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.     

Safety analysis of the reference research reactor MTR during reactivity insertion accident using the code MERSAT

The IAEA’s reference research reactor MTR-10 MW has been modeled using the code MERSAT. The developed MERSAT model consists of detailed representation of primary and secondary loops including reactor pool, bypass, main pump, heat exchanger and reactor core with the corresponding neutronics and thermalhydraulic characteristics. Following the successful accomplishment of the steady state operation at nominal power of 10 MW, reactivity insertion accident (RIA) for three different initial reactivity values of $1.5/0.5 s, $1.35/0.5 s and $0.1/1.0 s have been simulated. The predicted peaks of reactor power, hot channel fuel, clad and coolant temperatures demonstrate inherent safety features of the reference MTR reactor. Only in case of the fast RIA of $1.5/0.5 s, where the peak power of 133.66 MW arrived 0.625 s after the start of the transient, the maximum hot channel clad temperature arrives at the condition of subcooled boiling with the subsequent void formation. However, due to the strong negative reactivity feedback effects of coolant and fuel temperatures the void formation persists for a very short time so that thermalhydraulic conditions remained far from exceeding the safety design limits of thermalhydraulic instability and DNB. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermalhydraulic system codes.     

Thermal-hydraulic and neutronic analyses of the submersion-subcritical, safe space (S) reactor

The solid-core Submersion-Subcritical Safe Space (S⁴) reactor is cooled with He-28% Xe gas (molecular weight of 40 g/mole) and nominally generates 471 kWth for at least 7 years. To avoid single point failures in reactor cooling and energy conversion, the S⁴ reactor core is divided into three hydraulically independent, but neutronically and thermally coupled sectors. Each sector feeds a separate Closed Brayton Cycle (CBC) power conversion loop with separate heat rejection radiator panels. Detailed thermal-hydraulic analyses of the S⁴ reactor core are performed to ensure that the maximum fuel temperature during nominal operation stays below 1300 K. In addition, a neutronics analysis performed using MCNP 5 confirms that the S⁴ reactor satisfies the design reactivity requirements. These are at least $ 4 of cold clean excess reactivity, at least $ 2.25 of shutdown margin, and at least $ 1 subcritical in the worst-case of submersion and flooding, following a launch abort accident. Mass estimates of the S⁴ reactor design that meets both the thermal and the reactivity requirements are provided.     

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead-bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal-hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320-420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.     

Simulation of reactivity transients in a miniature neutron source reactor core

Computer simulation was carried out for reactivity induced transients in a HEU core of a tank-in-pool reactor, a miniature neutron source reactor (MNSR). The reactivity transients without scram at initial power of 3 W were studied. From the low power level, the power steadily increased with time and then rose sharply to higher peak values followed by a gradual decrease in value due to temperature feedback effects. The trends of theoretical results were found to be similar to measured values and the peak powers agreed well with experimental results. For ramp reactivity equivalent of clean core cold excess reactivity of 4 mk (4×10⁻³ Δk/k), the predicted peak power of 100.8 kW agrees favourably with the experimental value of 100.2 kW. The measured outlet temperature of 72.6 °C is also in agreement with the calculated value of 72.9 °C for the release of the core excess reactivity. Theoretical results for the postulated accidents due to fresh fuel replacement of reactivity worth 6.71 mk and addition of incorrect thickness of Be plates resulting in 9 mk reactivity insertion were 187.23 and 254.3 kW, respectively. For these high peak powers associated with these reactivity insertions, it is expected that nucleate boiling will occur within the flow channels of the reactor core.     

A spatial kinetic model for simulating VVER-1000 start-up transient

An accurate prediction of reactor core behavior in transients depends on how much it could be possible to exactly determine the thermal feedbacks of the core elements such as fuel, clad and coolant. In short time transients, results of these feedbacks directly affect the reactor power and determine the reactor response. Such transients are commonly happened during the start-up process which makes it necessary to carefully evaluate the detail of process. Hence this research evaluates a short time transient occurring during the start up of VVER-1000 reactor. The reactor power was tracked using the point kinetic equations from HZP state (100 W) to 612 kW. Final power (612 kW) was achieved by withdrawing control rods and resultant excess reactivity was set into dynamic equations to calculate the reactor power. Since reactivity is the most important part in the point kinetic equations, using a Lumped Parameter (LP) approximation, energy balance equations were solved in different zones of the core. After determining temperature and total reactivity related to feedbacks in each time step, the exact value of reactivity is obtained and is inserted into point kinetic equations. In reactor core each zone has a specific temperature and its corresponding thermal feedback. To decrease the effects of point kinetic approximations, these partial feedbacks in different zones are superposed to show an accurate model of reactor core dynamics. In this manner the reactor point kinetic can be extended to the whole reactor core which means “Reactor spatial kinetic”. All required group constants in calculations are prepared using the WIMS code. In addition CITATION code was used to calculate the flux, power distribution and core reactivity inside the core. To update the last change in group constants and resultant reactivity in point kinetic equations, these neutronic codes were coupled with a developed dynamic program. This study is applied on a typical VVER-1000 reactor core to show the reactor response in short time transients caused during start-up procedure.     

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Effects of core excess reactivity and coolant average temperature on maximum operable time of NIRR-1 miniature neutron source reactor

We appraised in this study the effects of core excess reactivity and average coolant temperature on the operable time of the Nigeria Research Reactor-1 (NIRR-1), which is a miniature neutron source reactor (MNSR). The duration of the reactor operating time and fluence depletion under different operation mode as well as change in core excess reactivity with temperature coefficient was investigated over a period of five years. Our result shows that there is a strong dependence of reactor operating time on core excess reactivity and temperature coefficient. It was observed in 2004 that with a cold core excess reactivity of 3.77 mk, at full-power flux of 1.0 × 10¹² n cm⁻² s⁻¹ the reactor operated for 5 continues hours. At half-power flux of 0.5 × 10¹² n cm⁻² s⁻¹ and under the same excess reactivity condition, the reactor reaches 8 h of operation. However, re-measurements done in 2009 shows that excess reactivity of the reactor has reduced to 2.80 mk, the operable time at full flux dropped to 3.5 h while that of half-power became 7 h. We also investigated the reactor's energy consumption within the period under study and found to be much more in 2008 compared to the previous years. We infer that the amount of fluence consumed and the excessive reactor usage in 2008 has contributed significantly to the reduction of the reactor's excess reactivity in that year. The results obtained here revels that for an MNSR with a clean core excess reactivity between 3.5 mk and 4.0 mk, 5 and 8 h are the maximum operable times under full and half-power flux conditions, respectively. Negative deviation from these optimum times is therefore an indication of a drop in excess reactivity and the need for beryllium shims addition.     

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.     

Reactivity control options of space nuclear reactors

This paper compares two ex-core control options of the gas-cooled Submersion Subcritical Safe Space (S^4) reactor with a fast neutrons energy spectrum: (a) rotating BeO drums with 120° thin segments of enriched B₄C in the BeO radial reflector; and (b) sliding segments in the BeO radial reflector. Investigated are the effects on the beginning-of-life (BOL) excess reactivity, reactivity depletion rate and operation life, and the spatial neutron flux distributions and fission power profiles in the core. Also investigated is the effect of reducing the thickness of the enriched B₄C segments in the control drums on the BOL excess reactivity, when one or two of the 6 drums are stuck in the shutdown position. Reducing the thickness of the B₄C segments from 0.5 mm to 0.238 mm, with one drums stuck in the shutdown position, increases BOL cold and hot-clean excess reactivity from +$1.71 and +$0.47 to +$2.38 and +$0.89, respectively. These reactivity values are almost identical to those of the reactor with one of the six reflector segments stuck open in the shutdown position. Results also showed that the control options made little difference in the reactor performance. The power peaking in the reactor core with sliding reflector segments is slightly lower and the spatial power profiles are relatively flatter. The operation life of the reactor with a sliding reflector segments control, when operating at a nominal thermal power of 471 kW, is only 22 full power days longer than with rotating drums control.     

Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

Thermal hydraulic design of a liquid salt-cooled flexible conversion ratio fast reactor

A 2400 MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCl₂ (30-20-50%) as coolant. The reference design uses a wire-wrapped, hexagonal lattice core, and is able to achieve a core power density of 130 kW/l with a core pressure drop of 700 kPa and a maximum cladding temperature under 650 °C. Four kidney-shaped conventional tube-in-shell heat exchangers are used to connect the primary system to a 545 °C supercritical CO₂ power conversion system. The core, intermediate heat exchangers, and reactor coolant pumps fit in a vessel approximately 10 m in diameter and less than 20 m high. Lithium expansion modules (LEMs) were used to reconcile conflicting thermal hydraulic and reactor physics requirements in the liquid salt core. Use of LEMs allowed the design of a very favorable reactivity response which greatly benefits transient mitigation. A reactor vessel auxiliary cooling system (RVACS) and four redundant passive secondary auxiliary cooling systems (PSACSs) are used to provide passive heat removal, and are able to successfully mitigate both the unprotected station blackout transient as well as protected transients in which a scram occurs. Additionally, it was determined that the power conversion system can be used to mitigate both a loss of flow accident and an unprotected transient overpower.     

基于TREND程序的石墨通道液态熔盐堆无保护反应性引入事故初步分析

采用自编系统分析程序TREND,基于液态点堆动力学模型,针对10 MW石墨通道液态熔盐堆的设计,研究分析不同反应性在阶跃引入和线性引入情况下10 MW石墨通道液态熔盐堆堆芯功率、石墨温度和堆芯出口熔盐温度的瞬态变化。结果表明,阶跃引入低于570pcm（1pcm=10?5）反应性,堆系统能在无保护的情况下安全运行;当单根控制棒失提引入约800pcm时,反应性引入速率不超过8pcm/s,反应堆能够利用自身的温度、功率负反馈特性有效地控制功率峰值和降低堆芯出口温度,保证反应堆在无保护情况下安全运行。因此,液态熔盐堆具有良好的固有安全性。      

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl₂ (LSFR), sodium (SFR), and supercritical CO₂ (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO₂ power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO₂ reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.     

Long operation life reactor for lunar surface power

The Pellet Bed Reactor (PeBR) with an operational life of 66 full-power years is developed for lunar surface power. It has Inconel X750 structure and vessel and would be launched unfueled then loaded with spherical fuel pellets (∼1.0 cm dia.) on the lunar surface after being placed below grade and surrounded with regolith. The pellets, comprised of ZrC-coated UC particles (∼850 μm in dia.) dispersed in ZrC matrix, are delivered to the lunar surface in subcritical canisters. The canisters are designed to remain sufficiently subcritical during launch and when submerged in wet sand and flooded with seawater in the unlikely event of a launch abort accident. The PeBR power system nominally generates ∼100 kW_e at a thermal efficiency of ∼21% and a reactor exit temperature of 910 K. It employs three separate closed Brayton cycle (CBC) loops each with a turbo-machine unit for energy conversion and two water heat pipes radiator panels for heat rejection. The reactor coolant and CBC working fluid is He-Xe binary gas mixture (40 g/mol). Estimates of the hot-clean excess reactivity and the full-power operation life are obtained using neutronics and fuel depletion analyses. In addition, estimates of the total radioactivity in post-operation PeBR, while being stored below grade on the lunar surface, are determined for up to 1000 years.     

Analysis of fuel management in the KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.     

Thermal hydraulic analysis of Syrian MNSR research reactor using RELAP5/Mod3.2 code

This paper presents the development and validation of a MNSR-RELAP5 model. MNSR is a 30 kW, light-water moderated and cooled, beryllium-reflected, tank in pool type research reactor. A RELAP5 model was set up to simulate the entire MNSR system. The model represents all reactor components of primary and secondary loops with the corresponding neutronic and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents are simulated.