Application of CANDLE burnup to block-type high temperature gas cooled reactor

The CANDLE burnup strategy, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed and without any change in their shapes, is applied to the block-type high temperature gas cooled reactor. If it is successful, a burnup control rod can be eliminated, and several merits are expected. This burnup may be realized by enriched uranium and burnable poison with large neutron absorption cross-section. With the fuel enrichment of 15%, gadolinium concentration of 3.0%, and fuel cell pitch of 6.6 cm, the CANDLE burnup is realized with the burning region moving speed of 29 cm/year and the axial half-width of power density distribution of 1.5 m. When the concentration of natural gadolinium is higher, the burning region moving speed becomes slower and the burnup becomes higher, though the effective neutron multiplication factor becomes smaller. When U-235 enrichment is higher, the effective neutron multiplication factor becomes larger, the speed becomes slower, and the burnup becomes higher. When the pitch is wider, the effective neutron multiplication factor becomes larger, the speed becomes faster, and the burnup becomes higher.     

铅铋冷却氮化物燃料小型模块化快中子反应堆堆芯物理特性分析

国际原子能机构(International Atomic Energy Agency,IAEA)认为小型模块化反应堆具有很好提高核能安全性、经济性和防止核扩散的能力,是未来核能最具发展前景的堆型之一。为适应未来核能发展的需求,提出了一种铅铋冷却氮化物燃料小型模块化反应堆(Small Modular Pb-Bi Cooled Reactor with Nitride Nuclear Fuel,SMPBN)设计方案,并利用PIJ组件计算程序和CITATION堆芯计算程序对SMPBN的物理特性和安全特性,包括反应性系数及其随燃耗变化、卸料燃耗、功率峰因子、燃料转换比和停堆余量等进行了深入分析。通过分析,认为SMPBN在20年寿期内,具有很好的燃料转换能力,不需要换料,反应性波动很小,反应性系数均为负值,具有固有安全性,符合国际上第四代反应堆的要求。     

Numerical studies of nuclear traveling waves in a supercritical water cooled fast reactor

An axial fuel shuffling strategy is proposed based on the mechanism of the nuclear fission traveling wave and implemented numerically in the calculation for a supercritical water cooled fast reactor (SCWFR). The ERANOS code is adopted to perform the neutronics and burn-up calculations, and the calculation scheme for axial fuel shuffling and coolant density coupling are set up. The parametric studies of a typical PWR with Th-U and U-Pu (²³⁵U instead of ²³⁹Pu) conversions by burn-up and k_eff calculations indicate that the breeding effects only exist in configurations with very low water content and the conversion or breeding becomes worse as the initial enrichment is increasing. The shuffling calculations for the 1-D SCWFR model described in this paper brought about some interesting results for a certain range of water content. The results indicate that the non-enriched fresh fuel is not possible for both Th-U and U-Pu conversions. As could be expected due to the η-values of the main fissile isotopes ²³³U and (²³⁵U, ²³⁹Pu), respectively, the Th-U conversion needs a lower enrichment, and results in a slightly higher burn-up than the U-Pu conversion. The asymptotic power density distribution of the Th-U conversion is broader and lower than that of the U-Pu conversion. By reducing the water volume fraction, an increased burn-up can be achieved with correspondingly reduced fuel shuffling speed and reduced initial enrichment. Furthermore, the steady state calculations for the asymptotic state show that the Th-U conversion is superior to the U-Pu one concerning SCWFR safety aspects, where the absolute value of the Doppler constant is larger and the coolant feedback is negative for the Th-U conversion, while the coolant feedback is positive for the U-Pu one.     

Simulation study on candle burnup applied to block-type high temperature gas cooled reactor

The CANDLE burnup is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. It can be applied easily to a block-type high temperature gas cooled reactor (HTGR) using an appropriate burnable poison with a high neutron absorption cross section mixed with uranium oxide fuel. In this study, natural gadolinium is used as burnable poison. In the present paper, the simulation of the burnup for the steady state and the startup is performed.

For the steady state simulation with direct solutions of steady state nuclide densities as inputs, the difference between the results of the steady state analysis and the simulation analysis is very small. It confirms that the steady state analysis is correct. When the initial core is constructed from easily available nuclides, the simulation result gives a reactivity change of 1.7% at a burnup time of 0.7 years.     

Thermohydraulic characteristics of coolant gas flow through a control rod channel in a very high temperature gas cooled reactor

A calculation code was developed to evaluate the thermohydraulic performance of a coolant flow through a control rod channel in a very high temperature gas cooled reactor (VHTR) and a high temperature engineering test reactor (HTTR). A one-dimensional flow network model was employed in the present calculation code. The calculated results agreed well with the experimental ones on the flow rate distribution and the total pressure loss in an isothermal coolant flow. The thermohydraulic characteristics of the HTTR control rod channel were evaluated by the code under various conditions, including the normal operating conditions of a HTTR.     

A constraint-based approach to intelligent support of nuclear reactor design

Constraint is a powerful representation to formulate and solve problems in design; a constraint-based approach to intelligent support of nuclear reactor design will be proposed in this paper. We will first discuss the features of the approach, and then present the architecture of a nuclear reactor design support system under development. In this design support system, the knowledge base contains constraints useful to structure the design space as object class definitions, and several types of constraint resolvers are provided as design support subsystems. The adopted methods of constraint resolution will be explained next in detail. The usefulness of the approach will be demonstrated using two design problems: design window search and multiobjective optimization in nuclear reactor design.     

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.     

Single-loop nuclear power plant with a pressurized-water reactor

Trust Tsentroénergomontazh. Translated from Atomnaya Énergiya, Vol. 71, No. 5, pp. 458–460, November, 1991.     

Safety assessment for computerized nuclear reactor protection systems: The Markov approach

In reactor protection systems based on minicomputers a central role is played by the diagnostic capability of selfchecking programs. It is thus of great importance to determine the efficiency that such programs must have with respect to fault detection in order to meet a certain reliability goal. Even though the content of this report is part of the safety study on a particular plant (Tapiro Research Reactor in service at C.S.N. Casaccia) it allows one to reach more general conclusions about the reliability of computerized protection systems. Another major aim of this paper is to point out the methodological difficulties met in the safety qualification of these systems.     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

Application of adjoint sensitivity analysis to nuclear reactor fuel rod performance

Adjoint sensitivity analysis in nuclear fuel behavior modeling is extended to operate on the entire power history for both Zircaloy and stainless steel cladding via the computer codes FCODE-ALPHA/SS and SCODE/SS. The sensitivities of key variables to input parameters are found to be highly non-intuitive and strongly dependent on the fuel-clad gap status and the history of the fuel during the cycle. The sensitivities of five key variables, clad circumferential stress and strain, fission gas release, fuel centerline temperature and fuel-clad gap, to eleven input parameters are studied. The most important input parameters (yielding significances between 1 and 100) are fabricated clad inner and outer radii and fuel radius. The least important significances (less than 0.01) are the time since reactor start-up and fuel burnup densification rate. Intermediate to these are fabricated fuel porosity, linear heat generation rate, the power history scale factor, clad outer temperature, fill gas pressure and coolant pressure. Stainless steel and Zircaloy have similar sensitivities at start-up but these diverge as burnup proceeds due to the effect of the higher creep rate of Zircaloy which causes the system to be more responsive to changes in input parameters. The value of adjoint sensitivity analysis lies in its capability of uncovering dependencies of fuel variables on input parameters that cannot be determined by a sequential thought process.     

氨气-水之间氘交换法纯化核反应堆重水

用氨气-水之间氘交换法与氨精馏法组合成的工艺流程提取核反应堆来的重水中的氕和氚,可以把重水的浓度提高到99.6％以上。这一方法比用氘-水交换法与氘精馏法组合成的工艺流程纯化核反应堆重水安全、易行。     

Concept of a pressurized water reactor cooled with supercritical water in the primary loop

A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation.     

The inherently-safe power reactor DYONISOS : Dynamic nuclear inherently-safe reactor operating with spheres

A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydrodynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h.     

Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid–structure-interaction coupling

A homogenisation method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a “reduced” numerical model accounting for inertial fluid–structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenisation techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to rector internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenisation approach to the case of rector internals is then exposed: it is shown that in such case, confinement effects can de modelled by a suitable modification of classical fluid–structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a “reduced” model with homogenised fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenisation approach is proved to be efficient from the numerical point of view and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted.     

Preliminary tests of particle image velocimetry for the upper plenum of a scaled model of a very high temperature gas cooled reactor

The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. During the design process multiple studies have been performed to develop safety codes for the reactor. Two major accidents of interest are the Pressurized Conduction Cooldown (PCC), and the Depressurized Conduction Cooldown (DCC) scenario. Both involve loss of forced coolant to the core, except the latter involves a pressure loss in the main coolant loop. During normal operation a circulator pumps the coolant into the upper plenum and down through the core, but following a loss of forced coolant the natural convection causes the flow to reverse to go through the core into the upper plenum. Computer codes may be developed to simulate the phenomenon that occurs in a PCC or DCC scenario, but benchmark data is needed to validate the simulations; previously there were no experimental test facilities to provide this. This study will cover the design, construction, and preliminary testing of a 1/16th scaled model of a VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets; the turbulent mixing of the jets was observed. After performing a sensitivity analysis the flow rate of a single pipe was extracted from the PIV flow field, and compared with an ultrasonic flowmeter and analytic flow rate. All the values lied within the uncertainty ranges, validating the test results.