This paper reports the development of HERON (Holistic Energy Resource Optimization Network), a newly-developed RAVEN (Risk Analysis Virtual ENvironment) plugin for grid and capacity optimization, to technoeconomic analysis in a deregulated market. A short description of the HERON plugin is provided, and the release process is described. HERON as a plugin enables RAVEN to perform stochastic technoeconomic analysis of grid-energy systems in a generic approach. The primary function of HERON is to generate the complex RAVEN workflows necessary to optimize component capacities under stochastic systems. HERON is capable of analyzing systems with complex components transferring a variety of commodities, including production components and varied markets. HERON is capable of optimizing high-resolution dispatch for such systems and guiding stochastic optimization algorithms in RAVEN for finding optimal component capacities. In particular, this document demonstrates the application of HERON to systems with deregulated markets. A system including a hyrdogen market, an electricity market, hydrogen storage, a hydrogen producer, and a nuclear power plant is considered. Stochastic histories for electricity prices at the electricity market are employed to perform stochastic analysis for ideal sizing of the hydrogen production facility and hydrogen storage unit. The impact of hydrogen market price and volatility of electricity price are also shown.

In assessing reduction of concrete strength under irradiated conditions, reference levels are introduced: 1×10 20 n/cm 2 for fast neutrons and 2×10 10 rad (2×10 5 kGy) for gamma-rays. Concrete structures are regarded as sound as long as irradiance levels accumulated during long-term operation are less than the reference levels. Most experimental investigations of irradiation effects on concrete were performed in the 1960’s and 1970’s. However, there is no good explanation of how concrete deteriorates under neutron and gamma-ray irradiation. Hilsdorf put the primal irradiation test data together to investigate effects of irradiance levels on residual strength ratio of concrete [1]. The reference levels were obtained from his paper. However, the test conditions in which the data quoted by Hilsdorf were obtained are very different from the irradiation and heat conditions usually found in a Light Water Reactor (LWR). This paper summarizes the interactions between radiation and concrete components and presents the results of gamma-ray irradiation tests on cement paste in order to provide a better understanding of the deterioration mechanisms of concrete under irradiation and heat conditions in LWRs.

Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR’s merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test with varied SSR geometry parameters, the transient test, and the simulation analysis of these steady state and transient tests. The steady state test results showed that the basic functioning principle such as the controllability of SSR water level by flow rate was maintained in the possible range of geometry parameters. The transient test results showed that the change rate of SSR water level was slower than the initiating parameters. The simulation analysis of steady state and transient test showed that the analysis method can simulate the height of SSR water level and its change with a good agreement. As a result, it is shown that the SSR design concept and its analysis method are feasible in both steady state and transient conditions.

Spherical fuel assembly with light water reactor technology has prominent advantages on safety and fuel cycle, for its large specific surface. However, larger specific surface brings larger pressure drop over the core comparing with traditional rod fuel assembly. A test facility was built to investigate the flow characteristics on a mono-size sphere packed bed with a diameter of 3mm. The Reynolds number varies from 52 to 665, and the flow pattern ranges from Darcy flow to turbulent flow. Pressure drop of the whole test part was measured with relevant components to ensure the expected inlet conditions. Results of tests with different inlet flow rates and temperatures were obtained. Numerical simulations were taken on the same packed structure and defined problem using CFD method. Quite good agreement was reached on pressure drop over the test part, and detailed flow information was also obtained. Furthermore, the scaling effects, including longitude pressure drop, cross section pebble bed size, and sphere diameter were analyzed using CFD method.

Small Modular Reactor (SMR) designs for the pressurized light-water nuclear steam supply system (NSSS) have instituted various and unique features in the pressure vessels and piping industry. Selection of materials and component geometries are creating an enviable opportunity for the respective design engineers to integrate and expand the use of the full complement of Subsections within Section III of the ASME Boiler and Pressure Vessel Code. This paper discusses the prescribed requirements contained within these tried and proven Code rules as may be applicable to the pressure vessels and piping components and assemblies that will make up the next generation of SMRs. A cross correlation between the current licensing regulations that specifically address the pertinent Codes and standards for the “traditional” pressurized light water reactors is outlined in terms of “equivalent” components within the SMR designs currently on the drawing boards. Specifically, at least one pathway to compliance with the current Codes and standards promulgated within 10 CFR 50.55(a) is developed and discussed herein. The authors also concede that the suggested pathway to compliance is only one of what will prove to be a multiple set of solutions to the crossing the U.S. domestic licensing tightrope.

All reactor designs take advantage of favourable inherent characteristics and compensate for unfavourable ones. Because of a very short reactor period for some postulated accidents, the Pressurized Water Reactor (PWR) design requires, and possesses, large negative values of fuel temperature and moderator temperature reactivity coefficients to ensure that rod ejection accidents can be compensated, and to stabilize reactivity transients from the operating state, which would otherwise be fairly rapid. In contrast, the CANDU design does not require strong negative feedback, given the small values of the reactivity coefficients around the operating point and the low reactivity worth of the control devices, both individually and collectively. Even for positive reactivity insertions near prompt critical, the rate of increase in reactor power in a CANDU reactor is inherently limited by its relatively long prompt neutron lifetime (about 40 times longer than that in a PWR), so that the reactor period is much longer and the rate of rise in power and enthalpy is much slower. Consequently, control and shutdown mechanisms are a practical and effective means for reducing total reactivity in the CANDU reactor. Although there are many international initiatives to align nuclear regulations and hence eliminate nation-specific requirements, many are still design-specific. The regulators who deal primarily with Light Water Reactor (LWR) designs tend to embed the LWR requirement of negative reactor reactivity coefficients in their regulations, whereby the regulations become very design-specific. In contrast, regulators who deal with various reactor designs typically favour a more technology-neutral approach — as in International Atomic Energy Agency (IAEA) standards, stating the safety goals to be achieved rather than defining reactivity coefficients. This paper presents a comparison of reactivity coefficients between typical modern LWRs and CANDUs. It discusses the relative importance of the reactivity coefficients in reactor safety, identifies major design differences and their influence on the type and value of the reactivity coefficients, and explains key features of the reactor operation and reactor behaviour in transients.

Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.

Boron carbide (B 4 C) is widely used as neutron absorbing control rod material in light water reactors (LWRs). It was also applied in all units of the Fukushima Dai-ichi nuclear power plant. Although the melting temperature of B 4 C is 2450 °C, it initiates local, but significant melt formation in the core at temperatures around 1250 °C due to eutectic interactions with the surrounding steel structures. The B 4 C containing melt relocates and hence transports material and energy to lower parts of the fuel bundle. It is chemically aggressive and may attack other structure materials. Furthermore, the absorber melt is oxidized by steam very rapidly and thus contributes to the hydrogen source term in the early phase of a severe accident. After failure of the control rod cladding B 4 C reacts with the oxidizing atmosphere. This reaction produces CO, CO 2 , boron oxide and boric acids, as well as significant amount of hydrogen. It is strongly exothermic, thus causing considerable release of energy. No or only insignificant formation of methane was observed in all experiments with boron carbide. The paper will summarize the current knowledge on boron carbide behavior during severe accidents, and will try, also in the light of the Fukushima accidents, to draw some common conclusions on the behavior of B 4 C during severe accidents with the main focus on the consequences for core degradation and hydrogen source term.

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.

The sodium-cooled fast reactor (SFR) is gathering worldwide attention for its features of the fast-spectrum reactor and closed fuel recycle system. This paper presents the modification of the ATHLET code for application to SFRs. The thermal-hydraulic computer code ATHLET is developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) in Germany for light water reactors analysis. In this paper, a sodium properties package is implemented in to the ATHLET code, and heat transfer correlations for sodium are also added for heat transfer prediction. To evaluate the capability of the modified code, the Phenix reactor, a SFR operated by French Alternative Energies and Atomic Energy Commission (CEA) from 1973 to 2009, is modeled. The scenario of transient from forced to natural convection is simulated and analyzed. The results are compared with the experimental data of the natural convection ultimate test of the Phenix facility. Results achieved so far indicate good applicability of the modified ATHLET code.

Considering a reasonable range of core meltdown accidents that can be postulated for GenIV sodium fast reactors, good safety margins exist for corium confinement and cooling inside the reactor vessel. Coolable conditions can be reached with the adoption of an ad-hoc device in the lower plenum, i.e. core catcher, capable to intercept the downward motion of the molten material and assure its cooling. Such device has to be designed to withstand to extreme thermal-mechanical conditions that rise as consequence of the large mechanical energy release and high temperature of molten corium. As this study has been carried out in the frame of the Collaborative Project on European Sodium Fast Reactor (CP ESFR) of the 7 th Framework Programme Euratom, on the basis of the postulated accident conditions assumed for a reference 1500 MWe pool-type sodium fast reactor, the present work provides a preliminary analysis of the thermal response of a possible core catcher placed within the vessel. The dynamic thermal behaviour of the corium-structure-coolant system is analyzed with the computer code CORIUM-2D, an original simulation tool developed by RSE - Ricerca Sul Sistema Energetico, with the aim to assess the thermal interaction among corium, structures and coolant under severe accident conditions in both Light Water Reactors (LWRs) and Liquid Metal Fast Breeder Reactors (LMFBRs). The results of the numerical simulations show that the steady-state coolable configuration of core debris and the structural integrity of main containment structures can be reached in a number of partial core meltdown situations.

Analyses were performed for flow blockage accidents postulated in a conceptual design of a 600 MWe demonstration sodium cooled fast reactor with 3 types of core designs, i.e., Uranium, *L-TRU (TRansUrium) and **M-TRU cores, using the MATRA-LMR-FB code. The analysis was addressed for the 6 sub-channel blockage which is a design basis event (DBE). The accidents for 24 and 54 sub-channel blockages were also analyzed to estimate the extent of the blockage size which could lead to sodium boiling or fuel melting. Three radial blockage positions were also taken into account in the analysis. In result, a higher maximum coolant temperature in the subassembly was obtained as the number of blocked subchannels increased. A recirculation region was usually developed right above the blockage for large blockage cases. The analysis results showed that a favorable safety margin was assured for the design basis event, i.e., the 6 sub-channel blockage accident. For the 24 and 54 sub-channel blockage cases, the peak cladding temperature limit was breached, and there was a case in which fuel melting could be threatened. * TRU extracted from LWR fuel ** Mixture of L-TRU and TRU extracted from self-recycled SFR fuel

Passive auto-catalytic recombiners (PARs) play a key role in the hydrogen mitigation strategy of European LWRs. In order to avoid possible threats related to hydrogen combustion, PARs are installed to remove hydrogen released during a loss-of-coolant accident. The possible impact of hydrogen explosions became evident during the reactor accident in Fukushima (Japan) in March 2011, where leaked hydrogen ignited and largely destroyed the upper part of the reactor building. The mitigation strategy is based and verified by computational accident assessments. Code validation against experimental data is vital in order to achieve reliable results.

Newly designed plants, e.g., next-generation light water reactor or ESBWR, employ a passive containment cooling system and have an enhanced safety with RHRs (Residual Heat Removal system) including active components. Passive containment cooling systems have the advantage of a simple mechanism, while materials used for the systems are too large to employ these systems to existing plants. Combination of passive system and active system is considered to decrease amount of material for existing plants. In this study, alternatives of applying containment outer pool as a passive system have been developed for existing BWRs, and effects of outer pool on BDBA (Beyond Design Basis Accident) have been evaluated. For the evaluation of containment outer pool, it is assumed that there would be no on-site power at the loss of off-site power event, so called “SBO (Station BlackOut)”. Then, the core of this plant would be uncovered, heated up, and damaged. Finally, the reactor pressure vessel would be breached. Containment gas temperature reached the containment failure temperature criteria without water injection. With water injection, containment pressure reached the failure pressure criteria. With this situation, using outer pool is one of the candidates to mitigate the accident. Several case studies for the outer pool have been carried out considering several parts of containment surface area, which are PCV (Pressure Containment vessel) head, W/W (Wet Well), and PCV shell. As a result of these studies, the characteristics of each containment outer pool strategies have become clear. Cooling PCV head can protect it from over-temperature, although its effect is limited and W/W venting can not be delayed. Cooling suppression pool has an effect of pressure suppressing effect when RPV is intact. Cooling PCV shell has both effect of decreasing gas temperature and suppressing pressure.

Scientific bases for corrosion control for water-cooled fusion reactors, such as ITER, are discussed in this paper. In the previous overview papers [1,2], the author has identified that ‘long cell action’ corrosion plays a pivotal role in practically all unresolved corrosion issues, irrespective of fission reactor types and operation. In trying to confirm the existence of radiation-induced ‘long-cell’ action (macro) corrosion cell in the primary cooling system of LWRs, the author attempted to theoretically reproduce the electrochemical potential differences demonstrated during experiments at the INCA Loop in Sweden and the NRI-Rez Loop in the Czech Republic [3,4]. Based on these knowledge bases, characterization of water chemistry for fusion was made through radiation- and electro-chemistry analysis following the methodology explained in the companion paper also presented at this conference [5]. Approximately 70 mV of potential difference is predicted between the high dose rate and the out-of-flux regions through the illustrative model examined in this paper, which supports the existence of the LGCA mechanism. This value is about 1/5 of the potential difference for a typical PWR. For further verification of this approach, ‘international benchmark in-pile tests’ are highly awaited. Further optimization studies as well as material testing under the influence of the LGCA mechanism should be performed to establish corrosion control.

A fusion-fission hybrid reactor with pressure tube assemblies is proposed for energy production in this paper. In the light water cooled blanket, spent nuclear fuel from 33 GWD/t LWR and natural uranium oxide are taken as driver fuel for neutron/energy multiplication aiming at increasing uranium utilization efficiency. Tritium self-sufficiency is obtained by Li 2 O. The home developed NAHR code is employed to simulate the neutronics behavior in the blanket. The plasma conditions and configuration of ITER are used to the neutronics calculations. The results show that the blanket can be operated for at least 5 years with good performance of acceptable energy multiplication and tritium breeding. The first neutron wall loading is lower than 0.57 MW/m 2 .

The main goal of the Generation-IV nuclear-energy systems is to address the fundamental research and development issues necessary for establishing the viability of next-generation reactor concepts to meet future needs for clean and reliable energy production. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a SuperCritical Water-cooled Reactor (SCWR), which continues the utilization of well-known light-water-reactor technologies. Research Centre Rez Ltd. has taken part in a large European joint-research project dedicated to Generation-IV light-water reactors with objectives to contribute to the fundamental research and development of the SCWRs by designing and building a test facility called “SuperCritical Water Loop (SCWL)”. The main objective of this loop is to serve as an experimental facility for in-core and out-of-core corrosion studies of structural materials, testing and optimization of suitable water chemistry for future SCWRs, studies of water radiolysis at supercritical conditions and nuclear fuels. This paper summarizes the concept of the SCWL, its design, utilization and first results obtained from non-active tests already performed within the supercritical-water conditions.

Dissimilar Metal Welds (DMWs) are generally applied to structural components such as reactor pressure vessel and pressurizer nozzles in nuclear power plants. A filler metal is used in the manufacture of DMWs in light water reactors (LWR) to join the low alloy steel pressure vessel nozzles and steam generator nozzles to nickel-based wrought alloy or austenitic stainless steel components. In recent years crackings have been observed in the welded joints of DMWs. Since there is the high susceptibility of heat affected zone (HAZ) and fusion zone (FZ) to stress corrosion cracking (SCC), a concern has been raised about the integrity and reliability in the joint transition zone. In this study, the dissimilar metal joints welded between Alloy 690, Ni-based alloy and Alloy 533 Gr. B (A533B), low alloy steel with Alloy 152 filler metal were investigated. Detail nano-structural and nano-chemical analysis were performed between Alloy 152 and A533B by using optical microscope, scanning electron microscope (SEM), transmission electron microscope (TEM), secondary ion mass spectrometry (SIMS) and 3-dimensional atom probe tomography (3D APT). It was found that in the weld root region, the weld was divided into different regions including unmixed zone in Ni-alloy, fusion boundary (FB), and the HAZ in the low alloy steel. The result of TEM, EDS and 3D APT analyses showed the non-homogeneous distribution of elements with higher Fe but lower Mn, Ni and Cr in A533B compared with Alloy 152, and the precipitation of carbides near the FB.

This paper presents the simulator of a high level nuclear power plant modelled in Simulink/MATLAB. It borrows also elements from SimHydraulics. The model is based on an energy balance across the main components in both the primary and secondary loops, however it does not include the reactor core kinetics/burn up and heat transfer/decay heat calculations yet. Further the simulator is not aimed to safety analysis. The primary components (both high and low pressure turbines including moisture separator, condenser, steam generator, etc.) have all been modelled using correlations and vendors data. The secondary side in particular models the transition of states (evaporation, condensation) based on look up tables. The model was validated with available data from a commercial nuclear plant design (the Westinghouse AP1000) and from a small reactor design (S.I.R., Safe Integral Reactor). Comparative values demonstrate that the model supports a number of different plant configurations and it can be successfully utilised in nuclear reactor preliminary design. Furthermore, it can be used to identify the consequences of various design choices. Future developments will include the reactor core physics using a 6-group delayed neutrons model as well as more detailed turbine and other components automotive, aerospace, power generation. In nuclear power two notable plant simulators are PCTRAN and PANTHER. PC-TRAN (6) is a PCbased reactor simulation software which was first designed for the Westinghouse AP600 and later expanded to the AP1000, Areva EPR and other light water reactors. PANTHER (Pitt Advanced Nuclear Training for Higher Education Reactor) (4) is a desktop AP1000 nuclear plant operations simulator developed at the University of Pittsburgh. Both simulators are not suited for the analysis of new nuclear reactors design. PC-TRAN in fact cannot run simulations of variants of a design in order to provide ‘before and after’ comparisons while PANTHER is mainly an educational simulator and it is unfinished. Further both simulators are not capable of modelling small modular reactors since they miss a model for once though steam generators. Here we present a scoping design tool for the preliminary design of nuclear reactors. The tool proves to be very valuable for the performance evaluation of new small modular reactors design. The model was built using MATLAB, Simulink and SymHydraulics as they provide already some validated components model and allow model configuration flexibility and accuracy.

A modified middle scaled creviced bent beam (CBB) test was conducted in order to investigate stress corrosion cracking (SCC) behavior around fusion boundary between nickel base alloy (Alloy82) and low-alloy steel (LAS) under high stress conditions in light water reactor (LWR) coolant environment. An Alloy182/LAS clad plate specimen was placed between the CBB jigs so that the tensile stress occurs in Alloy182 and the SCC cracks propagate from the Alloy182 surface to the fusion boundary. In pure water, only oxides and cracks propagating along the fusion boundary were observed. Meanwhile in Na 2 SO 4 injected water with electrical conductivity 0.3μS/cm, deep cracks propagated into the LAS. The higher conductivity enhanced the driving force of the SCC cracks. A notched specimen was also tested to investigate SCC propagating behavior in more severe stress conditions than those in a smooth surface specimen. When a notched specimen was used, even in pure water condition, some cracks penetrated the fusion boundary and oxides were observed. In the notched specimen, crack nucleation points are limited in the notch root and the stress relaxation hardly occurs due to multiple crack nucleations. Such severe stress conditions also contribute the cracks to propagate into the LAS. To discuss the crack propagation behavior in the vicinity of the fusion boundary, the cross section was observed with an optical microscope. The observation indicates that the crack propagating depth into the LAS and the diameter of oxides increase with increasing the Alloy182 thickness. Although an effect of applied strain and welding direction of specimen to SCC behavior was investigated, there was no significant difference in the morphology by them. Since 1% strain was applied to CBB specimen surface and much greater strain was observed at a notched specimen, elasto-plasticity should be considered to stress condition of a SCC crack tip. An ‘equivalent stress intensity factor, K J ’ was, therefore, proposed to describe a stress condition in elasto-plastic area. A good relationship was observed between depth of SCC crack in LAS and K J value. It was found that the proposed parameter K J was effective to describe the stress state at the crack tip from the elastic region to the plastic region up to approximately 1% strain.