According to the development of the concept of “zero failure” or “zero fuel element defect”, accepted in 2011, which consists in reducing the number of fuel elements that are depressurized in the process of operation to the reached level in the leading countries in nuclear energy (10 −6 –10 −5 defective fuel rods) and avoidance of fuel assemblies with non-hermetic cladding of fuel rods for further operation, including defects with a “gas leak” type, new promising fuels are being developed and introduced, including methods for justifying their safety. Thus, to ensure reliability and safety of new fuel types, it is necessary to provide procedures for monitoring current performance characteristics at all stages of the life cycle of fuel rods. In this paper, experience is given on the development and implementation of instrumentation and methods for monitoring of fuel rods with advanced types of nuclear fuel for VVER reactors that ensure the reliability, safety and competitiveness of technologies associated with the use of advanced fuel rod types, and the implementation of associated components, systems and equipment for monitoring and diagnostics. The features of the applied techniques are presented, and the new system of requirements for the implemented equipment created on their basis. This research continues, and the analysis of intermediate experimental data is carried out in this article.

The National Bureau of Standards reactor (NBSR) at the National Institute of Standards and Technology (NIST) is under conversion from high enriched uranium (HEU) to the low enriched uranium (LEU) schema under the Reduced Enrichment for Research and Test Reactors program (RERTR) as a part of the Global Threat Reduction Initiative (GTRI). The conversion of the high performance research reactors (HPRR) such as NBSR is a challenging task due to the high flux need (2.5 × 10 14 n/cm 2 -s for the NBSR), as well as other neutronics performance characteristics requirements without significant changes to the external geometrical configuration. One fuel candidate, the General Atomics (GA) UZrH LEU fuel, has showed particular promise in this regard. The TRIGA LEU fuel was initially developed in the 1980s with particular considerations for fuel conversion for high power regimes such as high density research and test reactors. This study performs a neutronics feasibility study of the UZrH LEU fuel schema for the NBSR, examining the accountability and sustainability of the TRIGA fuel when applying it to the NBSR conversion. To identify the best option to deploy the TRIGA fuel to NBSR in terms of key neutronic performance characteristic, the study is carried out with various considerations in the fuel dimensions, fuel rod layout configurations, and structure material selections. Monte Carlo based computational model is used to assist and facilitate the research procedure. The research findings in this study will determine the viability of the TRIGA fuel type for the NBSR conversion, and provide supporting data for future investigations on this subject.

The gas/liquid metal magnetohydrodynamic generator (G/LM-MHD) with the mixture of gas and liquid metal as working fluids shows a promising future due to recent development of liquid metal cooled nuclear reactors. Previous efforts on the G/LM-MHD energy conversion systems have predicted a higher efficiency than traditional thermodynamics cycle. However, most of the earlier studies focus on the conception designs, feasibility analysis and preliminary experiments, while less attention paid on some specific problems such as the bubble phenomenon in the two-phase flow. Therefore, this paper deals with numerical study on the performance characteristics of the gas/liquid metal two-phase flow in an ideal Faraday-type MHD channel, of which the geometry structure is 30 × 30 × 80 mm cuboid segmentary style. The conductive mixture fluid is composed of nitrogen as the gas phase and gallium as the liquid phase (N 2 /Ga). The temperature at the channel inlet is about 600 K considering the heat transfer after the mixing chamber, while the inlet velocity is around 10 m/s and gas volumetric void fraction is 50%. The external magnetic field is assumed as 4 Tesla adopting the superconducting technology, which seems essential for MHD industrial practice. Then the simulation is accomplished using a modified two-phase mixture model considering the electromagnetic influence. The simulation results show that the distribution of temperature changes much weaker than pressure and velocity, which agrees with earlier one-dimension analysis. On the other hand, the results characterizes clearly the increase of the void fraction close to the electrodes, which can explain intuitively the decrease of the power-generating capacity. Besides, the power output is predicted to reach maximum 22.5 kW while the voltage is 1.2 V and the power density can be 312.5 MW/m 3 which is far beyond traditional steam turbines. This study shows a promising future of the gas/liquid metal MHD generator for the small nuclear plants and power systems.

Station blackout accidents increasingly become the focus of research in the field of nuclear safety after Japan’s Fukushima nuclear plant accident in March 2011. Core decay heat under station blackout condition will be transferred by natural circulation occurring between core and passive heat exchanger for the nuclear plants incorporated passive safety design concept such as AP1000 or CAP1400. As a result, response of safety systems will differ in accident sequence and kind between passive safety plant and traditional plant. What is more, cooling capacity of passive heat exchanger (PHX) which takes on heat sink has significant effect on performance of natural circulation in passive safety system. The safety need that characteristics of passive safety plant should be verified through integral experiment facility makes scaling analysis important in design or modification of experiment facility. Furthermore, scaling analysis of natural circulation phenomena under station blackout accident plays an important role in design verification, safety review verification or thermo-hydraulic program development. It not only determines the similar similarity criteria between the nuclear power plant prototype and test facility, but also provides technical basis for selecting different experiment schemes. As a part of scaling analysis on natural circulation phenomena for station blackout, the cooling capacity of PHX in test facility should be scaled properly and reasonably with conservatism. Therefore, scaling of passive residual heat removal (PRHR) heat exchanger under station blackout accident is investigated analytically in this paper. The analytical model for natural circulation in passive heat exchanger is established based on the performance characteristics of PRHR system in passive plant. By proper hypothesis and simplification, the governing equations for PHX are normalized using steady-state solutions, initial or boundary conditions. The similarity criteria that should be preserved between PHXs in test facility and prototype are finally obtained from non-dimensionalized equations. Furthermore, the distortion analysis for PHE design is also investigated based on the similarity criteria for selected scaling factors and parameters. The safety analysis based on models of nuclear power plant prototype and test facility is conducted on transient performance of designed PHX with PHX of prototype. The results show that: heat source number is the dominant similarity criteria for PHXs design under SBO condition. Requirements of Richardson number and friction number could be satisfied by resistance adjusting on test loop. The performance of PHX designed following heat source number requirement can better represent the transient response characteristics of prototype under SBO condition.

An experimental research on performance characteristics of passive residual heat removal system (PRHRS) for the small modular reactor designed by Nuclear Power Institute of China (NPIC) under the station blackout accident was performed in the CREAS facility, which consists of the primary system, the secondary system, the passive safety injection system, the passive residual heat removal system, the overpressure protection system and the auxiliary system. The experimental results show that, after the station blackout accident, a stable two-phase natural circulation between the steam generators and the heat exchanger in the PRHRS was established with a mass flow of 0.4T/h, thus the heat from the primary system was removed to the water in the containment water tank (CWT). During this period, the core decay residual heat and the sensible heat were removed from the primary system by the PRHRS effectively. The cold water from the core makeup tanks was injected into the reactor pressure vessel for core cooling. The peaked primary pressure was 16.3MPa and less than relief valve opening pressure 16.9MPa. In addition, the average coolant temperature of the reactor core reduced below 483 K, and the reactor operated safely.

High temperature gas-cooled reactor combined with gas turbine cycle can make full use of the high outlet helium temperature of reactor. Compared to the Open Brayton Cycle, inventory regulation is a new regulation method for Closed Brayton Cycle, which can control the output power of the cycle through the change of pressure level and mass flow of the system. Taking the 10MW high temperature gas-cooled test reactor coupled with helium-turbine cycle (HTR-10GT) for example, a more practice-oriented resistance model is established in this paper for the analysis of influence of resistance characteristic on key components and closed Brayton cycle. When helium inventory drops from 100% to 30% of rated value, the mass flow rate of helium decreases, so the local pressure loss coefficient keeps constant and the pressure loss coefficient increases. As a result, the efficiency of closed cycle decreased about 15%. The performance characteristics of the turbomachinery and heat exchangers are also discussed in detail as key components in the power conversation unit (PCU).

This paper establishes a method for integrating real-time and historic plant data into thermal performance characterization tools. Many nuclear plants utilize data collection software that integrates with off the shelf spreadsheet programs. It is shown here that standard heat balance and condenser performance models can be constructed within the spreadsheet programs to utilize plant data in an effort to develop a strong understanding of overall Balance of Plant (BOP) performance. This method also allows the user to create operational scenarios to predict performance and to illustrate the potential for gain in electrical generation.

One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations for which existing, proven systems do not currently exist. These conditions require a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations can be extremely costly and time consuming. Current closure-welding technologies (fusion welding) are not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL) are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials. The performance characteristics of FSW, i.e., high weld quality, simple machine-tool equipment and increased welding efficiency, suggest that this new technology should be considered for radioactive materials packaging campaigns. FSW technology will require some development/adaptation for this application, along with several activities needed for commercialization. One of these activities will be to obtain approval from the governing construction code to use the FSW technology. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME B&PVC) will govern this work; however, rules for the use of FSW are not currently addressed. A code case will be required to define appropriate process variables within prescribed limits for submittal to the Code for review/approval and incorporation.

Spent radioactive ion-exchange resin (SIER) is a long-standing issue for radioactive waste management safety. Performance of radioactive waste form is featured with extra long-term and non-repairable. Calcium sulfoaluminate cement (SAC) was used in radioactive spent resins solidification in China. A prescription of X SAC cement + 0.5X waste resins (50%water hold) + 0.35X water was obtained first. In order to control the temperature rise caused by hydration of cement in 200L solidification matrix, various supplementary materials were tried. Based on compressive strength tests and center temperature rise, super powered zeolite was selected. In addition, more resins were added to reduce the center temperature rise. A superior combination was obtained as SAC 35wt.%, zeolite 7wt.% to mix 42wt.% of resins (50%water hold) with 16wt.% of water. The microstructures of hydrated Ordinary Portland Cement (OPC), SAC and SAC with different zeolite addition were compared by means of Scanning Electron Microscopy (SEM). From the SEM pictures, the structures of the needles or spines can be seen in SAC matrices and the needles structure of SAC change into flake structure gradually with more zeolite added. The simulated leaching tests showed that inclusion of zeolite in SAC reduced the leaching rates of radionuclides significantly. From 200L matrix test, the centre temperature curve was measured, and the highest temperature was lower than 90°C. No thermal cracks were found in the final solidified products. The effect of radiation on compressive strength and radiolysis gas generation was studied for cement solidified form of various content of ion exchange resin with Co-60 irradiator. Variation of compressive strength, as well as the compressive strength of the waste form both with and without irradiation all within the standards requirement under irradiation of 10 6 Gy. However, the data obtained for ion exchange resin shows that hydrogen generation under irradiation of 10 5 Gy reached up to 3.5% of the total gas generated. This implies that the radioactivity of spent ion exchange resin shall be limited for long term storage and disposal with High-Integrity-Container. Calculation demonstrates that cement solidification of spent radioactive ion exchange resin existing in China so far should not result in radiation stability concern. It is concluded that SAC is one of the preferential binding material for ion exchange resins, the resin loading can be up to 75 (vol%) (wet resin). It is recommended that the performance requirement for cement solidified radioactive form shall be amended and guidelines for performance characterization in certain detail should be established. Biodegradation of cement solidified resin waste would be a safety concern and shall be investigated. Modeling of leaching should be promoted.

On-Line monitoring evaluates instrument channel performance by assessing its consistency with other plant indications. Industry and EPRI [1] experience at several plants has shown this overall approach to be very effective in identifying instrument channels that are exhibiting degrading or inconsistent performance characteristics. On-Line monitoring of instrument channels provides information about the condition of the monitored channels through accurate, more frequent monitoring of each channel’s performance over time. This type of performance monitoring is a methodology that offers an alternate approach to traditional time-directed calibration. On-line monitoring of these channels can provide an assessment of instrument performance and provide a basis for determining when adjustments are necessary. Elimination or reduction of unnecessary field calibrations can reduce associated labor costs, reduce personnel radiation exposure and reduce the potential for miss-calibration. PEANO [2–9] is a system for on-line calibration monitoring developed in the years 1995–2000 at the Institutt for energiteknikk (IFE), Norway, which makes use of Artificial Intelligence techniques for its purpose. The system has been tested successfully in Europe in off-line tests with EDF (France), Tecnatom (Spain) and ENEA (Italy). PEANO is currently installed and used for on-line monitoring at the HBWR reactor in Halden. This paper describes the results of performance tests on PEANO with real data from a US PWR plant, in the framework of a co-operation among IFE, EPRI and Edan Engineering, to evaluate the potentials of PEANO for future installations in US nuclear plants.

Steam Dryers for extended power up-rated conditions, based on a design with proven performance record, have been designed by Westinghouse with performance characteristics meeting all utility requirements. Extensive CFD analysis has been used to design the new dryer intended for the BWR 3000 plants. The analysis has utilized previous knowledge, acquired from experimental work, on steam flow characteristics in the BWR 3000 with its asymmetrically placed steam line nozzles. A design has been developed that is predicted to both decrease vibration levels and to solve a water-level measurement problem. An extensive experimental verification of the design is presently done.