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Guest Editorial

ASME J of Nuclear Rad Sci. 2017;3(3):030301-030301-1. doi:10.1115/1.4036459.

The Nuclear Societies in Israel include the Israel Society for Radiation Protection (ISRP), the Israel Nuclear Society (INS), and the Israel Society for Medical Physics (ISMP). These three societies hold a biennial conference in which the scientific community in Israel dealing with nuclear science, technology, and engineering, as well as radiation uses and radiation protection, encounters. The 28th Conference of the Nuclear Societies in Israel (INS-28) was held between April 12 and 14, 2016, at the Dan Panorama Hotel in Tel Aviv, Israel. The INS-28 Conference included 25 plenaries, parallel and poster sessions, in which more than 160 studies in various scientific fields were presented.

Commentary by Dr. Valentin Fuster

Special Section Papers

ASME J of Nuclear Rad Sci. 2017;3(3):030901-030901-10. doi:10.1115/1.4035883.

Novel genetic algorithms (GAs) are developed by using state-of-the-art selection and crossover operators, e.g., rank selection or tournament selection instead of the traditional roulette (fitness proportionate (FP)) selection operator and novel crossover and mutation operators by considering the chromosomes as permutations (which is a specific feature of the loading pattern (LP) problem). The algorithm is applied to a representative model of a modern pressurized water reactor (PWR) core and implemented using a single objective fitness function (FF), i.e., keff. The results obtained for some reference cases using this setup are excellent. They are obtained using a tournament selection operator with a linear ranking (LR) selection probability method and a new geometric crossover operator that allows for geometrical, rather than random, swaps of gene segments between the chromosomes and control over the sizes of the swapped segments. Finally, the effect of boundary conditions (BCs) on the symmetry of the obtained best solutions is studied and the validity of the “symmetric loading patterns” assumption is tested.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030902-030902-11. doi:10.1115/1.4035693.

Knowledge of the nuclear power plants (NPPs) containment atmosphere composition in the course of a severe accident is crucial for the effective design and positioning of the hydrogen explosion countermeasures. This composition strongly depends on containment flows which may include turbulent jet mixing in the presence of buoyancy, jet impingement onto the stratified layer, stable stratification layer erosion, steam condensation on the walls of the containment, condensation by emergency spray systems and other processes. Thus, in modeling of containment flows, it is essential to correctly predict these effects. In particular, a proper prediction of the turbulent jet behavior before it reaches the stably stratified layer is critical for the correct prediction of its mixing and impingement. Accordingly, validation study is presented for free neutral and buoyancy-affected turbulent jets, based on well-known experimental results from the literature. This study allows for the choice of a proper turbulence model to be applied for containment flow simulations. Furthermore, the jet behavior strongly depends on the issuing geometry. A comparative study of erosion process for the conditions similar to the ones of international benchmark exercise (IBE-3) is presented for different jet nozzle shapes.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030903-030903-8. doi:10.1115/1.4035692.

Standards, guidelines, manuals, and researches refer mainly to the required protection of a nuclear power plant (NPP) containment structure (where the reactor's vessel is located) against different internal and external extreme events. However, there is no consideration regarding the man-made extreme event of external explosion resulting from air bomb or cruise missile. A novel integrated blast resistance model (IBRM) of NPP's reinforced concrete (RC) auxiliary facilities due to an external above ground explosion based on two components is suggested. The first is structural dynamic response analysis to the positive phase of an external explosion based on the single degree-of-freedom (SDOF) method combined with spalling and breaching empirical correlations. The second is in-structure shock analysis, resulting from direct-induced ground shock and air-induced ground shock. As a case study, the resistance of Westinghouse commercial NPP AP1000 control room, including a representative equipment, to an external above ground blast loading of Scud B-100 missile at various standoff distances ranging from 250 m (far range) till contact, was analyzed. The structure's damage level is based on its front wall supports' angle of rotation and the ductility ratio (dynamic versus elastic midspan displacement ratio). Due to the lack of specific structural damage demands and equipment's dynamic capacities, common protective structures standards and manuals are used while requiring that no spalling or breaching shall occur in the control room while it remains in the elastic regime. The engineering systems and equipments' spectral motions should be less than their capacity. The integrated blast resistance model (IBRM) of the structure and its equipment may be used in wider researches concerning other NPP's auxiliary facilities and systems based upon their specifications.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030904-030904-5. doi:10.1115/1.4036322.

Fly ash is widely used as a supplementary cementitious material in the production of cement and concrete, and improves durability and strength of the concrete. However, as for all materials of mineral origin, fly ash is a source for natural radioactivity; hence, its need for responsible use. The aim of this study is to investigate the radiation impact from fly ash as an additive to concrete compared against concrete without fly ash. For this purpose, eight concrete mixtures are experimentally tested, followed by a computation of the radiation dose when used as bulk material in building constructions. The results demonstrate an increase in the total radiation dose from around 0.8 mSv with no fly ash up to 0.92 mSv when fly ash is used. The increase mostly comes from external radiation, while the radon exhalation factor is reduced and sometimes even reduces the radon dose despite the higher radium content. The work has demonstrated that the impact from fly ash on the radiation exposure is limited when applied as a supplementary cementitious material. At the same time, fly ash provides real benefits to the quality and durability of the concrete. For this reason, exemption strategies for such applications should be developed.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030905-030905-8. doi:10.1115/1.4036431.

Advanced imaging systems, such as C-Arm machines, greatly improve physicians' diagnostic abilities and provide greater precision. Yet, these benefits come with a price of ionizing radiation exposure to medical teams and patients. Supplying proper training and skill improvement to operators on how to use this technology safely can help minimize risk of exposure. Previous studies on radiation knowledge among physicians and radiologists presented disturbing results of underestimated risk of exposure. The following research is based on an innovation in simulation-based training (SBT), a simulator using the Wizard of Oz (WOZ) concept that incorporates an online human trainer and was used for training emergency room (ER) physicians and ultrasound medical personnel. This research integrated WOZ technology with a radiation exposure formula for training to minimize unnecessary radiation exposure. The exposure formula presents real-time and overall exposure levels to operators based on their technique. The simulator also incorporates 3D animation graphics, enabling trainees to simulate the control of various factors. Image quality and the operator's radiation exposure levels are also animated, assisting trainees to focus on their exposure based on their device settings. Contrary to most previous studies, we measured radiation dose to the operator and quantified image quality accordingly. Validation was done on different C-Arm machines. Validation of learning outcomes was done using knowledge exams. Results from our knowledge exams presented significant improvement. The average result of knowledge exams given prior to training was 54%, whereas the average result after training was 94% (p < 0.001). Additionally, after a gap of 2–3 months, high retention was also found.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030906-030906-4. doi:10.1115/1.4036433.

Hyperthyroid and thyroid carcinoma patients treated with radioactive 131I (RAI) are a potential source of external and internal exposure to members of the public, to medical staff and especially to family members, who are in close contact with these patients. The relationship between radiation dose rates and various clinical parameters, including gender, age, thyroid size and weight, and iodine uptake, was assessed. Dose rates were measured on eight patients with thyroid carcinoma after total or subtotal thyroidectomy and on six patients with hyperthyroidism. All measurements were taken at 1 m from the patient on two levels—anterior to the neck and body center. Dose rates were measured at three or four times—at the time of RAI administration, and after 24 h, 48 h, and 168 h. Based on these measurements, retention curves were obtained. The effective half-life was derived by fitting an exponential equation and was estimated to be 12.8 h.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030907-030907-5. doi:10.1115/1.4035415.

The effect of incorporation of pozzolanic additives on the immobilization of cesium and strontium ions in cementitious pastes was investigated. Pastes containing Portland cement together with ground granulated blast furnace slag (GGBFS) (50%, 75%), metakaolin (MK) (10%, 20%), or silica fume (SF) (20%), either in its densified or raw form, were prepared. The transport properties of the immobilized ions through the paste were evaluated using leaching tests. Single differential thermal analysis (SDTA) was used to estimate the extent of the pozzolanic reaction and the pozzolanic reactivity of the different formulations. For strontium ions, the best immobilization system was the 20% raw silica fume (RSF) paste, characterized by the highest relative pozzolanity (RP). However, for cesium ions the most effective additive was the densified silica fume (DSF), reducing the apparent diffusion coefficient by two orders of magnitude compared to the unblended paste.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030908-030908-5. doi:10.1115/1.4035405.

Two main natural processes which control the migration of radioactive contaminants to the biosphere were studied in the Yamin Plain in order to evaluate the thickness of a cover layer needed for near-surface radioactive waste disposal facility. The first is the natural erosion of the cover layer, and the second is the infiltration during rain and runoff events. The erosion rate of the soil surface was studied by optical stimulation luminescence technique. It was found that during the last 14,000 years, the erosion rate was 0.3 mm/y which are 3 m for 10,000 years. The infiltration depth assessment was based on water content measurements and numerical modeling. It shows that under the most extreme rain event having an equivalent rain of 84 mm, infiltration depth was limited to 4.5 m. Taking into account the two processes, the effective cover layer thickness over 10,000 years should be at least 7.5 m thick.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030909-030909-8. doi:10.1115/1.4036458.

One of the preparation steps for a possible radiological attack is the capability of fast and effective decontamination of critical infrastructure. This study describes the implementation of a test plan at an intermediate scale (between bench scale and large scale or wide area) to evaluate decontamination procedures, materials, technologies, and techniques for removal of radioactive material from various surfaces. Two radioisotopes were tested: cesium-137 (137Cs) and the short-lived simulant to 137Cs, rubidium-86 (86Rb). Two types of decontamination hydrogel products were evaluated: DeconGel™ and Argonne SuperGel. Tests were conducted at the assigned Chemical, Biological, Radiological, and Nuclear (CBRN) Israel Defense Forces (IDFs) Home Front Command facility, and at the Nuclear Research Center Negev (NRCN), Israel. Results from these tests indicated similar removal and operational parameters for 86Rb and 137Cs, as expected from the chemical similarity of both elements. These results proved that the short-lived radioisotope 86Rb can be used in future experiments to simulate 137Cs. Results and conclusions from these experiments are presented and compared to results from laboratory-scale experiments performed on small coupons. In general, both hydrogel decontamination products may be used as a viable option to decontaminate large surfaces in a real-world event, removing between 30% to 90% of the contamination, depending on the surface type and porosity. However, both products may leave behind absorbed contamination that will need to be addressed at a later stage. Yet, the likelihood of resuspension through use of these products is reduced.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030910-030910-7. doi:10.1115/1.4036434.

Detection of microscopic fission track (FT) star-shaped clusters, developed in a solid state nuclear track detector (SSNTD) by etching, created by fission fragments emitted from particles of fissile materials irradiated by neutrons, is a key technique in nuclear forensics and safeguards investigation. It involves scanning and imaging of a large area, typically 100–400 mm2, of a translucent SSNTD (e.g., polycarbonate sheet, mica, etc.) to identify the FT clusters, sparse as they may be, that must be distinguished from dirt and other artifacts present in the image. This task, if done manually, is time consuming, operator dependent, and prone to human errors. To solve this problem, an automated workflow has been developed for (a) scanning large area detectors, in order to acquire large images with adequate high resolution, and (b) processing the images with a scheme, implemented in ImageJ, to automatically detect the FT clusters. The scheme combines intensity-based segmentation approaches with a morphological algorithm capable of detecting and counting endpoints in putative FT clusters in order to reject non-FT artifacts. In this paper, the workflow is described, and very promising preliminary results are shown.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030911-030911-6. doi:10.1115/1.4036698.

In this paper, a method is presented for the detection of special nuclear materials (SNMs) in shielded containers, which is both sensitive and applicable under field conditions. The method uses an external pulsed neutron source to induce fission in SNM and subsequent detection of the fast prompt fission neutrons. The detectors surrounding the container under investigation are liquid scintillation detectors able to distinguish gamma rays from fast neutrons by means of pulse shape discrimination method (PSD). One advantage of these detectors, besides the ability for PSD analysis, is that the analog signal from a detection event is of very short duration (typically few tens of nanoseconds). This allows the use of very short coincidence gates for the detection of the prompt fission neutrons in multiple detectors, while benefiting from a low background coincidence rate, yielding a low detection limit. Another principle advantage of this method derives from the fact that the external neutron source is pulsed. By proper time gating, the interrogation can be conducted by epithermal source neutrons only. These neutrons do not appear in the fast neutron signal following the PSD analysis, thus providing a fundamental method for separating the interrogating source neutrons from the sample response in the form of fast fission neutrons. This paper describes laboratory tests with a configuration of eight detectors in the Pulsed Neutron Interrogation Test Assembly (PUNITA). Both the photon and neutron signature for induced fission is observed, and the methods used to isolate these signatures are described and demonstrated.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030912-030912-3. doi:10.1115/1.4036435.

A multiline neutron source can be produced by using a metallic 232Th filter in conjunction with a white neutron source. The multiline spectrum consists of ∼20 relatively strong intensity lines ranging from 10 to 4000 eV. It is shown that the width of each neutron line of the spectrum is strongly dependent on the absorber thickness. This neutron source is useful for accurate cross section measurements with precise neutron energies. The optimum thickness of the 232Th absorber, which was found to yield a sharp multiline spectrum throughout the above energy range, was found to be ∼14 cm.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030913-030913-6. doi:10.1115/1.4035725.

This study suggests a new approach to diffusion bonding (DB) 316L stainless steel: a low-pressure procedure that includes a nickel interlayer. In this approach, relatively lower pressure is applied to the sample before the DB process, in contrast to the usual approach in which higher pressure is applied during the DB process. This new procedure was tested on mock-up 316L stainless steel tube-to-tubesheet joints, which simulated similar joints in coiled-tube heat-exchanger applications. This study confirms that the new procedure meets the overall success criteria, namely, a pull-out force exceeding the force required for tube rupture. It also shows that the DB joint is improved by the use of a Ni interlayer; the joint strength increased by approximately 33% for a 0.25 μm Ni interlayer and by approximately 18% for a 5 μm Ni interlayer. The joint cross sections were qualitatively examined using optical microscopy (OM) and scanning electron microscopy (SEM); the observations suggest that only portions of the interface were diffusion bonded, as a result of the low-pressure procedure and the surface roughness (due to the sample fabrication). The portions that were diffusion bonded, though, were sound, as characterized by the fact that the steel grains grew through the interface line to create a continuous metallographic structure.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030914-030914-3. doi:10.1115/1.4036432.

The reaction of fluoride ions with alumina was found to strongly depend on the concentration of fluoride ions in the aqueous solution. At low concentrations ([fluoride ions] < 0.1 mol/l in the case of potassium fluoride), the aqueous concentration of aluminum ions is relatively high as measured by using inductively coupled plasma optical emission spectroscopy (ICP-OES), and the aluminum oxide dissolves as a fluoride complex. At high concentrations of fluoride ([fluoride ions] > 0.5 mol/l in the case of potassium fluoride), a new structure is formed on the alumina surface involving fluoride, aluminum, potassium, and oxygen (in the case of potassium fluoride). The structure was characterized by using X-ray powder diffraction (XRD), scanning electron microscope (SEM), and energy-dispersive X-ray spectroscopy (EDS). The resulting structure improved the passivation of alumina, the solubility of aluminum ions decreasing compared to the untreated alumina. Aluminum surfaces that were fluoride-treated showed a better resistance to dissolution in both acidic and basic media.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030915-030915-3. doi:10.1115/1.4035568.

Nuclear facilities and in particular nuclear reactors are designed to withstand ground acceleration due to earthquake and to maintain the structures, systems, and components (SSC) safety function. The three main safety functions are shutdown, removal of residual heat, and indications of the first two. Achieving these three functions will be called a “safe path.” The Israel Research Reactor 2 (IRR2) has numerous diverse safe paths to ensure the availability of these three safety functions. Each safe path has an associated level of resistance to ground acceleration. A project for increasing the robustness of equipment to ground acceleration was initiated in order to improve the safety of the reactor and to comply with regulatory guidelines.

Topics: Safety , Robustness , Heat
Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030916-030916-2. doi:10.1115/1.4036137.

Recent recommendation, by the International Commission on Radiological Protection (ICRP), to reduce the dose limit to the lens of the eye by almost an order of magnitude, has increased substantially the need to monitor this dose, i.e., Hp(3), with an accurate dosimeter. Since such dosimeter has not yet been validated and fully implemented, present monitoring of the dose to the lens of the eye is based on the measurement of Hp(10) and Hp(0.07) values and using conservative assumptions, which lead to an overestimate of the required dose. A new method to estimate Hp(3) using measured values of Hp(10) and Hp(0.07) has been suggested, which is more accurate and less conservative. This method could be used for routine monitoring and also in cases where there is a need to reconstruct historical doses to the lens of the eye, such as in law court claims of workers that were diagnosed with cataract.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030917-030917-2. doi:10.1115/1.4036462.

The study aimed to determine how the effective dose (ED) in lumbar spine X-ray examinations is influenced by patient positioning considering the X-ray tube heel effect. The study used Monte Carlo simulation of the effective dose. Using the heel effect, positioning of the patient in the head to anode direction reduces the effective dose by 5% when compared with the head to cathode positioning.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):030918-030918-3. doi:10.1115/1.4036463.

In Israel, a single regulatory body for radiation protection does not exist. Instead, its responsibilities and functions are shared between five government ministries and agencies. Accordingly, the existing legal framework for radiation safety is of a very heterogeneous nature. It is made of laws, acts, orders, and regulations enacted during different periods, according to different principles. Moreover, some of the provisions of those legal instruments are obsolete or quote obsolete documents. The Standard for Radiation Protection (SRP) of the Israel Atomic Energy Commission (IAEC) was recently updated on the basis of the latest version of the International Atomic Energy Agency (IAEA) International Basic Safety Standards (BSS). It is proposed that the SRP of the IAEC serves as a model for a comprehensive framework law that would be structured in a similar manner, i.e., a division into three parts according to the three different types of exposure situation (planned, emergency, existing) defined by the International Commission on Radiological Protection (ICRP) and a subdivision of each part according to relevant exposure categories (occupational, public, medical). The adoption of such a structure would ensure that no aspect of radiation protection is left untreated.

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2017;3(3):031001-031001-7. doi:10.1115/1.4035549.

Corrosion behavior of Inconel 625 and Incoloy 800H, two of the candidate fuel cladding materials for Canadian supercritical water-cooled reactor (SCWR) designs, was evaluated by exposing the metals to supercritical water (SCW) in the University of New Brunswick’s flow loop. A series of experiments were conducted over a range of temperatures between 370 °C and 600 °C, and the corrosion rates were evaluated as the weight change of the materials over the exposure time (typical experiments measured the weight change at intervals of 100, 250, and 500 h, with some longer-term exposures included). Scanning electron microscopy (SEM) was used to examine and quantify the oxide films formed during exposure and the corrosion mechanisms occurring on the candidate metals. Data from in-house experiments were used to create an empirical kinetic equation for each material that was then compared to literature values of weight change. Dissolved oxygen concentrations varied between experimental sets, but for simplicity were ignored since the effect of dissolved oxygen has been demonstrated to be a minor secondary effect. Activation energies for the alloys were determined with Inconel 625 and Incoloy 800 H showing a distinct difference between the low-temperature electrochemical corrosion (EC) mechanism and direct high-temperature chemical oxidation (CO). The results were modeled using these separate effects showing dependence on the bulk density and dielectric constant of the supercritical water through the hydrogen ion concentration.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031002-031002-7. doi:10.1115/1.4035335.

Hitachi-GE developed a 300 MWel class modular simplified and medium small reactor (DMS) concept, and the DMS was originally designed for generating electricity only. In this study, the feasibility of a cogeneration DMS plant which supplies both electricity and heat is under investigation. The thermal performance of the DMS plant without or with low-, medium-, or high-temperature thermal utilization (TU) applications is evaluated by numerical simulations. The results show that the electricity generated reduces as the heating requirement of TU application becomes higher. Furthermore, the economic performance of the cogeneration DMS plant is compared with another two integrated systems: (i) DMS plus electric boilers and (ii) DMS plus natural gas boilers, for those three TU applications in Canada. The results illustrate that the DMS plus natural gas boilers system are most economic if there is no carbon tax, but with high-CO2 emissions (up to 180 kton per year). The cogeneration plant performs best as the carbon tax increases up to $40/ton. The cogeneration DMS plant is a promising scheme to supply both electricity and heat simultaneously in the economic-environmental point of view.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031003-031003-11. doi:10.1115/1.4035853.

This study investigates heat transfer characters of a volumetrically heated melt pool in LWR lower plenum. Experimental restrictions on prediction reliability are discussed. These restrictions include cooling boundary conditions, vessel geometries, and simulant melt selection on general and localized heat transfer. A survey of existing heat transfer correlations derived from individual experimental definitions is presented. The inconsistency in parameter definitions in Nu–Ra correlations is discussed. Furthermore, the discrepancy of upward Nu depending on the existence of crust is stressed. Several serials of experiments with different combinations boundary condition of external cooling and top cooling were performed in LIVE3D and LIVE2D facilities. The experiments were conducted with simulants with and without crust formation. The influences of cooling boundary conditions, the vessel geometry, and the simulant material on overall heat transfer as well as on heat flux distribution are analyzed. This paper provides own explanations about the discrepancies among the exiting heat transfer correlations and recommends the most suitable descriptions of melt pool heat transfer under different accident management strategies.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031004-031004-9. doi:10.1115/1.4035934.

Due to powerful geometry treatment capability, method of characteristics (MOC) currently becomes one of the best method to solve neutron transport equation. In MOC method, boundary condition treatment, complex geometry representation, and advanced acceleration method are the key techniques to develop a powerful MOC code to solve complex problem. In this paper, we developed a powerful MOC module, which can treat different boundary conditions with two methods. For problems with special border shapes and boundary condition, such as rectangle, 1/8 of square, hexagon, 1/3 of hexagon, 1/6 of hexagon problems with reflection, rotation, and translation boundary condition, the MOC module adopts periodic tracking method, with which rays can return to start point after a certain distance. For problems with general border shapes, the MOC module uses ray prolongation method, which can treat arbitrary border shapes and boundary conditions. Meanwhile, graphic user interface based on computer aided design (CAD) software is developed to generate the geometry description file, in which geometry is represented by “lines and arcs” method. With the graphic user interface, the geometry and mesh can be described and modified correctly and fast. In order to accelerate the MOC transport calculation, the generalized coarse mesh finite difference (GCMFD) is used, which can use irregular coarse mesh diffusion method to accelerate the transport equation. The MOC module was incorporated into advanced neutronics lattice code KYLIN-2, which was developed by Nuclear Power Institute of China (NPIC) and used to simulate the assembly of current pressurized water reactor (PWR) and advanced reactors, to solve the transport equation with multigroup energy structure in cross sections database. The numerical results show that the KYLIN-2 code can be used to calculate 2D neutron transport problems in reactor accurately and fast. In future, the KYLIN-2 code will be released and gradually become the main neutron transport lattice code in NPIC.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031005-031005-10. doi:10.1115/1.4035550.

The aim of this paper is to summarize authors' experience in adaptation of an existing plant-specific VVER-1000/V320 model for simulation of a rare example of a Kalinin 3 nuclear power plant (NPP) transient of “switching-off of one of the four operating main circulation pumps at nominal reactor power” with an asymmetric core configuration. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. Simulation results concerning fuel assembly (FA) power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system (ICMS). Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superficial description of the reference unit. In such a case, an approach based on a “generic” V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion. Some of the most important lessons learned are as follows. (1) individual characteristics of all the main circulation pumps and the reactor coolant loops are quite important for the quality of simulation and should be accounted for in the model; (2) variations in fuel assembly characteristics should be accounted for not only in terms of macroscopic cross section library but also in terms of local pressure loss coefficients and mixing factors in the case of mixed core loads; (3) comprehensive plant-specific model of dynamic response of instrumentation and control (I&C) systems is a necessity; dynamic characteristics of individual measurement channels (nuclear instrumentation, pressure, temperature) should be accounted for; and (4) comprehensive plant-specific model of balance-of-plant equipment, instrumentation, and control is a necessity. Above requirements impose a difficult task to comply with. Nevertheless, any individual nuclear power unit is supposed to maintain a detailed design database and data requirements for plant-specific model development should be considered.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031006-031006-7. doi:10.1115/1.4035974.

In the National Aeronautics and Space Administration (NASA) Design Reference Architecture 5.0 (DRA 5.0), fission surface power systems (FSPS) are described as “enabling for the human exploration of Mars.” This study investigates the design of a power conversion system (PCS) based on supercritical carbon dioxide (sCO2) Brayton configurations for a growing Martian colony. Various configurations utilizing regeneration, intercooling (IC), and reheating are analyzed. A model to estimate the mass of the PCS is developed and used to obtain a realistic mass-optimized configuration. This mass model is conservative, being based on simple concentric tube counterflow heat exchangers and published data regarding turbomachinery masses. For load following and redundancy purposes, the FSPS consists of three 333 kWe reactors and PCS to provide a total of 1 MWe for 15 years. The optimal configuration is a sCO2 Brayton cycle with 60% regeneration and two stages of intercooling. The majority of the analyses are performed in matlab, with certain data provided by a comsol multiphysics model of part of a low-enriched uranium (LEU) ceramic metallic (CERMET) reactor core.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):031007-031007-8. doi:10.1115/1.4036027.

Conceptual molten salt breeder reactor (MSBR) is under development in Bhabha Atomic Research Centre (BARC) with long-term objective of utilizing abundant thorium available in India. It is based on molten salts, which acts as fuel, blanket, and coolant for the reactor. LiF–ThF4 (77.6–22.4 mol %) is proposed as a blanket salt for Indian MSBR. A laboratory scale molten salt natural circulation loop (MSNCL) named molten active fluoride salt loop (MAFL) has been setup for thermal-hydraulic, material compatibility, and chemistry control studies. Steady-states and transient experiments have been performed in the operating temperature range of 600–750 °C. The loop operates in the power range of 250–550 W. Steady-state correlation given for natural circulation flow in a loop is compared with the steady-state experimental data. The Reynolds number was found to be in the range of 57–114. Computation fluid dynamics (CFD) simulation has also been performed for MAFL using openfoam code, and the results are compared with the experimental data generated in the loop. It has been found that predictions of openfoam are in good agreement with the experimental data. In this paper, features of the loop, its construction, and the experimental and numerical studies performed are discussed in detail.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2017;3(3):034501-034501-3. doi:10.1115/1.4035565.

Various questions can be examined when discussing safety in general. Among these, some key issues are the attitude toward risk and its acceptance, the ways of identifying, analyzing, and quantifying risks, and societal factors and public opinion toward risks. The identification and quantification of risks are central in the regulatory framework and decision making and will be the focus of this article. Various approaches have been used for safety analysis over the years. This paper will survey some of the central attitudes in the nuclear reactor regulation philosophy and discuss the historical background surrounding them. Among these, we mention the “defense-in-depth” approach, the design basis accident (DBA), and beyond design basis accident (BDBA) analyses and discuss the rather subjective nature of their associated decision making. We maintain that it has long been recognized that the natural approach that comes out of the scientific method of inquiry is the probabilistic one, which in today's tools is conducted through the probabilistic safety analysis (PSA) method. This approach unlike the deterministic one, which produced concepts like DBA and defense-in-depth, enables us to put risks into context and to compare different risks such as those posed by different activities in particular or by other industries in general. It has consequently been gaining wide acceptance in regulatory bodies around the world as an effective tool in the inspection and regulation of nuclear reactors. Yet, it is also recognized that despite significant development over the past few decades, PSA still suffers from some well-known deficiencies. Its main benefit at this point is its contribution to identification and prioritization of design features, maintenance, management, and quality assurance (QA) important to safety. PSA has mostly been used in the nuclear power industry, but in recent years it has also started to be incorporated in research reactor (RR) safety analysis, and we therefore cover the subject of PSA usage for this purpose as well.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(3):034502-034502-6. doi:10.1115/1.4036354.

Inspections of pressure tubes in CANDU® reactors are a key part of maintaining safe operating conditions. The current inspection system, the channel inspection and gauging apparatus for reactors (CIGAR), performs the job well but is limited by the fact that it can only inspect one channel at a time. A multidisciplinary team is currently developing a novel robotic inspection system. As part of this work, a Monte Carlo N-particle (MCNP) model has been developed in order to predict the dose rates that the improved inspection system will be exposed to and, from this, predict the component lifetime. This MCNP model will be capable of predicting in-core dose rates at any location within the reactor, and as such could be used for other situations where the in-core dose rate needs to be known. Based on estimates from this model, it is expected that at 7 days after shutdown, the improved inspection system could survive in core for approximately 7 h, providing it uses a tungsten shield 2.5 cm in thickness around the integrated circuit components. This is expected to be sufficient to perform a single inspection.

Commentary by Dr. Valentin Fuster

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