Fuel performance of hollow pellets for fast breeder reactors

Three fuel rods containing hollow mixed oxide (MOX) pellets of uranium and plutonium oxides were fabricated and irradiated at a high linear heat rate (LHR) to burn-up of nearly 30,000 MWd/t in the experimental fast rector, JOYO MK-II. After irradiation, one of the fuel rod pellets was examined by X-ray CT and conventional nondestructive and destructive methods.

Swelling rate was evaluated by both dimensional change and radial density distribution. There were no differences between both types of results and it was concluded that swelling rate can be examined in detail by the X-ray CT technique without dismantling the assembly. In addition, the swelling rate of hollow pellets was nearly the same as values reported for the fuel rods containing solid pellets. LHR was higher in the examined fuel rod containing hollow pellets than in the conventional fuel rod containing solid pellets, but fission gas release rates for both fuel rods were nearly the same.     

Cellular convection in vertical annuli of fast breeder reactors

In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.     

Fracture mechanics evaluation of the structural material for fast breeder reactors

Results of fracture mechanics investigations on austenitic steels used for LMFBRs (Liquid Metal Fast Breeder Reactors) are presented. A summary of reported tests on straight piping and elbows with through wall flaws is given which agree well with predictions made by using a plastic instability model. Crack growth experiments and calculations indicate that initial flaws will not extend significantly during service. Even if considerable crack growth is postulated cracks will penetrate the piping wall with a high safety margin to unstable crack configurations. Theoretical investigations of flawed structures under high strains show that the effect of crack size can be discussed similarly to the elastic range. The information demonstrate that with respect to the design requirements and operating conditions of LMFBRs a sudden rupture of the piping can be excluded. The integrity of the coolant boundary is given also in case of initial flaws.     

Use of nonuranium diluents for plutonium in large fast breeder reactors

The authors study the physical characteristics of fast breeder reactors with cylindrical and ringshaped cores, and also the characteristics of infinite lattices of heterogeneously-arranged large fuel cassettes distributed in the breeding-zone material.It is shown that there are certain reactors with optimum doubling period.Translated from Atomnaya Énergiya, Vol. 21, No. 2, pp. 84–92, August, 1966.     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

Inservice inspection of the reactor block of sodium-cooled fast breeder reactors

By using sodium as coolant special boundary conditions result for the inservice inspection (ISI) of fast breeder reactors. For that reason in general it is not successful applying the methods and equipment proved for the 151 of light water reactors.This report presents inspection methods and equipment developed for the ISI of the reactor block of sodium cooled fast breeder reactors. The survey takes into account the state of the art as well as some R&D-work at home and abroad. Entering into particulars the methods and equipment used for leak monitoring, the inspection of the reactor vessel wall, the inspection' of reactor internals above and below the sodium level, monitoring of structure home noise and the measurement of the gap between the reactor vessel and the guard vessel are described.     

Structural response of fast breeder reactors to hypothetical core disruptive accidents

Liquid metal-cooled fast breeder reactors (LMFBRs) so far have been analyzed for the consequences on the plant and the environment for hypothetical core disruptive accidents (HCDAs). To provide the appropriate analytical tools for this effort, analysis and codes are currently under development in several countries. They combine the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage stresses, strains, and deformations of the important components of the system, and the overall adequacy of the primary and secondary containments. The effort is partitioned into the structural analysis of (a) the core components, and (b) the primary system components beyond the core.The core mechanics effort covers the structural response of fuel pins, hexcans, fuel elements, and fuel element clusters to transient pressures and thermal loads. Two- and three-dimensional finite element codes are under development for these core components. The results of these analyses would permit evaluation of the adequacy of the heat removal process to continue following severe core component deformations. Also, these analyses are currently being combined with neutronics, for the core transition phase, to allow for the mass movements for realistic neutronic calculations.The primary system and containment program treats the structural response of the components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which in their combined form provide greater accuracy and longer durations for the treatment of HCDAs. More recently the codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. This step will permit treatment of the instabilities following slug impact, the ultimate reversal of the sodium slug with the rising bubble, the bubble break-up, and the calculations of sodium splillage and radioactive gases, if any, in the secondary containment. The extent of sodium spillage and sodium fires should be known for evaluation of the secondary containment. The mechanics of bubble migration are needed for radiological study of post-accident phenomena. More recently dynamic fracture mechanics considerations are being incorporated to remove arbitrary failure criteria imposed on components such as the core barrel and vessel.Most recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of the primary piping. The pulses are provided at the vessel primary piping interfaces of the inlet and outlet nozzles. The calculation includes the elbows and pressure drops along the components of the primary piping system. Pressures larger than the ones used as input at the inlet and outlet nozzles were observed. As expected, they occur far from the nozzles, in the pipe, where the pulses meet.Recent improvements to the primary containment codes include introduction of bending strength in materials, Lagrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. A further development involves the combination of a 2-D finite element code for the reactor cover with the 2-D finite-difference hydrodynamic code for continuous monitoring of stresses, strains, and deformations in the cover, as well as pressure changes in the hydrodynamic code. Substantial experimental effort is in progress in various countries on the response to energy releases of vessels and internals, piping systems, subassemblies, and subassembly clusters. These experimental results are being utilized for the verification or modification of the analyses and codes under development.     

Accelerator breeder reactors

In this work, we briefly describe the accelerator breeder reactors (ABR) and their possible uses both for production of energy and isotopes. This study indicates that ABRs can produce fuel, which would generate 2–15 times the electrical energy, ABRs consume. The energy gain depends on the type of ABR used. ABRs should also have several important advantages in safety over the modern breeder reactors. First, they have criticality less then 1 which makes an accident less likely. Second, they can be turned off any moment when their accelerator is turned off.     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

Investigation of material relocation phenomena during severe accidents in fast breeder reactors

The SIMBATH out-of-pile experiments simulate severe accidents in fast breeder reactors. In the tests the nuclear energy released is substituted by the exothermal energy of a thermite reaction. Single pin and small bundle experiments as well as freezing tests are performed. Material ejected from the fuel rod simulators in an early phase is finely dispersed. A portion penetrates the upper breeding zone without freezing. The bulk of molten material ejected afterwards leads to blockages in the colder zones of the bundle. Under these conditions bottled-up situations may occur in the SIMBATH experiments. Residual sodium may become entrapped. The current version of the computer code CALIPSO developed to interpret these experiments is verified by calculation of two single pin experiments. The computations show that the relocation mechanisms in the SIMBATH experiments are mainly controlled by expansion of noncondensible gases originally existing inside the pins. The contribution from fuel vapour pressure or from sodium evaporation due to fuel-coolant-interaction is of less importance during the first 100 ms after fuel pin failure.     

Studies of core disruptive accidents and licensing aspects of fast breeder reactors

This paper discusses the role of the core disruptive accident (CDA) in the safety evaluations and licensing of Liquid Metal Fast Breeder Reactors (LMFBR). Parametric studies of transient overpower (TOP) accidents based on calculations for SNR-300 using the HOPE computer code are presented. Major uncertainties in TOP analysis are identified and discussed with emphasis on the need for reliable fuel failure criteria. A series of calculations illustrating the possible behavior of the U.S. LMFBR demonstration plant following a loss-of-flow (LOF) accident without scram using the SAS-IIIA computer code are described. It is shown that for a beginning of life (BOL) core and end of equilibrium cycle (EOEC) core, the reactivity effects from sodium voiding and clad motion can lead to further sustained reactivity additions from subsequent fuel motion and FCI driven sodium voiding. In these calculations we have used the fuel enthalpy criterion which predicts clad failure around the core midplane. For the EOEC case these effects can add sufficient reactivity to take the system above prompt-critical (LOF driven TOP) and into hydrodynamic disassembly. For the BOL case the sodium void may not be sufficient to bring the system near sustained prompt-critical. However, clad motion appears to be effective in raising the reactivity to prompt-criticality. These results are based on clad failure dynamics modeling in SAS-IIIA. Further work is needed in the area of fuel-clad behavior under severe transients before definitive conclusions can be drawn regarding the applicability of current clad failure models at high clad temperatures (>1000°C). The potential significance of a new concept in CDA analysis called the “transition phase” is briefly mentioned.     

Core surveillance of liquid metal fast breeder reactors by delayed neutron signals

Concerning safe operation of LMFBRs, information is reviewed on delayed neutron signals from defective fuel pins including the results of recent investigations like: operation of KNK II with defective fuel pins, the inpile blockage experiments Mol 7C, and loop experiments with artificially defected fuel pins. The main result is that operation with cladding failures can be tolerated in principle under safety aspect, both improvements in the interpretation and comprehension of signals are required for the determination of admissible limits.     

Fast-fission tokamak breeder reactors

A fast-fission blanket around a fusion plasma exploits high neutron multiplication for superior breeding and high-energy multiplication to generate significant net electrical power. A major improvement over previous fast-fission blanket concepts is the use of mobile fuel, namely a pebble-bed configuration with helium cooling. Upon loss of coolant, the mobile fuel can be gravity-dumped to a separately cooled dump tank before excessive temperatures are reached. The pebble bed is also compatible with rapid fuel exchange and a low-cost reprocessing method. With the ignited tokamak plasma producing 620 MW of fusion power, the net electric power is 1600 MW_e and the annual fissile production exceeds 3 tonnes.     

The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod ejections with ejection times of 2, 10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR.The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used.The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of²³⁸U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30 s, an uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses.     

Exploring new coolants for nuclear breeder reactors

Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the cancelled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants.     

A feasibility study of ferro-boron as in-core shield material in fast breeder reactors

Shields around core and blankets form major part of reactor assembly in fast reactors as the incident neutron spectrum is hard with negligible thermal component and has anisotropic angular distribution. In this paper, a study is presented on the use of ferro-boron as neutron shield material in pool type fast reactors. The reference case chosen is the Prototype Fast Breeder Reactor (PFBR), a 500 MWe which is sodium cooled, pool type, mixed oxide (MOX) fuelled reactor, which is under construction at Kalpakkam, India. It is shown through 2D transport calculations, carried out using 175 neutron multigroup cross-sections, that this low cost material as shield is capable of satisfying the radiological safety criteria as well as the shields in the reference case. The secondary sodium activity and dose in steam generator building are marginally lower than the reference case. The total shield material weight will be lower by about 50 tonnes and the material cost lower by a factor 5 as compared to PFBR shields comprising of stainless steel and B₄C.     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

The paper describes radiation effects on 8₄C pellets used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Pellet swelling (Δ$\text{V/V}$) caused by irradiation was less than 1% in which crystal lattice swelling was less than 20%. Many microcracks, a main cause of pellet swelling, appeared in the irradiated pellets. The production of microcracks was related to graphite precipitation in the pellets before irradiation. Open pores which did not exist in the unirradiated pellets were formed in the irradiated ones. In a unit cell of B₄C, the α-axis elongated by 0.025 Å and the $\text{c-}\text{axis}$ shrank by 0.07 Å by irradiation. Moreover, we found three recovery stages which were from room temperature to 400°C, from 400 to 750°C and from 850 to 1100°C. The recovery mechanisms in the irradiated pellets are discussed in terms of the helium behavior.     

Development of thick wall 2.25Cr-1MoNiNb steel forgings for steam generators of fast breeder reactors

The large and thick forgings made of 2.25Cr-1MoNiNb steel are required for vessel material of steam generator (SG) of the fast breeder reactor SNR 300.In order to study the feasibility of 2.25Cr-1MoNiNb steel heavy section forging, chemical composition, melting practice and ingot making, hot working and heat treatment conditions were investigated. The following recommendations were derived: (1) 0.04% C, 0.10% ΔNb, (2) application of electro-slag remelting process, (3) grain refining by hot working, (4) two-step austenitizing at 1020°C.Based on these recommendations, the actual products such as hollow cylinders with maximum 290 mm thickness, solid bars with 420 mm diameter and forged plates with maximum 185 mm thickness could be supplied for application in the helical coiled SG of the SNR 300. Statistical analysis of the products showed the sufficient and isotropic material properties, which fulfill the requirements of the basic safety rules.