Newest Issue

Guest Editorial

ASME J of Nuclear Rad Sci. 2018;4(2):020201-020201-1. doi:10.1115/1.4039038.

This Special Section in this issue of the ASME Journal of Nuclear Engineering and Radiation Science is dedicated to selected papers from the 24th International Conference on Nuclear Engineering (ICONE-24) held in Charlotte, NC during June 26–30, 2016. The selected ICONE-24 papers were revised and reviewed to meet the journal standards and requirements. ICONE-24 was co-located with the 2016 Power and Energy Conference that brought together five concurrent ASME conferences and forums focusing on issues related to power generation, energy sources, and energy sustainability. Therefore, ICONE-24 benefited from the presence of a very large gathering of professionals in the energy sector. The ICONE conference is a premier global annual conference on nuclear engineering, science, and technology that has a quarter century history and is organized by the Nuclear Engineering Division (NED) of the ASME, the Japan Society of Mechanical Engineers (JSME), and the Chinese Nuclear Society (CNS). The first conference, ICONE-1, was held in 1991 in Tokyo with the JSME and ASME as sponsors. In 2005, the Chinese Nuclear Society formally joined as the third sponsor and hosted ICONE-13 in Beijing. The ICONE conferences are global significant events of choice for nuclear professionals to stay technically current, follow industry trends, promote professional development, and provide a platform for information exchange and dissemination. The contributions from industry, government, and academia lead to the success of ICONE as the quality and number of the technical presentations steadily increased. In addition to the archival paper publications, ICONE assists the nuclear industry in developing future generations of nuclear professionals to meet industry needs. This is achieved through an excellent student paper competition program that includes a significant number of students from around the globe, supported and partially funded by the conference sponsors (60 finalists in total consisting of 15 students from each of the following four regions: North America, Europe, China, and Japan with Asia at large).

Commentary by Dr. Valentin Fuster


ASME J of Nuclear Rad Sci. 2018;4(2):020901-020901-8. doi:10.1115/1.4038911.

A new experimental program using nontransfer (NTR) type plasma heating is under consideration in Japan Atomic Energy Agency (JAEA) to clarify the uncertainty on core-material relocation (CMR) behavior of boiling water reactor (BWR). In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm × 107 mm × 222 mm (height)). An excellent perspective in terms of applicability of the NTR plasma heating to melting high melting-temperature materials such as ZrO2 has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO2 fuel PHEBUS fission products (FP) tests. Furthermore, application of electron probe micro-analyzer (EPMA), scanning electron microscope (SEM)/energy dispersive X-ray spectrometry (EDX), and X-ray computed tomography (CT) led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the NTR plasma heating technology to the severe accident (SA) experimental study was obtained.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020902-020902-15. doi:10.1115/1.4038059.

The estimation of fission products (FPs) release from the containment system of a nuclear plant to the external environment during a severe accident (SA) is a quite complex task. In the last 30–40 yr, several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state of the art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, also a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus fission products test 1 (FPT-1) test employing the accident source term evaluation code (ASTEC) and MELCOR codes (respectively, ASTEC v.2.0 revision 3 patch 3 and MELCOR V2.1.6840 version). The analysis focuses on the stand-alone containment aspects of the test, and three different modelizations of the containment vessel have been developed showing that at least 15/20 control volumes (CVs) are necessary for the spatial schematization to correctly predict the test thermal hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020903-020903-13. doi:10.1115/1.4038223.

This paper deals with near past, ongoing, and planned R&D works on fission products (FPs) behavior in reactor cooling system (RCS), containment building and in filtered containment venting systems (FCVS) for severe accident (SA) conditions. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in simulation software. After being initiated in 2004, researches on iodine transport through the RCS are still ongoing and for containment, the last advances are linked to the source term (ST) evaluation and mitigation (STEM) OECD/NEA project. The objective is to improve the evaluation of ST for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major FPs: iodine and ruthenium. For ruthenium attention has been paid to study the amount and nature (gas/aerosol partition) of ruthenium species along the RCS. A follow-up, called STEM2, has started to reduce some remaining issues and be closer to reactor conditions. For FCVS works, the efficiencies for trapping iodine covering scrubbers and dry filters are examined to get a clear view of their abilities in SA conditions. Another part is focused on specific porous materials able to trap volatile iodine. Influence of zeolite materials parameters (nature of the counter-ions, structure, Si/Al ratio…) are tested as well as new kind of porous materials constituted by Metal organic Frameworks will also be looked at.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020904-020904-9. doi:10.1115/1.4037203.

Modernization of reactor instrumentation and control systems is mainly characterized by the transition from analog to digital systems, expressed by replacement of hardware equipment with new software-driven devices. Digital systems may share intelligence capabilities where except for measuring and processing information may also make decisions. State identification systems are systems that process the measurements taken over operational variables and output the state of the reactor. This paper frames itself in the area of control systems applied to state identification of boiling water reactors (BWRs). It presents a methodology that utilizes machine learning tools, and more specifically, a set of relevance vector machines (RVMs) in order to process the incoming signals and identify the state of the BWR in real time. The proposed methodology is comprised of two stages: in the first stage, each RVM identifies the state of the BWR, while the second stage collects the RVM outputs and decides about the real state of the reactor adopting majority voting. The proposed methodology is tested on a set of real-world BWR data taken from the experimental FIX-II facility for recognizing various BWR loss-of-coolant accidents (LOCAs) as well as normal states. Results exhibit the efficiency of the methodology in correctly identifying the correct state of the BWR while promoting real time identification by providing fast responses. However, a strong dependence of identification performance on the form of kernel functions is also concluded.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020905-020905-11. doi:10.1115/1.4039000.

The experience of severe accidents shows that reliable determination of technological process parameters is necessary but not always sufficient to avoid catastrophic consequences. The accident measures should be considered in a broader context that includes the human factor, organization of the nuclear technology, external influences, and safety culture. The anticipated transient without scram (ATWS) events were not considered in the original water water energy reactor (WWER) (Russian pressurized water reactors (PWR)) design basis accidents (DBA). The design extension conditions (DEC) scenarios progress in a context which is very uncertain and highly stressful for the operators. If a specific scenario requires some operators' actions as measures to mitigate, delay, or distribute the accident consequences, then the dynamics of accident context seem of primary importance for “best estimate” evaluations and enhancing the plant's capability. The paper presents the capacities of the performance evaluation of teamwork (PET) procedure for enhancing plant's capability for DEC based on best estimate context evaluation of human performance in ATWS events. The PET procedure is based on a thorough description of symptoms of various timelines and their context quantification. It is exemplified for different ATWS scenarios of the nuclear power plant (NPP) with WWER-1000 based on thermal-hydraulic simulations with RELAP5/MOD3.2 code and models.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020906-020906-8. doi:10.1115/1.4039288.

On Mar. 11, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the high temperature engineering test reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the “visual inspection of equipment and instruments.” The other is the “seismic response analysis” for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013, and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020907-020907-6. doi:10.1115/1.4037715.

Paks Nuclear Power Plant (NPP) identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of margins of existing severe accident management (SAM) facilities, and construction of some new systems and facilities. While developing the SAM strategy, the basic question was what is the sufficient margin above the design basis level of existing structures, systems, and components for avoiding the cliff-edge effects, and what level of or hazard should be taken for the design of new structures and systems dedicated for SAM. Paks NPP developed an applicable in the practice concept for the qualification of already implemented SAM measures and design the new post-Fukushima measures that are outlined in the paper. The concept is based on the generalization of the procedure and assumptions used in the definition of acceptable margins for seismic loads, analysis of the steepness of the hazard curves and features of the hazards. Justification of the definition of exceedance probability of the design basis effects for the design of SAM systems is given based on the first order reliability theory. The application of the concept is presented on several practical examples.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020908-020908-9. doi:10.1115/1.4038062.

A VTT Fukushima Daiichi Unit 3 (1F3) method for estimation of liquids and consequences of releases (MELCOR) model was modified to simulate the Fukushima Daiichi Unit 2 (1F2) accident. Five simulations were performed using different modeling approaches. The model 1F2 v1 includes only the basic modifications to reproduce the 1F2 accident. The model 1F2 v2 includes the same modifications used in 1F2 v1 plus the wet well (WW) improvement. In the 1F2 v3 model, the reactor core isolation cooling (RCIC) system logic was modified to avoid the use of tabular functions for the mass flow inlet and outlet. Because of this analysis, it is concluded that there is a strong dependency on parameters that still have many uncertainties, such as the RCIC two-phase flow operation, the alternative water injection, the suppression pool (SP) behavior, the rupture disk behavior and the containment failure modes, which affect the final state of the reactor core.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020909-020909-14. doi:10.1115/1.4038160.

Nuclear safety analysis and licensing criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that could pose a danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping postulated plant situations (or events) into categories that are defined according to their expected frequencies of occurrence (i.e., high-frequency, low-consequence events, and low-frequency, high consequence events). In plant licensing basis analyses, events in each category must be shown to yield consequences that remain within the limits that are specified for that category. To protect the integrity of this categorization scheme, events must not be allowed to develop into the more serious events that belong in other, higher-consequence categories. In other words, nuclear plant designs must not allow high-frequency, low-consequence events to degrade into high-frequency, high-consequence events. The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that could, and could not be used to demonstrate, for licensing purposes, that benign events are prevented from becoming serious accidents.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020910-020910-5. doi:10.1115/1.4038556.

In this paper, a new technique to handle solid radioactive materials inside a liquid matrix is presented. The conceptual design of the device profits of the experience and know-how gained in decontamination procedures. The proposed system makes use of an ejector for the suction of a water-highly radioactive swarf mixture from the purifier pool of the Italian E. Fermi nuclear power plant (NPP) and moving it in a suitable container for the subsequent conditioning. A dedicated circuit with an ejector to demonstrate the feasibility of the method was realized. A minimum inlet flow rate was found to have swarf suction. The feasibility of the method was demonstrated, even if it is required to homogenize the inlet mixture to avoid swarf packing conditions inside the ejector.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020911-020911-8. doi:10.1115/1.4037806.

The intermittency of renewable power generation systems on the low carbon electric grid can be alleviated by using nuclear systems as quasi-storage systems. Nuclear air-Brayton systems can produce and store hydrogen when electric generation is abundant and then burn the hydrogen by co-firing when generation is limited. The rated output of a nuclear plant can be significantly augmented by co-firing. The incremental efficiency of hydrogen to electricity can far exceed that of hydrogen in a standalone gas turbine. Herein, we simulate and evaluate this idea on a 50 MW small modular liquid metal/molten salt reactor. Considerable power increases are predicted for nuclear air-Brayton systems by co-firing with hydrogen before the power turbine.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020912-020912-9. doi:10.1115/1.4039076.

Recently, the nuclear industry has made a tremendous effort to assess the safety of nuclear power plants (NPP), as advances in seismology have led to the perception that the potential earthquake hazard in the U.S. may be higher than originally assumed. Due to the conservatism in the NPP design, structures and safety-related items are capable of withstanding earthquakes larger than the safe shutdown earthquake (SSE). One major aspect of conservatism in the design is ignoring the effect of soil-structure interaction (SSI), which results in conservative estimates of seismic demands for plant equipment. In this paper, a typical reactor building (RB) is chosen for a case study to investigate the potential benefit of accounting for SSI effects. A lumped mass stick model is first developed and analyzed with a fixed base configuration, using the free-field ground motion as input at the foundation level, as well as with a SSI configuration. Fragility analyses are then performed for the RB and one of its components to quantify the effects of the SSI on the overall seismic risk. In each case, a family of seismic fragility curves is developed. It is found that consideration of SSI effects in the analysis can improve the component fragilities, and potentially enhance the core damage frequency (CDF) of the plant.

Topics: Soil , Safety , Earthquakes , Design
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020913-020913-4. doi:10.1115/1.4038928.

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):020914-020914-7. doi:10.1115/1.4037116.

In this paper, the causes and the radiological consequences of the explosion of the Chernobyl reactor occurred at 1:23 a.m. (local time) on Apr. 26, 1986, and of the Fukushima Daiichi nuclear disaster following the huge Tsunami caused by the Great East Japan earthquake at 2.46 p.m. (local time) on Mar. 11, 2011 are discussed. The need for better severe accident management (SAM), and severe accident management guidelines (SAMGs), are essential in order to increase the safety of the existing and future operating nuclear power plants (NPPs). In addition to that, stress tests should, on a regular basis, be performed to assess whether the NPPs can withstand the effects of natural disasters and man-made failures and actions. The differences in safety preparations at the Chernobyl and Fukushima Daiichi will therefore be presented, as well as recommendations concerning improvements of safety culture, decontamination, and disaster planning. The need for a high-level national emergency response system in case of nuclear accidents will be discussed. The emergency response system should include fast alarms, communication between nuclear power plants, nuclear power authorities and the public people, as well as well-prepared and well-established evacuation plans and evacuation zones. The experiences of disaster planning and the development of a new improved emergency response system in Japan will also be presented together with the training and education program, which have been established to ensure that professional rescue workers, including medical staff, fire fighters, and police, as well as the normal populations including patients, have sufficient knowledge about ionizing radiation and are informed about the meaning of radiation risks and safety.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2018;4(2):021001-021001-10. doi:10.1115/1.4038823.

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code relap/scdapsim/mod4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021002-021002-8. doi:10.1115/1.4037898.

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021003-021003-13. doi:10.1115/1.4038557.

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021004-021004-7. doi:10.1115/1.4038772.

Point reactor neutron kinetics equations describe the time-dependent neutron density variation in a nuclear reactor core. These equations are widely applied to nuclear system numerical simulation and nuclear power plant operational control. This paper analyzes the characteristics of ten different basic or normal methods to solve the point reactor neutron kinetics equations. The accuracy after introducing different kinds of reactivity, stiffness of methods, and computational efficiency are analyzed. The calculation results show that: considering both the accuracy and stiffness, implicit Runge–Kutta method and Hermite method are more suitable for solution on these given conditions. The explicit Euler method is the fastest, while the power series method spends the most computational time.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021005-021005-6. doi:10.1115/1.4038899.

Detection of loss of coolant accident (LOCA) and generation of reactor trip signal for shutting down the reactor is very important for safety of a nuclear reactor. Large break LOCA (LBLOCA) is a typical design basis accident in all reactors and has attracted attention of the reactor designers. However, studies reveal that small break loss of coolant accident (SBLOCA) can be more severe as it is difficult to detect with conventional methods to generate reactor trip. SBLOCA in channel-type reactors is essential to consider as it may create stagnation channel conditions in the reactor coolant channel, which may lead to fuel failure, if the reactor is not tripped. Advanced heavy water reactor (AHWR) is a channel-type boiling water reactor, which may experience stagnation channel conditions in case of SBLOCA in feeder pipes. For initiating the trip signals and safe shut down of the reactor in such cases, a novel system comprising of acoustic-based sensors is incorporated in the reactor design. The system detects the peculiar sound of the steam leaked from the main heat transport system (MHTS) and generates reactor trip signal. The experimental demonstration of such new system is essential before its introduction in the reactor. The experimental demonstration of the stagnation channel break, its detection by acoustic-based sensors system, and reactor trip followed by generation of reactor trip signal was performed and presented in the paper. The experiment showed that the trip signal for AHWR can be generated within 5 s with acoustic sensor and 2 s by low flow signal and reactor trip can be ensured in 7 s following a LOCA.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021006-021006-5. doi:10.1115/1.4038774.

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235U mass and the total reactor mass.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021007-021007-4. doi:10.1115/1.4038771.

In depletion and transmutation calculation, it is important to solve detailed burnup chains with high computational accuracy and efficiency. This requires the good performance of the burnup algorithms. Nuclide inventory tool (NUIT) is a newly developed nuclide inventory calculation code, which is capable of handling detailed depletion chains by implementing various advanced algorithms. Based on the NUIT code, this paper investigates the accuracy and efficiency of the mini-max polynomial approximation (MMPA) method, and compares it with other burnup solvers in NUIT code. It is concluded that the MMPA method is numerically accurate and efficient for dealing with detailed depletion chains with extremely short half-lived nuclides.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021008-021008-6. doi:10.1115/1.4038653.

Solving the third-order simplified spherical harmonics method (SP3) equations is one of the key points in the development of advanced reactor calculation methods and has been widely concerned. The semi-analytical nodal method (SANM), based on transverse-integrated diffusion equation, has the advantages of high accuracy and convenience for multigroup calculation. Due to its advantages, the method is expected to be used in solving the SP3 equations. However, the traditional SANM is not rigorous since the expansion process does not take the special modality of the SP3 equations and their analytical solutions into consideration. There are two modalities of the SP3 equations, so there are two traditional SANM forms on solving the SP3 equations, and the differences between the two forms will be very important in further research on the SANM. A code is developed to solve the SP3 equations under the two different forms. After the calculation of the same benchmark, the difference between the two forms on solving the SP3 equations is found. According to the results, and in view of the special modality of the SP3 equations, points on a more rigorous SANM for solving the SP3 equations are discussed.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021009-021009-10. doi:10.1115/1.4038185.

This paper presents a novel methodology for generating radiation intensity maps using a mobile robotic platform and an integrated radiation model. The radiation intensity mapping approach consists of two stages. First, radiation intensity samples are collected using a radiation sensor mounted on a mobile robotic platform, reducing the risk of exposure to humans from an unknown radiation field. Next, these samples, which need only to be taken from a subsection of the entire area being mapped, are then used to calibrate a radiation model of the area. This model is then used to predict the radiation intensity field throughout the rest of the area that could not be directly measured. The performance of the approach is evaluated through experiments. The results show that the developed system is effective at achieving the goal of generating radiation maps using sparse data.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):021010-021010-5. doi:10.1115/1.4038058.

This paper studies the radiological consequences resulting from withdrawal of nuclear fuel element (FE) from a core of open pool type reactor during normal operation. The FE withdrawal accident may be occurred due to human error during routine transport of spent FE process or from failure of FE clamp during reactor normal operation. For both cases, the FE will move vertically upward toward pool water surface. In case of accidental failure of fuel clamp, a negative reactivity insertion in the core after FE withdrawal and the reactor will be shutdown. MCNP5 code was used in this study to calculate the radiation dose rate levels in the reactor hall and inside the control room during FE withdrawal. The results show that the operator in control room will receive dose rate lower than permissible dose rate limit when FE reaches at depths more than to 211 cm for vertical FE and at depths more than to 246 cm for horizontal FE.

Topics: Fuels , Control rooms , Water
Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2018;4(2):024501-024501-4. doi:10.1115/1.4038999.

The whole core model of China experimental fast reactor (CEFR) is established according to the parameters of China experimental fast reactor, which are given by technical publication from the International Atomic Energy Agency (IAEA-TECDOC-1531), and the physical parameters of CEFR are simulated with the Monte Carlo N-particle code (MCNP4a). The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design values, which successfully demonstrate the acceptable fidelity of the MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2018;4(2):024502-024502-3. doi:10.1115/1.4038336.

We adopted the verified transition state theory, which originates from the quantum chemistry approach to explain the anomalous plastic flow or plastic deformation for Si nanowires irradiated with 100 keV (at room temperature regime) Ar+ ions as well as the observed amorphization along the Si nanowire (Johannes, et al. 2015, “Anomalous Plastic Deformation and Sputtering of Ion Irradiated Silicon Nanowires,” Nano Lett., 15, pp. 3800–3807). We shall illustrate some formulations which can help us calculate the temperature-dependent viscosity of flowing Si in nanodomains.

Commentary by Dr. Valentin Fuster

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In