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### Editorial

ASME J of Nuclear Rad Sci. 2016;3(1):010201-010201-2. doi:10.1115/1.4035065.
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The ASME Journal of Nuclear Engineering and Radiation Science is celebrating its second year on the market. On this occasion, I would like to congratulate all members of our Journal Board, reviewers (their names are listed in this issue), authors, and readers. Below are the latest Journal statistics, changes, and plans for the future. Also, we would like to welcome the new Associate and Guest Editors:

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):010202-010202-1. doi:10.1115/1.4035080.
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The Journal Editorial Board would like to express its great appreciation to all the reviewers for their valuable comments and suggestions toward papers published in 2016 and papers to be published in 2017. Without your hard work and expertise, it would be impossible to run our Journal and to keep Journal papers at such a high international level!

Commentary by Dr. Valentin Fuster

### Review Article

ASME J of Nuclear Rad Sci. 2016;3(1):010801-010801-13. doi:10.1115/1.4035177.
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This paper presents a brief history of the ASME nuclear engineering division (NED) over the past 60 years. The technical interest of the Division naturally mirrored the main stages in nuclear-technology development and growth of the industry. This is reflected in how the NED evolved the technical content of its major international conferences, the role of ASME standards, and the value of international cooperation. The International Conference on Nuclear Engineering (ICONE) series of conferences is covered as it occupies a special place in the ASME and NED history. The paper covers the birth and growth of the Division, its leadership, publications, and its flagship student program. It concludes with activities the NED is working on for the distant future.

Commentary by Dr. Valentin Fuster

### Research Papers

ASME J of Nuclear Rad Sci. 2016;3(1):011001-011001-8. doi:10.1115/1.4032999.

A vacuum-induced salt transfer and storage (VISTAS) system is being evaluated to improve transfer and storage of molten electrorefiner (ER) salts at Idaho National Laboratory (INL). Salt is transferred by vacuum through a heated drawtube into a storage container. To control salt flow, a redundant level switch triggered by salt thermal conductivity and a preset temperature threshold activate a solenoid, stopping argon supply to the vacuum pump. A fail-safe cooling coil freezes the salt, halting its flow if the level switch malfunctions. The VISTAS system allows safe, timely salt transfer and reduces the storage footprint of current salt-removal methods.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011002-011002-8. doi:10.1115/1.4035046.

In this study, transient analysis code of SCWRs (TACOS), with the ability of simulating transients or accidents under both supercritical water (SCW) conditions and subcritical water conditions, has been developed with fortran 90 language, and simulation has been performed to the European SCWR fuel qualification test (FQT) system. The semi-implicit finite difference technique was adopted for the solution of coolant dynamic behavior in the loop. Furthermore, an illustration of numerical solution for the heat structure model and other models was presented. The code TACOS is then applied to simulate the Edward-O’Brian blow-down experiment to evaluate its capacity in simulating the fast blow-down progress. Therefore, the design basis accidents (DBAs) with the trans-critical transient were investigated for the SCWR-FQT system. The results by TACOS indicate that the SCWR-FQT with the existing safety system can be cooled effectively.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011003-011003-9. doi:10.1115/1.4035047.

Available computational fluid dynamics (CFD) predictions of pressure distributions in the vertical bypass flow between blocks in a prismatic gas-cooled reactor (GCR) have been analyzed to deduce apparent friction factors and loss coefficients for nuclear engineering systems and network codes. Calculations were performed for vertical gap spacings “$s$” of 2, 6, and 10 mm — representing 1, 3, and 5 mm in a GCR design, horizontal gaps between the blocks of 2 mm and two flow rates, giving a range of vertical gap Reynolds numbers $ReDh$ of about 40–5300. The present focus is on the examination of the flow in the vertical gaps. Horizontal gaps are treated in CFD calculations but their flows are not examined. Laminar predictions of the fully developed friction factor $ffd$ were about 3–10% lower than the classical infinitely wide channel. In the entry region, the local apparent friction factor was slightly higher than the classic idealized case, but the hydraulic entry length $Lhy$ was approximately the same. The per cent reduction in flow resistance was greater than the per cent increase in flow area at the vertical corners of the blocks. The standard $k–ϵ$ turbulence model was employed for flows expected to be turbulent. Its predictions of $ffd$ and flow resistance were significantly higher than direct numerical simulations (DNS) for the classic case; the value of $Lhy$ was about 30 gap spacings. Initial quantitative information for entry coefficients and loss coefficients for the expansion–contraction junctions between blocks is also presented. The present study demonstrates how CFD predictions can be employed to provide integral quantities needed in systems and network codes.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011004-011004-10. doi:10.1115/1.4034975.

We present a method to simultaneously pressurize fluid filled containers from outside and within, results of experiments with temporary 2 h of fluid precompression followed by overpressure removal before testing for cavitation strength and sensitivity to neutron radiation of multi-mL quantities of widely used unfiltered and undegassed liquids, such as water, ethanol, and dodecane (a surrogate jet fuel), enclosed within containers using glass, epoxy, and steel. We found that in contrast to prior methods involving laborious degassing and purification, a straightforward one-step approach using only a modest 2 h precompression treatment at a pressure of 0.7+ MPa enabled us, reproducibly, to reach directly the highest attainable “negative” (subvacuum) pressures attainable in our apparatus ($−0.7 MPa$)—enabling efficient sensitivity to neutron-type radiation. Cavitation strength results are explained on theoretical grounds. However, surprisingly using the technique of this paper, the 2-h precompressed (unfiltered, undegassed) fluid also retained memory of this property, after the overpressure was removed, even 3 months later—thereby suggesting that active cavitation nuclei suppression can be extended to long periods of time. Successful results for cavitation suppression (in the absence of ionizing radiation) through $−0.7 MPa$ were also attainable for fluids in simultaneous contact with a combination of glass, steel, and epoxy surfaces. The relative importance of cavitation strength retention at liquid–wall interfaces versus within the bulk of the fluids is reported along with implications for high-efficiency nuclear particle detection and spectroscopy, and nuclear fission water reactor safety thermal-hydraulic assessments for blowdown transients.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011005-011005-6. doi:10.1115/1.4033813.

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core temperature of the HTTR is too high. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011006-011006-14. doi:10.1115/1.4034569.

The chemical head loss experiment (CHLE) program has been designed to acquire realistic material release and product formation in containment under post-loss of coolant accident (LOCA) conditions and their impact on the measured head loss through the use of modified debris beds developed at the University of New Mexico (UNM). A full-scale water chemistry test was conducted under Vogtle containment chemistry conditions to determine the release of these materials and the resulting head loss response of the formed products within the emergency core cooling system (ECCS) under prototypical chemical conditions. The test was designed to investigate material corrosion with the presence of excess aluminum and a nonprototypical temperature profile (80 °C for 120 h) to promote the production of aluminum precipitates. The head loss measured within the first 72 h of the test either surpassed the operational limits of the equipment or caused a failure within the system. The increase in head loss is not attributed to the formation of in situ precipitates but to a physical reaction of the epoxy used in constructing the debris beds to the local chemistry during the early stages of the test.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011007-011007-7. doi:10.1115/1.4034477.

Small modular nuclear reactors (SMRs) are designed for long-term operation with minimum outages and for possible deployment in remote locations. To achieve this operational goal, the SMRs may require remote and continuous monitoring of performance parameters that contribute to operation and maintenance. This feature is also important in monitoring critical parameters during severe accidents and for postaccident recovery. Small integral light water reactors have in-vessel space constraints, and many of the traditional instrumentation are not practical in these systems. To investigate this issue, analytical and experimental researches were carried out using a flow test loop to characterize the relationship among process variables (flow rate and pressure) and pump motor signatures. The findings of this research are presented, with implications in relating electrical signatures to pump parameters. The relationship between the electrical signatures and the process variables is discussed with reference to the experimental results. The results of this work may be used for monitoring process variables in small modular reactor systems.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011008-011008-7. doi:10.1115/1.4034974.

For the protection of a coastal nuclear power plant (NPP) against external flooding hazard, the risks caused by natural events have to be taken into account. In this article, a methodology is proposed to analyze the risk of the typical natural event in China (Typhoon). It includes the simulation of the storm surge and the strong waves due to its passage in Chinese coastal zones and the quantification of the sequential overtopping flow rate. The simulation is carried out by coupling two modules of the hydraulic modeling system TELEMAC-MASCARET from ElectricitØ de France (EDF), the shallow water module (TELEMAC2D) and the spectral wave module (TOMAWAC). As an open-source modeling system, this methodology could still be enriched by other phenomena in the near future to ameliorate its performance in safety analysis of the coastal NPPs in China.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011009-011009-6. doi:10.1115/1.4032780.

The supercritical water reactor (SCWR) is one of the Generation IV designs. The SCWR is characterized by its high efficiency, low waste production, and simple design. Despite the suitable properties of supercritical water as a coolant, its physicochemical properties change sharply with pressure and temperature in the supercritical region. For this reason, there are many doubts about how changes in these variables affect the behavior of the materials to general corrosion or to specific types of corrosion such as stress corrosion cracking (SCC). Austenitic stainless steels are candidate materials to build the SCWR due to their optimum behavior in the light water reactors (LWRs). Nevertheless, their behavior under the SCWR conditions is not well known. First, the objective of this work was to study the SCC behavior of austenitic stainless steel 316 type L in deaerated supercritical water at $400°C/25 MPa$ and 30 MPa and $500°C/25 MPa$ to determine how variations in pressure and temperature influence its behavior with regard to SCC and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Second, the oxide layer formed at $400°C/30 MPa/<10 ppb O2$ was analyzed to gain some insight into these processes.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011010-011010-13. doi:10.1115/1.4032874.

For CANada Deuterium Uranium (CANDU) nuclear reactors, the characterization of the moderator thermal-hydraulic behavior under both normal and abnormal operating conditions constitutes an important safety issue. For normal operating conditions, the flow temperature distribution may produce changes on the heavy-water mass density, which in turn may affect the neutron moderation rate. Consequently, these variations influence the thermal neutron flux distribution in the reactor core. Therefore, it is fundamental to know all possible moderator flow configurations as well as the corresponding temperature distributions. In particular, any possibility of a dryout at the external wall of the Calandria tubes and consequently excessive temperature excursions must be prevented. Within this framework, this paper presents detailed two-dimensional (2D) numerical steady-state simulations for a wide range of flow conditions. Both the accuracy of the numerical approximations and the validity of some physical models used in computational fluid dynamic (CFD) codes are assessed. The numerical results are then used to construct a cartographical representation of moderator flows in CANDU-6 reactors. To support the existence of coherent flow asymmetries and eventually flow-structure oscillations, the present numerical results are also compared with the previous ones obtained using a porous medium-modeling approach.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011011-011011-5. doi:10.1115/1.4034571.

Deterministic and Monte Carlo methods are regularly employed to conduct lattice calculations. Monte Carlo methods can effectively model a large range of complex geometries and, compared to deterministic methods, they have the major advantage of reducing systematic errors and are computationally effective when integral quantities such as effective multiplication factor or reactivity are calculated. In contrast, deterministic methods do introduce discretization approximations but usually require shorter computation times than Monte Carlo methods when detailed flux and reaction-rate solutions are sought. This work compares the results of the deterministic code DRAGON to the Monte Carlo code Serpent in the calculation of the reactivity effects for a pressurized heavy water reactor (PHWR) lattice cell containing a 37-element, natural uranium fuel bundle with heavy water coolant and moderator. The reactivity effects are determined for changes to the coolant, moderator, and fuel temperatures and to the coolant and moderator densities for zero-burnup, mid-burnup [$3750 MWd/t(U)$] and discharge burnup [$7500 MWd/t(U)$] fuel. It is found that the overall trend in the reactivity effects calculated using DRAGON match those calculated using Serpent for the burnup cases considered. However, differences that exceed the amount attributable to statistical error have been found for some reactivity effects, particularly for perturbations to coolant and moderator density and fuel temperature.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011012-011012-6. doi:10.1115/1.4034711.

For industrial or domestic applications, the wide range of use of pleated filters makes the understanding of their airflow behavior a major issue for designer and users. In all industrial installations dealing with radioactive matter, the containment of pollutants must be ensured. High-efficiency particulate air (HEPA) filters are used as the last purification stage before the air is rejected in the environment. These filters can be used either alone, in the case of nonsensible installation, or coupled with other filtration devices disposed before it where contamination level could be important. The prediction of their pressure drop is very important in nuclear safety to be able to anticipate any dysfunction or rupture of these devices. It has been observed that geometry of the medium has an influence on the pressure drop of a pleated filter. In the case of HEPA filters, no convincing explanation has been brought to explain their airflow behavior. The pressure drop evolution of the filter during the clogging remains difficult to explain by assuming constant pleat geometry. Some studies show that deformation occurs during the filter use, which could induce an increase of the available volume in the pleat and a reduction of the efficient filtration surface. The increase in computation capacity introduces nowadays the possibility to perform complex simulation, taking into account the effect of fluids on sensible devices. This can be the case for simple structural analysis or for more complex analysis such as vibration induced by gas or fluid flow. It is mostly applied to avoid breaking or deformation of safety devices, and this can also be applied to anticipate the fluid behavior of some special devices such as filters. In classical filtration application, properties of the filter are coupled with particle deposition (e.g., changes in geometry and permeability depend on the thickness of the deposit). The studies concerning mechanical properties of filters are mainly performed for liquid filtration and clean filters. For pleated filters, the complexity of this kind of analysis remains the modification of the link between geometry, pressure drop, mechanical strength, and particle transport and accumulation inside the pleat. As a first approach, it has been chosen to combine an experimental and a numerical approach to improve the understanding of filter behavior. In this paper, the pleat deformation will be investigated using a direct nonintrusive laser measurement performed on a single pleat experiment. The rate of filtration surface lost will be estimated using these data and taken into account to evaluate the pressure drop against the filtration velocity. Results obtained show that the pleat deformation is an important parameter, which influences the geometry of the pleat.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011013-011013-4. doi:10.1115/1.4033812.

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR), which was constructed in Japan. The operating data of HTTR with burn-up to about 370 EFPD (effective full-power days), which are very important for the development of HTGRs, have been collected in both zero-power and powered operations. In the aspects of code validation, the detailed prediction of temperature distribution in the core makes it difficult to validate the calculation code because of difficulty in measuring the core temperature directly in powered operation of the HTTR. In this study, the measured data of the control rod position, while keeping the temperature distribution in the core uniform at criticality in zero-power operation at the beginning of each operation cycle were compared with the calculated results by core physics design code of the HTTR. The measured data of the control rod position were modified based on the core temperature correlation. At the beginning of burn-up, the trends of burn-up characteristics are slightly different between experimental and calculation data. However, the calculated result shows less than 50 mm of small difference (total length of control rod is 4060 mm) to the measured one, which indicates that the calculated results appropriately reproduced burn-up characteristics, such as a decrease in uranium-235, accumulation in plutonium, and decrease in burnable absorber.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011014-011014-8. doi:10.1115/1.4034573.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011015-011015-8. doi:10.1115/1.4034478.

This work focuses on the safety analysis of a typical pressurized water reactor (PWR) for reactivity-initiated transients. These transients result from withdrawal of six sets of groups of control rods that may occur under control systems or other faults. NEA/OECD PWR benchmark is considered for the study. A 3D space–time kinetics code, “TRIKIN” (neutronic and thermal-hydraulics coupled code) is used to account for local changes in the neutron flux. These local changes in the neutron flux affect the total reactivity, local power, and temperature distribution. The safety parameters are the usual 3D radial power distribution, flux tilt, axial heat flux for the peak channel, and radial peak central line temperature profiles over the horizontal plane. These safety parameters studied in the incident progression up to reactor SCRAM level. The minimum departure from the nucleate boiling ratio (MDNBR) has been investigated quantitatively for all six cases. The case that gives maximum drop in MDNBR at SCRAM level is identified and its consequences are discussed. The study is of high importance in revealing the importance of grouping of control rods’ configurations, providing insight in developing strategy for designing the configuration and reactivity worth of groups of control rods and local/global reactor control systems for large-size PWRs.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011016-011016-4. doi:10.1115/1.4033087.

The Balakovo Nuclear Power Plant operating in Russia has four operating units of a VVER-1000 type. The first unit was connected to the grid on Dec. 28, 1985, and put in commercial operation on May 23, 1986. According to modern requirements of the Russian Nuclear Regulatory Authority, the full-scope probabilistic safety assessment (PSA) must be performed to extend the license for operation beyond an initial 30-year lifetime of the unit. The paper presents the results of the Level 1 PSA covering internal initiating events for power and shutdown operational plant states, internal hazards (fires and floods), and external hazards, including natural and man-made events, in particular, seismic impact, and aircraft crash.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011017-011017-8. doi:10.1115/1.4033398.

As a nongreenhouse gas-emitting source, the benefits of nuclear as a main power-generation alternative are yet to be fully explored; part of the reason is due to the significant implementation costs. However, with cycle efficiencies of 45–50% in current studies, it can be argued that the long-term benefits outweigh the initial costs, if developed under the Generation-IV (Gen-IV) framework. The main objective of this study is to analyze the effects of pressure and temperature ratios (TRs) including sensitivity analyses of component efficiencies, ambient temperature, component losses and pressure losses on cycle efficiency and specific work. The results obtained indicate that pressure losses and recuperator effectiveness have the greatest impact on cycle efficiency and specific work. The analyses intend to aid development of the simple cycle recuperated (SCR) and intercooled cycle recuperated (ICR) cycles, applicable to gas-cooled fast reactors (GFRs) and very-high-temperature reactors (VHTRs), in which helium is the coolant.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011018-011018-6. doi:10.1115/1.4032599.

This paper describes the design and use of a new irradiation facility for the LVR-15 nuclear research reactor. The CHOUCA MT irradiation rig was produced in France according to a design of the ÚJV Group (ÚJV Řež and Research Center Řež). There are six heating sections situated along the rig, each instrumented and controlled by its own thermocouple. The rig’s insulation layers ensure a balanced temperature in irradiated specimens along its entire length. The specimen holder is 55.9 mm in diameter and 320 mm long. The CHOUCA MT rig can be repetitively irradiated in different positions within the reactor core, depending on irradiation condition requirements. The CHOUCA MT rig expands the possibilities of radiation research in the ÚJV Group.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011019-011019-7. doi:10.1115/1.4034568.

The iron cross-section in thermal regions influences the thermal neutron flux prediction in steel structural components of reactors and also in regions adjoining them. The thermal neutron flux level is proportional to pin power density in fuel. This quantity is an important criterion reflected in limits and conditions of reactor operation. The new power density evaluation shows notable, well distinguishable discrepancy between calculations realized using the CENDL-3.1 nuclear data library and experimentally determined pin power density in boundary rows of pins. All experiments were carried out in a water–water energetic reactor (VVER-1000) transport mock-up placed in the LR-0 reactor.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011020-011020-10. doi:10.1115/1.4034061.

Recent trends in nuclear reactor performance and safety analyses increasingly rely on multiscale multiphysics computer simulations to enhance predictive capabilities by replacing conventional methods that are largely empirically based with a more scientifically based methodology. Through this approach, one addresses the issue of traditionally employing a suite of stand-alone codes that independently simulate various physical phenomena that were previously disconnected. Multiple computer simulations of different phenomena must exchange data during runtime to address these interdependencies. Previously, recommendations have been made regarding various approaches for piloting different design options of data coupling for multiphysics systems (Seydaliev and Caswell, 2014, “CORBA and MPI Based “Backbone” for Coupling Advanced Simulation Tools,” AECL Nucl. Rev., 3(2), pp. 83–90). This paper describes progress since the initial pilot study that outlined the implementation and execution of a new distribution framework, referred to as “Backbone,” to provide the necessary runtime exchange of data between different codes. The Backbone, currently under development at the Canadian Nuclear Laboratories (CNL), is a hybrid design using both common object request broker architecture (CORBA) and message passing interface (MPI) systems. This paper also presents two preliminary cases for coupling existing nuclear performance and safety analysis codes used for simulating fuel behavior, fission product release, thermal hydraulics, and neutron transport through the Backbone. Additionally, a pilot study presents a few strategies of a new time step controller (TSC) to synchronize the codes coupled through the Backbone. A performance and fidelity comparison is presented between a simple heuristic method for determining time step length and a more advanced third-order method, which was selected to maximize configurability and effectiveness of temporal integration, saving time steps and reducing wasted computation. The net effect of the foregoing features of the Backbone is to provide a practical toolset to couple existing and newly developed codes—which may be written in different programming languages and used on different operating systems—with minimal programming effort to enhance predictions of nuclear reactor performance and safety.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):011021-011021-8. doi:10.1115/1.4034479.

The intercooled cycle (IC) as an alternative to the simple cycle recuperated (SCR) and intercooled cycle recuperated (ICR) is yet to be fully analyzed for the purpose of assessing its viability for utilization within Generation IV nuclear power plants (NPPs). Although the benefits are not explicitly obvious, it offers the advantage of a very high overall pressure ratio (OPR) in the absence of a recuperator. Thus, the main objective of this study is to analyze various pressure ratio configurations, the effects of varying pressure ratio including sensitivity analyses of component efficiencies, ambient temperature, component losses and pressure losses on cycle efficiency, and specific work of the IC, including comparison with the SCR and ICR. Results of comparison between the IC and the SCR and ICR derived that the cycle efficiencies are greater than the IC by $∼4%$ (SCR) and $∼6%$ (ICR), respectively. However, the pressure losses for IC are lower when compared with the SCR and ICR. Nonetheless, heat from the turbine exit temperature of the IC can be used in a processing plant including the possibility of higher turbine entry temperatures (TETs) to significantly increase the cycle efficiency in a bid to justify the business case. The analyses intend to bring to attention an alternative to current cycle configurations for the gas-cooled fast reactors (GFRs) and very-high-temperature reactors (VHTRs), where helium is the coolant. The findings are summarized by evaluating the chosen pressure ratio configurations against critical parameters and detailed comparison with the SCR and ICR.

Commentary by Dr. Valentin Fuster

### Technical Brief

ASME J of Nuclear Rad Sci. 2016;3(1):014501-014501-4. doi:10.1115/1.4035285.

This article conducts optimization and sensitivity analysis for the cold-end system of a nuclear power plant (NPP) with a sea water circulating system, using the minimum annual cost method by dynamic economical analysis. The following factors are taken into account for the optimization: the design of a condenser (cooling area and the parameters of cooling tubes), the scale of the cooling tower, the flow rate of circulation water, and the diameter of circulating pipes. In conclusion, the optimum scheme of the cold-end system is obtained for the double-back pressure turbine project. The variation trend of the annual cost with changing circulating water amount is also concluded. Finally, this article provides the sensitivity analysis of the feed-in tariff.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):014502-014502-4. doi:10.1115/1.4034570.

A spectrometric system was developed for spent fuel burnup evaluations at the LVR-15 research reactor, which employed highly enriched (36%) IRT-2M-type fuel. Such a system allows the measurement of fission product axial distribution by measuring certain nuclides, such as $Cs137$, $Cs134$, and their ratios, respectively. Within the paper, a comparison between experimental data provided by the spectrometric system and calculations in operational code called NODER is provided.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2016;3(1):014503-014503-7. doi:10.1115/1.4034680.

The key objective of the test blanket module (TBM) program is to develop the design technology for DEMO and future power-producing fusion reactors. The proposed first wall of the test blanket system (TBS) is a generalized concept for testing in ITER, an experimental fusion reactor under construction in France presently. The first wall of TBM (TBM FW) directly faces the plasma and is cooled by the first wall helium cooling system (FWHCS), which is considered as a critical component from an ITER safety point of view. The scope of this work comprises thermal hydraulic analysis of the FWHCS of a generalized TBS and the assessment of postulated initiating events (PIEs) on the ITER safety with the help of thermal-hydraulic code RELAP/SCDAPSIM/MOD4.0. The three reference accidents: in-vacuum vessel (VV) loss of coolant accident (in-vessel LOCA), ex-vessel LOCA, and loss of flow accident (LOFA) in FWHCS are selected for the safety assessment. This safety assessment addresses safety concerns resulting from FWHCS component failure, such as VV pressurization, TBM FW temperature profile, pressurization of port cell (PC) and Tokomak cooling water system vault annex (TCWS-VA), and passive decay heat removal capability. The analysis shows that the critical parameters are under the design limit and have large safety margins, in the investigated accident scenarios. A comparative analysis is also carried out with the previous results to validate the results.

Commentary by Dr. Valentin Fuster