Fuel irradiation of the first batches produced for the Chinese HTR-10

An irradiation test of four spherical fuel elements (SFE) had been performed in the Russian reactor IVV-2M. The elements were sampled randomly from the first and second product batches which were manufactured for the 10 MW high-temperature gas-cooled test reactor (HTR-10). The maximum burnup of the irradiated fuel elements reached 107,000 MWd/tU and the maximum fast neutron fluence was 1.31 × 10²⁵ m⁻². The release-to-birth rate ratio (R/B) did not increase significantly during irradiation. However, an in-pile heating-up test of element SFE 7 in Capsule 5 led to a failure of approximately 6% of the coated particles. After the test it was estimated that the fuel temperature had very likely been much higher than the intended 1600 °C.     

Manufacture and characteristics of spherical fuel elements for the HTR-10

The R&D of spherical fuel elements for the 10 MW high temperature gas-cooled reactor (HTR-10) started in 1986 in China. A process known as cold quasi-isostatic molding was used for manufacturing spherical fuel elements, and about 20,540 spherical fuel elements were produced in 2000 and 2001. Fabrication technology and graphite matrix materials were investigated and optimized. Cold properties of the spherical fuel elements met the design specifications. The mean free uranium fraction of 44 batches was 4.57 × 10⁻⁵. In-pile irradiation test results showed that irradiation did not lead to apparent change in linear dimensional, geometrical density, porosity and strength of matrix graphite samples. No cracks and blisters were observed in spherical fuel elements. This indicated that matrix graphite and spherical fuel elements of HTR-10 met the requirement of design specifications.     

Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor

The Chinese 10 MW high temperature gas-cooled reactor (HTR-10) attained its first criticality on December 21, 2000. The fabrication of the first fuel for the HTR-10 started in February 2000 at the Institute of Nuclear Energy Technology (INET), Tsinghua University. Up to September 2000, a total of 11 721 spherical fuel elements were successfully produced. The average free uranium fraction of the first fuel-determined by the burn-leach method-was 5.0×10⁻⁵. So far, the release rate R/B of the fission gas, measured in the irradiation test, shows that not a single particle in three irradiated spherical fuel elements failed as the results of the irradiation test carried out in Russia. This paper describes the design parameter, the fabrication technology and the performance data of the HTR-10 first fuel, and the production and quality control experiences obtained from the manufacture of the first fuel for the HTR-10.     

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S4) reactor

The He–Xe gas-cooled, S4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.     

Irradiation performance and modeling of HTR-10 coated fuel particles

The irradiation test of HTR-10 spherical fuel elements was carried out in the Russian IVV-2M research reactor with the irradiation temperature of 1000 ± 50 °C. After the burnup reached 100,000 MWd/t, the irradiation temperature was raised to a higher temperature. The high R/B levels observed during the normal irradiation test were due to manufacture defects of one to four coated particles. Post-irradiation examination indicated that at normal irradiation condition, the pyrolytic carbon (PyC) and silicon carbide (SiC) layers of particles kept their integrity. However, after irradiation at higher temperatures, several types of defects including radial and tangential cracks in SiC layers, cracks in buffer layers, and through coating failure were found, and the failure fraction reached 5.8 × 10⁻². These defects were most likely caused by the higher thermal stresses generated. In this study, PANAMA fuel performance code was used to estimate the heating temperature in the irradiation test. The calculated results showed that when the heating temperature is much higher than 1850 °C, the failure fraction of coated particle can reach the level of 1%.     

Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core

The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements, the helium coolant and the main components in the reactor. In the temperature calculation of fuel elements, several uncertain factors are considered carefully, including non-uniform burnup, power distribution deviation, manufacture deviation of fuel elements, graphite balls mixed with fuel balls in the core, calculation deviation of heat transfer and so on. In the flow distribution calculation, the conservative pebble bed core flow value is selected. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core.     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Gas-cooled fast breeder reactor fuel bundle irradiations in the BR2 helium loop

In the BR2 helium loop at Mol, Belgium, a 12-pin test fuel element of gas-cooled fast breeder reactor (GCFR) design and materials will be irradiated at a 500 W/cm maximum pin rating and a 700°C maximum cladding temperature to a target burnup of 60 MWd/kg (extension to 100 MWd/kg is intended). The design of the test element and the loop is described in detail. To fabricate the test element, parts of the GCFR fuel development had to be anticipated. Preliminary out-of-pile testing was successfully performed, and irradiation is scheduled to start in early 1977 and will be completed between mid-1978 and mid-1979, depending on the final burnup objective. GCFR operating conditions will be completely simulated except for the full size of the fuel element and the fast neutron flux. In combination with out-of-pile performance testing of full-size dummy elements and fast flux experience from the liquid metal fast breeder reactor program, the helium loop irradiation is regarded as an adequate basis for the design of a fuel element for a GCFR demonstration plant serving as the final test bed.     

Fast flux irradiation tests performed at high temperature

The first gas-cooled fast breeder reactor (GCFR) fast flux irradiation experiment [F-1(X094)] consists of seven fuel rods clad in 20% cold-worked 316 stainless steel. The rods are individually encapsuled, with sodium filling the gaps within the capsule walls. The rods are fueled with (15% Pu, 85% U)O₂ and have depleted UO₂ lower and upper axial blankets and charcoal to trap volatile fission products. The cladding i.d. temperature range covered by these rods is 570–760°C (1055–1400°F).The in-reactor performance of the fuel rods in the F-1 high-temperature experiment, which achieved a burnup of 121 MWd/kg (13.0 at.%) on the lead rod, is described. All rods in the experiment have remained intact. The results of interim examinations [at 25 and 50 MWd/kg (2.7 and 5.4 at.%)] of fuel and fission product behavior and transport and comparisons of observed results with LIFE-III code predictions are described.The F-3 experiment, which consists of ten encapsulated GCFR fuel rods with surface-roughened (ribbed) cladding, shares a nineteen capsule subassembly with Argonne National Laboratory. Temperatures are controlled over the range 675°C (1250°F) to 750°C (1380°F). Irradiation is in the core region of the EBR-II and thus permits achievement of a higher fluence-to-burnup ratio than that obtained in the F-1 experiment.Preliminary results of a planned interim examination at an exposure of 46 MWd/kg (4.9 at.%) burnup and a fluence of 5.2 × 10²² n/cm² show that cladding failures occurred in nine of the ten rods. Preliminary indications are that the failures are due to defects in the sodium bond between the fuel rod and the capsule.The tests completed and currently under way have been scoping in nature, and irradiation in EBR-II of GCFR prototypical fuel (pressure equalized) rods with ribbed cladding is required to provide the information needed for reactor design on effects of exposure to high fluence and burnup and on design reliability for a statistically significant number of rods. The design and the operating conditions for the F-5 experiment being prepared for this purpose are described.     

The OECD Halden reactor project fuels testing programme: methods, selected results and plans

The fuels testing programme conducted in the Halden reactor (heavy boiling water reactor (HBWR)) is aimed at providing data for a mechanistic understanding of phenomena, which may affect fuel performance and safety parameters. The investigations focus on implications of high burnup and address thermal property changes, fission gas release as influenced by power level and operation mode, fuel swelling, and pellet–clad interaction. Relevant burnup levels (>50 MWd kg⁻¹ U) are provided through long-term irradiation in the HBWR and through utilisation of re-instrumented fuel segments from commercial light water reactors (LWR). Both urania and MOX fuels are being studied regarding thermal behaviour, conductivity degradation, and aspects of fission gas release. Experiments are also conducted to assess the cladding creep behaviour at different stress levels and to establish the overpressure below which the combination of fuel swelling and cladding creep does not cause increasing fuel temperatures. Clad elongation measurements provide information on the strain during a power increase, the relaxation behaviour and the extent of a possible ratcheting effect during consecutive start-ups. Investigations foreseen in the programme period 2000–2002 include the behaviour of MOX and Gd-bearing fuel and other variants developed in conjunction with burnup extension programmes. Some LWR-irradiated fuel segments will undergo a burnup increase in the HBWR to exposures not yet achieved in LWRs, while others will be re-instrumented and tested for shorter durations.     

Design and experience of HEU and LEU fuel for WWR-M reactors

Thin-walled WWR-M5 fuel elements were designed and manufactured and have been used successfully for 16 years; they contain twice as much uranium-235 as the WWR-M2 and WWR-M3 fuel elements. The fuel elements have been optimized with regard to their neutron physics and thermal-hydraulic parameters and fuel consumption has been minimized. The mean specific power in the core of the WWR-M reactor was raised to 230 kW l⁻¹, the measured maximum volume thermal specific power was 900±100 kW l⁻¹ and the surface specific power was 136±15 W cm⁻². The WWR-M5 fuel elements enable the power of the WWR-M pooltype reactor to be raised to 30 MW while simultaneously increasing the number of cells in the core available for experimentation by a factor of approximately two and reducing fuel element consumption. Reactor tests of WWR-M fuel elements with reduced fuel enrichment (36 and 21%) were carried out for a meat uranium density up to 2–3 g cm⁻³. Conversion of WWR-SM-type reactors to these fuel elements did not lead to a loss in reactivity and enabled their power to be increased to 20–30 MW.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Neutronic and thermal hydraulic design of the graphite moderated helium-cooled high flux reactor

A new design concept for a high flux reactor was investigated, where a graphite moderated helium-cooled reactor with pebble fuel elements containing (²³⁵U, ²³⁸U)O₂ TRISO coated particles was taken as its design base. The reactor consists of an annular pebble bed core, a central heavy water region, and inner, outer, top, and bottom graphite reflectors. The maximum thermal neutron flux in its central heavy water region as high as 10¹⁵ cm⁻² s⁻¹ with thermal neutron spectral purity of more than two orders of magnitude and a large usable volume of more than 1,000 liters can be achieved by (1) diluting the fissile material in the core and (2) optimizing the core-reflector configuration. The in-core thermal-hydraulic analysis was done to obtain adequate information about the coolant flow pattern and pressure distribution, core temperature profile, as well as other safety aspects of the design. To protect the reactor during off-normal or accident events, a large margin of temperature difference between the operating fuel temperature and the fuel limit temperature is confirmed by lowering the coolant inlet and core rise temperatures.     

Symbiotic systems consisting of large-FBR and small water-cooled thorium reactors (WTR)

Small long life water-cooled thorium reactors (WTR; 30–300 MWth) have been investigated. For realizing thorium cycle of the reactors, a uranium–thorium mixture core is introduced to fast breeder reactors (FBR; 3000 MWth) to be a ²³³U producer. In the present study, two distinct metallic fuel pins, with natural uranium and thorium, are loaded into a large sodium-cooled FBR. The FBR itself is self-sustained by the plutonium produced in the uranium pins. Under the equilibrium burnup state, the FBR spent fuels are periodically discharged with a certain discharge rate and then separated. Some actinides are returned to the FBR core while ²³³U, which is discharged from the thorium pins, is utilized for the WTR fresh fuel. Fissile support capability is the main investigated parameter of the study. The system achieves higher support capability at higher burnup and lower power of the WTR, and shows that larger number of uranium pins is better for the FBR criticality while larger number of thorium pins and lower burnup give better support factor capability. For a symbiotic system consisting 3000 MWth FBR and 100 MWth WTRs, where discharged fuel burnup is 96 and 60 GWd/t for the FBR and WTRs, one FBR can support 5 WTRs.     

Plutonium credit in the low enriched uranium fuel of PARR-1

The amount of plutonium (Pu) isotopes and the resultant savings of ²³⁵U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations.The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 × 10⁻³Δk/k. Amount of ²³⁵U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.     

The analysis of the insertion of UO2 microspheres in the gap between pellet and clad as a new design solution to increase the burnup of the nuclear fuel

The insertion of UO₂ microspheres (eventually graphite coated) in the gap between pellet-clad is observed to decrease substantially the clad hoop plastic strain concomitantly with the elimination of the rim effect at high burnup if low enrichment is used for the microspheres. Taking into account the special features of the specialized finite element code ELFIN'90 for the behavior of fuel elements, it was possible to introduce this new type of material viewed as a granular media. The results of the new code version ELFIN'MS applications to a PHWR fuel for a power ramp irradiation history show that the hoop plastic strain is reduced by about three times in comparison to standard fuel, and that the ridge phenomenon disappears. To establish critical plastic strain limit for irradiated clad failure onset, quantitative evaluations of iodine chemisorbtion on graphite and at the surface of the irradiated zircaloy, are presented. The indications on technology procedure are also discussed. Therefore, the insertion of 2–3 layers of UO₂ microspheres of 100 μm diameter, graphite coated to retain corrosive fission products for clad and with the diameter greater than the design gap, can be considered a design solution to increase the burnup of nuclear fuel.     

High burnup performance of annular UO2 fuel with inter-pellet graphite discs

We have irradiated annular UO₂ fuel with inter-pellet graphite discs at linear powers of 62 and 44 kW/m to a maximum burnup of 775 MWh/kgU (32000 MWd/TeU). The combination reduced fission gas release by up to a factor of four compared with that for annular fuel alone, for absolute releases up to 45%. Disc compatibility with other fuel components was acceptable.     

Reactor physics ideas for large scale utilization of thorium in gas cooled reactors

The physics principles for maximizing the fertile to fissile conversion were used in developing reactor concepts for large scale utilization of thorium in thermal and fast reactors (Jagannathan & Pal, 2006; Jagannathan et al., 2008). It is recognized that these principles are very well suited for ‘He’ gas cooled reactors with graphite moderator since both helium gas coolant and the graphite moderator have low neutron absorption characteristics and thus gives better neutron economy. In this paper, these ideas are applied to the High Temperature Test Reactor (HTTR) core of Japan to assess its advantage over the present day gas cooled reactors. HTTR is helium cooled and graphite moderated system. Significant amount of thorium has been loaded in the HTTR core with some minimal changes in the existing core design. The modified design is called HTTR-M core.In the HTTR-M core, the fuel is changed from enriched UO₂ fuel to Pu in ThO₂ fuel. The locations of boron type burnable poison rods within each fuel assembly of HTTR are replaced by one cycle irradiated thoria rods. Also, the B₄C type control assembly around the HTTR core is replaced by fresh seedless thorium assembly. The fertile thoria assembly are scattered uniformly in the HTTR-M core. The equilibrium core of HTTR-M shows very small burnup reactivity swing. The core excess reactivity is ∼18 mk at BOC and reduces to 1 mk at 660 days. It is interesting to note that this small reactivity change is intrinsically achieved by the choice of seed and fertile dimensions and their contents without the use of burnable poison rods or mechanical control rods which are used in HTTR core. The burnup reactivity swing in the latter after using burnable poison is ∼100 mk. The fissile seed inventory ratio (FIR) in a fuel cycle is 0.90 as compared with 0.717 of HTTR core. Since ²³³U is a better fissile nuclide with highest ‘η’ value in thermal range, the above conversion ratio can be regarded as quite good.     

Analysis of relative release rates of radionuclides from the RBMK-1500 reactor fuel elements

Specific activities (concentrations) of fission products (FP) and activation products in spent fuel elements of the RBMK-1500 reactor were calculated using SCALE 5 computer code. Different burnup (5.1–21.0 MWd/kg) fuel assemblies were experimentally investigated. Activities of radionuclides present in the coolant water of storage cases of defective fuel elements were experimentally measured and analyzed. Experimental results provide a basis for a quantitative analysis of radionuclide release from spent fuel of the RBMK-1500 reactor. Relative release rates of radionuclides from the fuel matrix were assessed based on a comparison of experimental results with theoretical calculations. On the basis of analysis results released fission and activation products can be divided into several groups according to their release rates from fuel; this can be generalized for radionuclides with similar chemical properties.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.