低温精馏氢同位素分离技术及其应用

低温精馏是大规模分离氢同位素的有效手段,本文对该工艺涉及的气体纯化、制冷与真空、测量与控制、安全防护等关键技术进行简要介绍。对该工艺在重水生产、重水升级和除氚、聚变堆氘氚燃料循环、武器用氚等领域的国外应用做了回顾。分析了国内对该技术的需求,提出今后开展的研究方向。     

Pressure release of containments during severe accidents in Switzerland

As required by the Swiss Federal Nuclear Safety Inspectorate (HSK) all Switzerland's five nuclear power plants have to install a containment filtered venting system. The integrity of the containment (the last barrier for radioactive releases to the environment) can be threatened by overpressure due to inadequate heat removal. Design requirements have been provided for a specific class of severe accident scenarios. In general the capacity of the system is considered sufficient if it is able to vent the steam production corresponding to a decay heat level of 1% of the thermal reactor power. The mitigation capacity for the reduction of released radioactive material is specified by a retention factor of 1000 for aerosols to prevent or limit a long term ground contamination and a factor of 100 for elementary iodine for prevention or limiting of thyroid doses and to avoid short term evacuation. Besides existing requirements for design, maintenance and operation, additional claims such as passivity and operability at any pressure conditions inside the containment have to be met. Passivity implies that the system can be initiated after a severe accident without any operator action. The system also has to allow early manual venting. Various filtered venting systems are presently available. The nuclear power plants of Beznau, Gosgen, Leibstadt and Muhleberg have already selected such systems and already implemented them or are going to install them step by step. Beznau selected the Sulzer-EWI system which is using a water pool with nozzles-baffle plates and mixing elements to achieve the required filtration of the aerosols. In both Beznau units, the systems are installed and in standby mode. Gosgen, a pressurized water reactor as well as Beznau, is going to implement a filter system developed by Siemens-KWU, known as sliding pressure venting process, combining a venturi scrubber in a water pool and a mesh filter. The boiling water reactor of Leibstadt also selected the same system as Beznau while Müheberg choose the ABB system but not in the common design. The venturi pipes are thereby integrated in the water pool of the outer torus. The system in all five nuclear power plants is fully operable and in standby mode since December 1993.     

Swelling and gas release in oxide fuels during fast temperature transients

A previously reported intergranular swelling and gas release model for oxide fuels has been modified to predict fission gas behavior during fast temperature transients. Under steady state or slowly varying conditions it has been assumed in the previous model that the pressure caused by the fission gas within the gas bubbles is in equilibrium with the surface tension of the bubbles. During a fast transient, however, net vacancy migration to the bubbles may be insufficient to maintain this equilibrium. In order to ascertain the net vacancy flow, it is necessary to model the point defect behavior in the fuel. Knowing the net flow of vacancies to the bubble and the bubble size, the bubble diffusivity can be determined and the long range migration of the gas out of the fuel can be calculated. The model has also been modified to allow release of all the gas on the grain boundaries during a fast temperature transient.The gas release predicted by the revised model shows good agreement to fast transient gas release data from an EBR-II TREAT H-3 (Transient Reactor Test Facility) test. Agreement has also been obtained between predictions using the model and gas release data obtained by Argonne National Laboratory from out-of-reactor transient heating experiments on irradiated UO₂. It was found necessary to increase the gas bubble diffusivity used in the model by a factor of thirty during the transient to provide agreement between calculations and measurements. Other workers have also found that such an increase is necessary for agreement and attribute the increased diffusivity to yielding at the bubble surface due to the increased pressure.     

Incompatibility between zircaloy-2 and inconel X-750 during temperature transients

The incompatibility of Zircaloy-2 and Inconel X-750 has been investigated between 1000°C and 1200°C (1200°C being the currently allowable maximum temperature in the acceptance criteria for ECCS for water reactors). It has been found for the temperatures of 1000°C and 1200°C that oxide thicknesses of 2 and 30 μ respectively protect the Zircaloy-2 against attack by Inconel X-750.     

The release of fission gas from nuclear fuels during temperature transients

The fractional release of rare gas atoms from a spherical grain of uranium dioxide containing a uniform concentration of gas atoms and intragranular bubbles is calculated for short-duration temperature transients and post-irradiation annealing. The released fraction is shown to reach a small maximum value which is dependent only on the grain size, gas atom concentration and intragranular bubble distribution. The analysis therefore may be applied to any fuel material providing these parameters are known. The conclusion therefore is that during a brief temperature transient, only a very small fraction of gas will be released from grain interiors to grain boundaries. Consequently the majority of gas released into the free volume of a pin will come from gas already residing on grain boundaries and released by mechanical cracking; the fractional gas release is small and should not cause undue concern. The calculations do not cover the case of fuel melting.     

Verification of the from model for zircaloy oxidation during high temperature transients

As the sheath of nuclear fuel elements in a water cooled reactor is one of the principal barriers against the release of fission products into the primary heat transport system, the modelling of the different phenomena that influence the sheath integrity is important. The computer subroutine FROM was developed to model Zircaloy-4 oxidation during an arbitrary high temperature transient. The Zircaloy thickness can contain up to three regions separated by diffusion controlled moving boundaries. Interface oxygen concentrations are not necessarily maintained at equilibrium values.The comparison of FROM predictions of layer thickness with a large number of experimental results from Kermforschungszentrum Karlsruhe (KFK), Oak Ridge (ORNL) and Chalk River (CRNL), showed very good agreement. The model is also successful in prediction of oxygen concentration distributions.     

Fuel swelling due to retained fission gas and thermal effects during high temperature transients

Fuel swelling of previously irradiated pressurized-water-reactor-type fuel rods tested under power-cooling-mismatch conditions is due to retained fission gas and thermal effects within the film boiling region. In this paper empirical correlations for fuel swelling are presented, and mechanisms contributing to the observed swelling and the applicability of an analytical fission gas behavior computer code (GRASS-SST) to fuel swelling are evaluated. Major contributors to fuel swelling are fuel melting and expansion, expansion of solid fuel, fission gas bubble coalescence, fission gas diffusion to grain boundaries, and change in surface tension of fuel upon melting. The contributions to fuel swelling of solid fission products and the effects of cladding contraction and wall thinning on rod swelling are also included. The overall empirically-calculated fuel swelling values and the GRASS-SST code calculated values are compared with measured values. The agreement between measured and empirically calculated fuel swelling is generally close. Fuel swelling due to retained fission gas during the film boiling transient, as calculated by the GRASS-SST code, was found to be in good agreement with experimental results.     

Thermohydraulics in a high temperature gas-cooled reactor prestressed concrete reactor vessel during unrestricted core heatup accidents

The long term core and primary loop heatup of an HTGR subsequent to loss of all forced circulation has been analyzed using a modified version of the CORCON code. The results indicate that if the liner cooling system is operating, or can be restarted within about 60 h, safe cooldown can be achieved, but significant core damage will occur. Without functioning liner cooling system the core heatup will lead to PCRV concrete degradation and the resulting concrete gas releases will ultimately cause containment building failure after 6 to 10 days.     

Model to simulate oxidation kinetics of zircaloy-4 during temperature transients between 600 and 800°C

Numerical methods have been used extensively to solve differential equations in diffusion problems, particularly the finite difference method in its explicit version. A substantial modification is introduced here in the implicit finite difference method (parabolic interpolation in the interface) to allow for accurate tracing on the interface movement even in points non-coincident with the mesh. This procedure results in a powerful code, adequate to simulate any interface movement with good accuracy and low expense in calculation time.The code was employed to simulate the linear oxidation kinetics of Zircaloy-4 (Zry-4) in water vapor during various temperature transients: upward ramp, downward ramp and sawtooth.The fundamental assumption of this model is that the parabolic-to-linear transition is due to the interconnection of pores generated in the oxide together with the oxide itself. Moreover, the model assumes that on account of pore interconnection, two zones may be recognized in the oxide: a channeled region formed by stoichiometric oxide and a diffusion zone, separated by another moving interface. The hypothesis is formulated that the pore interconnection speed is the variable which controls the overall linear process.Experimental evidence suggests that under variable temperature conditions, Zr or Zry oxidation is not only temperature-dependent but also shows some memory effects. These are reflected in the model through the speed of pore interconnection which is allowed to be delayed with reference to the temperature variations.Although the results obtained with the model seem to be satisfactory, further experimental evidence is required for comparison with the predictions of the model.     

Estimation of fuel and clad temperature of a research reactor during dry period of de-fuelling operation

Heat transfer and fluid flow studies related to spent fuel bundle of a research reactor in fuelling machine has been carried out. When the fuel is in reactor core, the heat generated in the fuel bundle is removed by heavy water under normal reactor operation. However, during the de-fuelling operation, the fuel bundle is exposed to air for some period called dry period. During this period, the decay heat from fuel bundle has to be removed by air flow. This flow of air is induced by natural convection only. In this period, the temperatures of fuel and clad rise. If clad temperature rises beyond a certain limit, structural failure may occur. This failure can result into release of fission products from fuel rod. Hence the temperature of clad has to be within specified limit under all conditions. The objective of this study is to estimate the clad temperature rise during the dry period.In the CFD simulation, the turbulent natural convection flow over fuel and radiation heat transfer are accounted. Standard k-? model for turbulence, Boussinesq approximation for computing the natural convection flow and IMMERSOL model for radiation are used.The steady state and transient CFD simulation of flow and heat is performed, using the CFD code PHOENICS. The steady state analysis provides the maximum temperature the clad will attain if fuel bundle is left exposed to air for sufficiently long time. For safe operation, the clad temperature should be limited to a specified value. From steady state CFD analysis, it is found that steady state clad temperature for various decay powers is higher than the limiting value. Hence transient analysis is also performed. In the transient analysis, the variation of clad temperature with time is predicted for various decay powers. Safe dry time, i.e. the time required for clad to reach the limiting value, is predicted for various decay powers. Determination of safe dry time helps in deciding the time available to the operator to drop the bundle in light water pool for storage. The analysis is found useful in optimizing the de-fuelling process.     

Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.     

Design and development of a device management platform for EAST cryogenic system

EAST cryogenic system is one of the critical sub-systems of the EAST tokamak device. It is a large scale helium cryoplant, which adopts distributed control system to realize monitoring and control of the cryogenic process and devices. However, the maintenance and management of most field devices are still in the corrective maintenance or traditional preventive maintenance stage. Under maintained or over maintained problems widely exist, which could cause devices fault and increase operation costs. Therefore, a device management platform is proposed for a safe and steady operation as well as fault diagnosis and predictive maintenance of EAST cryogenic system.This paper presents the function design and architecture design of the cryogenic device management platform. This platform is developed based on DeltaV DCS and acquires monitoring data through OPC protocol. It consists of three pillars, namely device information management, device condition management, and device performance monitoring. The development and implementation of every pillar are illustrated in detail in this paper. Test results and discussions are presented in the end.     

Preliminary safety analysis on loss of flow accidents and external source transients for LBE cooled ADSR core

In this paper, preliminary safety studies on the 800 MWth accelerator-driven system (ADS) proposed by Xi'an Jiaotong university are presented. The system is a pool type facility coupling a proton accelerator with current in the range of 17–23 mA and a sub-critical core by means of a spallation target. The RELAP5/MOD3.3 code is selected as a base tool. In order to simulate the system, the point kinetics model is modified and the property of lead-bismuth is implemented to meet the requirement of ADS analysis. This paper focuses on the assessment of its response to the loss of flow events. The first part is originated from the failure of the pump and the second part derives from the significant flow blockage at a fuel assembly inlet. The reactivity insertion accidents are caused by the change of the proton beam current. The results show that the safety and criteria are satisfied and the system is tolerant to the loss of flow accidents and proton beam doubled accident and is sensitive to the external neutron changing.     

Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents

The high temperature engineering test reactor (HTTR) being constructed by the Japan Atomic Energy Research Institute is a graphite-moderated, helium-cooled reactor with an outlet gas temperature of 950 °C.Two independent vessel cooling systems (VCSs) of the HTTR cool the reactor core indirectly during depressurized and pressurized accidents so that no forced direct cooling of the reactor core is necessary. Each VCS consists of a water cooling loop and cooling panels around the reactor pressure vessel (RPV). The cooling panels, kept below 90 °C, cool the RPV by radiation and natural convection and remove the decay heat from the reactor core during these accidents.This paper describes the design details and safety roles of the VCSs of the HTTR during depressurized and pressurized accidents. Safety analyses prove that the indirect core cooling by the VCSs and the inherent safety features of the reactor core prevent a temperature increase of the reactor fuel and fission product release from the reactor core during these conditions. Furthermore, it is confirmed that even if VCS failure is assumed during these accidents, the reactor core and RPV can remain in a safe state.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

Thermodynamic consequences of leakages in a double containment during severe accidents

For a large nuclear power plant under normal operating conditions a leakage rate for the containment of 0.25 vol.%/day is admissible. During a successfully controlled LOCA leakages of the containment will be released through filters by the annulus^* air exhausting system into the environment. During a core melt accident a pressurization of the containment has to be expected, which could lead to a failure of the containment due to overpressurization. When openings in the containment steel shell occur before a catastrophic failure, a depressurization into the annulus takes place. The area of the openings determines the depressurization rate and the thermodynamic conditions in the annulus. Furthermore the behaviour of the components being necessary for accident mitigation is influenced too. This paper discusses the thermodynamic consequences of leaks in the containment shell of a German PWR during a core melt accident. The results of those calculations are the necessary boundary condition for the estimation of fission product retention in the annulus.     

Thermal hydraulic analysis of SMART for heat removal transients by a secondary system

SMART is an integral type reactor of 330 MW, which enhances its safety by adopting inherent safety design features. Thermal hydraulic characteristics of transients in heat removal by a secondary system for the SMART have been carried out by means of the TASS/SMR and MATRA codes. The primary, secondary, and passive residual heat removal systems RHRS of the SMART were modeled properly. Then, a set of transients for the whole system was investigated. The results of the analyses using the conservative initial and boundary conditions showed that the safety features of the SMART design carried out their functions well and there was a strong moderator temperature coefficient due to the soluble boron free reactor affected by the transient behavior. The natural circulation was well established in the primary and passive residual heat removal systems during the transients and was enough to ensure a stable plant shutdown condition after a reactor trip.     

Performance of a high purity Ge gamma detection system cooled by a cryogenic refrigerator

The energy resolution and thermal performance of a high purity Ge-detector system cooled by a cryogenic refrigerator is described. The mechanical vibrations of the cryomachine are damped effectively leading to a degradation in energy resolution of 200 eV only. Cool-down time is two hours and output pulse height stability is excellent between 82 and 94 K. For data display and spectrum analysis an economic personal computer-based system is used.     

Variation of the vapor volumetric fraction during flow and power transients

An exact solution has been obtained for the variation of the vapor volumetric (void) fraction during exponential flow and power transients using the method of characteristics. The ‘void propagation’ equation, as derived by Zuber and Staub [1] was used and extended to the more general case of non-saturated inlet conditions. Transient void fraction measurements made with a capacitance-type void gage show good agreement with the theory, indicating that the simplifying assumptions made in the analysis are valid over a fairly wide range of void fractions.