Prospects of using different clad materials in a material test research reactor – Part 3 – The dynamic behavior

The effects of using different clad materials on the dynamics of a material test research reactor were studied. For this purpose, the aluminum clad of an MTR was replaced separately with stainless steel-316 and zircaloy-4. Simulations were carried out to determine the reactor performance under reactivity insertion and loss-of-flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that during the fast reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 59.34 MW, while stainless steel-316 cladded attained 48.74 MW and zircaloy-4 cladded attained maximum power of 55.87 MW. During the slow reactivity insertion transient, Al cladded fuel attained the maximum reactor power of 12.38 MW, while stainless steel-316 cladded attained 12.23 MW and zircaloy-4 cladded attained maximum power of 12.34 MW. During the loss-of-flow transients, the reactor power of the stainless steel-316 cladded fuel remained slightly lower than the other two. The fuel temperature of stainless steel-316 and zircaloy-4 cladded fuels remained higher due to poor fuel–clad gap conductance.     

Effects of high density dispersion fuel loading on the dynamics of a low enriched uranium fueled material test research reactor

The effects of using high density low enriched uranium on the dynamics of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different properties affecting the reactor in different ways, fuels U–Mo (9w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to determine the reactor performance under reactivity insertion and loss of flow transients. Nuclear reactor analysis code PARET was employed to carry out these calculations. It is observed that during the fast reactivity insertion transient, the maximum reactor power is achieved and the energy released till the power reaches its maximum increases by 45% and 18.5%, respectively, as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved during the transient, by 27.7 K, 19.7 K and 7.9 K, respectively. The time required to reach the peak power decreases. During the slow reactivity insertion transient, the maximum reactor power achieved increases slightly by 0.3% as uranium density increases from 6.57 gU/cm³ to 8.90 gU/cm³ but the energy generated till the power reaches its maximum decreases by 5.7%. The temperatures of fuel, clad and coolant outlet remain almost the same for all types of fuels. During the loss of flow transients, no appreciable difference in the power and temperature profiles was observed and the graph plots overlapped each other.     

Conceptual design of compact supercritical water-cooled fast reactor with thermal hydraulic coupling

Fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650 °C. Fuel and cladding integrities, and flow-induced vibration were taken into account as design criteria. Designed fuel rod has the diameter of 7.6 mm and is arranged in the fuel assembly with pitch-to-diameter ratio of 1.14. New core arrangement for negative void reactivity is proposed by three-dimensional tri-z core calculation fully coupled with thermal hydraulic calculation, where ZrH layer concept is used for negative void reactivity. The core has high power density of 156 W/cm³ and its equivalent diameter is only 2.7 m for 1000 MWe class reactor core. High average core outlet temperature of 500 °C is achieved by introducing radial fuel enrichment zoning and downward flow in seed assembly. Small pressure vessel size and simplified direct steam cycle with higher thermal efficiency give an economical potential in aspect of capital and operating cost.     

Three-dimensional FE analysis of the thermal–mechanical behaviors in the nuclear fuel rods

In order to implement numerical simulation of the thermal–mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal–mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal–mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm³, the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm² K to 0.01 W/mm² K, while increases up to 54.7% when h decreases from 0.01 W/mm² K to 0.005 W/mm² K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal–mechanical behaviors in the fuel rod; when the gap size varies from 0.03 mm to 0.1 mm, the highest temperature in the pellet increases by 19.7%, and the maximum first principal stress at the outer pellet surface decreases by 17.4%; it is critical to optimize the gap size in order to reduce the pellet–cladding mechanical interaction and avoid their contact at early stage. This study lays a foundation for the further research on the irradiation-induced mechanical behaviors in the fuel rods.     

Influence of evaluated data of fission product poisons on criticality

The main objective of this research is to study the influence of cross-section differences of fission product poisons among various newly released evaluated cross-section libraries ENDFB-VI.8, JENDL3.2, JEF2.2, IAEA, ENDFB-VII and JEFF3.1 on criticality of an MTR type research reactor. The effect of cross-sections of poisons on the reactivity was studied with the help of WIMSD and CITATION codes. Various cross-section libraries were used in SARC (System for Analysis of Reactor Core) code for the fuel cycle analysis. It was found that the negative reactivity induced due to ¹³⁵Xe for the equilibrium core is around 36.00 mk whereas for ¹⁴⁹Sm it ranges from 6.65 to 7.06 mk. The three libraries (JENDL3.2, IAEA and ENDFB-VII) resulted in small increase in the Xenon worth as compared to the other three libraries. For Samarium, JEFF3.1 gives the highest worth whereas ENDFB-VI.8 gives the least worth among the six libraries.     

Measured and simulated reactivity insertion transients characteristics of NIRR-1

The power excursion characteristics of the Nigeria Research Reactor-1 (NIRR-1) under different reactivity insertion transients have been calculated using PARET/ANL 7.3 code. Experimental data of dynamic experiments performed in NIRR-1 facility during initial startup were used to benchmark the calculated data. For ramp insertion of total cold core excess reactivity of 3.77 mK, the reactor quickly reaches a maximum power of 82.3 kW in 390 s and decreases gradually due to the inherent safety features. The calculated maximum power for this insertion is in good agreement with a measured value of 79.4 kW recorded during the commissioning of NIRR-1. Similarly, moderator temperature for the hottest channel with respect to this transient was calculated to be 63.7 °C which is comparable to a measured value of 63.9 °C. Furthermore, step reactivity insertions of 2.23 mK and an approximately 0.32 mK, a regular criticality experimental check carried out at MNSR facilities by Safeguards Inspectors were accurately simulated.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

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.     

10 years of engineering and physics achievements by the Large Helical Device project

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 10²¹ m^?3 at the central temperature of 0.4 keV), high ion temperature (T_i of 5.2 keV at 1.5 × 10¹⁹ m^?3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.     

Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code

Assessment of fuel conversion from high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the Syrian MNSR reactor was conducted in this paper. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed to assess the possibility of fuel conversion from 89.87% HEU fuel (UAl₄–Al) to 19.75% LEU fuel (UO₂). The first model showed that 347 fuel rods with HEU fuel were required to obtain a reactor core with 5.17 mk unadjusted excess reactivity. The second model showed that only 200 LEU fuel rods distributed in the reactor core like the David star figure were required to obtain a reactor core with 4.85 mk unadjusted excess reactivity. The control rod worth using the LEU fuel was enhanced. Finally, the third model showed that distribution of 200 LEU fuel rods isotropically in the 10 circles of the reactor core failed to convert the fuel since the calculated core unadjusted excess reactivity for this model was 10.45 mk. This value was far beyond the reactor operation limits and highly exceeded the current MNSR core unadjusted excess reactivity (5.17 mk).     

Electrical and thermo-mechanical analysis of beam recovery system for megawatt power gyrotron

The paper presents the electrical and thermo-mechanical design of single stage beam recovery system for 120 GHz, 1 MW gyrotron. The electrical study shows that the cylindrical shape single stage beam recovery system enhances the efficiency by 66.26%. The maximum power deposited to collector in depressed collector operation is 0.48 MW for electronic efficiency, 30% and 1.44 MW for DC electron beam. The thermo-mechanical analysis has been performed to evaluate the water cooling system. The cooling system has capability of accommodating a peak wall loading, 0.9 kW/cm² at flow rate of 1500 l/min for safe operating time, 60 ms. Further, a high voltage analysis is also carried out to appraise the electric field distribution in the collector.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Performance analyses of VHTR plants with direct and indirect closed Brayton cycles and different working fluids

This paper investigated the performance of Very High Temperature Reactor (VHTR) power plants with helium working fluid and direct and indirect Closed Brayton Cycles (CBCs), and with binary mixture working fluids of He–Xe and He–N₂ (molecular weight of 15 g/mole) and indirect CBCs. Also investigated are the effects of using low- and high-pressure compressors with intercooling, versus a single compressor, using bleed cooling the reactor pressure vessel in direct CBC helium plants, and varying the reactor exit temperature from 700 °C to 950 °C on the plant thermal efficiency, cycle pressure ratio and the size of and number of stages in the turbine and compressor. Analyses are performed for a shaft rotation speed of 3000 rpm, reactor thermal power of 600 MW and a temperature pinch of 50 °C in the intermediate heat exchanger (IHX) for the indirect CBCs.     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

Prospects of using different clad materials in a material test research reactor – Part 2 – The reactivity feedback coefficients

The reactivity feedback coefficients of a material test research reactor using stainless steel-316 and zircaloy-4 as clad were calculated. For this purpose, the aluminum clad of an MTR was replaced with stainless steel-316 and zircaloy-4. Calculations were carried out to find the fuel temperature reactivity feedback coefficient, clad temperature reactivity feedback coefficient, moderator temperature reactivity feedback coefficient and moderator density reactivity feedback coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 38 °C to 50 °C, at the beginning of life, were maximum in magnitude for stainless steel-316 cladded fuel, followed by aluminum and least for the zircaloy-4 cladded fuel. The fuel temperature feedback coefficient increased in magnitude by 47.37% for stainless steel-316 and decreased by 4.72% for zircaloy-4 clad. The moderator temperature feedback coefficient increased in magnitude by 60.41% for stainless steel-316 and decreased by 3.03% for zircaloy-4 clad, while the moderator density feedback coefficient showed an increase in magnitude of 59.18% for stainless steel-316 and a decrease of 7.63% for zircaloy-4 clad. Zircaloy-4 gave a positive value for clad temperature feedback coefficient, while the others two did not have any clad temperature feedback coefficient.     

Heat loads on FTU liquid lithium limiter

The choice of the best material exposed to the plasma in a future reactor is still an open question. One of main requirements to be satisfied is the capability to withstand high heat loads, in the range 10–20 MW/m², during normal operations in a future reactor, as well as the peak power released by ELMs in H-mode operation. On FTU, since the end of 2005, we have started an innovative program having as main goal the possibility to expose a liquid surface to the plasma. The small wetted area, of the FTU three liquid lithium limiter units, does not allow to use it as main limiter for all the duration of the discharge so that it is always set in the shadow of the main toroidal limiter. In this condition, heat loads up to 2 MW/m² are normally withstood by the liquid lithium limiter without any surface damage and problems to the FTU operations. In order to increase the heat load on the liquid lithium limiter for a controlled limited period, the plasma column is shifted towards the liquid lithium limiter during the discharge. The surface temperature remains constant although the plasma column is pushed on the liquid lithium limiter. This saturation of the surface temperature can be understood considering the dependence of the evaporation rate versus the surface temperature between 250 °C and 550 °C that increases by five orders of magnitude. The evaporated lithium forms a strongly radiative cloud all around the three units limiting the power load on the surface. We do not observe any accumulation of lithium into the discharge as it can be also inferred from the time evolution of the Li III line growing up until the temperature is reaching the maximum value and then remaining almost constant.     

Thermal,structural, and radiological properties of irradiated graphite from the ASTRA research reactor – Implications for disposal

The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from ∼10¹⁷ to ∼10²⁰ n cm⁻² over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g⁻¹ and a Wigner energy accumulation rate of ∼7 × 10⁻¹⁷ J g⁻¹/n cm⁻². The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 10¹⁹ n cm⁻²) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 10¹⁹ n cm⁻²) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10⁻⁶ °C⁻¹. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The ¹⁴C activity in the inner thermal column graphite ranges from 6 to 467 kBq g⁻¹. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage.     

Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 °C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 °C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa √m occurred in room temperature tests when irradiation temperature was below 400 °C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa √m was measured when the irradiation temperature was above 430 °C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 °C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.     

Incorporation of lithium lead eutectic as a working fluid in RELAP5 and preliminary safety assessment of LLCS

The current work involves thermal hydraulic calculation of Lithium Lead Cooling System (LLCS) for the Indian test blanket module (TBM) for testing in International Thermonuclear Experimental reactor (ITER). It uses the RELAP portion of RELAP/SCDAPSIM/MOD4.0. Lithium-lead eutectic (LLE) has been used as multiplier, breeder and coolant in TBM. Thermodynamic and transport properties of the LLE have been incorporated into the code. The main focus of this study is to check the heat transfer capability of LLE as coolant for TBM system for steady state and the considered anticipated operational occurrences (AOO's), namely, loss of heat source, loss of primary flow and loss of secondary flow. The six heat transfer correlation (reported for liquid metals in the literature) has been tested for steady state analysis of LLCS loop and results are roughly same for all of them. A good agreement has been observed between the operating conditions of LLCS with those of RELAP5 calculations. Results from transient calculations show that a maximum temperature of 875 K is attained during a 300 s loss of primary flow (LLE).     

A model for coolant void reactivity evaluation in assemblies of CANDU cells

Coolant void reactivity is a very important safety parameter in CANDU reactor analysis. Here we evaluate the coolant void reactivity in a 2 × 2 heterogeneous assembly of CANDU cells using the code DRAGON. Since the current version of DRAGON can only treat the coolant void reactivity for a single CANDU cell, an approximate model for the geometry must be considered to perform assembly calculations in a 2 × 2 pattern. The model we propose consists of replacing the annular fuel pins by equivalent square fuel pins. The equivalence between annular fuel pins and square fuel pins is brought about by homogenizing the fuel plus its sheath and subsequently conserving the reaction rates between the two geometries using a SPH equivalence procedure. The approximate CANDU cells constructed using square pins were used to perform the transport calculations in 2 × 2 assembly patterns. In addition, the model was used to evaluate coolant void reactivity in 2 × 2 checkerboard voiding patterns. These calculations reflect more accurately the actual voiding situation being studied. This helps in assessing the effects due to the coupling of neutrons born in one cell to those born in the neighbouring cells.