Reactivity balance in the IPR-R1 TRIGA reactor

Considering that the power of the IPR-R1 TRIGA reactor, located at the Nuclear Technology Development Center, Brazil, will be increased from 100 kW to 250 kW, some experiments were done in order to evaluate the magnitude of the reactivity effects associated with the reactor operation. The core excess of reactivity obtained was 1.99 $, and the shutdown margin was 1.33 $. The reactivity needed to operate the IPR-R1 reactor at 100 kW was 0.72 $, mainly due to the prompt negative temperature coefficient. A significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels. The loss of reactivity due to xenon poisoning after 8 h of operation at 100 kW was around 0.20 $, and the highest reactivity loss value caused by a void inserted in the central thimble was 0.22 $. From the results obtained, it was possible to balance all the determined reactivity losses with the reactivity excess available in the reactor, considering the present and the future reactor power operation.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

Measurements of the isothermal,power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center – CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Measurements of the isothermal, power and temperature reactivity coefficients of the IPR-R1 TRIGA reactor

The aim of this paper is to present the experimental results of the isothermal, power and temperature coefficients of reactivity of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center - CDTN in Brazil. The measured isothermal reactivity coefficient, in the temperature range measured, was −0.5 ¢/°C, and the reactivity measurements were performed at 10 W to eliminate nuclear heating. The reactor forced cooling system was turned off during the measurements. When the reactor is at zero power there is no sensible heat being released in the fuel, and the entire reactor core can be characterized by a single temperature. The power coefficient of reactivity obtained was approximately −0.63 ¢/kW, and the temperature reactivity coefficient of the reactor was −0.8 ¢/°C. It was noted that the rise in the coolant temperature has contributed only with a small fraction to the observed negative effect of the reactivity. The power defect, which is the change in reactivity taking place between zero power and full power (250 kW), was 1.6 $. Because of the prompt negative temperature coefficient, a significant amount of reactivity is needed to overcome temperature and allow the reactor to operate at the higher power levels in steady state.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

MCNP程序在反应堆堆芯建模中的应用

用MCNP程序对清华大学试验核反应堆一号堆芯进行了建模,计算了正常棒位下的Keff值,计算结果与参考值吻合较好;提出了用MCNP进行反应堆堆芯建模的一般步骤和方法,此步骤和方法对研究其他各种反应堆堆芯的建模具有参考价值.     

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k_∞ for this methodology comparing the calculated and the experimental results.     

Determination of criticality safety MCNP5 calculation bias by using different libraries of cross section data

Application of different cross section libraries and different versions of Monte Carlo code MCNP has an influence on the calculation results and therefore determination of criticality safety calculation bias forms part of improving accuracy of simulations using computational systems and codes. In this paper, criticality calculations results are presented for 21 problems coming from the International Handbook of Evaluated Criticality Benchmark Experiments (International Handbook, 2007). All of these problems are related to VVER-440 reactors because of their extensive use in Slovakia. Three libraries of cross section data were investigated:

• •JEFF-3.1 General purpose library,
• •ENDF/B-VII library,
• •JENDL 4.0.

Calculations were provided with MCNP5-1.40 and MCNP5-1.60 transport codes. Two cluster systems situated at our Institute were used. Main purpose of this analyses was the determination of the bias which should be used in further simulations.     

零功率微堆的蒙特卡罗数值模拟计算

本文用蒙特卡罗程序MCNP/4B计算了中国原子能科学研究院零功率微堆的堆芯物理参数、铍反射层价值和中子实验孔道的部件反应性价值等数值,并与实验测量值比对.结果表明,两者符合一致,验证了理论计算模型的可靠性.     

Improvements of the reactivity devices modeling for the advanced CANDU reactor

In the context of the ACR™ (Advanced CANDU Reactor), 3D transport calculations are required in order to simulate the reactivity devices located perpendicularly to the fuel channels. The computational scheme that is usually used for CANDU-6 and ACR reactors is based on a simplified supercell geometry in which the fuel clusters and devices are replaced by annuli. Recently, an exact modeling of 3D supercell configurations was introduced within the framework of the ACR calculations. However, with such a model, fine meshing requirements lead to problems that are very demanding in terms of computational resources.

In this paper, we present improvements introduced in the ACR context to reduce the cost of the 3D supercell calculations. Two avenues of investigations are reported. First, the introduction of an accelerated characteristics method permits to reduce the computational burden of such calculations involving a large number of regions. In addition, contrarily to CANDU-6 supercell configurations, the ACR 3D geometry is prismatic and consequently a special tracking procedure can be used. This approach introduces no approximation and is significantly faster than the general 3D tracking technique. Thanks to these modifications in the computational procedure, 3D supercell calculations with a level of mesh discretization comparable to 2D cell configurations become affordable for industrial applications.     

The Monte Carlo codes MCNP and MCU for RBMK criticality calculations

The modern Monte Carlo codes MCNP and MCU have been established as important tools to determine the neutronic behavior of reactor cores. For a comparison of their capabilities, detailed representations of seven critical experiments performed at the Russian Research Center ‘Kurchatov Institute’ were developed, identical in geometry and material composition for each code. Despite the different philosophy of code development, especially in the process of generating cross-section tables, the calculated isotopic reaction rates and the flux distributions are in excellent agreement. Effective multiplication factor and void reactivity effect agree well with experiment. Additional uniform lattice calculations confirm the equivalent potential of MCNP and MCU, but exhibit significant differences to results achieved with transport codes like WIMS-D4.     

Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be ¹³³Ba, ⁶⁰Co and ¹⁵²Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jo?ef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. ¹³³Ba, ⁴¹Ca) are not included in the IAEA and EU basic safety standards.     

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.     

Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10¹⁷, 6.1 × 10¹⁶ (E < 0.55 eV), 7.6 × 10¹⁶ (E > 0.1 MeV) and 4.0 × 10¹⁶ (E > 1 MeV) m⁻² s⁻¹. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Accident scenarios of the TRIGA Mark II reactor in Vienna

The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequences to the environment. The destruction of the cladding of the most activated fuel element, the destruction of all fuel elements and a plane crash were considered scenarios in that report. The calculations were made in 1978 with the software program named STRISK. In this paper, the program package PC Cosyma was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were calculated with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. This paper focuses on two accident scenarios: the destruction of the cladding of the fuel element with the highest activity content and the case of a large plane crash. The current accident scenarios show good agreement with the calculations from 1978, hence no technical modifications in the safety report of the TRIGA reactor Vienna were necessary. Even in the very worst case scenario – complete destruction of all fuel elements in a large plane crash – the expected doses in the Atominstitut's neighborhood remain moderate.     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.