ASME Conference Presenter Attendance Policy and Archival Proceedings

2017;():V007T00A001. doi:10.1115/ICONE25-NS7.

This online compilation of papers from the 2017 25th International Conference on Nuclear Engineering (ICONE25) 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presenter attendance policy, if a paper is not presented at the Conference by an author of the paper, the paper will not be published in the official archival Proceedings, which are registered with the Library of Congress and are submitted for abstracting and indexing. The paper also will not be published in The ASME Digital Collection and may not be cited as a published paper.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

Fuel Cycle, Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management

2017;():V007T10A001. doi:10.1115/ICONE25-66036.

The radioactive activity of spent nuclear fuel is high, and the transportation safety is concerned by public and specialist. The periodic radiation shielding performance measurements of spent fuels package is important content to ensure transportation safety of spent fuels. The radiation shielding performance of package must meet the requirements of “Regulations for the safe transport of radioactive material” (GB11806-2004). However, some of the problems and difficulties reflected in practice need to be solved, such as the measurements results of neutron radiation level of spent fuels package outer are not always reliable. In this paper, the periodic shielding performance measurements of one type of spent fuel transportation package are presented. The monitoring results of using both the neutron multi-sphere spectrometer and portable neutron measurement instrument are compared, and the Monte Carlo simulation is done to verify the measurements results. Some factors are discussed, and an optimized scheme is recommended.

Commentary by Dr. Valentin Fuster
2017;():V007T10A002. doi:10.1115/ICONE25-66081.

Legacy radioactive waste streams from the Cold War still exist and newly generated waste streams from nuclear power plants and research institutes go untreated and expose environmental hazards at many nuclear sites. The nature of the waste is extremely diverse, depending upon the source or the process from which it originated. The most problematic waste streams include complex liquids such as organic (tri-butyl-phosphate TBP) solutions contaminated with Pu and U isotopes, mixed sludge types, high acid radioactive waste, H-3 tritium contaminated organic and aqueous streams, etc. Environmental and economic challenges exist for the treatment and disposal of such waste streams.

A proven technology that has been applied to LRW on a global basis provides a low-cost solution to legacy streams and small volume, highly complex LRW frequently generated from routine NPP operations, in nuclear laboratories and during decommissioning. The engineered polymer technology from Nochar, USA, is capable of solidifying standard and highly complex LLW and ILW waste streams for interim or final storage, or for incineration.

Commentary by Dr. Valentin Fuster
2017;():V007T10A003. doi:10.1115/ICONE25-66155.

Project management literature has, until now, mainly focused on new build and only in the last decades the issues of decommissioning (mega) projects has arisen. To respond to this changing environment, project management will need to understand the challenges of decommissioning projects. Decommissioning projects within Oil & Gas, Chemical and Nuclear sectors are characterized by high costs, long schedules and uncertainty-based risks. The budget for Nuclear Decommissioning Projects and Programmes (NDPs) are subject to well publicized increases and, due to their relatively recent emergence, complexity and variety, key stakeholders lack a full understanding of the key factors influencing these increases.

Benchmarking involves “comparing actual or planned practices [...] to identify best practices, generate ideas for improvement” [1] and offers significant potential to improve the performance of project selection, planning and delivery. However, even if benchmarking is the envisaged methodology to investigate the NDPs characteristics that impact on the NDPs performance, until now, it has only been partially used and there is a huge gap in the literature concerning benchmarking NDPs.

This paper adapts a top-down benchmarking approach to highlight the NDPs characteristics that mostly impact on the NDPs performance. This is exemplified by a systematic quantitative and qualitative cross-comparison of two major “similar-but-different” NDPs: Rocky Flats (US) and Sellafield (UK). Main results concern the understanding of the alternatives of the owner and/or the contractors in relation to (1) the physical characteristics and the end state of the nuclear site, (2) the governance, funding & contracting schemes, and (3) the stakeholders’ engagement.

Commentary by Dr. Valentin Fuster
2017;():V007T10A004. doi:10.1115/ICONE25-66279.

Due to the potentially dangerous properties of radioactive material, it is during the transport that the process of nuclear energy and technology uses are prone to nuclear and radiation accidents. Radioactive material hence must be transported with reasonable containers to achieve heat dissipation, confinement of radioactive material, radiation shielding and prevention of nuclear criticality. The key to transport safety lies in the designing and manufacturing quality of the transport containers. Therefore, the safety supervision for transport containers of radioactive material is a guarantee for the environment and the public from nuclear and radiation hazards, also is international general practice.

As the most authoritative international organization, International Atomic Energy Agenda (IAEA) draws up and regularly revises safety regulation ‘Regulation for the Safe Transport of Radioactive Material’, which proposes technical indicators for transport containers of radioactive material and responsibility of competent authorities. According to the transport modes, other international organizations, such as International Maritime Organization, International Civil Aviation Organization, International Air Transport Association, United Nations Economic Commission for Europe, enacted related transport safety regulations based on actual needs.

This paper introduces the administrative licensing approval process for the transport containers of radioactive material in China and the research on competent authority and approval procedure in American, Russia, France, Canada, Germany and Great Britain. In China, National Nuclear Safe Administration (NNSA) is responsible for the licensing approval for the transport containers of radioactive material, including designing, manufacturing, using and transporting of transport containers. NNSA also organizes and formulates relevant administrative regulations and approval procedures, and has issued administrative regulation ‘Regulation on the Safe Management for the Transport of Radioactive Material’ and a series of administrative rules, management procedures, guide, technical documents and so on. These regulations established the sort management of radioactive materials and the responsibility for competent authority, and also stipulated approval and supervision for transport and transport containers of radioactive materials. While some other countries, such as America, certifies the transport containers of radioactive material to achieve the control.

The domestic and overseas research into administrative licensing approval processes for transport containers is in view of the increasing transport of radioactive material among countries and the requirement of international transport. Transport containers with material of high potential risk, such as spent fuel, need to obtain the transport approval from the competent authority of transit or arrival country. Therefore, the research on domestic and other countries licensing management of transport containers of radioactive material, which is not only beneficial to improving the transport safety management of radioactive material in China, but also can promote international transport campaigns of radioactive material..

Commentary by Dr. Valentin Fuster
2017;():V007T10A005. doi:10.1115/ICONE25-66336.

In this paper, the feasibility of continuous countercurrent extractor (CCE) applied in extraction of U scraps rather than mixer-settler was studied. Effects of rotation Reynolds number, residence time on extraction and reverse extraction efficiency were investigated respectively. During the process, impurity content in raw material had a slight effect on extraction. Under the optimum conditions of O/A phase ratio of 1.2 and rotation Reynolds number of 13824, extraction degree of scrap pellet solution and alkalescence dreg lixivium were good with the residence time of 19 min and 26 min respectively.

Under the best reverse extraction conditions of A/O phase ratio of 1.2, rotation Reynolds number of 23128 and residence time of 19 min, two kinds of raw material liquid mentioned as above could get better reverse extraction efficiency, while uranium concentration of raffinate phase and impurity content of reverse extract phase satisfied the technological requirement. This study verified the feasibility of CCE developed for purification of U scraps, identified the operation parameter, and provided an experimental basis for industrial expansion in the future.

Commentary by Dr. Valentin Fuster
2017;():V007T10A006. doi:10.1115/ICONE25-66417.

Tritium (H-3) can discharge mainly in the form of HT and/or HTO as gaseous and/or liquid waste into the environment from the nuclear power plant, and participate in the cycle among hydrosphere, atmosphere and biosphere, which would lead to the long-term radiological effects on organisms. Thus, in the daily operation of the nuclear power station, tritium became one of the most concerned nuclides in the source term analysis. In high temperature gas-cooled reactors (HTGRs), tritium was mainly generated by the ternary fission reaction of heavy nuclei in the fuel and neutron activation reaction of impurities like lithium-6 (Li-6), lithium-7 (Li-7) and boron-10 (B-10) in the graphite matrix, carbon bricks, etc. Tritium would be resorted completely in the intact tristructural-isotropic (TRISO) coated particles, while tritium in the graphite can diffuse into the primary loop depending on the local temperature. In the helium purification system of a typical HTGR, the molecular adsorber can adsorb the tritium in the primary coolant, and then the tritiated water was formed from the regeneration and desorption process of the molecular adsorber. Meanwhile, since the high permeability of tritium at a high temperature, it can permeate into the secondary loop through stainless steel heat exchange tube from the primary loop, and entered into the environment with leakage of the secondary water. Therefore, it was very important to analyze the production, transport and release mechanism of tritium for the estimation of the inventory and distribution of tritium in a nuclear power plant. With the rapid development of nuclear energy and the commercial application of HTGRs, tritiated water treatment technologies attracted more attention in the field of radioactive nuclear waste. Current paper will introduce and summarize general tritiated water treatment technologies, including water distillation, tritium sorbent process, palladium membrane reaction (PMR), and combined electrolysis catalytic exchange (CECE).

Commentary by Dr. Valentin Fuster
2017;():V007T10A007. doi:10.1115/ICONE25-66429.

JMCT-S[1] 3-D high resolution Monte Carlo neutron and photon transport code is proposed for shielding design due to its significant advantages such as complex geometry modeling, accurate calculation method, high efficient parallel computing and visible mass data analysis. To demonstrate its feasibility and benefits in engineering, this paper presents the verification of JMCT-S code based on primary shielding analysis for CAP1400 nuclear power station.

Firstly, by employing JMCT-S code this paper completes the modeling and simulations of CAP1400 nuclear power station with full-core pin-by-pin and reactor pressure vessel internals inside of primary shield wall[2]. Secondly, neutron fluence rates are calculated based on the equilibrium cycle information. Particularly, the full-core neutron source is calculated from 3-D power distribution, fissionable nuclide fractions, fission spectrum, particle numbers and energy released per fission. Finally, JMCT-S code is verified by comparing results with those of the MCNP[3] code.

According to the calculation results, JMCT-S code has similar accuracy to the MCNP code. The relative error of neutron fluence rate in active core is within ±1.5%, and deviation of neutron fluence rate for fast neutron fluence rate on inner surface of reactor pressure vessel is less than ±10%. Consequently, JMCT-S code is reliable in shielding design of reactor plant which lays the foundation for usage of JMCT-S code.

Commentary by Dr. Valentin Fuster
2017;():V007T10A008. doi:10.1115/ICONE25-66430.

As China’s first nuclear power plant connected to the grid, the first Qinshan nuclear power plant is approaching the decommissioning period. Other nuclear power plants also turn into the preparation phase of decommissioning in succession. In order to facilitate decommissioning, source survey is conducted during the pre-decommissioning phase, which can provide radioactive inventory, contamination distribution, species and quantities of nuclides.

The internals of the reactor work under the most severe radiation environment. During the reactor operation, the materials of internals are irradiated by high-energy neutrons. So activated nuclides are generated due to the neutron capture reaction, which are the main radioactive waste to be treated during decommissioning. In this paper, the neutron irradiation and the generated activation source of the internals for pressurized water reactors (PWR) are studied and analyzed.

Firstly, core modeling was carried out, and the neutron transport calculation is performed to obtain three-dimensional distribution of the neutron flux.

Secondly, according to the three-dimensional distribution of the material composition and the neutron flux rate of the reactor, the activation calculation is carried out to obtain the activation source.

Commentary by Dr. Valentin Fuster
2017;():V007T10A009. doi:10.1115/ICONE25-66445.

In this study, the applicability of Monte-Carlo code PHITS(1) to the equipment design of sampler and detector in the radiation monitoring system was evaluated by comparing calculation results with experimental results obtained by actual measurements of radioactive materials.

In modeling a simulation configuration, reproducing the energy distribution of beta-ray emitted from specific nuclide by means of Fermi Function was performed as well as geometric arrangement of the detector in the sampler volume. The reproducing and geometric arrangement proved that the calculation results are in excellent matching with actual experimental results. Moreover, reproducing the Gaussian energy distribution to the radiation energy deposition was performed according to experimental results obtained by the multi-channel analyzer.

Through the modeling and the Monte-Carlo simulation, key parameters for equipment design were identified and evaluated.

Based on the results, it was confirmed that the Monte-Carlo simulation is capable of supporting the evaluation of the equipment design.

Commentary by Dr. Valentin Fuster
2017;():V007T10A010. doi:10.1115/ICONE25-66462.

Spent fuel dry storage technology is one of the most important intermediate storage technologies for spent fuel, because of its high security, good economic and easy to expand the scale. This article aims at designing a spent fuel dry storage cask which can contain 21 FA300 spent fuel assemblies.

The spent fuel dry storage cask is designed as concrete cask structure, which has the advantages of low manufacturing cost and simple manufacturing technology. Ventilation channels are designed for heating transfer, because the concrete is not a good thermal conductivity material. And labyrinth structure is designed for the ventilation channel to reduce the cavity streaming.

Radiation sources in spent fuel assemblies are mainly produced from fission products, actinides and their daughters located inside the effective fuel region, and other activation products in structure materials, which are calculated by ORIGEN.

The source and geometry of this problem are complex, and this is a real world deep penetration and streaming problem. Discrete ordinate method has great advantage in solving the deep penetration problem. Based on three-dimensional discrete ordinate code TORT, radiation shielding design method for spent fuel dry storage cask is studied, including main shield cask, cover lid, and ventilation channel. The results show that this spent fuel dry storage cask containing 21 FA300 spent fuel (cooling time: 10 years) assemblies can satisfy the requirement of dose rate limits in GB18871.

Commentary by Dr. Valentin Fuster
2017;():V007T10A011. doi:10.1115/ICONE25-66508.

Dry storage is one of the ways to store spent fuel in the middle of the reactor, which can effectively alleviate the pressure of the storage on the spent fuel pool of nuclear power plant. This paper tries to combine the site of dry storage facilities and the design characteristics to explain and discuss the safety evaluation method under the accident condition, from the mechanical analysis, critical safety, the decay heat removal, the shielding design and so on. Then according to the operating procedures and the accident condition that may be occurred, put forward some possible ways of monitoring and measures of safety protection should be added.

Commentary by Dr. Valentin Fuster
2017;():V007T10A012. doi:10.1115/ICONE25-66540.

In order to recommend a pool scrubbing model to be applied in investigating the aerosol pool scrubbing effect of spent fuel pool (SFP) during the period of CAP1400 containment overpressure release, different pool scrubbing models are studied in this paper. At first, the thermal hydraulic processes and aerosol retention mechanisms considered in five mainstream pool scrubbing models are compared briefly, through which SPARC-B/98 model (the updated version of SPARC-90) is identified as the most comprehensive one. Subsequently, ACE aerosol pool scrubbing experiments are simulated using SPARC-B/98 and SPARC-90 model respectively, whose results could contribute to assess the conservatism of the two models. At last, considering the lack of experimental decontamination factor (DF) data under CAP1400 SFP pool scrubbing boundary conditions, sensitivity study involving these special conditions is performed with SPARC-B/98 and SPARC-90 model, in which injected gas flow rate, steam volume fraction in injected gas and pool temperature are selected as the analyzed parameters. The results show that SPARC-B/98 model is more conservative and applicable under SFP pool scrubbing boundary conditions and are advised to be used in analyzing the SFP aerosol pool scrubbing effect.

Topics: Aerosols
Commentary by Dr. Valentin Fuster
2017;():V007T10A013. doi:10.1115/ICONE25-66558.

For passive nuclear power plants, the radioactive aerosols in containment are removed by natural processes after accidents. These processes have high reliability. However, it is very complicated to analyze the behaviors of aerosols in natural processes. The dominant processes include coagulation, sedimentation, diffusionphoresis, and themophoresis. The main work of this paper is summarized as follows: (1) To determine methods of analysis radioactive aerosol natural removal coefficient in containment. (2) To complete comparative analysis natural removal processes to AP1000 and CAP1400 after LOCA. (3) A comprehensive sensitivity analysis is carried out for a number of parameters, including containment free volume, sedimentation area, aerosol size, packing fraction, mass ratio of radioactive and nonradioactive aerosol, and steam condensation rate etc, (4) To complete comparative analysis of natural removal processes between core meltdown and non-meltdown accident sequence after LOCA. The results show that, (1) Removal coefficient by sedimentation of CAP1400 is larger than AP1000, removal coefficient by diffusionphoresis and themophoresis of CAP1400 also smaller than AP1000. (2) In general, the greater the containment free containment, the smaller the aerosol natural removal coefficient, and the greater aerosol size, the more the amount of aerosol removed by sedimentation, in the case of other parameters unchanged. (3) The removal coefficients have few differences after core meltdown and non-meltdown accidents.

Commentary by Dr. Valentin Fuster
2017;():V007T10A014. doi:10.1115/ICONE25-66598.

Typically, the concentration of natural radioactive aerosols in the containment vessel of nuclear power plant is low and in equilibrium. But when a serious nuclear accident occurred, the massive radioactive aerosols would be rapidly released. In order to ensure the integrality of the containment, the pressure inside the containment must be reduced by reducing the concentration of the aerosol. It can cause a serious damage to the atmospheric environment if such radioactive aerosol directly release. In this paper a new mesoscopic impactor filter has been developed which not only can filtrate and collect the aerosol particles but also can decrease the flowing resistance of gas. This paper intends to make numerical simulation to study the regularity of the deposition of aerosols under the laminar condition at different working condition in mesoscopic impactor filters. The 3D model of the filter was established with the commercial software of ICEM CFD and the meshes were divided accurately. The gas phase uses the laminar model and the particles use DPM (Discrete Phase Model). The detailed modeling method is given and the simulation results are analyzed.

Commentary by Dr. Valentin Fuster
2017;():V007T10A015. doi:10.1115/ICONE25-66611.

Development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima Daiichi Nuclear power Station (Fukushima NPS). In this study, the adsorption properties for nine kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Modified Chabazite, Phillipsite, Erionite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of sodium salts, boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ions (Cs+, Sr2+, Eu3+, I, UO22+, Am3+ and NpO2+), and (2) multi-nuclide adsorption properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. The distribution coefficient (Kd, ml/g) and uptake (R, %) were estimated by batch method using NaI (Tl) scintillation counter, ICP-AES and AAS.

Zeolites with different crystal structures have the diversity of the adsorption selectivity for various radioactive nuclides. Chabazite, mordenite and clinoptilolite with lantern or tunnel structure were very effective for the adsorption of monovalent Cs+ ions even in real seawater. Zeolite A and X with three-dimensional cage structures were effective for the adsorption of divalent Sr2+ and Co2+ ions under the practical condition (30% diluted seawater). Zeolite L was effective for the adsorption of Eu3+ ions under the practical condition. As for I adsorption, Ag-zeolites are found to be effective, and the uptake (%) of I (NaI in pure water) for Ag-zeolites was estimated to be above 98% in pure water. As for actinoid adsorption, the distribution profile, Kdvs pH, had a maximum depending on the hydrolysis pH. Zeolite A, Zeolite L and Zeolite X showed an excellent adsorption property for UO22+, Am3+ and NpO2+, respectively.

Selective adsorption tendencies of different zeolites were evaluated for 26 elements referred to ALPS. Comparing the uptake results for different zeolites, the following tendency of adsorbability was observed. Mordenite had adsorption selectivity for monovalent alkali metal ions of Rb+ and Cs+. Zeolite A and X exhibited relatively high adsorption selectivity for divalent ions of Sr2+ and Co2+. Zeolite L had adsorption selectivity for trivalent lanthanide ions such as Ce3+ and Eu3+. These tendencies were the same as those without boric acid.

Thus, the zeolites with diverse adsorption selectivity are effective for the multi-nuclide decontamination of radioactive contaminated water.

Commentary by Dr. Valentin Fuster
2017;():V007T10A016. doi:10.1115/ICONE25-66693.

The present paper provide an available FEA method for the structural analysis and seismic design of pipelines coated with flexible shielding materials, in nuclear power plant, subjected to earthquake action. The modal and spectral analysis are carried out by using FEM software ANSYS. By comparing the numerical simulation results of dynamical behavior of the pipelines with or without flexible shielding materials, the effects of the flexible shielding material on the seismic performance of the pipeline system in nuclear power plant are discussed. Finally, a local supporting constraint for reducing seismic effects is put forward, which can significantly decrease the maximum stresses in the pipelines coated with flexible shielding materials.

Commentary by Dr. Valentin Fuster
2017;():V007T10A017. doi:10.1115/ICONE25-66713.

Based on the PWR nuclear power plant, the sources of radioactive noble gas from core fission and the model of the gas movement in the waste gas system are analyzed, focusing on the delay state analysis of the isotopes of Kr and Xe creeping through the activated carbon. This paper proposes a design method of delay bed mass calculation based on system flow rate, dynamic adsorption coefficient and the holdup time back calculated from the annual radioactive release of waste gas system. Finally the section and dimensions of delay beds are determined based on the flow velocity recommended in ANSI/ANS 55.4. This calculation method lays the theoretical foundation of the waste gas system design and the dimension design for activated carbon delay bed.

Topics: Delays
Commentary by Dr. Valentin Fuster
2017;():V007T10A018. doi:10.1115/ICONE25-66945.

Corrosion product in the primary coolant is formed mainly by structure material corrosion and activated corrosion product. Under normal condition, source term of corrosion product maintains at a low level while its radioactive concentration reaches a peak value after the primary is injected with hydrogen peroxide during shutdown oxygenation operation. Following the procedure of corrosion product elimination, analysis has been made on effect which accumulated corrosion product exerts on RCV pre-filter and mixed bed demineralizer. Different primary coolant source terms and the cold shutdown peak value of corrosion product have been considered in the analysis. Under the assumption that dissolved and particle dissolved product has a fixed ratio of 1:1, the surface dose rate of RCV pre-filter varies as the different primary coolant varies. Nuclides proportion has been listed and charted based on the calculation. If cold shutdown condition has been taken in primary coolant source term calculation, then the RCV pre-filter surface dose rate will reach 345Sv/h; If primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 42.2Sv/h; If 1/3 primary coolant design value has been taken in source term calculation, then the RCV pre-filter surface dose rate will reach 14.1Sv/h. For the waste resin in RCV mixed bed demineralizer, the corrosion product leaves a small contribution (less than 5%) if the cold shutdown is not considered in primary coolant source term calculation; and a contribution of over 90% if the cold shutdown is considered. The analysis provides a radiation safety analysis basis for NPP solid waste.

Commentary by Dr. Valentin Fuster
2017;():V007T10A019. doi:10.1115/ICONE25-66950.

Advanced Thermal Reactor (ATR) FUGEN is the heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. The commercial operation of FUGEN started on Mar. 1978 and terminated on Mar. 2003 and the decommissioning of FUGEN has been carried out since the decommissioning plan was approved in 2008.

In order to perform the decommissioning work such as dismantling and decontamination safely and reasonably, technology development for the decommissioning has been promoted actively.

This paper describes a part of technology development as follows.

(1) Technology development on reactor dismantling

The reactor of FUGEN is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum which have relatively high activity concentration by operation for 25 years.

With consideration of these characteristics, the reactor will be dismantled under water remotely in order to shield the radiation and prevent dust from migrating from water to air generated by the cutting considering the usage of zirconium alloy which is likely to be oxidized. In addition, laser cutting method whose features are fast cutting speed and less secondary waste in cutting will be applied for reactor dismantling.

However, laser cutting method has no experiences to be applied to dismantlement of reactor facilities. Therefore, laser cutting for actual dismantled objects in air was demonstrated in controlled area in FUGEN using laser cutting system composed of articulated robot and laser cutting head. As a result, safety and applicability of laser cutting system was confirmed. From now on, primary cutting work in air, cutting demonstration with a relatively high dose rate and mock-up test in water for dismantling the actual reactor will be carried out.

(2) Technology development on investigation of contamination

It is necessary to evaluate radioactive inventory in the facilities accurately in order to reflect the evaluated data to dismantling plan appropriately. Therefore, the investigation of the contamination for the facilities has been carried out for safe and reasonable decommissioning work.

The in-situ simple investigation method for the contamination of inner pipes which is mostly dominated by Co-60 is started to develop using the portable NaI(Tl) spectrometer. This method complements conventional investigation method to take samples from the pipes and to analyze them by radiochemical method to figure out the contamination of the whole facility.

Commentary by Dr. Valentin Fuster
2017;():V007T10A020. doi:10.1115/ICONE25-67007.

In this paper, air-immersion, ground deposition, ingestion and inhalation of airborne radioactive effluent released from nuclear power plant under normal operating conditions is studied according to the atmospheric diffusion and ground deposition patterns and parameters that are suitable for the environmental characteristics of the nuclear power plant site, and the public living habits and food chain parameters around the site. Based on the Gaussian plume model, with a radius of 80 kilometers we divide 1, 2, 3, 5, 10, 20, 30, 40,50,60,70,80 km concentric circles around the nuclear power plant site. The 16 compass azimuth axial are the sector center-line, forming a total of 192 sub-regions, atmospheric diffusion of radionuclides is simulated in the assessment area of the region. The annual average atmospheric dispersion factor is calculate by using hourly observation data of wind direction, rainfall and atmospheric stability of the meteorological tower and the ground station, taking into account the ground reflection during transmission, the the decay of the radionuclide, and the loss brought by the wet and dry settling that caused by gravity and rain washing. The airborne radioactive effluent is deposited on the ground or plant surface by dry settling and wet settling in the process of atmospheric environment changing and diffusion. Radioactivity of per unit area brought about by dry settling and rain fall settling is described by the deposition coefficient and deposition speed. The long-term ground deposition factor and ground annual concentration in the evaluation area were calculated under the situation of airborne radioactive effluents in the nuclear power station mixing emission, and the calculated result of radionuclide concentration in the air and soil was compared with the natural background value and the actual monitoring value. Based on the radionuclide deposited on the ground and air through the terrestrial food radioactive transfer mode, together with a large number of environmental surveys data on the population distribution, agriculture, farming, animal husbandry and people’s living and eating habits in the 80km around nuclear station, combing with the actual situation of nuclear power station, the calculation model is amended accordingly. Using reasonable dose mode to calculate the maximum individual and entire public effective dose of the residents in the assessment area, and the results will be compared with other human activities. By comparing the calculated results of radionuclide concentration and radiation dose, it provide quantitative reference information for us understanding the influence of nuclear power station on the surrounding radiation environment, and to meet the requirements of nuclear power plant influence on surrounding environment and people under normal operating conditions.

Commentary by Dr. Valentin Fuster
2017;():V007T10A021. doi:10.1115/ICONE25-67059.

With the introduction of third generation of nuclear power AP1000, Westinghouse uses the mobile device (a mobile wastewater treatment device 6 units shared) radioactive waste system design concepts. This design not only simplifies the process of nuclear island waste system; saves equipment layout space; improves equipment utilization; while increases the use of new technologies lifetime of the plant and the possibility of flexibility. This paper introduces the first AP1000 unit (Sanmen, Zhejiang Province) by using the advanced mobile device technology and application of wastewater treatment under the condition of the primary coolant source level. At the same time, the paper also discusses the periodic system inspection and the strategy of maintenance. In addition, the paper further expands the application direction of the mobile waste processing aspects, such as: decommissioning of nuclear facilities; enhancing the facility decommissioning radioactive liquid waste purification capability. Another example: After the Fukushima accident, people pay more attention to accident-mitigation-design and hope to accelerate the development of emergency radioactive liquid waste processing devices. Thus, in addition to strengthening the nuclear power plant inherent defense in depth and resistance emergency capability, mobile waste treatment device or combination device special regional settings can be made to improve and enhance the ability to get more diversified emergency response.

Commentary by Dr. Valentin Fuster
2017;():V007T10A022. doi:10.1115/ICONE25-67143.

Uranium hexafluoride is the intermediate material of uranium fuel enrichment process, which is widely used in uranium conversion plant, uranium enrichment plant and nuclear fuel element plant[1]. Because of its active chemical properties and its radioactive and chemical toxicity, great importance should be attached to the uranium hexafluoride release accident. This paper describes the possible leakage scenarios for uranium hexafluoride accident. And the general step of the evaluation for uranium hexafluoride leakage accident release source term is given, as well as an application example for the feed facility of a gaseous diffusion plant.

Commentary by Dr. Valentin Fuster
2017;():V007T10A023. doi:10.1115/ICONE25-67154.

Excessive gamma rays will be emitted when a nuclear power plant is under the refueling overhaul, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding material with a “sandwich structure” was developed. This kind of material has been adopted in a mega-kilowatt pressurized water reactor (PWR) nuclear power plant as the first demonstration project. Based on the characteristics of the new flexible shielding material, the process design program of the product application was founded, and the new material was compared with the traditional lead shielding material. At last, the installation and removal time together with the shielding effect were analyzed according to the first demonstration application.

Commentary by Dr. Valentin Fuster
2017;():V007T10A024. doi:10.1115/ICONE25-67283.

MSR (Molten salt reactor) is a reactor with the fission material dissolved in the fluoride salt as the fuel, which continuously circulates through the primary loop. The liquid fuel makes the online refueling and reprocessing be possible, consequently, more fuel cycle options would be presented due to the introduction of various refueling fuel type and reprocessing mode (continuous / batch reprocessing) comparing with that of the solid fuel reactor systems. It is important to evaluate all the possible fuel cycle options and screen out the promising ones to narrow the R&D activities of TMSR (Thorium-based Molten-Salt Reactor nuclear energy system) program. In this study, we firstly established a screening and decision-making framework, then conducted an initial evaluation work and identified the potential promising fuel cycle options. Synthesizing the goal of TMSR program and technology readiness, we proposed a “Three-steps” fuel cycle development strategy with the aim for gradually increasing the thorium resource utilization while considering the challenges of the reprocessing technology.

Commentary by Dr. Valentin Fuster
2017;():V007T10A025. doi:10.1115/ICONE25-67358.

In order to meet the sustainable development demand for energy, developing nuclear power actively has become an important means for the country to improve energy supply pattern and change the energy structure. The normal operation of the nuclear power station brings enormous economic and social benefits, but also accompanied by airborne radioactive effluent and liquid radioactive effluent discharge. In order to estimate the effect of radioactive effluent release during the normal operation of nuclear power plant on radiation dose of the environment and the public, in this article, domain knowledge analysis and domain knowledge modeling will be carried out on the theoretical model of evaluation calculation. Then, combined with actual business logic, advanced software development techniques and mature design ideas will be adopted to realize the field component of radiation dose evaluation in nuclear power plant. Establish the corresponding domain component library, and thus to achieve the building of radiation dose evaluation software system.

The main content of this paper will be divided into two parts to elaborate, First part is the domain knowledge analysis and modeling of radiation dose evaluation model of nuclear power plant radiation environment, the other part is based on radiation dose evaluation component library software system researching, designing and achieving.

In the domain knowledge analysis of the radiation dose estimation model, due to the mutual independence of the airborne radioactive effluent and the liquid radioactive effluent in the evaluation mode, this paper will analyze and model the domain knowledge separately. The airborne radioactive effluent is divided into four parts: air immersion external exposure, ground deposition external irradiation, ingestion inhalation and inhalation in air. The main contents of the analysis include atmospheric diffusion suitable for environmental characteristics of nuclear power plant site, ground deposition factors, food chain data, and lifestyle habits around the site. In the field of knowledge analysis of liquid radioactive effluent, there are four ways: external marine activity irradiation, coastal deposit sediment irradiation, and irradiation of seafood. The analysis mainly includes dilution and diffusion conditions in the surrounding sea area of nuclear power plant, the radionuclide in the seawater of the receiving water body, the shore sedimentation factor and the transfer model of the radionuclide in the seafood.

Based on the detailed analysis and research on the radiation dose evaluation of nuclear power plant, and the designing of the domain model by adopting mature software design idea and the advanced software development technology, the domain component of the radiation dose evaluation field is realized, and build the corresponding domain component library. This can provide a reliable and usable domain component library for the radiation dose evaluation of each nuclear power plant, improve the maintainability of the radiation dose evaluation system of the nuclear power plant, the comprehensibility of business logic, and the reusability of the evaluate module, so as to meet the calculation demands on radiation dose effects of public caused by radionuclide release in normal operation of the nuclear power plant.

Commentary by Dr. Valentin Fuster
2017;():V007T10A026. doi:10.1115/ICONE25-67372.

Tomographic gamma scanning (TGS) method is one of the most advanced non-destructive assay (NDA) methods. TGS method can determine quantitatively with high accuracy transuranic nuclides in heterogeneously distributed media with medium- and high-density, and is thus widely used to assay the location and quantity of selected radioisotopes in scraps and wastes within sealed containers. In this paper, a prototype of tomographic gamma scanner which we designed is introduced. The TGS system is composed with four parts: external source with front collimator, radioactive material drum turning table, HPGe γ detector assembly including back collimator and cooling system, and computer. Successful implementation of the work has broken through the difficult problems or restraints to the development and applications of TGS, it will be applied widely to the non-destructive assay of nuclear materials within sealed container in the nuclear safeguards, radwaste measurement and arms control fields.

Commentary by Dr. Valentin Fuster
2017;():V007T10A027. doi:10.1115/ICONE25-67382.

A novel type of flexible composite with tungsten powder filled in polymers (TPFP) homogeneously was developed as an alternative to lead (Pb) based radiation shields used in nuclear industries. TPFP had a density in the range of 4 to 11.3 g/cm3, which can be tailor-made according to the applications. In addition to the advantage of lower toxicity over Pb-based shielding, TPFP can be formed into various shapes, such as pipe shields, pipe wraps, safe floor shields, blankets, etc. The mechanical properties and attenuation of γ-ray was investigated for the developed TPFP.

Commentary by Dr. Valentin Fuster
2017;():V007T10A028. doi:10.1115/ICONE25-67522.

It’s necessary to establish a high-precision analytical method of Gas Chromatography (GC) for achieving the rapid detection of trace hydrogen and methane in cycling process gas of fusion reactors. The analysis of H2, CH4 with concentrations from 0.5ppm to 5000ppm were detected through the application of Discharge Ionization Detector (DID) in high purity helium and the column of molecular sieve, whose parameters was 15m×0.53mm×50μm.

The studies have shown that the retention time of H2, CH4 were 36s and 71s, respectively, their relatively standard deviation (RSD) of area responses were severally below 1.0% and 0.5%, which illustrated good repeatability. Separately, the detection limit (DL) were no more than 10ppb and 2ppb of H2 and CH4 with high column efficiency, while the column temperature was 40°C and the flow rate was 15mL/min. Meanwhile their values of R2 were all larger than 0.999, which included a better linearity relationship between area response and concentration. This method has already applied to the on-line measurement of molecular sieve adsorption, the analysis of electrochemical hydrogen pump and methane decomposition, etc.

Commentary by Dr. Valentin Fuster
2017;():V007T10A029. doi:10.1115/ICONE25-67554.

Since Westinghouse Savannah River Company (WSRC) of America first applied PUREX process in 1954, PUREX process is always the top priority in nuclear fuel reprocessing plant. And this process is based on liquid to liquid extraction with TBP as the extractant. TBP is irreplaceable in the development of PUREX process in nuclear fuel reprocessing, its advantages are well recognized. However TBP does have some disadvantages such as formation of red oil, which will appear in the system of high nitric acid concentration and heavy metal nitrate, once the red oil forms, it can lead a exothermic runaway decomposition in reasonable conditions, such as exceeding a certain temperature (typically 130°C) or high acid concentration. If gas products and energy released from the decomposition reaction could not be exported in time, it will lead to vessel overpressure and caused violent explosion accidents. By now, it has happened 6 times so-called red oil explosion accidents worldwide, resulting in different degrees of equipment and construction damage and environmental contamination. From 1953 to now, research related to red oil has never stopped. WSRC, Hanford Company, Oak Ridge National Laboratory and Los Alamos National Laboratory of America have conducted many studies, as well as some research institutions from Russia, UK, France and India. Defense Nuclear Facilities Safety Board of America issued a technical report in 2003, preventive measures for red oil explosion were established in this report, and these measures provided good practice experience and reference for other countries, and the temperature condition (⩽130°C)and nitric acid concentration (⩽10M)for preventing red oil explosion are employed in some countries which has built the reprocessing plant. Nevertheless, research conclusions and knowledge of red oil vary from country to country. Especially, Kumar and Smitha etc. conducted several experiments in adiabatic condition in recent years, and investigation on stability of TBP - nitric system was made, the results indicated that the red oil runway reaction will happen even in lower temperature and lower nitric acid concentration in contrast with the reported value, and they thought it would need a further study to assess the validity of present preventive measures, and to rebuild the safety limits for preventing red oil explosion in the operation of nuclear fuel reprocessing plants. In this paper, related research results of red oil explosion accidents were combed, and the characters of study work of different periods were summarized, and definition, formation conditions of red oil, pathway of runaway reaction, control and preventive measures for preventing red oil explosion of different countries were analyzed and compared, as well as the new viewpoints of recent literatures. And some research ideas for future investigation based on present work were also proposed.

Commentary by Dr. Valentin Fuster
2017;():V007T10A030. doi:10.1115/ICONE25-67626.

The new licensing standards were further improved by taking into account of lessons learned from the Fukushima-Daiichi nuclear accident, and countermeasures against severe accidents were newly required as regulatory items, where severe accidents were defined as serious accidents that occur under conditions exceeding design bases. Organic solvent fire in cell was defined as one of the severe accidents in nuclear fuel reprocessing facilities, which should be investigated, in order to establish methods for evaluating effectiveness of the countermeasures.

One of the combustibles in the fire accident at reprocessing facilities is the organic solvent composed of 30% tributyl phosphate (TBP) and 70% dodecane. When the solvent burns, aerosol of soot and radioactive substances are released inside the facility. The aerosol causes a clogging of high-efficiency particulate air filters (HEPA filters) in a ventilation system of the facility, which increases a differential pressure of the filters.

We have performed combustion tests simulating the fire accident. As one of interesting results of the tests, we observed, when most of dodecane in the solvent was burned out, a rapid increase in a differential pressure of a HEPA filter, which may cause its rupture. We also found a small amount of RuO4 release from the burning solvent, which can pass through HEPA filters due to its volatility. These phenomena should be adapted in the effectiveness evaluations of the countermeasures against the fire accident.

Commentary by Dr. Valentin Fuster
2017;():V007T10A031. doi:10.1115/ICONE25-67665.

The technology of radioactive liquid wastes containing boron is a hot spot of radioactive waste disposal. The radioactive borate waste is solidified by conventional cement solidification in our country. In order to dispose the radioactive boron-containing waste more securely and efficiently, this work focuses on the development of high-efficiency cement solidification. In this work, the borate in radioactive liquid wastes containing boron is changed into polyborate and solidified by high efficiency formula to increase the waste containing rate. The high efficiency cement solidification formula was gotten in lab, its indicators are as follows: The containment of boron is 49.5%;The fluidity is 375mm;The compressive strength is 22MPa;The resistance of leachability (Sr, Cs,Co) is better than the requirement of national standard.

Topics: Solidification , Boron
Commentary by Dr. Valentin Fuster
2017;():V007T10A032. doi:10.1115/ICONE25-67673.

For nuclear fuel reprocessing plants, some materials with high activity have been collected in the product container in the process of spent fuel reprocessing. In order to ensure the staff’s occupational radiation safety, an analysis of product container radiation shielding and source terms is crucial to determine the total activity limit and the activity limit range of different rays while reprocessing radioactive materials in the product container, so as to determine the working time limitation. A basic calculation result shows that generally alpha particles are not taken into consideration when calculating its occupational radiation effect for product container. In order to reduce the radiation exposure caused by neutron and gamma ray, radiation nuclides such as actinides and activated products which have a high neutron and gamma yield, and also the total nuclide amount should be controlled. In addition, the purification of nuclide emitting pure beta ray such as 90Sr, 90Y, 106Ru, 106Rh should be strictly supervised in order to restrict the beta ray exposure and its bremsstrahlung effect.

Commentary by Dr. Valentin Fuster
2017;():V007T10A033. doi:10.1115/ICONE25-67717.

The first hot test have been completed at China Reprocessing and Radiochemistry Laboratory in September to December 2015. According to the technology process and radiation protection requires, the radioactive aerosol monitoring and airborne radioactive effluence monitoring was taken in workplace. During the hot test, the maximum concentration of α and β radioactive aerosol in controlled area I was 9.66Bq/m3 and 1.33×103Bq/m3, and the concentration of α and β radioactive aerosol in supervised area and controlled area II was under the authorized limits, which was 1.0–6.0×10−3Bq/m3 and 1.0–7.0×10−2Bq/m3. The leakage of radioactive aerosol from sealing chamber was not found. The concentration of airborne radioactive effluence in supervised area reached the discharge standard, which was α: under 7.0×10−3Bq/m3 and β: under 3.0×10−2Bq/m3. For the dissolution process, the concentration of 85Kr and 129I was significant parameters to monitored.

Topics: Aerosols , China
Commentary by Dr. Valentin Fuster
2017;():V007T10A034. doi:10.1115/ICONE25-67719.

With the development of information and computer technology, the Digital Instrumentation and Control (I&C) System has been widely used in nuclear power plants, which leads the tendency of NPPS’ construction and rebuilding on digital I&C system. As an approximate approach, conventional fault tree approach has been used quite often in the analysis of nuclear power plants’ Probability Safety Assessment (PSA), which combine with system components’ failure modes in order to modeling the digital system’s failure. However, for the reason that conventional fault tree approach has a great disadvantage on analyzing the reliability of digital I&C system, which may not be able to fully describe the dynamic behavior of digital I&C system with significant hardware/software/human action process interaction, multi-failure modes and logic loops, it cannot carry on effective modeling and evaluation of digital I&C system. Therefore it is necessary to establish some dynamic approaches to modeling digital I&C system. As a new probability safety analysis method, Dynamic Flowgraph Methodology (DFM) can model the relationship between time sequence and system variables because of its dynamic property. Therefore, DFM can be used to analyze the impact of software failure, hardware failure and external environment, which are closely related to the reliability of the whole system. In the first place, this paper introduces the theoretical basis, model elements and the modeling procedures of DFM and demonstrates how Dynamic Flowgraph Methodology (DFM) can be applied to Reactor Protection System with interactions between hardware/software and physical properties of a controlled process. Meanwhile, in this case, DFM and fault tree methodologies are both used to conduct the PSA for the same top event by calculating the probability of it and finding out the prime implicants of DFM and minimal cutsets of conventional fault tree. During the process of analysis, we mainly evaluate the reliability of reactor trip function of Reactor Protection System (RPS) by using DFM and conventional fault tree approach and mainly focus on modeling the four-way-redundant voting logic and the reactor trip breaker logic. Finally, through the comparison of this two methods and model results, it is concluded that there is a distinct advantage of DFM over conventional fault tree approach by using multi-logic to fully display the fault mode and utilizing decision table to describe the interaction between software and hardware. In general, conclusion can be drawn that, as a dynamic approach, Dynamic Flowgraph Methodology could be more accuracy and effective than conventional fault tree approach in analysis, ensuring the reliability and safety of the whole digital I&C system.

Commentary by Dr. Valentin Fuster
2017;():V007T10A035. doi:10.1115/ICONE25-67728.

Single event effect occurs when a single energetic particle penetrates sensitive nodes and deposits enough charge by ionization in semiconductor devices. It has become a major reliability concern for spaceflight. Single event effect characteristics analysis methods based on simulation are presented for typical circuit elements in spacecraft power systems. The failure mechanism and impact factors of single event burnout in trench power MOSFETs are investigated through numerical simulation. The broadening, quenching and capture of single event transients in combinational logics are analyzed by circuit simulation based on a coupled single event transient injection method. A behavioral modeling is introduced to predict single event effect sensitivity of the parallel to serial conversion circuit. The results show that ion-induced holes can spread efficiently to turn on the parasitic bipolar junction transistor and result in stronger carrier multiplication when the ion strikes at the center of the gate region in trench power MOSFETs. Increasing the depth of P+ plugs, decreasing the doping concentration of source regions and using lower drain bias voltages can mitigate single event burnout susceptibility. Using the same load capacitance for each stage inverter and selecting a suitable gate-width ratio are recommended to prevent single event transient broadening in inverter chains. Moreover, the reset pin of D flip-flops in the parallel to serial conversion circuit should be hardened by design according to the behavioral modeling results.

Commentary by Dr. Valentin Fuster
2017;():V007T10A036. doi:10.1115/ICONE25-67740.

A series of nanofiltration membranes were prepared by interfacial polymerization of piperazine and terephthaloyl chloride on the surface of polyacrylonitrile (PAN) ultrafiltration membranes. ZnO nanoparticles were incorporated in the active separation layer to modify the performances of the membranes. The preparation conditions as the monomer concentration, dosage of nano-ZnO particles and the reaction time on removal of a simulated radioactive nuclide Co (II) were investigated.

Fourier transform infrared in attenuated total reflection mode verified the formation of polyamide on the PAN ultrafiltration membrane. The scanning electron microscope images showed that the nano-ZnO particles can homogeneously fixed on the membrane surface. The retention of Co (II) increased with increasing the dosage of nano-ZnO in the range of 0∼0.03 g. Further adding more nano-ZnO, the rejection rate of Co (II) first decreased and then increased. The concentration of piperazine and terephthaloyl chloride showed similar effect on removal of Co (II) ion. 5 minutes polymerization time was sufficient to form an active separation layer on the substrate membrane which changed the separation mechanism from ultrafiltration to nanofiltration. The separation performance of NF3 prepared by the following conditions was optimum: 0.03g nano-ZnO, 0.6 wt% piperazine, 0.5 wt% terephthaloyl chloride, and the reaction time was 15 min. The rejection rates of 1000 mg/L Na2SO4 and Co2+ in CoCl2 solution were 90% and 75% respectively. The Co (II) removal rate can be increased to nearly 90% by using ethylenediaminetetraacetic acid disodium salt. Increasing the operation pressure or the feeding concentration of Co (II) can also improve the performances of the membranes in this experiment.

Topics: Membranes
Commentary by Dr. Valentin Fuster
2017;():V007T10A037. doi:10.1115/ICONE25-67742.

The main control room (MCR) ventilation system has been designed to maintain habitability of the control room envelope both under normal condition and accident condition. The system adopting dual air intakes adds one more air intake for accidents at suitable position. During accidents, the air intake with lower contamination will be selected and the other with higher contamination will be isolated, to reduce the amount of radioactive substances entering MCR extremely and enhance the habitability of MCR envelop. This paper is devoted to research on the impact of switching time interval (STI) for dual intakes on workers in main control room during accidents. As the contamination condition varies, the switching action will be happened. Switching time interval (STI) referred in this paper means the time between two switching actions. When accidents occur, the air intake will operate and switch between two intakes automatically. The action of switching will be influenced by several parameters: the meteorological conditions of the site, the response features of the monitoring instruments and the source term released to the environment after accidents. Analysis of these parameters and their sensitivity analysis are performed, which show that the ventilation system cannot afford too frequent switching actions resulted from instantaneous sudden changes of intake’s activity. That’s the reason why it is necessary to set a minimum STI which means the contamination of one intake have to be lower than the other intake and this dominant position should be kept longer than the minimum STI, if not, the switching action will not be happened. As it is essential to set a minimum STI to prevent frequent switching of system, the analysis of its impact on the atmospheric relative concentrations and the doses of the workers in main control room are performed on basis of specific site meteorological condition and the response characteristic of dose monitoring instruments. Three kinds of accident release conditions are considered, which are relief valve release, containment leakage and elevated funnel release. The atmospheric relative concentrations and the doses of the workers in MCR are evaluated for every case and compared with the dose limits. Finally an acceptable minimum STI of dual air intakes is recommended.

Commentary by Dr. Valentin Fuster
2017;():V007T10A038. doi:10.1115/ICONE25-67807.

High level waste (HLW) originating from reprocessing of spent fuel commercial power reactors contains more than 40 different elements. Vitrification into borosilicate glass at 1100 ∼1200°C is the process of choice. It is routinely used to immobilize the radioactive waste constituents in a chemically stable matrix for a final geological disposal.

Melting process for commercial HLW glasses are variants of two basic designs. (1) The joule-heated ceramic-lined melter (JHCM) originally developed in the United States in 1973 and used in several nuclear sites in the world. (2) The hot-walled induction melter (HWIM), developed in France starting in 1962 and used in France and UK. These technologies, while effective, do pose limitations in waste form compositions and throughput rates. Particularly HLW originating from commercial spent fuel reprocessing usually contains noble metals elements such as ruthenium (Ru), rhodium (Rh), and palladium (Pd) which require special attention when this waste is vitrified.

Recent advances to both of these baseline technologies are beginning to be used with large gains to ensure waste form flexibility, throughputs, and noble metals compatibility. The next generation JHCMs use a steeply sloped bottom and a subsidiary-heating bottom drain to allow these noble metals particles to be effectively flushed from the melter with higher waste loadings. Similar melters are being installed near Guangyuan/Sichuan province, China by German consortium team and being developed for the second K-Facility melter at Rokkasho by Japan Nuclear Fuel Limited (JNFL). As another example the advanced JHCMs will be installed in the Hanford WTP project having large glass pool surface area with rapid bubbling. Significant improvements on induction melters have also been implemented. AREVA recently installed a cold-crucible induction melter (CCIM) in combination with a rotary calciner at La Hague in France. This melter uses radio frequency induction to power the glass melt itself and water cooling of the outer surface maintains a frozen glass shell (skull) as the glass contact material. Because no permanent refractories or embedded electrodes are used, this design allows for high-temperature operation and can tolerate more corrosive melts, and uses a water-cooled, motor-driven mechanical stirrer to comply with noble metals behavior.

This paper highlights some of these advances and suggests potential advantages and disadvantages of these next generation melter technologies comparing advanced JHCM with updated CCIM. In conclusion, these melters have made the technologies of choice for new HLW vitrification projects around the world.

Commentary by Dr. Valentin Fuster
2017;():V007T10A039. doi:10.1115/ICONE25-67831.

The cavity streaming is the neutron beam from the reactor core through the tunnel, which is between the external surface of the pressure vessel and the shield inner surface.

Reactor cavity streaming calculation is a typical deep penetration problem with complex geometry. The accurate calculation of neutron radiation streaming is a key problem to the reactor shielding calculation, for which the Monte Carlo method and the discrete ordinate method are two popular methods. The speed of discrete ordinate method calculation is fast, but it is hard to describe the complex pile of cavity; the Monte Carlo method can accurately describe the complex geometry, it has a high calculation precision, but with a low direct simulation efficiency. Based on a pressurized water reactor nuclear power plant, this paper presents a detailed model realized by Monte Carlo code, with continuous energy points cross section libraries. The neutron flux density distribution of PWR reactor cavity streaming can directly be calculated by a three-dimensional simulation. For such an actual deep penetration problem, a variety of variance reduction techniques are studied, an effective variance reduction technique is used to obtain results with small statistic errors for a Monte Carlo simulation, which effectively solves the problem of large-scale deep penetrating convergence difficulty, the cavity radiation streaming calculation and analysis are completed. The result shows that the Monte Carlo method can be used as a powerful tool to solve the problem of cavity streaming leakage.

Commentary by Dr. Valentin Fuster
2017;():V007T10A040. doi:10.1115/ICONE25-67832.

Fukushima nuclear accident has aroused concern about ionizing radiation damage to the marine environment. It is subjective to assess the accident’s effects on the marine ecological environment during different scholars, because of different understanding of nuclide release quantity, dose estimation models and parameters. To solve this problem, a gradually progressive research approach is designed based on a coastal nuclear power plant in study, which is “from influencing factors analyzing to biological dose assessment, and to uncertainty analysis”. First, the factors affecting the biological dose assessment will be analyzed, and then the concerns of the various factors and their impact on the results of the assessment will be discussed. Second, the biological dose will be assessed based on appropriate dose mode after selecting representative species and analyzing critical exposure pathway. Finally, the uncertainty of radionuclide release quantity, dilution factor, concentration factor, dose conversion factor will be analyzed. The study will provide reliable scientific bases to identify the factors impact biological dose assessment effectively, and improve the accuracy of the dose evaluation.

Commentary by Dr. Valentin Fuster
2017;():V007T10A041. doi:10.1115/ICONE25-67859.

According with the requirements of HAF-102[1], the purpose of radiation protection optimization is to achieve occupational exposures as low as reasonably achievable (ALARA). More than 80% of the collective exposure doses are received during the outage for PWRs. Activated corrosion products, especially Co-60, deposited on the surface of reactor coolant system are the main causes of these doses. Co-60 is the principle contributor to out-of-core radiation fields in PWRs, which is a high energy gamma emitter with a 5.3 year half-life period. The contamination by Co-60 plates out in plant piping systems and produces long term high radiation areas in the plant. Stellite, a hard facing alloy trade, which has high Co-59 content (approximately 60%), is identified as one of the most important source of Co-60 in reactors, whereas the contribution of Stellite to cobalt are always not paid enough attention to. Cobalt is released through ware and corrosion of Stellite. When cobalt is released in systems with a flow path to the reactor, it can become activated by neutron to Co-60.

This paper analyzes the Stellite material release into the primary coolant. Based on the above considerations, the impact of Stellite replacement on plant contamination by Co-60 is calculated and analyzed with the use of activated corrosion product source term calculation program. The primary conclusions are as follows: (1) Reducing the quantity of Stellite material used during the design and building of PWRs can significantly reduce the Co-60 deposition on primary circuit. (2) The replacement of Stellite material has a positive impact on PWR contamination by Co-60, but the outcome is not so significant. Based on the above analysis, recommendations are made. Stellite material replacement project can be used for the operating reactors to reduce the Co-60 contamination. But for the design and building of new reactors, reduction of the use of Stellite material is the best choice.

Commentary by Dr. Valentin Fuster
2017;():V007T10A042. doi:10.1115/ICONE25-67899.

In this study, a simulated ammonium diuranate filtrate (ADUF) which was obtained by adding uranyl nitrate to a 35 g/L ammonium nitrate solution to adjust the uranium concentration to about 50 mg/L was treated by a nanofiltration process. Experiments were carried out on a plate membrane testing device with a trans-membrane pressure (TMP) range of 0.5 ∼ 3.0 MPa, a crossflow velocity range of 10 ∼ 50 cm/s and a temperature range of 5 ∼ 35 °C. The results show that NF270 membrane has good rejection property for uranium and excellent permeability for ammonium nitrate. The ammonium nitrate concentration in the permeate is about 32 g/L which means the reject ratio of ammonium nitrate is only about 10%. Though NF270 membrane shows good uranium rejection property, the corresponding permeate flux is very high. When the trans-membrane pressure is 1.5 MPa, the uranium reject ratio is 96.8% and the the corresponding permeate flux is about 80 L/(m2·h). It indicates a bright application prospect of nanofiltration process in the treatment of ADUF.

Topics: Uranium
Commentary by Dr. Valentin Fuster
2017;():V007T10A043. doi:10.1115/ICONE25-67917.

The active magnesium oxide (AMO) was synthesized by homogeneous precipitation method with microwave and characterized by scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), X-ray diffraction (XRD) and infrared spectroscopy (FTIR). Adsorption of Sr(II) by AMO was investigated under the effect of AMO dosages, pH of solution, temperature and contact time and analyzed the kinetics and thermodynamics characteristics. The results showed that AMO has very good adsorption capacity on Sr(II) in aqueous solution,When pH of solution is 8.0, the solid-liquid ratio is 0.25 g·L−1, initial Sr(II) concentration is 50mg·L−1, the contact time is 80 min at 298K, the removal rate and adsorption capacity reached 98.29% and 187.5 mg·g−1, respectively. Kinetic and thermodynamic results indicate that adsorption behavior of Sr(II) by AMO fitted well with pseudo-second-order model and the Freundlich isothermal model. Adsorption thermodynamic parameters showed that the process of adsorption is spontaneous and endothermic.

Commentary by Dr. Valentin Fuster
2017;():V007T10A044. doi:10.1115/ICONE25-67922.

The activated MgO was synthesized by microwave homo-precipitator method and characterized by SEM, EDS and FT-IR methods. It was used to adsorption of U(VI) from aqueous solution with batch system. The paper discussed the effect of pH, temperature, contact time, adsorbent dose and initial U(VI) concentration on the adsorption. The results showed that activated MgO has good adsorption capacity for U(VI), the removal rate and equilibrium adsorption capacity reached 83.5% and 84.04mg·g−1 at pH 5.0, 15mg dose and 313K,respectively. The adsorption kinetics of U(VI) onto activated MgO were better fitted with pseudo-second-order kinetic.The adsorption isotherm data were fitted well to Freundlich isotherm model.The thermodynamic parameters showed that the adsorption process is endothermic and spontaneous.

Commentary by Dr. Valentin Fuster
2017;():V007T10A045. doi:10.1115/ICONE25-67926.

In order to ensure the thermal safety of the spent fuel stored in an underground vertical shaft, an air current heat dissipation is simulated in CFD way using ANSYS FLUENT code. Forced convection heat dissipation is focused in the research. The air current in the shaft and the temperature distribution on the surface of the spent fuel canister are calculated. The result confirms the reliability and security of the spent fuel dry storage. Finally, based on the calculating result, a support structure is designed, and the storage position of the spent fuel canister in vertical shaft is discussed, to optimize the decay heat removal of the spent fuel, and to ensure the temperature measuring point is set in a reasonable position.

Commentary by Dr. Valentin Fuster
2017;():V007T10A046. doi:10.1115/ICONE25-67960.

In the process of decommissioning of the nuclear facilities, abandoned and contaminated equipments or devices often need to be dismantled and cut into proper pieces in order to facilitate subsequent treatment and disposal. Since the nuclear facilities were places where radioactive operations were frequently executed, gloveboxes should be such typical abandoned and contaminated devices. Usually, gloveboxes were cut into proper pieces by kinds of tools which was chosen depending on the thickness of stainless steel from different parts of gloveboxes. This traditional cutting was laborious and high concentration of harmful aerosols and gases would be created during the cutting operation. In order to develop more advanced cutting and disintegration ways, a cutting and disintegration device was developed using abandoned and contaminated gloveboxes as operation objects in this work. During the design of this device, the operation convenience, operation exactitude and the protection for operators were fully considered. Also, the hot verification test was carried out. Based on the verification test results, the cutting and disintegration device was reliable and could meet the design requirements, which precisely executed various movements required in the hot verification test. Due to the application of remote operation and advanced cold cutting as the main cutting way in the development of this cutting and disintegration device, radioactive aerosols and harmful gases created during operation of this device were obviously declined compared to past cutting devices, which was of great importance to the health of operators. This work can provide technical support for the development of other similar devices applied in nuclear facility decommissioning.

Commentary by Dr. Valentin Fuster
2017;():V007T10A047. doi:10.1115/ICONE25-67965.

In 2015, Office of Nuclear Energy of the U.S. Department of Energy established a research and development project at Pacific Northwest National Laboratory to evaluate the process control capability to produce a specific uranium/plutonium product from used commercial nuclear fuel using an aqueous co-decontamination separations process. The process is controlled using on-line instrumentation supported by a dynamic process model. The new program is called the CoDCon project.

The result of the study will be a quantitative measure of the current capability to produce a specific U/Pu product, using U-IV to reduce plutonium to Pu-III and prepare a mixed product, the composition predicted by a dynamic model, measured with on-line instrumentation and controlled by the adjustment of process variables. This approach would be an alternative to the use of the PUREX aqueous separations process that produces separate plutonium and uranium products that are later blended to prepare the desired mixture. Since plutonium is always accompanied by uranium, the project will provide a safeguards-by-design tool for possible use in future commercial separations plant designs. The effectiveness of the design tool will be quantified using a methodology published by B. Cipiti, et al1.

The paper describes the process and the control system used by the project and provides details on the current status of the research and development program. Since this advanced process control system may have international applications, arrangements will be described for possible foreign participation in the project.

Commentary by Dr. Valentin Fuster
2017;():V007T10A048. doi:10.1115/ICONE25-67973.

A silica-based ammonium molybdophosphate (AMP/SiO2) was successfully applied to remove Cs(I) from radioactive waste water in our previous study. The structure characterization of irradiated AMP/SiO2 was carried out by X-ray diffraction (XRD). The effect of the initial concentration of metal ions and acidity concentration for removal Cs(I) by irradiated AMP/SiO2 were investigated. No obvious difference was observed for AMP/SiO2 with 300 kGy absorbed dose from the XRD patterns, except the slightly shift of (110) peak toward a small angle. Relatively large distribution coefficient (8×103cm3/g) of Cs(I) was obtained in 3M HNO3 by irradiated AMP/SiO2. The maximum adsorption capacity of AMP/SiO2 irradiated at 300 kGy was estimated as 22.4 mg/g. The loss of molybdenum from AMP/SiO2 in 3 M HNO3 with 200 kGy absorbed dose is less than 5%.

Commentary by Dr. Valentin Fuster
2017;():V007T10A049. doi:10.1115/ICONE25-68003.

AREVA is the only global player in this D&D and Waste Management domain which can propose its huge feedback in Europe and worldwide for all kind of nuclear facilities (from mining, chemical, fuel and recycling plants, Research Reactors to Nuclear Power Plants to legacy waste retrieval).

To be owner and operator of sites in dismantling phases with a lot of waste to retrieve, treat and package for final storage, gives AREVA a global view=on important key parameters of this kind of projects for more than 40 years.

Specificities of D&D and associated Waste projects are multiple:

Complexity of operations to perform in buildings not designed for these tasks.

Big components to decontaminate and to cut.

Innovation, skills and Research and Development needs to solve and anticipate projects issues.

Safety and security for workers and populations.

Environmental issues to solve (contaminated soil to sort for example, or water to decontaminate).

Risks management during all project life using lessons learnt.

Costs optimization from the most important project phase: scenario and studies of the operations.

• Amount and type of waste (alpha with contamination, gamma with irradiation, high volume of very low level waste with or without free release opportunities).

• Interfaces with waste transports and storage facilities.

• ...

Commentary by Dr. Valentin Fuster
2017;():V007T10A050. doi:10.1115/ICONE25-68004.

AREVA has been running since decades nuclear reprocessing and recycling installations in France. Several industrial facilities have been built and used to this aim across the time.

Following those decades and with the more and more precise monitoring of the impact of those installations, precise data and lessons-learned have been collected that can be used for the stakeholders of potential new facilities.

China has expressed strong interest in building such facilities. As a matter of fact, the issue of accumulation of spent fuel is becoming serious in China and jeopardizes the operation of several nuclear power plants, through the running out of space of storage pools. Tomorrow, with the extremely high pace of nuclear development of China, accumulation of spent fuel will be unbearable.

Building reprocessing and recycling installations takes time. A decision has to be taken so as to enable the responsible development of nuclear in China. Without a solution for the back end of its nuclear fuel cycle, the development of nuclear energy will face a wall.

This is what the Chinese central government, through the action of its industrial CNNC, has well understood. Several years of negotiations have been held with AREVA. Everybody in the sector seems now convinced.

However, now that the negotiation is coming to an end, an effort should be done towards all the stakeholders, sharing actual information from France’s reference facilities on: safety, security, mitigation measures for health protection (of the workers, of the public), mitigation measures for the protection of the environment.

Most of this information is public, as France has since years promulgated a law on Nuclear transparency. China is also in need for more transparency, yet lacks means to access this public information, often in French language, so let’s open our books!

Topics: Recycling , China
Commentary by Dr. Valentin Fuster

Mitigation Strategies for Beyond Design Basis Events

2017;():V007T11A001. doi:10.1115/ICONE25-66015.

Even after 6 years since the accident, the exact accident progression for each unit and location of core debris has not been clarified. Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside of the Primary Containment Vessels (PCV). In spite of their effort, the observed environmental data do not support the existence of a large radiation heat sources attributable to the molten core at the bottom of the Primary Containments.

Under this situation the author has conducted a forensic engineering study (i.e., different fields of science work together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident. Through this study the author found that the environmental contamination and public exposure could have been substantially mitigated should the following vulnerabilities have been removed before the accident:

(1) The threat of hydrogen generation through “radiation-induced electrolysis”.

(2) Potential threat of “internal hydrogen explosion” in the suppression pools. The atmosphere on top of the suppression pool water (cover gas) should have been nitrogen.

(3) Potential threat of “internal hydrogen explosion” in pipes which had occurred in the case of the Hamaoka Unit 1 accident.

(4) The leak rate of PCV should have been testable at its design pressure. Intrinsic safety factor of the containment flanges against effluent leakage should have been rated as a 3 for functional integrity of the PCV.

(5) Spread of hydrogen gas from vent lines through duct networks connected to SGTS. The hardened vent line should be independent and provided with filters for Dry Well venting. No rupture disks should have been installed in the vent lines.

Commentary by Dr. Valentin Fuster
2017;():V007T11A002. doi:10.1115/ICONE25-66090.

It is more and more concerned after Fukushima Daiichi about spent fuel pool severe accident management, and it is also required by the regulatory body of NSC (Nuclear and Radiation Safety Center). During such a beyond design basis event, spent fuel assembly becomes over-heated and finally damaged. In order to prevent and stop SFP (spent fuel pool) severe accident process, strategies should be developed when existing procedures are no longer effective or available. This paper gives a development approach of CPR1000 NPP (nuclear power plant) SFP severe accident management guidelines. According to this development approach, the framework of SFP severe accident management guidelines has been developed. To fulfill the strategies and framework for SPF severe accident guidelines, calculation of typical accident cases and determination analysis are carried out using PSA method, based on investigation and survey of international severe accident management procedures research status, considering the characteristics of CPR1000 NPP and SFP design. Then related actions and proper times are proposed, which would help to develop detailed severe accident management guidelines (SAMG) of SFP.

Commentary by Dr. Valentin Fuster
2017;():V007T11A003. doi:10.1115/ICONE25-66110.

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

Topics: Heat , Emergencies
Commentary by Dr. Valentin Fuster
2017;():V007T11A004. doi:10.1115/ICONE25-66199.

The extreme hazards such as large floods, earthquakes, fires or explosion may be lead to an extensive damage or large scale incident in NPP, for instance, loss of command and control, loss of most instruments or equipment which are used for accident mitigation. The typical situations are loss of control room and alternate shutdown capabilities, or loss of AC and DC power, or all of them. Extreme Hazards Mitigation Guideline (EHMG) is developed as the coping strategy for the above extensive damage situations according to the requirements of the characteristic of CPR1000 NPP and the modifications implemented after Fukushima accident. The technical scheme and frame of EHMG includes two modules: EH-IRs (initial response) and EH-MGs (mitigation guideline). And then based on the typical accident analysis, the critical diagnostic parameters and mitigation strategies were made for the EHMG. EH-IRs contains the following contents: Rebuilding the command and control, Off-Site and On-Site Communications, Initial Operational Response Actions, Initial Damage Assessment. EH-MGs contains the following strategies: Manually operating auxiliary turbo pump, Manually depressurizing SGs and injecting into SG, Alternate makeup to RCS, Controlling containment conditions, Alternate Makeup to Spent fuel pool, Alternate Makeup to water storage tank, Containment flooding and so on. The strategies in EHMG can cover both prevention and mitigation phases of severe accident. The first EHMG in China has been validated and implemented in Hong Yan He NPP in September, 2016.

Commentary by Dr. Valentin Fuster
2017;():V007T11A005. doi:10.1115/ICONE25-66326.

This paper studies the mitigation strategies for combustible gases in a NPP. Based on the PSA results and engineering experiences, one typical case and one bounding case are chosen for the study of the control systems for the combustible gases. The combustible gases control strategies for AP1000 and EPR are also discussed. The studies find that, the combustible gases control system should be designed along with containment heat removal systems. The containment heat removal system will cool the containment and condense the vapor in the containment, which will facilitate the effectiveness of the igniters. The advantage of the passive recombiners system is the ability to eliminate hydrogen beyond the flammability limit. But the recombination rate is slow, and at the early stage of the accidents, the hydrogen will still be at a high level. In this condition, the operation of the containment heat removal systems should be designed carefully to eliminate the negative hydrogen risks.

Commentary by Dr. Valentin Fuster
2017;():V007T11A006. doi:10.1115/ICONE25-66431.

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.

Commentary by Dr. Valentin Fuster
2017;():V007T11A007. doi:10.1115/ICONE25-66433.

In order to deal with the nuclear severe accidents, the severe accident management systems are popularly considered and developed at home and abroad recently. A severe accident management system usually includes these functional parts: accident monitor, accident diagnosis, accident simulation, accident prognosis and SAMG support. Here, the accident diagnosis part is mainly concerned, and three nuclear accident diagnosis methods are introduced here, including BP neural network method, SDG expert diagnosis technique and artificial diagnosis method, which are also applied to a severe accident management system developed by us. In this paper, firstly, the severe accident management system developed by us will be introduced briefly. Then, three accident diagnosis methods for nuclear power plant (NPP) are showed and described in detail. At last, two cases including LOCA and SGTR accidents are used for the verification of these accident diagnosis methods and some analyzing results and conclusions are given. The results show that the three diagnosis methods are very useful for the accident diagnosis of NPP, which can diagnose the accident type accurately and offer much information or support to the severe accident management system and operators. The paper offers some reference significance for the research of accident diagnosis methods and the development of severe accident management system.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2017;():V007T11A008. doi:10.1115/ICONE25-66451.

Fukushima Daiichi accident shows that extreme external events may lead to extensive damage in nuclear power plants (NPPs), which may not only affects equipment, but also damage the control and monitor ability of control rooms (including human factor, habitability, control panel and so on). Accidents-related procedures of NPPs in China are Emergency Operation Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs), both of them rely on control and monitor ability of control rooms. Therefore, current structure of accident management procedures can’t work effectively in the scene of extensive damage. The development of Extensive Damage Mitigation Guidance (EDMGs) may strengthen the defense in depth. In relation to this, an Integrated EDMG has been developed to deal with extensive damage resulted from fire (or explosion) or extreme hazards, while the Nuclear Energy Institute (NEI) EDMG only considers fire and explosion. Besides the extended scope, Integrated EDMG develops a comprehensive and detailed mitigation method which is applicable for extensive damage. In this paper, the characteristics of extensive damage will be analyzed first, and then the structure of Integrated EDMG is developed based on the characteristic analysis. According to the emergency response analysis and accident management research, detailed introduction of the Integrated EDMG and its sub-guidances are discussed.

Commentary by Dr. Valentin Fuster
2017;():V007T11A009. doi:10.1115/ICONE25-66605.

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

Topics: Accidents
Commentary by Dr. Valentin Fuster
2017;():V007T11A010. doi:10.1115/ICONE25-66773.

A typical 1000MW pressurized-water reactor (PWR) unit model of China’s living nuclear power plant (NPP) units is built based on MAAP4[1] in this paper. Different severe accidents cases caused by different LOCA area on hot leg of primary loop are studied. And different mitigation measures are focused to evaluated their effectiveness. The study indicates that during the accident, the larger broken area LOCA case caused the more severe rector core damaged. However, it is important to inject water into the reactor core in good time. And that can mitigate the severe accident progress effectively.

Commentary by Dr. Valentin Fuster
2017;():V007T11A011. doi:10.1115/ICONE25-66969.

In-Vessel Retention (IVR), which arrests relocated molten core materials in the vessel during severe accident, has been singled out as an appealing accident management approach to many reactors. The heat transfer imposed by in-vessel corium is a vital part for IVR success considering the difficulty of significantly altering ex-vessel CHF. For a given decay power, corium pool configuration determines the heat flux profile along the vessel wall, which may produce uncertainties associated with IVR strategy. In this paper, a thermodynamic tool is employed to study the corium pool configurations by analyzing the possible interaction among relocated corium, zircalloy cladding and core internals. The results reveal the immiscibility gap phenomena under high temperature which separates molten materials into oxidic and metal phase in the lower head. The oxidic phase is quite stable and its density is only slightly changed by various accident scenarios. The metal phase is relatively unstable and its density is susceptible to the condition of cladding oxidation degree and crust integrity. The corium pool configurations in the lower head are determined based on the results of thermodynamic analysis and phase density comparison. Both two-layer and three-layer corium pools are likely to be formed under different accident scenarios. CAP1400 has intentionally increased the mass of lower core support plate, which is a beneficial design change to prevent possible focusing effect if material infiltration through crust is assumed to be impossible.

Commentary by Dr. Valentin Fuster
2017;():V007T11A012. doi:10.1115/ICONE25-67008.

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.

Commentary by Dr. Valentin Fuster
2017;():V007T11A013. doi:10.1115/ICONE25-67012.

In-containment hydrogen, produced during severe accident, can largely challenge the integrity of containment, which has been studied for many years. However, Fukushima Dachii accident has shown the possibility of explosion outside containment due to leaked hydrogen, which was never considered before. In the paper, ex-containment hydrogen analysis is conducted with and without mitigation measures based on the newly-built CAP1400 MAAP4 containment model. Middle annulus is screened as the only location for the possible hydrogen accumulation. Due to steam condensation, hydrogen concentration can be unexpectedly high enough to exceed that in the containment, but either igniters or PARs can mitigate its risk even under conservative conditions. The possibility of containment failure due to hydrogen leakage is very low as long as mitigation measures are functional.

Commentary by Dr. Valentin Fuster
2017;():V007T11A014. doi:10.1115/ICONE25-67034.

Radiation monitoring instruments (KRT) is important to decide emergency response level (EAL) in accident situation. Emergency response drills is more and more significant after Fukushima Daiichi severe accident. This article develops a code to simulate radiation monitoring instruments data for emergency response exercises. A part of calculated input of the code comes from MAAP calculated results included source terms and thermo hydraulic data. KRT simulation code runs with MAAP calculation results and output KRT simulated value at the same time. This article gives a way to prove the result of KRT simulation code analyzed is correct and matched through simulating an emergency response exercise scenario.

Commentary by Dr. Valentin Fuster
2017;():V007T11A015. doi:10.1115/ICONE25-67121.

While hydrogen release into large compartment from confined compartment, hydrogen diffusion flame is easy to occur. There is intense heat radiation effect on electrical penetration from diffusion flame. Aim to evaluate the influence of diffusion flame on electrical penetration, systematic method is constructed, including computing view factor of electrical penetration, assessment of hot spot of containment vessel and research of heat transfer for electrical penetration. Research results give theory basis for determining location of venting which can generate hydrogen diffusion flame. The method can be extended to use in the influence evaluation of personnel hatch and equipment hatch in the containment vessel.

Commentary by Dr. Valentin Fuster
2017;():V007T11A016. doi:10.1115/ICONE25-67123.

The contribution provides information about the development of a system for visualization of NPP severe accident progress. This visualization is under development in cooperation of UJV Rez, a.s. and Czech Technical University in Prague. The project is supported by the Technology Agency of the Czech Republic and is planned to be solved from 2015 to 2017. The visualization uses results of an analytical code MELCOR for evaluation of the NPP severe accident progress. The visualization firstly reads MELCOR results, transforms them to a suitable format for quick processing and provides graphical screens with reactor components that could demonstrate the progress of the evaluated severe accident. The visualization can even provide parallel presentation of more different scenarios of the severe accident. The system is planned to be used for training of NPP staff to handle severe accidents. In the first year of the project solution (2015), the software for MELCOR data transformation, next for providing information about transformed data were developed. In the following year (2016), software for creation of graphical screens with reactor components and software for severe accident progress presentation is creating. In the final year of the project (2017), thorough testing is going to be carried out, and the applicability of the visualization for a practical use during a NPP staff training is going to be verified.

Commentary by Dr. Valentin Fuster
2017;():V007T11A017. doi:10.1115/ICONE25-67176.

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

Commentary by Dr. Valentin Fuster
2017;():V007T11A018. doi:10.1115/ICONE25-67248.

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology.

This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions.

Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

Commentary by Dr. Valentin Fuster
2017;():V007T11A019. doi:10.1115/ICONE25-67317.

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident.

If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower.

TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout).

This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment.

Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied.

(1) RPV heat removal by restoring the RHR system

(2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger

(3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers

(4) RPV heat removal using the clean up water system (CUW)

By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.

Commentary by Dr. Valentin Fuster
2017;():V007T11A020. doi:10.1115/ICONE25-67520.

Hydrogen combustion including deflagration and detonation could become a significant threat to the integrity of containment vessel or reactor building in a severe accident of nuclear power stations. In the present study, numerical analyses were carried out for the ENACCEF No.153 test to develop computational techniques to evaluate the flame acceleration phenomenon during the hydrogen deflagration. This experiment investigated flame propagation in the hydrogen-air premixed gas in a vertical channel with flow obstacles. The reactingFoam solver of the open source CFD code, OpenFOAM, was used for the present analysis. Nineteen elementary chemical reactions were considered for the overall process of the hydrogen combustion. For a turbulent flow, renormalization group (RNG) k-ε two-equation model was used in combination with wall functions. Three manners of nodalization were applied and its influences on the flame propagation acceleration were discussed.

Commentary by Dr. Valentin Fuster
2017;():V007T11A021. doi:10.1115/ICONE25-67598.

During a severe accident when a corium is ejected from the reactor vessel into the containment, one of the mitigations strategy to contain radioactivity from discharging in to environment and to keep containment under design pressure is to use containment flirted venting system (CFVS). In this paper the key phenomena related to the operation and performance in the components of the CFVS system, pool venture scrubber, cyclone separator, particulate filter, and molecular sieve filters were identified. Based on these phenomena a scaling analysis was performed that was based on system level and local phenomena level scaling. For system level scaling boundary flow, mass flux, pressure and temperature were preserved. On the local phenomena scaling various phenomena were considered. Scaling analysis was also carried out for scrubber system, cyclone and filter. Utilizing a scaled down model from 50 nozzles in the prototype to three nozzles in the scaled model, the flow parameters for the model facility scrubber were obtained. Using these parameters, the governing non-dimensional parameters were obtained for prototype and model facility. The scaling ratios for all the relevant parameters are summarized in this paper.

Topics: Testing , Containment
Commentary by Dr. Valentin Fuster
2017;():V007T11A022. doi:10.1115/ICONE25-67678.

Severe accident has become one of the main directions of research since the crisis at Fukushima plants in Japan, including release of radioactive material, accessibility analysis for staff and evaluation of consequence. This paper, mainly for design basic accident and severe accident, makes calculation of migration and release of radioactive material after accident by considering the different building, also combine the on-site operation requirement of worker after accident, analyzes dose rates of typical zone for staff and evaluate the exposure caused by radioactive material. The main results of the paper supply reference and basis for person accessibility research after design basis accident and severe accident.

Commentary by Dr. Valentin Fuster
2017;():V007T11A023. doi:10.1115/ICONE25-67800.

Hydrogen risk in the spent fuel compartment becomes a matter of concern after the Fukushima accident. However, researches are mainly focused on the hydrogen generated by spent fuels due to lack of cooling. As a severe accident management strategy, one of the containment venting paths is to vent the containment through the normal residual heat removal system (RNS) to the spent fuel compartment, which will cause hydrogen build up in it. Therefore, the hydrogen risk induced by containment venting for the spent fuel compartment is studied for advanced passive PWR in this paper. The spent fuel pool compartment model is built and analyzed with integral accident analysis code couple with the containment analysis. Hydrogen risk in the spent fuel pool compartment is evaluated combining with containment venting. Since the containment venting is mainly implemented in two different strategies, containment depressurization and control hydrogen flammability, these two strategies are analyzed in this paper to evaluated the hydrogen risk in the spent fuel compartment. Result shows that there will not be significate hydrogen built up with the hydrogen control system available in the containment. However, if the hydrogen control system is not available, venting into the spent fuel pool compartment will cause a certain level of hydrogen risk there. Besides, suggestions are made for containment venting strategy considering hydrogen risk in spent fuel pool compartment.

Commentary by Dr. Valentin Fuster
2017;():V007T11A024. doi:10.1115/ICONE25-67858.

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2.

In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Topics: Steel , Accidents , oxidation
Commentary by Dr. Valentin Fuster
2017;():V007T11A025. doi:10.1115/ICONE25-68005.

In this study, severe accident analyses were conducted for the Advanced Boiling Water Reactor (ABWR) using a detailed nodalization and the impact of MAAP nodalization on the severe accident analysis was investigated. The results of the analysis obtained using a detailed nodalization shows that the PCV temperature is more challenging in comparison to the results of the original nodalization. However, it was found that the structural temperature of the PCV head flange can be maintained below the ultimate temperature of the head gasket when passive mitigation systems are successful, even if no reactor well injection is conducted.

Commentary by Dr. Valentin Fuster
2017;():V007T11A026. doi:10.1115/ICONE25-68012.

New nuclear safety objectives and principles are being studied in main nuclear power countries and organizations after Fukushima Dai-ichi nuclear accident, to further improve the safety level of nuclear power plants (NPPs). Based on International Atomic Energy Agency (IAEA) Specific Safety Requirements (No.SSR-2/1), “Safety of Nuclear Power Plants: Design” (HAF102-2016) is issued in China. The concept “design extension condition (DEC)” is put forward, which is intend to enhance the plant’s capability to withstand accidents that are more severe than Design Basis Accidents (DBA). DEC could include conditions without significant fuel degradation (DEC-A in this paper) and conditions with core melting (DEC-B in this paper), e.g. severe accident. In this paper, the DEC-A and its application was discussed preliminarily, firstly, the development and connotation was introduced, then the identification of DEC-A, and the safety analysis principles of DEC-A were mainly described. This study may play a valuable role for implementation of new nuclear safety requirements in China.

Commentary by Dr. Valentin Fuster
2017;():V007T11A027. doi:10.1115/ICONE25-68020.

In view of the 1F decommissioning project, the Institute of Applied Energy (IAE) has been analyzing the course of the accident using the SAMPSON code with an aim to investigate its progression in detail. For 1F Unit-2, certain discrepancies between measurement values and analysis results still exist. For example, although three pressure peaks occur after a manual activation of a safety relief valve (SRV), its mechanism is yet unclear. This study seeks to elucidate the mechanism for the three pressure peaks using a new presumption for the relocation of the debris through the lower part of the core. The current results could reproduce the basic tendency for each peak. However, some deviations between simulation results and measured values indicate the necessity for further improvement of the thermal-hydraulic model used in SAMPSON.

Commentary by Dr. Valentin Fuster

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