During the course of 42 years of irradiated operations in the Hot Fuel Examination Facility at the Idaho National Laboratory (INL), a hot cell window had never been replaced. Recently, slow deterioration of a window seal resulted in mineral oil leaking at a rate of over a liter per month from the window tank unit on through the protective A-slab seal and into the hot cell. A hot cell window consists of both a steel tank unit with five slabs of glass of varying thicknesses with the remaining free space filled with clear mineral oil, and a thinner protective interior A-slab of glass. The repair solution was to remove and replace the A-slab window followed by replacing the window tank unit in two distinct phases. The facility original A-slab design was a leak tight barrier and a frame that was “L” shaped with a gasket between the glass and the window flange. Problems with the gasket adhering to the glass and the window flange resulted in pulling the glass from the frame during initial installation activities. Due to the adhesion problem, the gasket was changed to a dust seal during commissioning the facility. Over time, the window tank unit mineral oil leak flowed through this dust seal. Replacing the leaking tank unit necessitated the need for a new leak tight boundary as well as provide a method to drain the accumulated oil behind the A-slab until the tank unit could be replaced. These criteria led to a new A-slab design to be installed. Initially, removal and replacement of the A-slab was performed in the main cell (hot side) to reestablish a leak tight barrier. Transfers of the windows and removal of the bolts/reinstallation of new bolts were all performed with specialized equipment designed for remote operations in a hazardous environment using remote manipulators and cranes. Removal and replacement of the window tank unit via the operating corridor (cold side) was scheduled during a facility outage to accommodate availability of contract service personnel who specialize in hot cell windows. Due to the complexity of the replacement task, approximately 30% of the personnel on site were involved in the window replacement. Engineering, facility operations and radiation control personnel were primary contributors with electricians, carpenters and the analytical laboratory personnel contributing, as well. The multi-year installation program was safely concluded allowing the facility to resume full operations with the window properly sealed.

The aim of this work is to show a preliminary investigation on the propagation of electromagnetic fields generated by wireless technologies inside a nuclear facility or power plant. First, a survey of currently proposed wireless technologies for nuclear facilities and plants has been carried out. Then, for selected scenarios, the electromagnetic field propagation has been studied by means of electromagnetic simulation tools, and the presence of the nuclear environment has been simulated by properly modeling environmental parameters and engineered barriers. The choice of the proper simulation techniques and tools is mandatory in order to simulate the effect of the realistic environment on the propagation. Accordingly, the feasibility of wireless technologies application at nuclear facilities can be assessed on the basis of results achieved from simulated scenarios. The goal is to analyze, for selected scenarios, possible issues due to the propagation of an electromagnetic field in presence of simplified barriers mimicking the real nuclear environment. This approach can provide indications on how to deploy potential benefits of wireless technologies in a nuclear environment, evaluating pros and cons of the investigated technologies.

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I&C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make basement for evaluating the reliability and safety of DCS.

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.

Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO 2 cycles can also be designed for operation under direct dry air cooling. This paper presents the results of the coupled control analysis of a sCO 2 cycle for a 100 MWe sodium-cooled fast reactor. The plant control mechanisms were investigated and optimized for load following operation.

The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MW th nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.

Supercritical carbon dioxide Brayton cycle power converters can benefit advanced nuclear reactors, as well as small modular reactors, by reducing the plant cost and increasing plant electrical output. The sCO 2 cycles can also be designed for operation under direct dry air cooling. The paper presents the results of the coupled control analysis of a sCO 2 cycle for a 100 MWe sodium-cooled fast reactor under changing ambient air temperatures. The optimum plant operation modes are identified.

Over the course of the last several years the Canadian Nuclear Safety Commission (CNSC) has engaged with numerous vendors and potential licenses of small modular reactors (SMR) technology. This paper describes why Canada, and the CNSC, is of such interest to the international SMR community for prelicensing engagement and potential licensing of SMRs. It discusses what an SMR is and what potentially differentiates them from standard nuclear power plants (NPP). Readiness activities for the potential licensing of SMRs are described as well as modifications being made to the CNSC’s existing regulatory framework to facilitate the same, without reducing safety. The role of the CNSC’s discussion paper (DIS-16-04, Small Modular Reactors: Regulatory Strategy, Approaches and Challenges) and how feedback received on it helped confirm the CNSC’s modifications to be undertaken to the regulatory framework, as well as areas requiring further clarity, are highlighted. Finally, The CNSC Vendor Design Review (VDR) process is described as well as its part in ensuring a state of readiness to evaluate a licence application.

The purpose of the European Utility Requirements (EUR) Organisation is to actively develop and promote harmonised requirements for new mid- and large-size LWR NPPs that are proposed for construction in Europe. The harmonisation which is sought by the fourteen member utilities of the EUR Organisation, aims at delivering the safest and most competitive designs based on common requirements shared across Europe. The harmonised requirements are presented in the EUR document. This consists of an extensive set of requirements covering all aspects (safety, performance, competitiveness) and all parts of a NPP (Nuclear and Conventional Islands). It can be used by Utilities (e.g. for assessment of the GEN3 designs proposed by vendors, technical reference for call for bids) and by Vendors (e.g. understanding of customer’s expectation of GEN3 NPPs, facilitating the licensing process). The presentation will describe the main outcomes of the last 3 years of EUR Organisation activities (roadmap 2016–2018) and the challenges for the coming near future, in the following three fields. First, completion of Revision E of the EUR Document was achieved in December 2016 and issued in July 2017, followed in October 2017 by a training course that was attended by 90 participants. The presentation will describe the most significant updates including revision of the safety requirements to align to the most recent European and International safety standards issued by WENRA and IAEA, lessons learned from the Fukushima accident, including re-evaluated Seismic and External Natural Hazards approach and updated international standards (e.g. for I&C and for European Grid code). Revision E also includes feedback from previous design assessments of NPPs. Future possible development of the EUR Document (and of the assessment process) will be considered within the EUR Organisation ‘Roadmap’. Assessment of new designs is the second main technical activity. The assessment of the KHNP EU-APR (European version of APR1400) was completed in 2017 and an assessment of the Russian AEP’s VVER-TOI is planned to complete in 2018. These assessments are against EUR Revision D. The first assessment against Revision E (of the CGN HPR1000 “Hualong” design) is planned between 2018 and 2020. The presentation will recall the EUR design assessment objectives and process and the outcomes and progress of the different assessments. The third topic is the interaction between the EUR and other stakeholders, in particular other international organisations (ENISS, WNA/CORDEL, WENRA, EC, IAEA, EPRI/URD) with the aim of promoting Industry Requirements and influencing prospective regulation where appropriate. The presentation will describe how the EUR Organisation is connected to these stakeholders and how it presents Utility requirements to the wider nuclear industry.

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the only Generation III.7 reactor incorporating Fukushima lessons learned and complying with Western European Nuclear Regulation Association (WENRA) safety objectives. It is about twice safer than any existing Gen III.5 reactors. It has 7-day grace period for SBO and SA without containment venting. It enables no evacuation and no long-term relocation in SA. It, however, is based on the well-established proven ABWR. The NSSS and TI are exactly the same as those of the existing ABWR. The iB1350 only enhanced the ABWR safety by adding an outer well (OW) as additional PCV volume, built-in passive safety systems (BiPSS) for SA, DEC systems and an APC shield dome over the containment. The BiPSS include an isolation condenser (IC), an innovative passive containment cooling system (iPCCS), in-containment filtered venting system (IFVS), and innovative core catcher (iCC). All the BiPSS are embedded and protected in the containment building against APC. No specialized safety features remote from the R/B are necessary, which reduces plant cost. The primary system has only one integrated RPV. There are no SGs, no pressurizer, no core makeup tanks, no accumulators, no hot legs, and no cold legs. The iB1350 consists of only one integrated RPV and passive safety systems inside the containment building. This configuration is simpler than the simplest large PWR and as simple as SMR. While SMR have rather small outputs, the iB1350 has 1350 MWe output. It is simple, large and economic. As for the safety design it has an in-depth hybrid safety system (IDHS). The IDHS consists of 4 division active safety systems for DBA, 1 or 2 division active safety systems for DEC and the built-in passive safety systems (BiPSS) for SA. The IDHS is originally based on the four levels of safety that have provided an explicit fourth defense level against devastating external events even before 3.11. It also can be explained along with WENRA Defense in Depth (DiD). It is said that independence between DiD levels are important. However, there are many exceptions for independence between DiD levels. For example, SCRAM is used in DiD2, DiD3a and DiD3b. Any DiD that allows exceptions of independence of DiD levels is fake. The iB1350 is rather based on the three levels of safety proposed by Clifford Beck (AEC, 1967). There is complete independence between level 2 (core systems) and level 3 (containment systems) without any exceptions of independence. DiD without exceptions of independence is a real DiD. Only passive safety reactors can meet the real DiD.

Nuclear heating reactor is a new type of power plant that uses nuclear energy as heat source. Low temperature nuclear heating reactor should be the forerunner and main force for developing nuclear heating plant in China. Due to the lower water temperature required by the heating system, this dedicated, non-power generating nuclear reactor works at low temperatures and pressures with inherent safety features. The design, construction and operation of the nuclear heating reactors in various countries in the world were reviewed in this paper, and China’s new demonstration nuclear heating project and NHR-200 low-temperature heating reactor which would be used was discussed in the paper. We put forward the developing route and suggestion for the development of low-temperature heating reactor in China.

Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture. All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage: - brittle and ductile analysis of hypothetical large flaw - leak before break approach - break exclusion concept - incredibility of failure of high integrity components - end of fabrication acceptable defect - in-service inspection performance - acceptable standards in operation - Long Term Operation (LTO) All these requirements needs a procedure, an analysis method with material properties and criteria. After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.

Engineering feasibility of the beam window is one of the design issues in the accelerator-driven system (ADS). This study aims to perform the coupled analysis for the feasible beam window concept. To mitigate the design condition, namely to reduce the required proton beam current, subcriticality adjustment rod (SAR) was installed to the ADS core. The burnup analysis was performed for the ADS core with SAR and the results indicated that the maximum proton beam current during the burnup cycle was reduced from 20 to 13.5 mA. Based on the burnup analysis result, the coupled analysis; particle transport, thermal hydraulics and structural analyses, was performed. As the final result, the following design; the hemisphere shape, the outer radius = 180 mm, the thickness at the top of the beam window = 1.5 mm, and the factor of safety for the buckling = 3.8, was presented. The buckling pressure was almost same as the previous one and more feasible beam window concept was presented through this study.

The Steam Generator Tube Rupture (SGTR) postulated event constitutes one of the most hazardous safety issues for Gen IV pool reactors, cooled by heavy liquid metals. This accidental scenario is characterized by quick water flashing when in contact with primary coolant liquid metal, causing pressure wave propagation, cover gas pressurization in the reactor main vessel as well as possible tube rupture propagation, vapour dragged through the core, oxides precipitation and consequent slugs and plugs formation. The design phase of Gen IV MYRRHA reactor addressed the SGTR scenario issues in the framework of MAXSIMA project, supported by the European Commission. This research activity was fully executed at ENEA CR Brasimone, where a new test section was designed, assembled, instrumented and implemented in the large scale pool facility CIRCE. It was supported by the execution of preliminary and detailed pre-tests analysis performed adopting SIMME-III and -IV code, respectively. This paper details the test section main features, able to host four full scale portions (each one constituted by 31 tubes) of the MYRRHA Primary Heat eXchanger (PHX), for carrying out four independent SGTR experiments. A couple of tests investigated the tube rupture at middle position between two spacer grids of the bundle. The other two tests analysed instead the rupture near the bottom tube plate. Auxiliary systems were adopted for reaching primary (Lead Bismuth Eutectic alloy, LBE) and secondary (water) coolant initial conditions in accordance with MYRRHA design. Water was injected at 16 bar and 200°C in LBE at 350°C under an argon cover gas at about atmospheric pressure. The experimental results of the first test (middle rupture), in terms of CIRCE vessel pressurization, vapour flow path through tube bundle and tubes deformation, are presented. The post-test analysis was performed by SIMMER-IV code adopting the 3D Cartesian code version. The whole main vessel of CIRCE facility and implemented test section were modelled conserving heights and flowing areas. The experimental initial conditions were successfully matched by numerical results as well as the vessel pressurization and temperature time trends in the tube bundle following the SGTR. An important engineering feedback, for MYRRHA designer, was the evidence of rupture propagation absence. Moreover, the effectiveness of implemented safety devices, rupture disks, was evaluated and characterized for pressure relief feedbacks. A wide series of high quality measured data (pressure, temperature, strain and mass flow rate) was acquired and constitutes a database enlargement for future codes validation and possible new model development.

The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on fusion power P fusion = 200 MW, fusion power gain Q plasma = 1 – 5, tritium breeding ratio TBR>1.0, neutron DPA requirement ∼10 dpa. Phase II emphasizes DEMO validation, which means Q plasma > 10, P fusion > 1 GW, e.g. 1.5 GW. It is required that one blanket design can cover the operation of both phases of CFETR from the viewpoint of saving construction cost and reducing waste inventory. However, fusion power in Phase-II is 4–6.5 times larger than those in Phase-I, and this also causes the great challenge facing the thermal-hydraulics design of the blanket. A new version of water cooled ceramic breeder (WCCB) blanket for both phases is proposed for CFETR, based on a trade-off considering on TBR, release tritium temperature in breeder zone, and removal heat capability of coolant. This design continues to employ the mixed breeder of Li 2 TiO 3 and Be 12 Ti as tritium breeder and primary neutron multiplier, and a few Be as supplement of multiplying neutrons, Reduced Activation Ferritic/Martensitic steel as structural material, tungsten as plasma facing material. Pressurized water of 15.5 MPa is chosen as coolant with 285 °C inlet/325 °C outlet temperature. Main design change is that it employs two independent coolant systems in the blanket cooling components. For Phase I, one coolant system is only used and hoped to improve the breeder zone temperature higher than tritium release temperature. For Phase II, all of two coolant systems are put into using to ensure the material temperature less than the allowable limit. In this paper, the WCCB blanket design work is presented and its feasibility is investigated from the aspect of neutronics and thermo-hydraulics.

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO 2 due to the behavior CO 2 below its critical conditions.

Chinese Fusion Engineering Test Reactor (CFETR) is a new test Tokamak device which is now being designed in China to make the transition from the International Thermonuclear Experimental Reactor (ITER) to the future Fusion Power Plant (FPP). Breeding blanket is the key component of fusion reactor which is mainly responsible for the tritium self-sufficiency and fusion energy conversion. In the past few years, three kinds of blanket conceptual design schemes have been proposed and tested in parallel for CFETR Phase I, in which the helium cooled solid breeder (HCSB) blanket concept is acknowledged as the most promising one. However, nowadays, the design phase of CFETR has gradually changed from I to II aiming for the future DEMO operation condition, the main parameters of which are quite different from the previous one. Therefore, it’s necessary to perform conceptual design and various analyses for the HCSB blanket under the new working condition. In this work, firstly, a new conceptual design scheme of HCSB blanket for Phase II is put forward. Then, the radial build arrangements, of the two typical blanket modules are optimized by using the NTCOC. This work can provide valuable references for further conceptual design and neutronics/thermal-hydraulic coupling analyses of the HCSB blanket for CFETR Phase II.

The Japan Atomic Energy Agency (JAEA) has been conducting the research and development (R&D) on accelerator-driven subcritical system (ADS) as a dedicated system for the transmutation of long-lived radioactive nuclides. To foster the R&D of ADS, the Transmutation Physics Experimental Facility (TEF-P) in the J-PARC project has been planned to build by JAEA [1]. The TEF-P is used minor actinide (MA) fuel which has large decay heat, so during the failure of the core cooling system, the evaluation of the core temperature increase is important. This study aims to evaluate the natural cooling characteristics of TEF-P core and to achieve a design that does not damage the core and the fuels during an accident (the failure of the core cooling system). The experiments using mockup device was performed to validate the heat transfer characteristics in the empty rectangular lattice tube. It was obtained that the actual heat transfer coefficient of empty rectangular lattice tube was about 2.2 times larger than the theoretical free convection model. It was also confirmed that the insertion of any block into the empty rectangular lattice tube could achieve the higher heat transfer coefficient. Using the heat transfer coefficient obtained by experiment results, thermal analysis was performed by the three-dimensional heat transfer analysis. As a result, the calculation results showed that the maximum core temperature will be 294 °C which is less than the design criterion of temperature, 327 °C. It was presented that the design condition which the core temperature will be below the design criterion during the failure of the core cooling system through this study.

Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. New results of the application of the nodal diffusion code, DYN3D-MG, to model the steady state operation of molten salt fast reactors based on circulating fluoride salt, are presented here. The cross-section set was prepared using the Monte Carlo particle transport code, SERPENT. Full core comparisons between SERPENT and DYN3D-MG demonstrate that the molten salt fast reactor steady state operation can be modelled by DYN3D-MG within the required accuracy. The work is an important initial step for the development of a coupled CFD/neutron physics code system. This code system will finally be applied for the investigation of an innovative material protection scheme to avoid the contact of the liquid salt with the structural material.