar angle performs slightly better than Gauss-Legendre and Tabuchi polar angles and more than three polar angles are not necessary. In addition, parameter sensitivity study shows that coarse parameters are prone to introduce error to the neutron flux but not k-eff.

Commentary by Dr. Valentin Fuster
2016;():V001T02A021. doi:10.1115/ICONE24-60505.

The duplex pellets under a “Low-Interact” (LOWI) nuclear fuel design, which consist of an outer enriched annulus and a depleted or natural core, can provide lower center temperature and reduced probability of pellet-clad mechanical interact (PCMI). Analysis and experiments were done in 1970s to examine the benefits and cost of LOWI design for water-cooled reactors. Results showed that the additional economic cost of this design should not be neglected in spite of the benefits. However, due to the improvement of nuclear fuel fabrication technology in the past 30 years, the benefits of LOWI design become more significant, especially when the potential of other methods to elevate the power density and overcome the constraints on ramp rates in power reactors is running out. In order to evaluate the feasibility of deploying the LOWI fuel in commercial and research reactors, neutronics and thermal calculations are made to figure out the performance of duplex UO2 pellets in particular reactors. It is shown that the center temperature of pellet has been greatly reduced without any change on assembly and core geometry, which means the opportunity of less fission gas production, higher power density and more adequate safety margin. A mechanical analysis of a typical LOWI design is also done. The challenges on duplex pellet manufacture are also discussed. Several fabrication techniques are presented to show the potential of cutting the cost of pellet production.

Topics: Fuels , Water
Commentary by Dr. Valentin Fuster
2016;():V001T02A022. doi:10.1115/ICONE24-60583.

In this paper, the occurrence mechanism of blistering was studied and development processes of blistering were summarized. In addition, a thermal-mechanic-material coupling analysis code, named FROBA-PLATEs (Fuel Rod Behavior Analysis for PLATEs), was developed for plate-type fuel with the consideration of burnup effect. FROBA-PLATEs code was applied to perform the behavior analysis of a dispersion-plate-type fuel. Significant phenomena, including fission gas release and matrix damage, were simulated and key parameters, such as temperature profile, stress and strain profile, were obtained. Most important of all, the starting time of blistering was gained according to the deformation of cladding. The result indicates that: blistering happened at high burnup stage; power density and thickness of cladding are sensitive parameters for blistering. Reducing the power density or enlarge the thickness of cladding can delay or prevent blistering. Furthermore, the influence of blistering on thermal-hydraulic performance was preliminarily investigated by CFD simulation. The simulation result indicates that blistering results in deterioration of heat conduction in the fuel plate.

Commentary by Dr. Valentin Fuster
2016;():V001T02A023. doi:10.1115/ICONE24-60627.

Activation products are the primary radiation source in the maintenance of the reactor pressure vessel (RPV) of High temperature reactor pebble-bed module (HTR-PM). In order to properly plan the maintenance and reduce the related occupational exposure, it is important to correctly evaluate the activation of the impurities in the metal material of HTR-PM’s RPV. In this study, activation of the material of HTR-PM’s RPV is performed using both FLUKA program and experimental formula. Based on the impurity control limit in the technical specification, various impurity nuclides are taken into account. The specific activities of activated radionuclides calculated by the two methods after 40 years irradiation and at the shutdown time are compared. It is demonstrated that FLUKA and experimental formula are in agreement. Also primary contributory activation products at 30 days, 1 year, 5 years and 10 years after the shutdown time are listed.

Commentary by Dr. Valentin Fuster
2016;():V001T02A024. doi:10.1115/ICONE24-60659.

COre and System INtegrated Engine for design and analysis (COSINE) is developed by State Nuclear Power Software Development Center (SNPSDC), which is an integrated nuclear engineering code package. A lattice physics code named COSLATC is an essential part included in COSINE. COSLATC is a multi-group two-dimensional transport code. Pin cells and assemblies of Pressurized Water Reactor (PWR) can be calculated by COSLATC. It is used to calculate few group constants and nuclei densities for core simulator.

In order to make sure the quality of the COSLATC, a strategy of verification & validation (V&V) are discussed and applied to the COSLATC. Firstly a V&V requirement analysis is performed. A test matrix considering the variety of fuel enrichments, materials, geometric and working conditions is generated. Then the corresponding benchmark is collected and classified. Finally, the numerical results of COSLATC and the reference values are compared and analyzed. According to the validation strategy discussed above, the preliminary validation is carried out.

The benchmark provides consistent and comprehensive tests for high burnup (approx. 70GWd/t) fuels of PWR. By comparison with k-infinity and isotopic composition in the UO2 and MOX fuels pin cell problems or assembly problems benchmark, the result shows that the calculated results from COSLATC code agree well with reference results. COSLATC will be an important lattice calculation code in the scientific calculation and engineering application.

Topics: Physics
Commentary by Dr. Valentin Fuster
2016;():V001T02A025. doi:10.1115/ICONE24-60661.

Due to powerful geometry treatment capability, Method Of Characteristics (MOC) becomes the most popular method to solve neutron transport equation. However, boundary conditions always restrict the MOC method’s widely application. Most of the current neutronics lattice codes based on MOC can only be used to solve one or two specific geometrical shapes. In this paper, we developed a powerful MOC module, which can treat different geometrical shapes with two methods. For special geometrical shapes, such as rectangle, 1/8 of square, hexagon, 1/3 of hexagon, 1/6 of hexagon, the MOC module adopts special trajectory layout and angle quadrature set, which can reduce the computation time. For other general geometrical shapes, the MOC module use ray prolongation method, which can treat arbitrary geometry shapes and boundary conditions but need much computation time. This MOC module was incorporated into advanced neutronics lattice code GALAXY, which developed by Nuclear Power Institute of China. The numerical results show that the GALAXY code can be used to calculate 2D neutronics problems with rectangle, hexagon, and other complicated geometry shapes accurately. In future, the GALAXY code will gradually become the main neutroncis lattice code in NPIC.

Commentary by Dr. Valentin Fuster
2016;():V001T02A026. doi:10.1115/ICONE24-60747.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T02A027. doi:10.1115/ICONE24-61031.

The used fuel inventory of the United States commercial nuclear fleet has been accumulating since the inception of nuclear reactors. In order to understand the mass and composition of the used fuel inventory, a nuclear fuel cycle simulation package (Cyclus) is used with a reactor modeling tool (Bright-lite). The parameters for the simulation are obtained as historical operation and burnup data for every reactor in the US fleet, taken from the U.S. Energy Information Administration. The historical burnup data is used to calculate the fuel enrichment of every reactor at every refueling. Discharged uranium inventories calculated by the software are shown to closely match the reference data. The total mass of three major actinide groups are presented as they build up over time. In addition, the evolution of the plutonium composition in discharged fuel is also presented, illustrating Cyclus’ ability to track the composition of material flowing through a large, evolving reactor fleet over decades.

Commentary by Dr. Valentin Fuster
2016;():V001T02A028. doi:10.1115/ICONE24-61093.

The thermal conductivity is one of the most important properties for UO2. The influences of microstructure are especially important for UO2 due to the severe structural changes under irradiation conditions. In this study, we have investigated the thermal conductivity of UO2 with different microstructures using Finite Element Method. The thermal conductivity increases with increasing grain size. The grain size distribution has obvious influence on the thermal conductivity especially when there are pores in the polycrystal. The influences of porosity and pore size are very sensitive to the position of the pores. The results obtained in this study are useful for prediction of property changes of UO2 fuel in pile and important to gain some design guidance to tune the properties through the control of the microstructure.

Commentary by Dr. Valentin Fuster
2016;():V001T02A029. doi:10.1115/ICONE24-61130.

The purpose of this study is to analyze the hydraulic characteristics of adapter plate with angled trailing edge in the bottom nozzle design for the self-reliant nuclear fuel development plan. Hydraulic behavior and pressure drop were calculated by using Computational Fluid Dynamics, which is an ANSYS CFX program. Several models for analysis were selected according to factors influencing the hydraulic. Each analysis model was classified by the depths of angled trailing edge on bottom nozzle flow plate with respect to the reference design. Bottom nozzle flow plate with angled trailing edge showed the lower pressure drop. Different from conventional design, angled trailing edge ensure flow area changing smoothly along the FA longitude, and nozzles with this feature have lower pressure drop. Angled trailing edges also provide a passage of flow exchange, making the flow more evenly distributed. Bottom nozzles with angled trailing edge also show better performance than the former design in hydraulic balance.

Topics: Design , Nozzles
Commentary by Dr. Valentin Fuster
2016;():V001T02A030. doi:10.1115/ICONE24-61135.

In this paper, a new flux expansion nodal method for hexagonal-z geometry is presented to solve multi-group neutron diffusion equations. In each three dimensional node and each group, the intra-nodal flux is approximated by the linear combination of exponential functions and orthogonal polynomials up to the second order. The coefficients are obtained by the weighted residual methods and the coupling conditions of the nodes, which satisfy the continuity of both the zero- and first-order moments of fluxes and currents across the nodal surfaces. A series of benchmark problems including the three dimensional cases are used to test this method. The numerical results verify that it is a rather accurate and efficient for the estimation of the eigenvalue and power distribution.

Topics: Geometry
Commentary by Dr. Valentin Fuster

Plant Systems, Structures, Components and Materials

2016;():V001T03A001. doi:10.1115/ICONE24-60010.

Most of the large components in the thermal, traditional and nuclear power plants such as pressurized vessels and pipes are operating at elevated temperatures. These temperatures and stress are high enough for creep to occur. For variety of reasons many of these power plants are now operating beyond their design life time. It is -known fact that as the high temperature components aged the failure rate normally increases as a result of their time dependent material damage. Further running of these components may become un-safe and dangerous in some cases. Therefore, creep assessment of the high temperature components of these plants is essential for their safe operation. Mainly for economic reasons these components have to be creep assessed as they are in service. However, assessing the creep strength for these high temperature components as they are in service, it can be challenging task, especially when these components are operating under extremely high temperature and/or stress. This paper introduces newly invented, small creep test specimens techniques. These new small types of specimens can be used to assess the remaining life times for the high temperature components, using only small material samples. These small material samples can be removed from the operating components surface, without affecting their safe operation. Two of the high temperature materials are used to validate the new testing techniques.

Commentary by Dr. Valentin Fuster
2016;():V001T03A002. doi:10.1115/ICONE24-60043.

Even though in the past 30 years, Beyond Design-Basis Event (BDBE) Design for nuclear power plant (NPP) has been considering as one of the design commitment for the safety and function goals, often time a compromise inevitably take the place when conduct the detail engineering design. There is several reasons lead to this situation: (1) the lacking of thorough investigation and research on this subject, (2) the need for clarity and recommendations from industrial code and standard practice, (3) the need for clearer and specific regulation and regulatory requirements, (4) the consideration of economy.

By understanding the above situations, this paper is contributed to investigate the determination of Beyond Design Basis Seismic (BDBS) design for new NPPs in current timeframe. Due to the complexity of this new emerging subject, this paper, as the carrier of preliminary (phase 1) results, is attempting to summarize the authors’ latest research progress from technical and design practice point of view.

The preliminary research on BDBE determination is focusing on the investigation of the applicability of PSHA method on this subject. Following aspects are considered in this phase 1 report:

• The definition of BDBE from current nuclear power engineering perspectives.

• The discussion of the applicability of PSHA on BDBE determination.

• Suggested procedures for BDBE determination etc.

• The recommendation of approaches for SSC modeling and analysis under BDBE.

This paper will also cover some topics on the design criteria for beyond design basis earthquake loading conditions. A more detailed discussion on beyond design basis loading’s combinations and design criteria will be investigated in the phase 2 report of this research.

Commentary by Dr. Valentin Fuster
2016;():V001T03A003. doi:10.1115/ICONE24-60051.

There are more than 400 reactors in operation to generate electricity in the world, most of them are pressurized water reactors and boiling water reactors, which generate great amount of spent fuel every year. The residual heat power of the spent fuel just discharged from the reactor core is high, it is required to store the spent fuel in the spent fuel storage pool at the first 5 years after discharged from the reactor, and then the spent fuel could be moved to the interim storage facility for long term storage, or be moved to the factory for final treatment.

In the accident of the Fukushima in 2011, the spent fuel pool ruptured, which led to the loss of coolant accident, it was very danger to the spent fuel assemblies stored in the pool. On the other hand, the spent fuel stored in the dry storage facility was safe in the whole process of earthquake and tsunami, which proved inherent safety of the spent fuel dry storage facility.

In china, the High Temperature gas cooled Reactor (HTR) is developing for a long time in support of the government. At the first stage, HTR-10 with 10MW thermal power was designed and constructed in the Institute of Nuclear Energy Technology (INET) of Tsinghua University, and then the High Temperature Reactor-Pebble bed Modules (HTR-PM) is designed to meet the commercial application, which is in constructing process in Shandong Province. HTR has some features of the generation four nuclear power plant, including inherent safety, avoiding nuclear proliferation, could generate high temperature industrial heat, and so on.

Spherical fuel elements would be used as fuel in HTR-PM, there are many coating fuel particles separated in the fuel element. As the fuel is different for the HTR and the PWR, the fuel element would be discharged into the appropriate spent fuel canister, and the canister would be stored in the appropriate interim storage facility. As the residual power density is very low for the spent fuel of HTR, the spent fuel canister could be cooled with air ventilation without water cooling process. The advantage of air cooling mode is that it is no need to consider the residual heat removal depravation due to loss of coolant accident, so as to increase the inherent safety of the spent fuel storage system.

This paper introduced the design, arrangement and safety characteristics of the spent fuel storage well of HTR-PM. The spent fuel storage wells have enough capacity to hold the total spent fuel canisters for the HTR-PM. The spent fuel storage facility includes several storage wells, cold intake cabin, hot air discharge cabin, heat shield cylinders, well lids and so on. The cold intake cabin links the inlets of all the wells, which would be used to import cold air to every well. The hot air discharge cabin links the outlets of all the wells, which would be used to gather heated air discharged from every well, the heated air would be discharged to the atmosphere through the ventilating pipe at the top of the hot air cabin. The design of the spent fuel storage well and the ventilating pipe could discharge the residual heat of the spent fuel canisters in the storage wells, which could ensure the operating safety of the spent fuel storage system.

Commentary by Dr. Valentin Fuster
2016;():V001T03A004. doi:10.1115/ICONE24-60081.

Nuclear energy is a challenging and ambitious choice for space power system in contrast to solar and chemical fuel. It is able to realize high power and long operating time simultaneously to meet the need of potential applications. Aiming at the thermodynamic performances of the regenerative Brayton cycle with two-stage compression, the paper is objective to get a set of reasonable and competitive operating parameters for the design of the space nuclear power system. Thermodynamic process calculation is applied to analyze the relations of cycle efficiency and influence factors including compression ratio, gas temperature at cold side and hot side, recuperator efficiency, system pressure. The mass estimate model is established to calculate total mass and specific mass of the system with the variation of such design parameters. The calculating results using MATLAB code show that the optimal compression ratio of single compressor varies between 1.2 and 2 along with the other parameters. Either decreasing the cold side temperature or increasing the hot side temperature contributes to enhance the cycle efficiency to about 50%. When the recuperator efficiency changes from 60% to 98%, an ideal heat exchange efficiency, the efficiency corresponding to the optimal compression ratio increase from 35.8% to 52%. But the total mass will also rise from 9.1 tons to 29 tons. It is concluded that the system with cold side and hot side temperature of 450 K and 1300 K, recuperation efficiency of 80% is capable to obtain the maximum cycle efficiency of 36% and the system mass of 10.2 tons. Supposing a space nuclear power system with thermal power of 5 MW, the specific mass is only 5.8 kg/kWe, which indicates obvious technical and economic advantages.

Commentary by Dr. Valentin Fuster
2016;():V001T03A005. doi:10.1115/ICONE24-60092.

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Commentary by Dr. Valentin Fuster
2016;():V001T03A006. doi:10.1115/ICONE24-60111.

Structure design of supercritical light water cooled reactor (SCWR) faced some problem such as materials selection for their special application environment. SCWR vessel internal component materials not only service with irradiation damage, but also various temperatures. The temperature of reactor coolant inlet is 280 °C while outlet is 550 °C, in addition, the thermal expansion of internal components materials and reactor pressure vessel (RPV) components are different. Thus, reducing thermal stress in SCWR internal components and keeping inlet/outlet pipe seal are contradictory. The authors designed a SCWR internal components structure with combined application of 316Ti and N10242. The mechanical properties and irradiation performances of these materials were proved to be available, and finite elements calculation suggests that this new structure can evidently reduce the thermal stress of SCWR internal components.

Commentary by Dr. Valentin Fuster
2016;():V001T03A007. doi:10.1115/ICONE24-60128.

The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear Air-Brayton Combined Cycle systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by Co-Firing when generation is limited. The rated output of a nuclear plant can be augmented by several hundred per cent by Co-Firing. The incremental hydrogen to electricity efficiency can far exceed that of hydrogen in a stand-alone gas turbine.

Topics: Carbon , Power grids
Commentary by Dr. Valentin Fuster
2016;():V001T03A008. doi:10.1115/ICONE24-60141.

Specific heat capacity of irradiated and un-irradiated Zr-4 alloy in the temperature range from 40°C to 500°C was measured by DSC (Differential Scanning Calorimetry) in Hot Cell of NPIC for the first time in China. The irradiated discal specimen was sampled from Zr-4 alloy plate, and the dose was about 18μSv/h. The measurement was proceeded on DSC device (404 F1 type), and the data of specific heat capacity was obtained by relative method made use of Proteus analytical software. Discharge air exhausted to hot cell by vent in DSC. The test result indicated that the test method was confirmed feasible and appropriate, then provided an effective approach for thermal parameter measurement for irradiated specimens.

Commentary by Dr. Valentin Fuster
2016;():V001T03A009. doi:10.1115/ICONE24-60171.

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

Topics: Steel , Missiles
Commentary by Dr. Valentin Fuster
2016;():V001T03A010. doi:10.1115/ICONE24-60174.

GH3128, which is one of the domestic Ni-base super alloy, shows good performances under high temperature. Preliminary research has shown that this alloy is promising as structural material for Intermediate Heat Exchangers (IHX), which are the key components of Very High Temperature Reactor (VHTR). In this paper, the tensile properties of GH3128 are tested by the uniaxial tension experiments, and then the fractographic evaluation is conducted by using the scanning electron microscope (SEM). In the paper, the variation tendency of tensile strengths of GH3128 as well as plasticity is described along with the variation of temperature. The results also show that the tension fracture mode will gradually transform from ductile fracture to cleavage-intergranular compound fracture with the increasing of temperature, and the transition of fracture mechanism corresponds well with the variation of alloy plasticity.

Commentary by Dr. Valentin Fuster
2016;():V001T03A011. doi:10.1115/ICONE24-60216.

Feasibility of Rankine cycle improvement for LWR power plant is studied. In LWR, the main steam is saturated steam, thus steam turbine is mostly operated in wet region. Therefore, the moisture separation plays a key role to achieve better turbine plant efficiency. Typical reheating configuration for present LWR plant is single reheating after HPT by MSR.

On the other hand, it is well known that the increasing of numbers of moisture separation or reheat improves the turbine plant efficiency. Double reheat cycle is common for fossil power plant and also has been proposed for LWR. However, double reheat cycle has not been adopted to actual nuclear power plant yet since the double reheat configuration makes heating and drain system complicated, and larger MSR shell and diameter piping are required due to the increase of steam volume flow.

In this paper, improved Rankine cycle with high pressure moisture separator (HP-MS) and improved low pressure loss MSR (LP-MSR) is studied in order to increase the plant efficiency. Using CFD analysis and scale model test, we confirmed pressure loss and moisture separation characteristic of LP-MSR. The performance of improved Rankine cycle with this new MSR was calculated and the results show that the electrical output of typical LWR Plant may increase 1% or more from single reheat cycle.

Commentary by Dr. Valentin Fuster
2016;():V001T03A012. doi:10.1115/ICONE24-60228.

A portion of the Service Level (SL) I coating systems inside the AP1000®1 containment become submerged during a design basis loss-of-coolant accident (LOCA) as a result of containment flood-up. The Design Basis Accident (DBA) qualification of SL I coatings for the AP1000 did not initially include post-DBA submergence conditions.

Submergence testing was performed using a standard test (Reference 2). Test articles were carbon steel coupons coated with one of two coating SL I coating systems; an untopcoated inorganic zinc system (Carboline Carbozinc® 11 HSN2) and a system consisting of inorganic zinc and an epoxy (Carboline Carboguard® 890N3) topcoat. Half of the inorganic zinc samples tested were irradiated to a total accumulated dose of approximately 1×109 rads prior to submergence testing.

Autoclaves were used to simulate the post-LOCA environment inside of the AP1000 containment. The working fluid was a boric acid solution buffered with trisodium phosphate (TSP) to a pH of approximately 7.8. The pressure and fluid temperature inside of the autoclaves was regulated to begin at ambient conditions followed by a pressurization and heat up following pressure and temperature conditions calculated for a cold leg break LOCA for the AP1000 plant, plus an added 10% margin, and the slowly decrease over the 30 day test period to saturation temperature at about 20 psia, simulating a cool-down of the AP1000 plant. Coolant samples of the autoclave inventory were taken and evaluated for dissolved chemical species at specific time intervals during the test.

Following exposure to the submerged test environment, each coupon was visually inspected immediately following removal from the autoclaves and was inspected again several days after completion of the test cycle for signs of blistering, rusting, lifting, peeling, discoloration, softening, etching, wrinkling, cracking, swelling, dissolving, delamination, and changes in gloss relative to their observed pre-test condition.

Commentary by Dr. Valentin Fuster
2016;():V001T03A013. doi:10.1115/ICONE24-60334.

The condition monitoring of the feedwater pump in secondary circuit is critical to the safe operation of the nuclear power plant. This article presents a fault diagnosis method of feedwater pump by using parameter-optimized support vector machine (SVM). While the fault features of feedwater pump are reflected from the power spectrum of the vibration signals, we trained and diagnosed the fault feature table with support vector machine. The optimal penalty factor C and kernel parameter γ of support vector machine are selected by grid search and k-fold cross validation. Then the faults are diagnosed by the SVM model under the optimal parameters. Diagnostic results show that the parameter-optimized SVM method achieves higher diagnostic accuracy than the PNN method, exhibiting superior performance to effectively diagnose the faults of feedwater pump.

Commentary by Dr. Valentin Fuster
2016;():V001T03A014. doi:10.1115/ICONE24-60409.

In this article, the effect of bolt clamping force and constraint arrangement on structural strength of bolted joint was investigated by finite element method (FEM) prior to hardware tests. This study developed a numerical simulation to predict the deformation behavior and detect potential failure modes. In achieving it, a three dimensional (3D) detailed model of bolted joint was constructed. FE dynamic simulation was used to simulate the structural behavior of the bolted joint by gradually applying tension force on the ends. The numerical simulations were conducted with the torque of 0.3N.m, 3.0N.m, 6.0N.m at different tensile levels in several frictional states between contact plates. In order to determine the critical friction force on the plates, three kinetic frictions between bolt and plate hole were employed in the FE calculation to detect the shear stress on the bolt. Finally, the structural behavior of the bolted joint was analyzed in terms of stress distribution, deformation state by varying clamping force and frictional coefficient.

Commentary by Dr. Valentin Fuster
2016;():V001T03A015. doi:10.1115/ICONE24-60516.

Density functional theory calculations have been used to calculate the ground state structure and oxygen and hydrogen adsorption properties of the pure and doped-iron nanoclusters. Small atomic clusters containing two to six atoms have been considered and a single Fe atom has replaced by a minor element i.e. Zr, Ti, and Sc. Doping of a minor element increases the cluster stability and octahedron Fe5Zr is the most stable structure within this study. Zr- and Sc-doped clusters have the highest oxygen and hydrogen adsorption energy. The electronic structure shows a strong hybridization between the metal 3d and oxygen 2p orbitals with a small contribution from metal 4s and 3p orbitals. Additionally, H s and metal 4s states form a new peak below the Fermi energy and a small modification is observed for 3d orbitals near the Fermi level. A small amount of Zr- and Sc-doping into the Fe-based alloys might improve the oxide film adherence.

Topics: Hydrogen , Iron , Oxygen
Commentary by Dr. Valentin Fuster
2016;():V001T03A016. doi:10.1115/ICONE24-60599.

Among the countermeasures to reduce the impacts of cable fires, installing automatic fire detection and suppression equipment may be applicable for cable trays due to the overcurrent incident considering a single failure of the circuit breaker. To establish a method to construct such systems effectively, we applied two types of automatic fire suppression systems using gaseous and bubble foam agents. For cable tray fire suppression tests, we used the overcurrent test equipment installed in the high power short-circuit testing facility in the CRIEPI (Central Research Institute of Electric Power Industry) in Japan. Several series of tests subjected to 2kA-class overcurrent of six times the allowable current were performed using FR (Flame Retardant)/ Non-FR high-/low-voltage power cables. During these tests, to reproduce the cable fire, one target cable was continuously energized by ring transformers, which confirmed the reliable fire suppression capability of the cooling and suffocation.

Topics: Cables , Construction , Fire
Commentary by Dr. Valentin Fuster
2016;():V001T03A017. doi:10.1115/ICONE24-60630.

A numerical simulation method of fretting wear on heat-transfer tube under random excitation is proposed in this paper. Linear springs are used to simulate the in-plane support of anti-vibration Bars (AVBs) to the U-bend tube. The contact forces between the tube and the AVBs are calculated by using FEM random excitation analysis program. And the in-plane relative displacements subjected to turbulence between the tube and the AVBs at each contact points are calculated. The workrate in Archard equation can be calculated by multiplying the contact force, the relative displacement, coefficient of friction and the modal frequency. By using this method, a representative tube is studied. And the effect that the pre-tightening forces between the tube and AVBs had on the fretting wear and the fluid-elastic stability is discussed.

Commentary by Dr. Valentin Fuster
2016;():V001T03A018. doi:10.1115/ICONE24-60673.

The AP1000® Containment Vessel (CV) is a freestanding steel containment designed to protect the public from radiation release. The CV consists of 2 ellipsoidal heads connected by a cylindrical shell and is constructed of carbon steel. The AP1000 plant design has four large penetrations (two airlocks and two equipment hatches) located in approximately the same quadrant of the circumference of the shell which imposes asymmetric effects in the shell structure. The CV is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Subsection NE.

Traditionally, the local and global stability of freestanding steel containments have been designed by use of formulae using conservative assumptions based on an axisymmetric structure. ASME Code Case N-284 “Metal Containment Shell Buckling Design Methods, Class MC Section III, Division 1” outlines methodology for satisfying the stability of the CV using two approaches. Section 1710 provides a stress based buckling approach using detailed formulae that assumes an axisymmetric structure. The second approach provides guidance and acceptability based on a linear bifurcation analysis (2D (1720) or 3D (1730)). Due to the asymmetric structure of the CV, the 3D linear bifurcation method delivers the most accurate results.

The methodology and assumptions implemented by Westinghouse to qualify stability of the CV via Code Case N-284 are outlined. Also, the procedure to properly amplify the stresses as required by N-284 is included as justification of the methods used. This justification was thoroughly investigated by the Nuclear Regulatory Commission (NRC) and deemed acceptable.

Commentary by Dr. Valentin Fuster
2016;():V001T03A019. doi:10.1115/ICONE24-60674.

The seismic probabilistic risk assessment (SPRA) for a nuclear power plant involves the estimation of fragility curves for plant equipment. The seismic qualification of an equipment based on testing requires the equipment to continue to function when subjected to a specific test response spectrum (TRS). Broad banded ground motions have been found to cause more damage to equipment than the filtered narrow banded excitations. As a result, the definition of acceleration capacity used in the fragility models use clipped response spectra for both test response spectrum (TRS) and required response spectrum (RRS). The main purpose of the clipping factors is to convert a narrow banded response spectrum to a broad banded spectrum. The broadband correction factor and the modal interaction correction factors together contribute to the definition of clipping factor. The current study involves reconciliation with previous research by generating the mean response factor for different waveforms and subsequently the root-mean-square (RMS) severity ratio as a function of bandwidth. This ratio can be estimated for real earthquakes from their peak-to-rms values and the peak spectral values. In addition it can be shown that in case of real narrow banded earthquakes, this ratio is even lower and therefore the clipping would be greater. The modal interaction correction factor which considers the effect of interaction between different modes in case of broad banded time histories has also been investigated. The primary objective of this work is to study the existing Conservative Deterministic Failure Margin (CDFM) and Probabilistic approaches for estimating these factors as per the guidelines of EPRI [2] and apply the same to real life ground motions. It has been observed that the recommended practices are based on studying the behavior of random ground motions generated artificially for different bandwidths and center frequencies. The present study aims towards a more realistic fragility estimation of equipment by studying the spectral response of equipment based on actual ground motions. The purpose is to evaluate clipping factors that are consistent with Seismic Probabilistic Risk Assessment.

Commentary by Dr. Valentin Fuster
2016;():V001T03A020. doi:10.1115/ICONE24-60703.

In this paper the differences of floor response spectra (FRS) resulting from different ground response spectra are discussed. These spectra include the site effects which are quantified via site response analysis. This response is generated by wave propagation from the base rock through the overlying soil layers to the surface. The influences of the different layers and the corresponding dynamic soil properties are considered by using wave propagation analysis.

The paper then discusses the results obtained from seismic input at different depths conditions. Similar results might be expected, because the depth of the input spectra is adjusted for each layer. However, in comparing the floor response spectra of these calculations, significant differences are observed and therefore interpreted.

The paper is completed with the explanation of these significant differences and also with comparable floor response spectra.

Commentary by Dr. Valentin Fuster
2016;():V001T03A021. doi:10.1115/ICONE24-60714.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T03A022. doi:10.1115/ICONE24-60736.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T03A023. doi:10.1115/ICONE24-60768.

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.

Topics: Storage tanks
Commentary by Dr. Valentin Fuster
2016;():V001T03A024. doi:10.1115/ICONE24-60826.

Corrosion behavior of 9 %Cr ferritic/martensitic (F/M) P92, E911 and EUROFER steels was investigated in flowing (2 m/s) Pb-Bi with 10−7 mass%O at 450 and 550 °C for up to 8766 and 2011 h, respectively. The steels show mixed corrosion modes simultaneously revealing protective scaling, accelerated oxidation and solution-based attack. At 450 °C, the accelerated oxidation resulted in a metal recession averaging 6 μm (± 2 μm) after ∼8766 h while local solution-based corrosion attack ranged from ∼40 to 350 μm. At 550 °C, the accelerated oxidation resulted in a metal recession of about 10 μm (± 2 μm) after ∼2011 h. Solution-based corrosion attack appears more regularly at 550 °C, with a maximum depth ranged from ∼90 to 1000 μm. Incubation time for solution based attack is 500–2000 h for 450 °C and < 300 h for 550 °C. The EUROFER steel showed more severe metal recession via both oxidation and solution-based corrosion in comparison with P92 and E911 steels. The possible effect of alloying and structure on the corrosion response of 9 %Cr F/M steels is discussed.

Commentary by Dr. Valentin Fuster
2016;():V001T03A025. doi:10.1115/ICONE24-60845.

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.

Topics: Steel , Corrosion , Oxygen
Commentary by Dr. Valentin Fuster
2016;():V001T03A026. doi:10.1115/ICONE24-60854.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T03A027. doi:10.1115/ICONE24-60894.

The author recently found that there should exist a “radiation-induced electrolytic (RIE)” mechanism in the reactor water inducing severe interaction between structural materials and their environments in aged LWRs. 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.

These results are indicating that the in-core potential is approximately 0.1/0.4volt higher, in BWR(NWC)/PWR water chemistry respectively, when compared to the out-core regions. Through modeling studies, it was found that the concentrations of (DH)/(DO) for PWR/BWR(NWC) are higher/lower respectively, in the in-core region compared with the out-of-core region. These solute species in high concentrations should spontaneously decompose at the out-of-core region, enabling control of their water chemistry.

This mode of corrosion cell has been dismissed in the nuclear community considering that the transport of ions with flow is insignificant due to high purity of reactor water. Part 1 of this paper focuses on how the RIE phenomena are prompted although the reactor water is kept in high purity. The stable molecular species in the reactor water flow transport the valence electrons. They are released at the cathodic in-core region and are recovered at the anodic out-of-core region. Thus estimated potential differences have been benchmarked with the published in-pile test results for both PWR- and BWR water chemistry environments as explained in Part 2 of this series (1).

Commentary by Dr. Valentin Fuster
2016;():V001T03A028. doi:10.1115/ICONE24-60895.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T03A029. doi:10.1115/ICONE24-60962.

Steel-plate composite (SC) walls and associated connections can be designed based on the provisions of Appendix N9 to AISC N690s1. 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. AISC is currently in the process of developing a design guide to further enable the use of this specification. This design guide will explore the provisions of this specification in detail and discuss different possible design methodologies.

SC wall details at the beginning of the design process are based on typical plant layouts, shielding requirements and prevalent practices. The spacing of tie bars and steel anchors in SC wall needs to ensure the faceplate does not undergo buckling before steel yielding. The steel anchor additionally need to be spaced to ensure that (i) interfacial shear failure does not occur before out-of-plane shear failure, and (ii) the yield strength of the faceplates is developed over the development length. The tie bars need to have sufficient tensile strength to prevent splitting failure of SC walls. The elastic analysis of the SC walls is performed using a finite element analysis. The analysis needs to consider cracked transformed stiffnesses and equivalent material properties. The analysis will be conducted for operating thermal and accident thermal load combinations. The individual demands and the combination of demands need to be compared with the available strengths. The SC walls need to be adequately detailed for openings, meet construction and fabrication tolerances, and satisfy the Quality Assurance and Quality Control requirements. The designed SC walls needs to be safe for impactive and impulsive loads.

SC wall panel may need to be (i) anchored to basemat, (ii) connected to another SC wall panel, or (iii) connected to RC slab. The SC connections can be designed as full strength connection or over strength connections. The connection needs to have a well-defined force transfer mechanism. The connection required strength is determined from the design demands of SC walls and the connection design philosophy. The available strength is determined from the individual strengths of connectors participating in the force transfer mechanism.

Commentary by Dr. Valentin Fuster
2016;():V001T03A030. doi:10.1115/ICONE24-60994.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T03A031. doi:10.1115/ICONE24-61008.

The process and effluent radiation monitoring system can provide information about kinds of the radionuclide and activity concentrations, which is indispensable, important, and peculiar for a nuclear power station compared to a regular thermal power station. On the basis of knowledge about pressurized water reactors (PWRs) and high temperature gas-cooled reactors (HTGRs), the process and effluent radiation monitoring system of HTR-PM has been designed. It mainly contains several monitoring channels for concerned process systems, certain important areas, and gaseous and liquid effluents. For the coolant is helium and spherical fuel elements containing tristructural isotropic (TRISO) coated particles are adopted in HTR-PM, the source terms are different from those of PWRs. Not only fission or activation products in the gaseous or liquid form are monitored, but also the radioactive dust in the primary loop is sampled for analysis. The tritium (H-3) and carbon-14 (C-14) are taken a key consideration, which will be sampled in the primary loop, in certain important areas, in the secondary loop, and in the gaseous effluent in the stack. Design features of the process and effluent radiation monitoring system of HTR-PM are introduced and discussed compared to those of PWRs.

Commentary by Dr. Valentin Fuster
2016;():V001T03A032. doi:10.1115/ICONE24-61097.

This paper addresses a mixed mode driven cracking and its integrity assessment for applications in aging nuclear power piping. Following our earlier discussion on the use of mode-I based criteria in the current R6-method-based practice of integrity assessment, case studies conducted using finite element analysis are conducted and examined: (1) A plate with a single and multiple central crack(s) under tension; (2) A full-scale laboratory test of a straight pipe with an obliquely inserted crack in a dissimilar metal weld. Our results confirm the following earlier observations: For cases when mixed mode loading conditions are significant, (i) the fracture initiation predicted by using J-integral based mixed mode cracking criteria can approximately be achieved by an “effective stress intensity factor” based approach; (ii) it is not conservative to use a purely mode-I based criterion for the evaluation of the fracture failure assessment for typical problems of mixed mode driven cracking; (iii) The effect of multiple cracks can be significant and an assessment by only examining one crack, which is a common practice today, may not be fully conservative.

Commentary by Dr. Valentin Fuster
2016;():V001T03A033. doi:10.1115/ICONE24-61101.

Axial compression allowable stress for pipe supports and restraints based on linear elastic analysis is detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF. The axial compression design by analysis equations within NF-3300 are replicated from the American Institute of Steel Construction (AISC) using the Allowable Stress Design (ASD) Method which were first published in the ASME Code in 1973. Although the ASME Boiler and Pressure Vessel Code is an international code, these equations are not familiar to many users outside the American Industry. For those unfamiliar with the allowable stress equations, the equations do not simply address the elastic buckling of a support or restraint which may occur when the slenderness ratio of the pipe support or restraint is relatively large, however, the allowable stress equations address each aspect of stability which encompasses the phenomena of elastic buckling and yielding of a pipe support or restraint. As a result, discussion of the axial compression allowable stresses provides much insight of how the equations have evolved over the last forty years and how they could be refined.

Topics: Pipes , Compression
Commentary by Dr. Valentin Fuster
2016;():V001T03A034. doi:10.1115/ICONE24-61106.

Hard facings are used in a number of different types of components generally to provide improved wear, erosion, and corrosion resistance to the substrate. A typical application is to a valve seat which may be subjected to high wear loads and high flow rates such that a hard wearing surface is required.

Traditionally, a common application method has been to weld deposit the hard facing onto a softer substrate, e.g. the main body material of the component. This can result in poor quality micro-structures, e.g. cast/inhomogeneous structures being created that may not provide the required wear/erosion/corrosion resistance. This can be a particular issue for Nuclear applications where the hard facing is cobalt based. If the facing deteriorates and releases cobalt based debris, the debris (crud), if it becomes activated, can contribute significantly to the overall plant radiation activation level. Also, depending upon the geometry, deposition process, the expansion coefficient difference between the facing and substrate, and the cooling rate, a detrimental tensile residual stress can be generated. This can cause cracking of the facing which, depending upon the in-service loading conditions, could lead to crack propagation into the substrate, potentially threatening the structural integrity of the component; a leak or structural failure could result.

One application method to address the potential poor micro-structure that may be created from weld deposition methods, is to use a Hot Isostatically Pressed (HIPed) hard facing. This is where a wear surface is created from the solid state, rather than using a localised melting route, by the HIP consolidation of powder produced by a gas atomisation process. Homogenous, isotropic, finer grained, and defect free microstructures can be created via this method which generally exhibit improved resistance to wear/erosion/corrosion. Also, where the HIP consolidated facing is HIP bonded to the substrate, a beneficial compressive residual stress is created that can arrest any crack propagation into the substrate if cracking of the facing was to occur.

For Nuclear applications it is imperative the material quality of a hard facing is assured. It is essential that care is taken particularly in the production of the powder to ensure it does not contain undesirable defects/contamination.

This paper presents: the benefits that can be achieved with creating component hard facings using the HIP process, the key defects/contamination that may be present, and the quality control measures recommended to assure the material quality.

Commentary by Dr. Valentin Fuster
2016;():V001T03A035. doi:10.1115/ICONE24-61128.

Research and development of three-dimensional vibration simulation technologies for nuclear facilities is one mission of the Center for Computational Science and e-Systems of the Japan Atomic Energy Agency (JAEA). A seismic intensity of upper 5 was observed in the area of High-Temperature Engineering Test Reactor (HTTR) at the Oarai Research and Development Center of JAEA during the 2011 Tohoku earthquake. In this paper, we report a seismic response analysis of this earthquake using three-dimensional models of the HTTR building. We performed a parametric study by using uncertainty parameters. Furthermore, we examined the variation in the response result for the uncertainty parameters to create a valid 3D finite element model.

Commentary by Dr. Valentin Fuster
2016;():V001T03A036. doi:10.1115/ICONE24-61139.

The AP 1000 reactor coolant pump design included the selection and scrutiny of a Flywheel Retainer Ring Material to ensure its performance during its service life. This paper summarizes the results of an autoclave slow strain rate test (SSRT) program undertaken to establish the Environmentally Assisted Cracking (EAC) susceptibility of AP 1000 RCP Flywheel Retainer Ring (A 289 18Cr-18Mn Steel) material under simulated RCS environment. The tests program represented “beyond design basis” condition of leakage in the Alloy 625 hermetically sealed flywheel can. There is very little service experience or test data available in the literature on the performance of A289 material under exposure to primary water environment. The current program is undertaken to support the ACRS request. The testing was done under two phases. The first phase consisted of preliminary screening testing at 1×10−6/sec. The second phase represented slower strain rate at 2×10−7/sec. Both phases accompanied control sample tests under nitrogen. The primary coolant environment was maintained close to 1000 ppm boron as boric acid and 2 ppm lithium as lithium hydroxide which corresponded to a 300?C pH value of 7.0. The concentration of hydrogen was maintained at 25 cc/kg and the test temperature was maintained at 260 to 280?F. A number of SSRT parameters were employed in assessing the material susceptibility to EAC. The fracture morphologies were correlated by Scanning Electron Microscopy. Overall, the 18Cr-18Mn material SSR tests demonstrated sufficiently adequate resistance to SCC in primary water. Details of test parameters and test results will be presented.

Commentary by Dr. Valentin Fuster
2016;():V001T03A037. doi:10.1115/ICONE24-61141.

This paper reviews the role of fundamental factors such as stress, temperature and material microstructure in contributing to the PWSCC susceptibility in Alloy 600 components of operating PWRs. The effect of residual stresses from fabrication, cold work and welding as well as cyclic work hardening are considered. The influence of fabrication practice and the resulting microstructural parameters such as grain size, carbide distribution and ghost boundaries are considered. The paper considers the development of deterministic and probabilistic PWSCC susceptibility models to predict failures in the components and explain how these are applied to operating units. Investigation of service failures in the base metal and welds are considered.

Commentary by Dr. Valentin Fuster
2016;():V001T03A038. doi:10.1115/ICONE24-61159.

Object of the study relates to passive safety systems of cooling, heat removal and thermal protection that operate as independent evaporation-condensation (EC) systems and could maintain required thermal conditions of the technological systems of nuclear power cycle. Reliability of the passive systems is provided by absence of moving parts and by their operation based on physical laws of nature, i.e., without any intervention of staff, power supply, and control signals.

One of the main features of these systems is their ultimate heat transferring ability. There are hydrodynamic limitations of heat transferring ability connected with provision of coolant circulation in vapor-condensate lines of transportation zone that could be combined into two groups: 1) the crises depending upon quantity and distribution of liquid phase; and 2) the crises affected by hydrodynamic interaction of liquid and vapor phases. The authors undertook investigation of various thermophysical factors limiting this ability, determined and analyzed its regularities, which depend upon thermodynamical conditions, transport ability of capillary structures (if any), and the interaction of vapor and liquid flows of HP coolant.

Heat transferring ability of a model of EC system of passive surface cooling and thermal shielding under the conditions of heat supply from radiating surface of reactor simulator to heat pipes as the elements of two-row screen was investigated.

The analysis and calculations made by the authors proved the possibility to create an efficient passive evaporation EC system of surface cooling and thermal shielding of reactor unit. Such a system has a number of advantages as compared with known active safety systems (e.g., autonomy, higher reliability, and operational safety), does not require emergency water resources, compressed air systems, numerous valves, etc.).

The tests were performed at vertical orientation of HP evaporation zone (condensation zone was above evaporation zone) as a part of the double-row screen. The heat pipe was tested at its location in each of two rows and at two options of condensation zone: vertical and inclined in transportation zone at 20° to horizon.

It was found that only insignificant circumferential nonisothermality of heat-pipe surface under steady one-side heat supply in evaporation zone took place. Quite satisfactory agreement of the experimental and predicted values of heat flux transmitted by heat pipes of two-row screen was obtained.

The investigation proved efficiency and reliability of EC system of surface cooling and thermal shielding of the reactor equipment.

Commentary by Dr. Valentin Fuster

I&C, Digital Controls, and Influence of Human Factors

2016;():V001T04A001. doi:10.1115/ICONE24-60048.

Automated state identification systems facilitate reactor monitoring and control of nuclear systems by consolidating information collected by deployed sensors. In the current paper, we present the use of relevance vector machines (RVM) for real-time state identification of boiling water reactors (BWR). In particular, RVM models utilize the incoming signals of interest and identify in real time the state of the BWR either as normal or as one of the transition states. Each of the RVM models is assigned to a single signal; it receives the measured value at each instance and outputs the identified BWR state. The state that has been designated by the majority of the signals is displayed to the human operator as the identified BWR state. The proposed methodology is applied and tested on a set of signals taken from the FIX-II experimental facility that is a scaled representation of a BWR.

Commentary by Dr. Valentin Fuster
2016;():V001T04A002. doi:10.1115/ICONE24-60055.

The industry response to the Nuclear Regulatory Commission (NRC) Order EA-12-049 is based on a set of Diverse and Flexible Coping Strategies (commonly referred to as FLEX) for beyond design basis external events as described in NEI 12-06. The Pressurized Water Reactors Owners Group (PWROG) developed generic guidance for response to these Beyond Design Basis External Events (BDBEE), called FLEX Support Guidelines (FSGs). These guidelines are referenced from the plant Emergency Operating Procedures (EOPs) when it is determined that an event exhibits certain beyond design basis characteristics such as an Extended Loss of all AC Power (ELAP). These generic FLEX Support Guidelines provide a uniform basis for all PWRs to implement the FLEX guidance in NEI 12-06 that was endorsed by the NRC to maintain core, containment and spent fuel cooling.

The PWROG generic FSGs include guidance in FSG-7, “Loss of Vital Instrumentation or Control Power” for obtaining information for key plant parameters in an ELAP event. The key parameters were selected based on industry guidance and plant specific implementation. This set of key parameters will allow the licensed operators to have vital instrumentation to safely shutdown the core and maintain the core in a shutdown condition, including core, containment and spent fuel pool cooling. These parameters are used in the EOPs as well as the FSGs that are designed to mitigate a beyond design basis event. The requirements of NEI 12-06, as implemented through the FSGs, enhance both availability and reliability of instrumentation by requiring diverse methods of providing DC power for instrumentation and control as well as protection of instrumentation from the beyond design basis event.

The subsequent implementation of this guidance at the Byron Station has proven to also be beneficial for diagnosis of severe accident conditions (where core cooling could not be maintained). The same parameter values that are needed to verify core, containment and spent fuel cooling prior to core damage are also needed to diagnose severe accident conditions. Guidance provided within FSG-7, as implemented at the Byron Station, contains several layers of diverse methods to obtain parametric values for key variables that can be especially useful when the environmental qualification is exceeded for the primary instrumentation that provides this information. The methods range from the use of self-powered portable monitoring equipment to the use of local mechanical instrumentation. The FSG-7 guidance is referenced from the Byron Severe Accident Management Guidance (SAMG) to either obtain parameter information during a severe accident or to validate the information that is available from the primary instrumentation.

Commentary by Dr. Valentin Fuster
2016;():V001T04A003. doi:10.1115/ICONE24-60075.

An accelerator driven sub-critical (ADS) system consists of an accelerator, a sub-critical reactor, and a spallation target located at the centre of the reactor core. In this paper, we report the conceptual design of the control system for ADS system, which will integrate two big nuclear facilities, an accelerator and a reactor, into an overall system for the first time in the world. Because varied expectations on redundancy, diversity, availability, reliability, communication speed and latency are required for both accelerator and reactor, at least six systems have been designed for an ADS system, i.e. a main control system, three protection systems, a timing system, and a data archiving system. The main control system is designed to monitor and control the overall ADS system. In order to exactly couple the high-energy beam from the accelerator to the spallation target located in the reactor, many instrumentations are required in the main control system, such as instrumentations for beam monitoring and beam switching-off, instrumentations for ex-core neutron monitoring, instrumentations for in-core temperature and neutron monitoring, instrumentations for monitoring and controlling several cooling loops, and Instrumentations for control rod system. In order to enhance the safety of the ADS system during the operation, three protection systems, a reactor protection system, an accelerator’s machine protection system and a personnel protection system, have been designed. Independent network is used for each system. Furthermore, both redundant network and hardwired communication are used in each protection system to provide extremely reliable communication link. Finally, the communication network with parallel redundancy protocol has been studied for the main control system. In comparison with a network without redundancy, the parallel redundancy protocol can improve dramatically the reliability of a communication network by up to forty thousand times.

Commentary by Dr. Valentin Fuster
2016;():V001T04A004. doi:10.1115/ICONE24-60082.

In this paper, we study the electronics in the instrumentation and control (I&C) systems for an accelerator driven sub-critical (ADS) system, where a target located at the centre of a sub-critical reactor core is bombarded by the protons from an accelerator. In comparison with a commercial reactor used in nuclear industry, more control electronics are required to exactly couple the high-energy beam from the accelerator to the spallation target in the reactor core. There is a strong drive to utilize standard commercial-off-the-shelf devices to minimize cost and development time. In order to improve the reliability of I&C systems, redundancy architecture has been considered by adding more electronic devices. In comparison with I&C system without redundancy, the dual redundancy architecture improves the reliability of the system by 20000 times. Then, we study the potential application of electronics devices, such as the preamplifiers for detectors, in the reactor building by shielding them with shielding materials. Since the most effective neutrons in creating radiation damages are those fast neutrons with the energy of more than 0.1 MeV, we have proposed a sandwich shielding method to reduce the neutron-induced radiation effects, in which the first and third layers are made of polyethylene and the second layer is made of heavy metal, e.g. tungsten. Simulation results with GEANT4 code have indicated that the shielding with a 30 cm-thick sandwich can increase the expected lifetime of electronics by 1258 times, and can reduce the soft errors caused by single event upsets by 5400 times.

Commentary by Dr. Valentin Fuster
2016;():V001T04A005. doi:10.1115/ICONE24-60125.

Since the digital instrument and control (I&C) system was used in nuclear power plant (NPP), software Verification and Validation (V&V) is becoming more and more important for ensuring the safety function to be correctly implemented by the safety system. According to the classification of different safety functions of I&C system, with the use of digital technology, software classification needs further discussion to determine. And different classification of software performs different testing. Software V&V processes can be divided into different stages by reference to the standard IEEE 1012-2004, and each stage focuses on different contents. In China, software V&V on the NPP has been started from LingAo Phase II project and strongly done on other CPR1000 projects, which is a Generation II + pressurized water reactor. Based on the practice about YangJiang units 5 and 6 projects, combined with the relevant laws and regulation standards, this study summarizes the characteristics of the independent third party software V&V, analyzes the key points and methods of V&V activities in the software development process. As a result, it is also benefit to the design, operation and maintenance of safety I&C System as technical reference.

Commentary by Dr. Valentin Fuster
2016;():V001T04A006. doi:10.1115/ICONE24-60148.

The U-tube Steam Generator (UTSG) of AP1000 Nuclear Power Plant (NPP) is the crucial component transferring heat from the primary loop to the secondary loop to make steam. The UTSG of AP1000 NPP is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it is difficult and challenging to well control the water level of AP1000 UTSG by tuning the PID controller parameter in a traditional way, especially when the system is undergoing a sharp transient. To achieve better control performance, the Particle Swarm Optimization (PSO) algorithm was applied for the parameter optimization of the AP1000 UTSG feedwater control system in this study. First, the mathematical model of AP1000 UTSG was established and the objective function was developed with the system constraints considered. Second, the simulation platform was built and then the simulation was conducted in MATLAB/Simulink environment. Finally, the optimized parameters were obtained and the feedwater control system with optimized parameters was simulated against that without optimized. The simulation results demonstrate that optimized parameters of AP1000 UTSG feedwater control system can significantly improve the water level control performance with smaller overshoot and faster response. Therefore, the PSO based optimization method can be applied to optimizing AP1000 UTSG feedwater control system parameters to provide much better control capabilities.

Commentary by Dr. Valentin Fuster
2016;():V001T04A007. doi:10.1115/ICONE24-60274.

This article designs the water level of steam generator control system applying PCH system modeling and control principle of Energy-Shaping.

The basic principles of port-controlled Hamiltonian (PCH) system modeling and control principle of Energy-Shaping are provided, including passivity, dissipation and so forth. Meanwhile, port-controlled Hamiltonian (PCH) system, energy-shaping principle, matching method and its corresponding realization and stability analysis are presented. The water level of steam generator is accomplished; the controlled object is mathematically modeled with principle modeling method. Then the mathematic model is translated into PCH model. the proper Hamiltonian function is finely selected. The water level equilibrium point of steam generator system is respectively determined; the PCH controller based on energy-shaping method is ultimately obtained.

The proposed modeling and control methods provide a potential idea for the other control methods applied method on steam generator system. Moreover, the proposed methods can be put into industrious process control.

Commentary by Dr. Valentin Fuster
2016;():V001T04A008. doi:10.1115/ICONE24-60406.

High Temperature gas-cooled Reactor Pebble-bed Module (HTR-PM) is a high temperature gas-cooled reactor demonstration plant with a structure of two modules feeding one steam turbine. Compared with the structure of a single reactor feeding a turbine, there are more devices shared between these two modules. When they are operated, the shared components are prone to introduce collisions or even logical deadlocks for different technical processes. The future commercial HTR-PM plants are supposed to comprise more modules for a larger turbine, thus the collision problem and potential deadlocks introduced by the shared components will become severer, which may impact the efficiency or even the functions of the related systems. Therefore, how to design suitable control policies in the distributed control system (DCS) to relieve the collisions and avoid the deadlocks during using these shared devices is a new and also a very important problem. In this paper, the classifications of the shared devices are first addressed, and then how to identify the shared objects of nuclear power plant (NPP) is proposed. Furthermore, a general model for the control logic design is proposed, taking into consideration the collision avoidance, time delay and fairness. Moreover, a scheme for deadlock avoidance is proposed. The example of how to apply the schemes to relieve the conflicts and deadlocks in the processes of using the shared devices in fuel element cycling system is illustrated.

Commentary by Dr. Valentin Fuster
2016;():V001T04A009. doi:10.1115/ICONE24-60408.

Redundancy is widely used for enhancing a system’s overall availability. As an HTR demonstrated plant, a high temperature gas-cooled reactor pebble-bed module (HTR-PM) now is under construction in China and the construction will be completed around 2017. In HTR-PM, there are many devices and device groups used in a redundant way to guarantee the high availability of the related functions, especially the functions shared by two reactors during the entire life time. It is very important and necessary to determine their reliabilities as well as how to make a decision about the related maintenance policies to enhance their availabilities. In this paper, typical redundant styles in the HTR-PM are summarized and demonstrated. Accordingly, the theoretical models, which are able to describe the reliabilities of the redundant systems, are proposed based on Markov chain model. Moreover, for a specific redundant structure, the relationship between the availability and the maintenance period is analytically addressed. Based on the model, we address that: as the digital monitoring and control technologies are widely used in nuclear power plants, monitoring methods targeting for decreasing maintenance costs and meanwhile increasing the availabilities for different redundant styles are very beneficial.

Commentary by Dr. Valentin Fuster
2016;():V001T04A010. doi:10.1115/ICONE24-60483.

The demonstration construction of the Chinese design of modular high temperature gas cooled reactor (MHTGR), named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM) is now under construction. HTR-PM employs a layout of two nuclear steam supply system (NSSS) modules coupled to one steam turbine. The verification and validation (V&V) of HMIs is important for HTR-PM, as the first two-modular design. The V&V program of HMIs for HTR-PM is introduced in this paper. The V&V activities may include the static and dynamic verifications. The basic layout and static HMI usabilities will be verified on the static stage, and the task support based on operation procedures will be evaluated in the dynamic V&V phase. A verification platform is built to make verification and validation of the HMIs. The HMIs of HTR-PM will be improved according to the V&V result.

Commentary by Dr. Valentin Fuster
2016;():V001T04A011. doi:10.1115/ICONE24-60535.

Nuclear safety is one of the key issues for a nuclear power plant (NPP). Digital instrumentation and control (I&C) systems have been employed gradually in the newly-built and upgraded NPPs, while the reliability of software brings great challenges to the Probability Risk Assessment (PRA) of NPPs. Software testing is regarded as one of the most important methods to guarantee the quality of safety software. The testing data can then be adopted to assess the coding quality by reliability modelling. As the variety of digital I&C systems, software modelling methods corresponding to particular I&C system, as well as a model which is suitable in all situations, are both expected.

The Reactor Protection System (RPS) in High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated commercially in China. As the designer, we also took part in the software testing work of this digital I&C system. In this paper, we gave a comprehensive introduction to the software testing and reliability modelling research of RPS in HTR-PM, including the objective, tools, methods, testing strategy, organizational structure, and the implementation phases.

During the testing experience of safety software of RPS in HTR-PM, we collected the software abnormal reports which could be employed for the reliability analysis to evaluate the quality of the safety software. We introduced the data mining and reliability modelling research according to the abnormal reports. Different characteristics of faults could be used for software reliability modelling, such as software version, fault severity, test stage, submission date, debugging data, and so on. In the end we introduced a software modelling method based on severity analysis of the abnormal reports.

The work we showed in this paper can contribute to improve the process of testing and reliability analysis for other digital I&C systems in NPPs.

Commentary by Dr. Valentin Fuster
2016;():V001T04A012. doi:10.1115/ICONE24-60657.

At present, simulation research on condenser mostly focuses on the one of power plant. While little attention is paid on the marine one. The marine condenser is different from the one of power plant. One difference between them is that marine condenser limited by the hull space is much smaller than the one of power plant. Another difference is that marine condenser usually runs in condition that steam load changes greatly and frequently. In addition, the temperature of cooling water changes with seasons and maritime space. When the ship makes an ocean-going voyage, the temperature of cooling water varies obviously, which has a negative effect on the control system. Therefore, simulation research to the dynamic characteristic of marine condenser is carried on by this paper.

The vacuum and the condensate supercooling degree are the main operating parameters for a condenser. In this paper, a mathematical model for marine condenser is established. The dynamic variation of each parameter is monitored by this model. Then an expert control system is designed. The condenser vacuum and supercooling degree are controlled by the flow rates of cooling water, air-blend and steam jet for deaerating. The deaerating steam bubbles which are not absorbed completely by condensate water, have a negative impaction on the condenser vacuum. The deaerating steam bubbles heating the condensation water to be exactly saturated is calculated. Referring to it, the fluctuation of condenser vacuum is filtered. So the negative influence on condenser control system can be remitted. The temperature of water replenishing is usually lower than that of condensate water in condenser heat well. And the water is replenished into the heat well part. So water replenishing would increase the supercooling degree of condensate water.

This paper adopts the simulation experiment method to gain the expert knowledge of condenser control. According to power level, inlet temperature of cooling water, the messages of operating modes and sensors, the control system regulates the regulating valves for turbo circulating pump of cooling water, air extractor and steam jet deaerator to change the flow rates of cooling water, air-blend and steam jet for deaerating, so as to achieve automatic control of condenser vacuum and supercooling degree. Finally, two PID controllers are added as well, to counteract the adverse effect, enhance anti-interference ability and optimize control performance of intelligent control system. Simulations of varying load and cooling water temperature are done. Results show that the intelligent control system could effectively control condenser vacuum and supercooling degree under different power level and variable temperature of cooling water condition.

Commentary by Dr. Valentin Fuster
2016;():V001T04A013. doi:10.1115/ICONE24-60672.

A novel two-stage feature extraction scheme is proposed in this paper for eliciting discriminant information contained in the data from various nuclear power plant (NPP) sensors to facilitate event identification. Based on the idea of sensor type-wise block projection, the primal features can be extracted without losing the intrinsic structure contained in the multi-sensor data. The features are then subject to further dimensionality reduction by selecting the sensors that are most relevant to the events under consideration. Results from detailed experiments with data generated from a simulator of Taiwan Maanshan NPP illustrate the efficacy of the proposed scheme.

Commentary by Dr. Valentin Fuster
2016;():V001T04A014. doi:10.1115/ICONE24-60815.

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.

Topics: Sensors
Commentary by Dr. Valentin Fuster
2016;():V001T04A015. doi:10.1115/ICONE24-60848.

Study of availability in radiation environment is presented for video monitoring system. Testing experiment has been completed by seven kinds of Image sensor module which includes three kinds of digital image sensor module and four kinds of analog image sensor module. Radiation accident condition was simulated by γ-ray ionizing radiation environment where the dose rate at 16.63Gy/h 20.20Gy/h and 58.30Gy/h. Availability has been studied by analyzing real time monitoring image quality parameters captured in γ-ray exposure environment. The primary image quality parameters include average gray level of the dark image and synthetic brightness of the color image. The most suitable image sensor module has been selected by image quality parameters for comparison before and after irradiation. Experimental results show that, digital camera has minimum background noise. The radiation resistance of CMOS image sensors is better than CCD image sensors. Therefore, digital video monitoring system with CMOS image sensor has the best image quality parameters and slightest effect in γ-ray ionizing radiation environment where the dose rate is less than 58.30Gy/h. Meanwhile, adequate light could reduce noise interference reduced by γ-ray for all types of video monitoring system. However, digital signal processing integrated circuit board has been destroyed when the accumulated dose reach to 88.40Gy, but the CMOS image sensor integrated circuit board has normal working parameters. As a conclusion, digital video monitoring system with CMOS image sensor could be used for real-time monitoring in radiation accident condition, but the digital signal processing integrated circuit board needs to be hardened by radiation hardened technology.

Commentary by Dr. Valentin Fuster
2016;():V001T04A016. doi:10.1115/ICONE24-60879.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T04A017. doi:10.1115/ICONE24-60897.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T04A018. doi:10.1115/ICONE24-61024.

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.

Commentary by Dr. Valentin Fuster
2016;():V001T04A019. doi:10.1115/ICONE24-61034.

Thermoacoustic Power Sensor (TAPS) technology offers the potential for self-powered, wireless measurement of nuclear reactor core operating conditions. TAPS are based on thermoacoustic engines, which harness thermal energy from fission reactions to generate acoustic waves by virtue of gas motion through a porous stack of thermally nonconductive material. TAPS can be placed in the core, where they generate acoustic waves whose frequency and amplitude are proportional to the local temperature and radiation flux, respectively. TAPS acoustic signals are not measured directly at the TAPS; rather, they propagate wirelessly from an individual TAPS through the reactor, and ultimately to a low-power receiver network on the vessel’s exterior. In order to rely on TAPS as primary instrumentation, reactor-specific models which account for geometric/acoustic complexities in the signal propagation environment must be used to predict the amplitude and frequency of TAPS signals at receiver locations. The reactor state may then be derived by comparing receiver signals to the reference levels established by predictive modeling. In this paper, we develop and experimentally benchmark a methodology for predictive modeling of the signals generated by a TAPS system, with the intent of subsequently extending these efforts to modeling of TAPS in a liquid sodium environment.

Commentary by Dr. Valentin Fuster

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