For a few years, AREVA and the Commissariat a` l’Energie Atomique (CEA) have been conducting an extensive R&D program on V/HTR fuels with the objective is to optimize the TRISO fuel coatings produced in a Spouted Bed Chemical Vapor Deposition (SB-CVD) reactor. Numerical simulation models of this SB-CVD process have been developed in this work, describing physical and chemical phenomena occurring in high temperature spouted bed reactors. These models have been used to link external operating conditions (gas flow rate, precursor concentration, temperature, etc.) to local deposition conditions (concentration and temperature fields, deposition rate profiles, etc...). The adopted strategy has been to develop simplified models based on a process engineering approach, which require low computational efforts but can handle complex chemical systems and provide relatively accurate predictions. A model based on a hydrodynamic stream tube formulation (pseudo 2D model) and including a complete description of heat and mass transfer has thus been developed. Bed hydrodynamic has been described using high temperature correlations developed in the frame of this work. Radiation and heat transfer at reactor walls, which are of key importance for an accurate description of coupled transfer phenomena, have been implemented in the model formulation. The heterogeneous and homogeneous chemical mechanisms involved in the SB-CVD process have been first selected from the literature, then developed and reduced according to the main reaction paths. The pseudo 2D model developed and the specific high temperature correlations used have been validated by in-situ pressure and temperature profile measurement in 2 ″ and 3 ″ diameter SB-CVD reactors and finally by measurements of deposition rates on coated ZrO 2 and UO 2 kernels. The results presented in this paper show that the model is capable of handling rather large chemical schemes and combines simplicity and relatively good accuracy; hence it can be used for preliminary design and optimization of HTR fuel coating fabrication. Calculations have shown that particular attention must be paid to the heat transfer description in high temperature spouted bed reactors. A forthcoming work will focus on further model validation by varying the experimental conditions and using different SB-CVD furnace sizes and configurations. In addition further analyses and optimization studies of the chemical mechanisms involved are planned, which aim to increase the model accuracy and reliability. A better understanding of the SB-CVD process through accurate modeling will be very helpful for the optimization of coating deposition parameters on an industrial scale and for the design and scale up of large SB-CVD reactors.

Safety demonstration tests on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) were conducted to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the reactor core and primary cooling system transient data for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 100% rated power level in July, 2005. This paper simulates the reactor transient behaviour during the test by using the THERMIX code system. The reactor power transition and a comparison with the test result are presented. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shutdown after the stop of the helium circulator and keeps subcritical till the end of the test. Due to the loss of forced cooling, the residual heat is slowly transferred from the core to the Reactor Cavity Cooling System (RCCS) by conduction, radiation and natural convection. The thermal response of this heat removal process is investigated. The calculated and test temperature transients of the measuring points in the reactor internals are given and the differences are preliminarily discussed. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature is always lower than 1230 °C which is the limited value at the first phase of the HTR-10 project. The simulation and test results show that the HTR-10 has the built-in passive safety features, and the THERMIX code system is applicable and reasonable for simulating and analyzing the helium circulator trip ATWS test.

This paper describes the design process followed by Westinghouse Electric South Africa for the insertion of hot internals into the Reactivity Control System (RCS) and Reserve Shutdown System (RSS) Units Under Test (UUTs) at the Helium Test Facility (HTF) at Pelindaba, South Africa. The aim of the UUTs is to allow the validation of the high temperature operation of the RCS and RSS systems for implementation into the proposed Demonstration Power Plant of the PBMR. The units use electrical heaters to obtain pebble-bed reactor thermal conditions for both the control rods and small absorber spheres (SAS) under a pressurized helium environment. Design challenges include providing for strength under elevated temperatures (900°C maximum); pressure boundary integrity (9MPa maximum); separation of different volumes (representing core barrel, reactor citadel and other Reactor Pressure Vessel (RPV) volumes); thermal protection of carbon steel vessels by using thermal insulation; allowing for diverse thermal expansion coefficients of different materials; allowing for depressurization events within the insulation and internals; having access for temperature, pressure, stress and proximity sensors and electrical wiring through high pressure penetrations; and provision for assembly of the hot internals both on and off-site. Several thermal analyses using Computational Fluid Dynamics (CFD) were performed to evaluate both worst-case and operational conditions of the UUTs. Factors that were considered include thermal insulation properties, heat transfer modes (internal radiation, external radiation and natural convection, forced internal convection for cooling) and operating pressure (ranging from 1 to 9MPa). The thermal design uses elements originally proposed for hot gas duct design. The results obtained show that the proposed design satisfies ASME VIII requirements of the pressure boundary and that all challenges are successfully met.

For the Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant (DPP) several neutron flux measurements are made, both within the Reactor Pressure Vessel (RPV) and outside the RPV. The measurements within the RPV are performed by the Core Structures Instrumentation (CSI) system. While those outside the RPV are performed by the Nuclear Instrumentation System (NIS). The PBMR has a long annular core with a relative low power density, requiring flux monitoring over the full 11 M of the active core region. The core structures instrumentation measures the neutron flux in the graphite reflector. Two measurement techniques are used; Fission Chamber based channels with high sensitivity for initial fuel load and low power testing and SPND channels for measurements at full and near full power operation. The CSI flux monitoring supports data acquisition for design Verification and Validation (V&V), and the data will also be used for the characterization of the NIS for normal reactor start-ups and low power operation. The CSI flux measurement channels are only required for the first few years of operation; the sensors are not replaceable. The Nuclear Instrumentation System is an ex core system that includes the Post Event Instrumentation. Due to the long length of the PBMR core, the flux is measured at several axial positions. This is a fission chamber based system; full advantage is taken of all the operating modes for fission chambers (pulse counting, mean square voltage (MSV), and linear current). The CSI flux monitoring channels have many technical and integration challenges. The environment where the sensors and their associated signal cables are required to operate is extremely harsh; temperature and radiation levels are very high. The selection and protection of the fission chambers warranted special attention. The selection criteria for sensors and cables takes cognizance of the fact that the assemblies are built in during the assembly of the reactor internal structures, and that they are not replaceable. This paper describes the challenges in the development of the monitoring systems for the measurement of neutron flux both within the RPV and the ex core region. The selection of detector configuration and the associated signal processing will be discussed. The use of only analogue signal processing techniques will also be elaborated on.

Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3, 4]. In the framework of the European RAPHAEL project, this paper presents the results of the preliminary investigations carried out on the VHTR design. These studies aimed at understanding the physical aspects of the annular core and to identify the limits of a standard block type VHTR with regard to a degradation of its passive safety features. Analysis was performed considering various geometrical scales: fuel cell and fuel column located at the core hot spot, 2D and 3D core configurations including the coupling between neutronic and thermal-hydraulic. From the thermal analysis performed at the core hot spot, the capability to reduce the maximum fuel temperature by modifying the design parameters such as the fuel compact and the fuel block geometry was assessed. The best performances are obtained for an annular fuel compact geometry with coolant flowing inside and outside the fuel compact (ΔT > 50°C). The reliability of such design option should however be addressed with respect to its performance during the LOFC transient (the residual decay heat will be evacuated by radiation during the transient instead of conduction through graphite). As far as the fuel element geometry is concerned, a gain of approximately 50°C can be achieved by making limited changes on the fuel compact distribution in the prismatic block: reduction of the number of fuel compact in the outer ring of the fuel element where the average ratio between coolant channels and fuel compact is smaller. On the other hand, the adopted modifications should also be evaluated with respect to the maximum temperature gradient achieved in the fuel (amoeba effect). In the end, calculations performed on the full core configuration taking into account the thermal feedback showed that the radial positioning of the fuel elements allows to reduce significantly the power peaking factor and the maximum fuel temperature. The gain on the fuel temperature, which varies during the core irradiation, is in the range 100 – 150°C. Several modifications such as the increase of the bypass fraction and the replacement of a part of the graphite reflector by material with better thermal properties were also addressed in this paper.

The HTR TRISO particle consists of a fissile kernel and surrounding layers, whose density and thickness are among the key fuel parameters. Destructive methods i.e. the sink float method and image analysis of ceramography, were developed in the past and are still used to evaluate these particle parameters. Although exhibiting great accuracy, these methods generate effluents and wastes, and are extremely cost/time consuming. In the framework of the AREVA NP HTR R&D program, the development of nondestructive evaluation methods as alternatives to destructive methods is carried out and aims at a new HTR Fuel QC strategy. In this scope, an innovative method was developed to automatically measure particle layer density and thickness from X-Ray Phase Contrast Imaging (PCI). First tested at the European Synchroton Radiation Facility (ESRF), this method was then applied to a custom built industrial demonstrator. Comparisons between the density and thickness values obtained by the developed method and their corresponding values obtained with destructive methods justify progressing to the validation phase. Particle samples were selected among the particle batches that were characterized by destructive methods. Layer density and thickness were determined by the X-Ray based technique on the industrial demonstrator as well as at the ESRF. Correlation levels obtained from this benchmark demonstrated that both parameters can be confidently measured by the developed method. Additionally, it is important to stress that this technique provides the opportunity to directly determine buffer density on finite particles as opposed to the sink float method. Thanks to its accuracy, its rapidity and its absence of waste generation, it is planned to implement the X-Ray thickness and density measurement method on the French lab scale fuel line. It was also decided to enter the characterization work package of the IAEA Coordinated Research Project 6 in order to benchmark the AREVA NP method with foreign techniques and materials.

The Very High Temperature Reactor (VHTR) has been selected by the U.S. as the Generation IV technology for the Next Generation Nuclear Plant (NGNP), and both the U.S. and Japan have been developing VHTR concepts based on a prismatic, block-type core design. For these VHTR concepts, the primary coolant (helium) inlet temperature is expected to be in the range 490°C to 590°C and the outlet temperature is expected to be in the range 850°C to 950°C. Passive safety is one of the fundamental requirements for the VHTR, and the VHTR is designed to be passively safe even during Loss of Coolant Accidents (LOCAs) and Loss of Flow Accidents (LOFAs). For the VHTR, these two transient events are referred to as a Low-Pressure Conduction Cooldown (LPCC) and High-Pressure Conduction Cooldown (HPCC), respectively. During both events, the decay heat is conducted through the graphite to the vessel. The heat is transferred from the vessel by thermal radiation and natural convection to a passive Reactor Cavity Cooling System (RCCS). In this paper, we describe parametric studies of LPCC and HPCC events using a 30-degree sector, 3-dimensional ANSYS model of the VHTR, which includes a detailed radiation exchange model between the RPV and RCCS.

The electrical utility in South Africa (Eskom) plan to construct a first of a kind Pebble Bed Modular Reactor (PBMR). It has been recognized that there is a need to adapt the licensing process for the PBMR to ensure that credible and effective licensing process be developed and implemented for this technology. This paper will briefly outline the regulatory framework within South Africa, explain the licensing process adopted and present the challenges that the South African National Nuclear Regulator (NNR) was facing in developing and implementing the licensing process and how these are being addressed. The paper will discuss the update of the regulatory framework and the gaps identified in terms of regulatory requirements needed for such a project. The scope of the regulatory assessment for the licensing of the PBMR is based on the licensing requirements and criteria defined by the NNR in regulatory documents that expand on the current legislative requirements. In addition guidance is provided on selected issues in regulatory guidance documents and position papers. The requirements comprise, besides the general requirements to respect good engineering practice and the ALARA and defense-in-depth principle, specific risk criteria and radiation dose limits. These are categorized for normal operation and operational occurrences as well as for design basis events and beyond design basis events for workers and the public. Additional requirements and recommendations are stipulated by the NNR on safety important areas like quality and safety management, qualification of the nuclear fuel and the core structures, core design, verification and validation of computer codes, source term analysis and others. Selected NNR Position Papers have been developed to elaborate and provide further clarification on NNR requirements. For preparation of the PBMR safety case so-called Key Licensing Issues have been defined and agreed with the applicant. Discussions relating to these Key Licensing Issues allow important nuclear safety aspects identified for the PBMR demonstration plant to be clarified in advance of the safety case submittal.

Currently, two composites types are being developed for incore application: carbon fiber carbon composite (CFC), and silicon carbide fiber composite (SiC/SiC.) Irradiation effects studies have been carried out over the past few decades yielding radiation-tolerant CFC’s and a composite of SiC/SiC with no apparent degradation in mechanical properties to very high neutron exposure. While CFC’s can be engineered with significantly higher thermal conductivity, and a slight advantage in manufacturability than SiC/SiC, they do have a neutron irradiation-limited lifetime. The SiC composite, while possessing lower thermal conductivity (especially following irradiation), appears to have mechanical properties insensitive to irradiation. Both materials are currently being produced to sizes much larger than that considered for nuclear application. In addition to materials aspects, results of programs focusing on practical aspects of deploying composites for near-term reactors will be discussed. In particular, significant progress has been made in the fabrication, testing, and qualification of composite gas-cooled reactor control rod sheaths and the ASTM standardization required for eventual qualification.