Guest Editorial

ASME J of Nuclear Rad Sci. 2017;3(2):020301-020301-2. doi:10.1115/1.4035852.

The discovery of fission is roughly seven decades old, and the potential for large-scale power production from fission is now well established. Currently, nearly 430 nuclear power plants (NPPs) produce electricity around the globe contributing to around 11.2% of the total electricity production. With the present concern over the green house gas generation by burning fossil fuel and the rapid growth of electricity generation capacity in Asian and African nations, particularly in China and India, nuclear power capacity is expected to grow substantially in the near future. Among the operating nuclear power plants (NPPs), the water-cooled designs are the dominant ones due to simplicity in their design and economic competitiveness. After 25 years of Chernobyl accident when resurgence in the nuclear power program worldwide was imminent, the accident at Fukushima gave a serious blow to the confidence in nuclear safety, particularly under conditions of extreme natural disasters, such as earthquake and tsunami.

Commentary by Dr. Valentin Fuster

Special Section Papers

ASME J of Nuclear Rad Sci. 2017;3(2):020901-020901-9. doi:10.1115/1.4035856.

The pressurized heavy water reactor (PHWR) technology was conceived in Canada and has moved to several nations for commercial production of electricity. Currently, 49 power reactors operate with PHWR technology producing nearly 25 GWe. The technology is flexible for adopting different fuel cycle options which include natural uranium, different mixed oxide (MOX) fuel, and thorium. The technology has made substantial improvement in materials, construction, and safety since its inception. PHWRs have demonstrated excellent performance historically. Their safety statistics are excellent. Indian PHWRs also have shown economic competitiveness even in small sizes, thus providing an ideal design for new entrants. While the technology features of PHWRs are available even in textbooks, the objective of this paper is to highlight the historical development and salient features, and innovations for further improvement in operation, safety and economics. Thus, this paper shall serve as a curtain raiser for the special issue “Pressurized Heavy Water Reactors (PHWRs) Safety: Post Fukushima.”

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020902-020902-13. doi:10.1115/1.4035435.

The technology of pressurized heavy water reactors (PHWRs) which was developed with prime objectives of using natural uranium fuel, implementing on power fuelling, utilizing mined uranium most effectively, and achieving excellent neutron economy has demonstrated impressive performance in terms of high capacity factors and an impeccable safety record. The safety features and several technology advancements evolved over the years in which Indian contributions that are considerable are briefly discussed in the first part of the paper. Unique features of PHWR such as flexibility of fuel management, distribution of pressure boundaries in multiple pressure tubes (PTs), and a large inventory of coolant-moderator heat sink in close proximity of the core provide inherent safety and fuelling options to these reactors. PHWRs, in India have demonstrated to have the advantage of lower capital cost per megawatt even in small size reactors. Low burn up associated with natural uranium fuel, higher level of tritium in the heavy water coolant, and a slightly positive coolant void coefficient in present generation PHWRs have all been addressed in the design of advanced heavy water reactor (AHWR). The merit of adopting closed fuel cycle with partitioning of minor actinides in reducing the burden of radio-toxicity of nuclear waste and of deploying light water reactors (LWRs) in tandem with PHWRs in the evolving nuclear fuel cycle in India are also discussed.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020903-020903-11. doi:10.1115/1.4035726.

The International Atomic Energy Agency (IAEA) organized a coordinated research project (CRP) on “Benchmarking Severe Accident Computer Codes for Heavy Water Reactors (HWR) Applications,” (IAEA TECDOC Series No. 1727), and the activity was completed in 2012. This paper summarizes the results from the CRP: the selection of a severe accident sequence, definition of appropriate geometrical and boundary conditions, benchmarking code analyses, comparison of the code results, evaluation of the capabilities of existing computer codes to predict important severe accident phenomena, and suggestions for code improvements and/or new experiments to reduce uncertainties.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020904-020904-8. doi:10.1115/1.4035416.

The paper overviews the analytical studies performed at Politehnica University of Bucharest on the analysis of late phase severe accident phenomena in a Canada Deuterium Uranium (CANDU) plant. The calculations start from a dry debris bed at the bottom of calandria vessel. Both SCDAPSIM/RELAP code and ansys-fluent computational fluid dynamics (CFD) code are used. Parametric studies are performed in order to quantify the effect of several identified sources of uncertainty on calandria vessel failure: metallic fraction of zirconium inside the debris, containment pressure, timing of water depletion inside calandria vessel, steam circulation in calandria vessel above debris bed, debris temperature at moment of water depletion inside calandria vessel, calandria vault nodalization, and the gap heat transfer coefficient.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020905-020905-8. doi:10.1115/1.4035784.

The objective of current study is to develop and verify computer models to accurately predict the behavior of reactor structural components under operating and off-normal conditions. Indian pressurized heavy water reactors (PHWRs) are tube type of reactors. The coolant channel assemblies, being one of the most important components, need detailed analysis under all operating conditions as well as during postulated conditions of accidents for its thermomechanical behavior. One of the postulated accident scenarios for heavy water moderated pressure tube type of reactors, i.e., PHWRs, is loss of coolant accident (LOCA) coincident with loss of emergency core cooling system (LOECCS). In this case, even though the reactor is tripped, the decay heat may not be removed adequately due to low- or no-flow condition and inventory depletion of primary side. Initially, this will result in high temperature of the fuel pins. Since the emergency core cooling system (ECCS) is presumed to be not available, the cooling of the fuel pins and the coolant channel assembly depends on the moderator cooling system, which is assumed to be available. Moderator cooling system is a separate system in PHWRs. In PHWRs, the fuel assembly is surrounded by pressure tube, an annulus insulating environment, and a concentric calandria tube. In this postulated accident scenario, a structural integrity evaluation has been carried out to assess the modes of deformation of pressure tube—calandria tube assembly in a tube type nuclear reactor. The loading of pressure and temperature causes the pressure tube to sag (by weight of fuel bundle) and/or balloon (by internal pressure) and come in contact with the outer cooler calandria tube. The resulting heat transfer could cool and thus control the deformation of the pressure tube thus introducing interdependency between thermal and mechanical contact behavior. The amount of heat thus expelled significantly depends on the thermal contact conductance (TCC) and the nature of contact between the two tubes. Deformation of pressure tube creates a heat removal path to the relatively cold moderator. This, in turn, limits the temperature of fuel for a sufficiently long period and ensures safety of the plant. The objective of this paper is to provide insights into this thermomechanical behavior by computational studies and to understand the role of underlying parameters (such as material constants, thermal contact conductance, and boundary conditions) that control the tube deformation and further damage progression. The deformation characteristics of the pressure tube have been modeled using finite-element-based program. Experimental data of pressure tube material, generated for this research work, were used in modeling and examining the role of nonlinear stress–strain laws in the finite-element analyses.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020906-020906-8. doi:10.1115/1.4035851.

The Nuclear Regulatory Commission (NRC) has considered revision of 10-CFR-50.46C rule (Borchard and Johnson, 2013, “10 CFR 50.46c Rulemaking: Request to Defer Draft Guidance and Extension Request for Final Rule and Final Guidance,” U.S. Nuclear Regulatory Commission, Washington, DC.) to account for burn-up rate effects in future analysis of reactor accident scenarios so that safety margins may evolve as dynamic limits with reactor operation and reloading. To find these limiting conditions, both cladding oxidation and maximum temperature must be cast as functions of fuel exposure. To run a plant model through a long operational transient to fuel reload is computationally intensive, and this must be repeated for each reload until the time of the accident scenario. Moreover for probabilistic risk assessment (PRA), this must be done for many different fuel reload patterns. To perform such new analyses in a reasonable amount of computational time with good accuracy, Idaho National Laboratory (INL) has developed new multiphysics tools by combining existing codes and adding new capabilities. The parallel highly innovative simulation INL code system (PHISICS) toolkit (Rabiti et al., 2016, “New Simulation Schemes and Capabilities for the PHISICS/RELAP5-3D Coupled Suite,” Nucl. Sci. Eng., 182(1), pp. 104–118; Alfonsi et al., 2012, “PHISICS Toolkit: Multi-Reactor Transmutation Analysis Utility—MRTAU,” PHYSOR 2012 Advances in Reactor Physics Linking Research, Industry, and Education, Knoxville, TN, Apr. 15–20.) for neutronic and reactor physics is coupled with the reactor excursion and leak analysis program—three-dimensional (RELAP5-3D) (The RELAP5-3D© Code Development Team, 2014, “RELAP5-3D© Code Manual Volume I: Code Structure, System Models, and Solution Methods,” Rev. 4.2, Idaho National Laboratory, Idaho Falls, ID, Technical Report No. INEEL-EXT-98-00834.) for the loss of coolant accident (LOCA) analysis and reactor analysis and virtual-control environment (RAVEN) (Alfonsi et al., 2013, “RAVEN as a Tool for Dynamic Probabilistic Risk Assessment: Software Overview,” 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID, May 5–9, pp. 1247–1261.) for the probabilistic risk assessment (PRA) and margin characterization analysis. For RELAP5-3D to process a single sequence of cores in a continuous run required a sequence of restarting input decks, each with different neutronics or thermal-hydraulic (TH) flow region and culminating in an accident scenario. A new multideck input processing capability was developed and verified for this analysis. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical pressurized water reactor (PWR).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020907-020907-14. doi:10.1115/1.4035434.

Severe accidents are of increasing concern in the nuclear industry worldwide since the accidents at Fukushima Daiichi (March 2011). These events have significant consequences that must be mitigated to ensure public and employee safety. Filtered containment venting (FCV) systems are beneficial in this context as they would help to maintain containment integrity while also reducing radionuclide releases to the environment. This paper explores the degree to which filtered containment venting would reduce fission product releases during two Canada Deuterium Uranium (CANDU) 6 severe accident scenarios, namely a station blackout (SBO) and a large loss of coolant accident (LLOCA) (with limited emergency cooling). The effects on the progression of the severe accident and radionuclide releases to the environment are explored using the Modular Accident Analysis Program (MAAP)–CANDU integrated severe accident analysis code. The stylized filtered containment venting system model employed in this study avoids containment failure and significantly reduces radionuclide releases by 95–97% for non-noble gas fission products. Filtered containment venting is shown to be a suitable technology for the mitigation of severe accidents in CANDU, maintaining containment integrity and reducing radionuclide releases to the environment.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020908-020908-6. doi:10.1115/1.4035336.

The thickness at which the calandria vault floor in a generic CANDU 6 nuclear reactor may collapse while undergoing molten core–concrete interaction (MCCI) was studied using an approximate analytical model and a finite-element model. It was confirmed that the collapse criterion of 0.45 m floor thickness that is currently used in severe accident analyses is adequate. The estimated timing of collapse is subject to uncertainty of several hours.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020909-020909-8. doi:10.1115/1.4035691.

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):020910-020910-7. doi:10.1115/1.4035783.

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

Commentary by Dr. Valentin Fuster

Research Papers

ASME J of Nuclear Rad Sci. 2017;3(2):021001-021001-9. doi:10.1115/1.4035564.

A new numerical model for stratified two-phase flows with wavy interface is derived in this study. Assuming an equilibrium condition between turbulent kinetic energy, potential energy, and surface energy, the turbulent length scale in the inner region of a two-layer turbulence approach can be described by a statistical model to account for the influence of the waves. The average wave number, which is an input parameter to this model, is taken from wave spectra. They can be predicted from a Boltzmann statistic of turbulent kinetic energy. The new turbulence model is compared with the two-phase k–ϵ turbulence model. Time-averaged flow properties calculated by the new approach, such as velocity, turbulence, and void profiles, are shown to be in good agreement with experimental data.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021002-021002-12. doi:10.1115/1.4034572.

An integrated chemical effects test (ICET) was designed and executed to investigate the corrosion of materials in a hypothetical post-loss of coolant accident (LOCA) environment for pressurized water reactors (PWRs) and the resulting effects on the measured head loss in three vertical columns through multiconstituents debris beds. The head loss columns were isolated approximately after 72 h, as the measured head loss in all three columns approached or surpassed the maximum limit of the differential pressure (DP) cells. Additional bench scale tests were carried out to investigate the cause of high head loss in the three columns. Combination of epoxy agglomeration and adhesion to fiber resulted in subsequent blockage of the flow through the debris bed with no chemical precipitation was concluded as the most reasonable cause of high head loss observed in the test. The test continued thereafter up to 30 days as an integrated chemical effects test using the corrosion tank only. The results presented in this article demonstrate trends for zinc, aluminum, and calcium release that are consistent with separate bench scale testing and previous integrated tests conducted under trisodium phosphate (TSP)-buffered post-LOCA environmental conditions. In general, the total and filtered samples showed almost identical concentration of all metals (Al, Ca, Si, and Mg) except zinc which clearly indicate that no precipitation occurred. The release rate and maximum concentrations of the released materials were slightly different than the separate effect testing as a result of different experimental conditions (temperature, surface area-to-water volume ratio) and/or the presence of other metals and chemicals in the integrated test.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021003-021003-9. doi:10.1115/1.4035466.

A pipe-wall thinning measurement is a key inspection to ensure the integrity of the piping system in nuclear power plants. To monitor the integrity of the piping system, a number of ultrasonic thickness measurements are manually performed during the outage of the nuclear power plant. Since most of the pipes are covered with an insulator, removing the insulator is necessary for the ultrasonic thickness measurement. Noncontact ultrasonic sensors enable ultrasonic thickness inspection without removing the insulator. This leads to reduction of the inspection time and reduced radiation exposure of the inspector. The inductively-coupled transducer system (ICTS) is a noncontact ultrasonic sensor system which uses electromagnetic induction between coils to drive an installed transducer. In this study, we investigated the applicability of an innovative ICTS developed at the University of Bristol to nuclear power plant inspection, particularly pipe-wall thinning inspection. The following experiments were performed using ICTS: thickness measurement performance, the effect of the coil separation, the effect of the insulator, the effect of different inspection materials, the radiation tolerance, and the measurement accuracy of wastage defects. These initial experimental results showed that the ICTS has the possibility to enable wall-thinning inspection in nuclear power plants without removing the insulator. Future work will address the issue of measuring wall-thinning in more complex pipework geometries and at elevated temperatures.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021004-021004-7. doi:10.1115/1.4034567.

In selecting the materials for the Canadian supercritical water-cooled reactor (SCWR), the effects and extent of stress corrosion cracking (SCC) on candidate alloys of construction, under various operational conditions, must be considered. Several methods of applying stress to a corroding material are available for investigating SCC and each have their benefits and drawbacks; for simplicity of the experimental setup at University of New Brunswick (UNB), a constant load C-ring assembly has been used with Inconel 718 Belleville washers acting as a spring to deliver a near-constant load to the sample. To predict the stress at the apex of the C-ring, a mechanistic model has been developed to determine the force applied by the spring due to the thermal expansion of each component constrained within a fixed length when the temperature of the assembly is increased from ambient conditions to SCWR operational temperatures. In an attempt to validate the mechanistic model, trials to measure the force applied by the washers as the assembly thermally expanded were performed using an Instron machine and an environmental chamber. Accounting for the thermal expansion of the pull rods, the force was measured as temperature was increased while maintaining a constant displacement between the platens holding the C-ring. Results showed the initial model to be insufficient as it could not predict the force measured through this simple experiment. The revised model presented here considers the thermal expansion of the C-ring and all the components of the testing apparatus including the tree, backing washers, and Belleville washers. Further validation using the commercial finite element (FE) package abaqus is presented, as are preliminary results from the use of the apparatus to study the SCC of a zirconium-modified 310 s SS exposed to supercritical water.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021005-021005-6. doi:10.1115/1.4035334.

A coupled neutronics/thermal-hydraulics (N/T) three-dimensional code system SNTA is developed for supercritical water-cooled reactor (SCWR) core steady-state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydraulic subchannel code. The appropriate outer iteration coupling method and self-adaptive relaxation factor are proposed for enhancing convergence, stability, and efficiency of coupled N/T calculation. The steady-state analysis for the CSR1000 core is applied to verify SNTA. The results calculated by SNTA agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC, where the neutronics modules are based on the finite-difference method. The numeric results show that SNTA can be applied to SCWR core steady-state analysis and core concept design.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021006-021006-8. doi:10.1115/1.4035331.

A series of austenitic alloys (800H, H214, I625, 310S, and 347) with different surface finishes were exposed to supercritical water (SCW) at 550 °C and 2.5 × 107 Pa for 120 h, 260 h, and 450 h in a static autoclave with an initial level of dissolved oxygen of 8 ppm. Indentation with a hardness indenter was used for assessment of oxide adhesion. This was compared with the results of a similar test on SCW-oxidized ferritic alloys. Delamination in all the tested ferritic alloys was insufficient for quantification of the results but allowed for qualitative comparison within this group. In the set of austenitic alloys, oxide on stainless steel (SS) 347 exfoliated during cooling from 550 °C, and from the remaining four alloys, only oxide on H214 delaminated, which made the qualitative comparison across the whole group impossible. Energy dispersive X-ray spectroscopy (EDX) revealed that under delaminated external Cr2O3 on H214 alloy, there was a submicron thick layer of Al-rich oxide. To investigate a possible oxide spallation on austenitic samples during exposure, mass loss obtained through descaling was compared with mass gain due to SCW exposure. The results indicated that the applied descaling procedure did not, in most cases, fully remove the scale. Apart from one case (SS 347 with alumina surface finish), there was no clear indication of oxide spallation.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021007-021007-10. doi:10.1115/1.4035332.

The potential for high turbine entry temperature (TETs) turbines for nuclear power plants (NPPs) requires improved materials and sophisticated cooling. Cooling is critical for maintaining mechanical integrity of the turbine for temperatures >1000 °C. Increasing TET is one of the solutions for improving efficiency after cycle optimum pressure ratios have been achieved but cooling as a percentage of mass flow will have to increase, resulting in cycle efficiency penalties. To limit this effect, it is necessary to know the maximum allowable blade metal temperature to ensure that the minimum cooling fraction is used. The main objective of this study is to analyze the thermal efficiencies of four cycles in the 300–700 MW class for generation IV NPPs, using two different turbines with optimum cooling for TETs between 950 and 1200 °C. The cycles analyzed are simple cycle (SC), simple cycle recuperated (SCR), intercooled cycle (IC), and intercooled cycle recuperated (ICR). Although results showed that deterioration of cycle performance is lower when using improved turbine material, the justification to use optimum cooling improves the cycle significantly when a recuperator is used. Furthermore, optimized cooling flow and the introduction of an intercooler improve cycle efficiency by >3%, which is >1% more than previous studies. Finally, the study highlights the potential of cycle performance beyond 1200 °C for IC. This is based on the IC showing the least performance deterioration. The analyses intend to aid development of cycles for deployment in gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs).

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):021008-021008-10. doi:10.1115/1.4035566.

This paper presents the results of experimental investigations into two-phase mass transport in a coarse packed bed representing the Canada Deuterium Uranium (CANDU) end shield. This work contributes to understanding of phenomena impacting in-vessel retention (IVR) during postulated severe accidents in CANDU reactors. The air barbotage technique was used to represent boiling at the calandria tubesheet surface facing the inner cavity of the end shield. Qualitative observations of the near-wall two-phase region were made during air injection. In addition, flow visualization was carried out through the addition of dye to the water. Air flow rate, shielding ball diameter, and cavity dimensions were varied within relevant ranges; and the impact of these parameters on the near-wall region was identified. A brief review of the relevant knowledge base is presented, allowing demonstration of the applicability of the test parameters. The observed phenomena are compared to published results involving similar geometries with capillary porous media.

Commentary by Dr. Valentin Fuster

Technical Brief

ASME J of Nuclear Rad Sci. 2017;3(2):024501-024501-7. doi:10.1115/1.4035690.

In Korea, pressurized heavy water-cooled reactors (PHWR) account for 17% of operating units and have taken an important role in providing national energy supply. The recent biggest issue in domestic PHWR community was the continued operation of the Wolsong-1 CANada Deuterium Uranium (CANDU) plant, which has recently been approved to operate for 10 more years after a 30 year design life. In relation to this issue, various actions from both post-Fukushima lessons and Wolsong-1 (WS1) stress test results are being taken. In KAERI R&D, the following topics are studied to support the basis for these actions. First, PHWR severe accident issues such as (1) primary heat transport system (PHTS) overpressure protection capability, (2) containment overpressure protection capability, and (3) bypass source term are evaluated. Second, a computer tool (called MAAP–ISAAC) has been modified and updated to support analyzing Wolsong severe accident issues. Third, a decision supporting tool, called Severe Accident Management Expert (SAMEX)–CANDU, has been developed to aid emergency response experts under severe accident conditions.

Commentary by Dr. Valentin Fuster
ASME J of Nuclear Rad Sci. 2017;3(2):024502-024502-6. doi:10.1115/1.4035333.

The accelerator-driven subcritical reactor system (ADS) is a kind of nuclear reactor which can burn minor actinide waste products produced from conventional reactors with inherent safety features. In this paper, the thermal-hydraulic model and a corresponding program for a 10 MW helium-cooled experimental ADS are presented. Through the analysis of the heat transfer mechanism in ADS, the thermal-hydraulic model of ADS was built, in which the solid domain is simulated with three-dimensional heat conduction model and the fluid domain is simulated with the one-dimensional quasi-static model. In order to analyze the transient characteristics of ADS with cooling system, a RELAP5–TRCAP coupling model for the cooling system was established, in which the decay heat of the target and core is considered. The results of steady condition and transients demonstrate the effectiveness of the transient model.

Commentary by Dr. Valentin Fuster

Book Review

ASME J of Nuclear Rad Sci. 2017;3(2):026501-026501-2. doi:10.1115/1.4035327.

REVIEWED BY JOVICA R. RIZNIC, Ph.D., PP.Eng., Technical Specialist, Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9, Canada.

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