Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride saltcooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant,as well as the activation of coolant and impurities,were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system,and the activity of radionuclides accumulated in the coolant during normal operation was calculated.The release rate (release fraction per unit time) of fission products was calculated with STACY code,which is modelled mainly based on the Fick's law,while the activation of coolant and impurities was calculated with SCALE code.The accumulation of radionuclides in the coolant has been calculated with a simplified model,which is generally a time integration considering the generation and decay of radionuclides.The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown.In operation condition,health-impacts related nuclides such as 3H,and 14C originate from the activation of lithium and coolant impurities including carbon,nitrogen,and oxygen.According to the calculated effective cross sections of neutron activation,6Li and 14N are the dominant 3H production source and 14C production source,respectively.Considering the high production rate,3H and 14C should be treated before being released to the environment.     

A review of revaporisation behaviour of radioactive Cs deposits and its impact on the source term in severe nuclear accidents from Phébus FP results and single effect testing

ABSTRACT

Revaporisation of the fission products deposited in the primary circuit of a reactor was identified as a possible late source of fission product release during a severe accident: e.g. loss of coolant accident (LOCA). Subsequent testing has shown that revaporisation is very likely to occur given a breach of the reactor and is an important contributor for the source term release to the containment and biosphere. The first part reviews the revaporisation mechanisms of Cs and other volatile or semi-volatile fission products transported in the primary circuit that were derived from the Phebus FP and associated programmes. The second part examines the separate effects testing to determine the high temperature chemistry of volatile and semi-volatile fission products (I, Mo, Ru) and structural materials (Ag, B), as well as atmospheric effects that substantially affect the source term. Finally, it examines Cs data from reactor accident sites that is providing additional knowledge of longer-term fission product chemistry. The results have been summarised in the form of a table and schematic diagram. This accumulated knowledge and experience has important applications in minimising contamination during decommissioning and site remediation techniques, as well as improving SA simulation codes and raising nuclear safety.     

Modeling and simulation of corrosion product activity in pressurized water reactors under power perturbations

An improved methodology is presented for simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor (PWR) under power perturbations. Using time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for ²⁴Na, ⁵⁶Mn, ⁵⁹Fe, ⁶⁰Co and ⁹⁹Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to ⁵⁶Mn. Flow rate has been assumed to follow power changes and different types of power perturbations are introduced after the equilibrium activity has been achieved. In particular the effects of linear changes in reactor operating power and power peaking on the corrosion product activity of the primary coolant have been studied.     

Sensitivity analysis of fission product activity in primary coolant of typical PWRs

A model has been developed for static and dynamic activity analysis of the fission product activity (FPA) in primary coolant of typical pressurized water reactors (PWRs). It has been implemented in the FPCART based computer program FPCART-SA. For long steady power operation of reactor, the computed values of normalized static sensitivity have been compared with the corresponding values obtained by using the dynamic sensitivity analysis. The normalized sensitivity values for the reactor power (P), failed fuel fraction (D), coolant leakage rate (L), total mass of coolant (m) and the let-down flow rate (Q) have been calculated and the values: 1.0, 0.857, −2.0177 × 10⁻⁶, 2.349 × 10⁻⁴, −2.329 × 10⁻⁴ have been found correspondingly for Kr-88 with the dominant value of FPA as 0.273 μCi/g.     

Stochastic simulation of fission product activity in primary coolant due to fuel rod failures in typical PWRs under power transients

During normal operation of PWRs, routine fuel rods failures result in release of radioactive fission products (RFPs) in the primary coolant of PWRs. In this work, a stochastic model has been developed for simulation of failure time sequences and release rates for the estimation of fission product activity in primary coolant of a typical PWR under power perturbations. In the first part, a stochastic approach is developed, based on generation of fuel failure event sequences by sampling the time dependent intensity functions. Then a three-stage model based deterministic methodology of the FPCART code has been extended to include failure sequences and random release rates in a computer code FPCART-ST, which uses state-of-the-art LEOPARD and ODMUG codes as its subroutines. The value of the ¹³¹I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found to have good agreement with corresponding experimental values for time dependent ¹³⁵I, ¹³⁵Xe and ⁸⁹Kr concentrations in primary coolant measured during EDITHMOX-1 experiments.     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

Real-Time High-Sensitivity Fuel Failure Detection for HTGR

A real-time high-sensitivity fuel failure detection (FFD) method has been developed, where a wire precipitator radiation detector measures noble-gas fission products (FPs) released from a High Temperature Gas-Cooled Reactor (HTGR). By changing the reference counting rate of the precipitator between the normal state and the failed fuel state in real time in response to reactor operation conditions, i.e. reactor power, fuel temperature, coolant-gas flow rate and so on, fuel failure with an extremely low failure fraction (Release-to-Birth ratio <5×10^?6) can be detected. The reference counting rate is obtained by adding an operational tolerance to the background counting rate that is estimated by a diagnostic equation. The diagnostic equation consists of a release equation for estimating the release rate of noble-gas FPs, a gas circulation equation for calculating concentrations of noble-gas FPs in the primary coolant system and a response equation for determining the detection efficiency of the wire precipitator. The feasibility of the method was evaluated by irradiation experiments using gas swept capsules and the Oarai Helium Gas Loop (OGL-1) in the Japan Material Testing Reactor (JMTR). The background counting rate was estimated with an error of about 20% in real time by the diagnostic equation.     

A general model for predicting coolant activity behaviour for fuel-failure monitoring analysis

A mathematical treatment has been developed to predict the release of volatile fission products from operating defective nuclear fuel elements. The fission product activity in both the fuel-to-sheath gap and primary heat transport system as a function of time can be predicted during all reactor operating conditions, including: startup, steady-state, shutdown, and bundle-shifting manoeuvres. In addition, an improved ability to predict the coolant activity of the ¹³⁵Xe isotope in commercial reactors is discussed. A method is also proposed to estimate both the burnup and the amount of tramp uranium deposits in-core. The model has been validated against in-reactor experiments conducted with defective fuel elements containing natural and artificial failures at the Chalk River Laboratories. Lastly, the model has been benchmarked against a defective fuel occurrence in a commercial reactor.     

Core temperature transient in the early phase of core uncovering at TMI-2

This paper describes a best-estimate analysis of the initial core boil-down and heat-up transient at Three Mile Island Unit (2) on 28 March 1979. This transient began shortly after all reactor coolant pumps were secured (100 min after reactor trip) and was terminated by a period of sustained high pressure injection of emergency cooling water, starting at 202 min.

The analysis is primarily directed to understanding the progression of core damage, rather than providing a detailed characterization of the core end-state condition. The latter objective can be achieved only after vessel head removal and visual examination.

The thrust of the present effort has been to: (1) develop a core coolant mixture level (dry-out level) calculation which satisfies the boundary conditions implied by various instrument responses and system operational characteristics; (2) couple the level calculation with a core heat-up modelto simulate the accumulation of thermal damage in the exposed, upper regions of the core; (3) compare calculated gross damage to the core with measurements of hydrogen and fission product releases subsequent to the accident.

Results indicate that:

1. (i) Observed containment hydrogen levels were due to Zircaloy/stainless steel corrosion that occurred during the period of core uncovering between the de-activation of the loop A reactor coolant pump (100 min after trip) and sustained operation of the high pressure injection system 100 min later. Appreciable zircaloy oxidation probably commenced at 150 min after trip, and continued at a high rate until the sustained high pressure injection at 202 min caused a major core quench.

2. (ii) There was some potential for fuel liquefaction. Calculations imply that peak fuel temperatures did not exceed the UO₂ pellet melting temperature, but 30% of the fuel was exposed to temperatures where liquid U---Zr---O alloys could have formed.

3. (iii) A substantial fission product release was obtained from fuel over-heating; however, an apparent disparity between the expected fission product release by calculation and the high range of fission product estimates obtained from plant measurements suggests that a significant release fraction may have originated from powdered or rubbilized fuel during cooldown. Additional gas releases may have developed from hot spots which persisted after core quench.

4. (iv) Steam temperatures in the upper plenum, at the outlet nozzle elevation, were generally below 900°C (1650°F) although this value was probably exceeded for a few min during the partial fuel quench caused by activation of the loop 2B reactor coolant pump, at 174 min after trip. The metal-work in the upper plenum, above the upper tieplate did not experience appreciable heating.

Thermal damage to the fuel and consequential weakening and mechanical disruption of the core was essentially complete 230 min after turbine trip.     

Prediction of Nongaseous Fission Products Behavior in the Primary Cooling System of High Temperature Gas-Cooled Reactor

In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs) are released mainly from fuel with failed coatings and are transported in the primary cooling system with the primary coolant during normal operation. In that case, condensable FPs plateout on the inner surface of components in the primary cooling system. On the other hand, since the HTGRs use helium gas as primary coolant, the primary coolant is not activated itself and very small amount of corrosion products is generated. Then, γ-ray emitted from the FPs becomes main source in shielding design of the HTGRs, and not only release amount from fuel but also plateout distributions of the FPs should be properly evaluated. Therefore, prediction of plateout behavior in the primary cooling system of HTGRs was carried out based on the calculation result of plateout distribution in High Temperature Engineering Test Reactor. Before the calculation, analytical model was verified by comparison with experimentally obtained plateout distributions and the applicability of the model to predict the plateout distributions in the primary cooling system of HTGR was certified.

This report describes the predicted result of plateout distribution in the primary cooling system of HTGR together with the verification result of the analytical model.     

反应堆燃料元件破损当量诊断方法

破损当量是衡量反应堆燃料元件破损严重程度的重要指标,但破损当量无法直接测量,在决策应用中不具有可操作性,需要建立与破损当量对应的可监测指标。本文结合实践经验,分析确定了可用于燃料元件破损诊断的典型核素,建立了反应堆一回路冷却剂中裂变产物核素活度浓度与燃料元件破损当量之间的传递关系;给出了一回路冷却剂取样分析实验方法,并指出实验过程中应注意的问题;建立了采用监测一回路冷却剂中典型裂变产物核素活度浓度诊断破损当量的方法,并分析了诊断中不确定度的主要影响因素。本研究为反应堆燃料元件破损当量诊断提供了技术方法。     

Heat and fission product transport in molten core material pool with crust

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the reactor vessel during a severe accident. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool is estimated by product of the mass concentration and energy conversion factor of each fission product. Twenty-nine elements are chosen and classified by their chemical properties to calculate heat generation rate in the pool. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis is performed for heat and fission product transport in a molten core material pool during the Three Mile Island Unit 2 (TMI-2) accident. The pool is assumed to be a partially filled hemisphere, whose change in geometry is neglected during the numerical calculation. Calculated results indicate that the peak temperature in the molten pool is significantly lowered, since a substantial amount of the volatile fission products is released from the molten pool during progression of the accident. The results may directly be applied to the existing severe accident analysis codes to more mechanistically determine the thermal load to the reactor vessel lower head during the in-vessel retention.     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Isotope separation methods for self-consistent nuclear energy system

It is known that for transmutation of fission products(FPs) in the concept of self-consistent nuclear energy system(SCNES) based on fast neutron reactor it is necessary to apply isotope separation of some FPs to keep neutron balance (to decrease parasitic capture of neutrons by stable isotopes). It is a question whether such FPs isotope separation can be feasible or not within amount of nuclear fission energy production. So it is necessary to consider isotopic content of FPs after fast reactor and to choose energetically appropriate isotope separation method for each radioactive FPs taking into account safe radioactivity level of FPs. In this paper we discuss about isotope separation method for SCNES. Isotopic composition of FPs was calculated using tables of fission yields from ²³⁹Pu fission. It isshown that concentrations of radioactive isotope in the main FPs to be isotopically separated are significant and vary from 2% in ruthenium up to 74% in iodine. We consider new isotope separation methods developed recently such as plasma separation process (PSP) based on selective ion cyclotron resonance heating and atomic vapor laser isotope separation (AVLIS) as a possible candidates. It seems to be energetically profitable to combine various methods to achieve desired separation characteristics. Since the most of FPs have a high initial concentration of radioactive isotope, PSP method seems to be a good candidate for first stages of separation process. We consider the main parts of energy expenditure in one PSP module and its separation characteristics. Estimations of energy consumption in multistage isotope separation process of FPs give maximum value 100keV/fiss. using PSP only and 3MeV/fiss. using AVLIS only. We can significantly decrease these values using AVLIS after PSP when concentration of target isotope in separation cascade will become sufficiently low. We can affirm that energy consumption in isotope separation of FPs is less than 60 MeV of electricity per one fission in nuclear reactor.     

Determination of I activity in the RBMK-1500 main circulation circuit

The purpose of this work was to evaluate the content of difficult to measure isotope ¹²⁹I in the RBMK-1500 reactor fuel-to-clad gap and reactor main circulation circuit (MCC) coolant. To determine fission product (FP) release from the defective fuel, the methodology proposed by Lewis and Husain for the CANDU reactor primary coolant activity prediction was applied. The determined effective diffusion coefficient D′ = 1.2E−09 s⁻¹ of iodine in the RBMK-1500 fuel is higher than the one evaluated for the CANDU fuel 6.8E−10 s⁻¹. Results show that the method developed by Lewis and Husain can be applied for the RBMK-1500 fuel gap and reactor main circulation circuit coolant activity prediction.     

Kinetic simulation of fission product activity in primary coolant of typical PWRs under power perturbations

Kinetic simulations of fission product activity in primary circuits of a typical PWR under power transients, has been performed. A detailed two-stage model-based methodology has been developed and implemented in a computer coder FPCART which uses LEOPARD and ODMUG codes as subroutines. For normal constant power operation, results for over 39 fission products show that the activity due to fission products in fuel region of PWRs is dominated by ¹³⁴I which is followed by ¹³⁴Te and ¹³³I. The value of the total fission product activity in fuel region predicted by FPCART code has been found to agree with-in 0.36% range with the corresponding values found by using the ORIGEN-2.0 code. The predictions of FPCART code have also been found in good agreement with the corresponding values found in ANS-18.1 Standard as well as with some available power-plant operation data with 2.4% deviation in the value of specific activity of the dominating fission product ¹³⁴I. The saturation value of the fission product activity in coolant depends strongly on the fuel-clad gap escape rate coefficient () and approaches a maximum value with increasing value of . During power transients, the FPCART predictions have been found in good agreement with the corresponding experimental measurements of ¹³¹I specific activity for Beznau and Surry PWRs.     

反应堆主回路卸压对压力容器外裂变产物释放的影响分析

本工作耦合建立了600 MW压水堆核电厂热工水力、裂变产物行为和放射性后果评价的分析模型,选取SB-LOCA、SGTR、SBO和LOFW等4个高压熔堆事故序列,研究了主回路卸压对压力容器外裂变产物释放的影响,包括主回路卸压对压力容器外裂变产物释放的缓解效应和其他负面影响。分析表明：实施主回路卸压可缓解高压熔堆事故序列下压力容器外的释放,但卸压工况下事故早期安全壳内的气载放射性活度较基准工况下的大。相关分析结论可作为严重事故管理导则制定的技术基础。     

Fundamental aspects of defective nuclear fuel behaviour and fission product release

An in-reactor research program with individual, purposely defected, nuclear fuel elements has provided a fundamental understanding of the physical processes of fission product release from defective fuel. On the basis of these experiments, an analytical model has been developed to describe the release of radioactive iodine and noble gas from defective fuel into the primary coolant. An analytic treatment has also been used to model the low-temperature release of fission products from small particles of uranium-bearing compounds (uranium contamination) deposited on in-core surfaces. As a result of this study, a methodology is established whereby release from surface uranium contamination can be distinguished from that resulting from fuel pin failure. Application of this work to power reactor operation is discussed.     

LaBr_3(Ce)γ谱仪在燃料元件破损监测中的应用研究

为了实现用LaBr_3(Ce)γ谱仪实时监测压水堆燃料元件的破损,对该谱仪系统在燃料元件破损监测中的几个关键问题进行了研究。通过实验测试与蒙特卡罗(MC)模拟计算,提出了使用LaBr_3(Ce)γ谱仪测量一回路冷却剂中裂变产物~(135)Xe和~(88)Kr的活度浓度来判断燃料元件是否发生破损的方法,并对该方法进行了验证。对某反应堆一回路冷却剂进行测量的结果表明,基于LaBr_3(Ce)γ谱仪的燃料元件破损监测方法可有效避免监测中的干扰因素的影响,降低了定量测量中的不确定度。