Transient Elongation of a Fresh Fuel Rod under Reactivity Initiated Accident Conditions

In-pile experiments of fresh fuel rods under reactivity initiated accident (RIA) conditions have been performed in the Nuclear Safety Research Reactor at the Japan Atomic Energy Research Institute in order to understand the basic pellet cladding mechanical interaction (PCMI) behavior. Rapid fuel pellet expansion due to a power excursion would cause radial and longitudinal deformation of the cladding. This PCMI could be one of the possible incipient failure modes of an embrittled cladding of a high burnup fuel under the RIA conditions.

Basic PCMI behavior was studied by measuring cladding deformation of a fresh fuel rod without complicated irradiation effects. The transient elongation measurements of the fuel with two kinds of gap width indicated not only PCMI-induced cladding elongation, but also reduction of the pellet stack displacement by the cladding constraint. In the tests under a high-pressure and high-temperature condition simulating an operation condition of BWRs, additional ridge-type cladding deformation was generated due to the axial collapse of the cladding. A preliminary analysis for interpretation of the tests was made using a computer code for the transient analysis of fuel rods, FRAP-T6.     

Behavior of Irradiated ATR/MOX Fuel under Reactivity Initiated Accident Conditions

Pulse irradiation experiments with irradiated ATR/MOX fuel rods of 20MWd/kgHM were conducted at the NSRR in Japan Atomic Energy Research Institute to study the transient behavior of MOX fuel rod under reactivity initiated accident conditions. Four pulse irradiation experiments were performed with peak fuel enthalpy ranging from 335 J/g to 586 J/g, resulted in no failure of fuel rods. Relatively large radial deformation of the fuel rods due to pellet-cladding mechanical interaction occurred in the experiments with peak fuel enthalpy above 500 J/g. Significant fission gas release up to 20% was measured by rod puncture measurement. The generation of fine radial cracks in pellet periphery, micro-cracks and boundary separation over the entire region of pellet were observed. These microstructure changes might contribute to the swelling of fuel pellets during the pulse irradiation. This could cause the large radial deformation of fuel rod and high fission gas release when the pulse irradiation conducted at relatively high peak fuel enthalpy. In addition, fine grain structures around the plutonium spot and cauliflower structure in cavity of the plutonium spot were observed in the outer region of the fuel pellet.     

Fission Gas Induced Cladding Deformation of LWR Fuel Rods under Reactivity Initiated Accident Conditions

Behavior of irradiated fuel rods under power burst conditions by accidental reactivity insertion in light water reactors (LWRs) has been studied in the Nuclear Safety Research Reactor (NSRR). In the experiments, cladding hoop deformation, which reached up to about 10%, was much larger than that of the fresh rods. The current LWR fuel behavior analysis codes, which only take account of the thermal expansion of the fuel pellets for the deformation calculation, under-predicted the plastic deformation of the cladding to be less than about 1%. Fission gas release during the pulse irradiation tests reached as high as 22% in the NSRR irradiated fuel tests. In order to describe these test results, a model of grain boundary fission gases to cause the cladding deformation has been developed and installed in a fuel behavior simulation code, FRAP-T6. In the model, the over-pressurized gases by the pulse irradiation cause grain boundary separation and stress the cladding during the tests. The model assumes that the gases remain in the fuel during the early part of pulse irradiation and are released to the open volume in the rod after the cladding deformation. The model, in combination with a fuel thermal expansion model, GAPCON, which was validated through fresh fuel tests, reproduces the NSRR test results reasonably well.     

Experimental Study on the Fuel Behavior during Reactivity Accident at Power Operation Condition

In the Nuclear Safety Research Reactor (NSRR) program of Japan Atomic Energy Research Institute (JAERI), the fuel behavior in reactivity initiated accidents (RIAs) has been studied through irradiation tests with simulated power burst using fresh or preirradiated test fuel rods. In order to investigate possible influence of the difference of initial temperature profile in the fuel pellet on the fuel failure behavior, two tests were conducted with fresh fuel rods for RIAs at power operation using the newly developed NSRR operation mode and the results were compared with the results of previous irradiation tests which were for RIAs at zero power.

In the tests for RIAs at power, the reactivity of 2.0$ or 2.3$ was inserted rapidly after the linear heat rate of the test fuel rod was kept constant at 39kW/m for 5s. It has been shown through this study with fresh fuel rods that the fuel enthalpy of the failure threshold for RIAs at power is the same as that for RIAs at zero power and that the failure mechanism is the same as that of RIA at zero power. It has been clarified that there is no obvious influence of initial temperature distribution on the fuel behavior during RIAs in case of fresh fuels. The evaluation method of fuel enthalpy with which the fuel failure threshold is described was also studied.     

Effect of Fuel Pellet Eccentricity with Cladding on Fuel Rod Thermal Behavior under Reactivity Initiated Accident Condition

Fuel rod behavior under Reactivity Initiated Accident (RIA) conditions has been studied in the Nuclear Safety Research Reactor (NSRR), JAERI. In the experiments, cladding thermal behavior was observed to be influenced by the fuel pellet eccentricity to produce large azimuthal temperature variation in the cladding. The maximum azimuthal cladding temperature difference was measured to be as large as 150°C by thermocouples attached to opposite sides of the cladding around the circumference, though the thermocouples did not always detect the maximum temperature difference around the circumference. The actual temperature differences in the fuel rods subjected to less than 290 cal/g?UO₂ were estimated to be 350°C at maximum based on metallographies. A simple calculation considering gap conductance variations also showed that the maximum temperature difference became 350°C under fully eccentrical condition in the fuel rod subjected to 260 cal/g?UO₂. Moreover, as the rod damage such as cladding deformation, melting and failure occurs unevenly around the circumference due to the fuel pellet eccentricity in general, the fuel pellet eccentricity should influence the fuel rod failure under RIA conditions.     

Transient Behavior of Low Enriched Uranium Silicide Plate-Type Fuel for Research Reactors during Reactivity Initiated Accident Conditions

Abstract

This paper describes the results of transient experiments using a low enriched uranium silicide miniplate fuel for research reactor. The pulse irradiation was performed in the Nuclear Safety Research Reactor (NSRR) at the Japan Atomic Energy Research Institute (JAERI).

The results obtained in this study are summarized as follows :

(1) The tested fuel plates were damaged with energy depositions above 94 cal/g·fuel, but remained intact below 82cal/g·fuel. A failure threshold should therefore exist between these two values.

(2) Four of the fuel plates that showed peak cladding surface temperatures <330°C were damaged by the thermal stress during quenching. These damaged fuel plates revealed small intergranular cracks that propagated perpendicularly to the axial direction of the plate, from the Al cladding surface to the fuel core, without significant dimensional changes. On the other hand, when peak cladding surface temperatures were >400°C, the test fuel plates were damaged mainly by melting of the Al cladding, accompanying significant dimensional changes.

(3) The thermal stress of the damaged fuel plates calculated on the basis of the maximum transient temperature drop during quenching was greater than the tensile stress that occurred during fabrication.     

Transient Cooling Process of Fuel Rod in Reactivity Initiated Accident

Out-of-pile experiments were performed with Zircaloy-4 rods in subcooled water environment to study the basic phenomena occurring in the transient cooling process undergone by a fuel rod during a reactivity-initiated accident (RIA) affecting a light water reactor (LWR). The experimental results show that the cooling process of the fuel rod during an RIA can be divided into three phases separated by the quenching temperature T_q and the rewetting temperature T_q .

It is also noted from the experimental results that with increasing degree of subcooling, T_q tends to rise to levels far exceeding the maximum liquid superheat temperature of water; T_q , on the other hand, is little affected by the cooling water temperature, and remains close to that of the maximum superheat temperature.

Numerical calculations indicate conclusively that radial heat transfer to coolant water is the dominant factor that governs the transient cooling process in an RIA affecting the cold start-up of a BWR, rather than the axial heat conduction through rod which is considered to be the basic mechanism of cooling that governs the reflooding process during a LOCA.     

Effects of Coolant Flow on Light Water Reactor Fuel Behaviors during Reactivity Initiated Accident

This paper describes the in-pile experimental results to study the influences of coolant flow on fuel behaviors under reactivity initiated accident (RIA) conditions performed in the Nuclear Safety Research Reactor (NSRR). A single PWR type test fuel rod was irradiated by a large neutron pulse in the NSRR to simulate a prompt power excursion of RIA's. The effects of coolant flow were studied at a coolant flow velocity of 0.3~1.8m/s and a coolant temperature of 20~90°C under the atmospheric pressure. It was found that the cooling conditions had considerable influences on fuel thermal behaviors under prompt heat-up. The increase of coolant flow velocity and subcooling enhanced heat transfer coefficient at cladding surface during film boiling, which resulted in large decrease of maximum cladding temperature and film boiling duration, and consequently in the increase of fuel failure threshold energy. The data tendencies are summarized and the influences of coolant flow are discussed with some computer analyses.     

Effects of Rod Pre-Pressurization on Light Water Reactor Fuel Behavior during Reactivity Initiated Accident Conditions

Previously pressurized (pre-pressurized) fuel rod tests recently performed in the Nuclear Safety Research Reactor (NSRR) investigate the effects of initial internal pressure on fuel rod behavior during reactivity initiated accident (RIA) conditions. A single PWR type fuel rod was contained within a waterfilled, ambient temperature and ambient pressure capsule. The fuel rod was then heated by the pulsing operation of the NSRR.

Results from the tests show that the effect of pre-pressurization was significant for the fuel rods with initial internal pressure of 0.8 MPa and above, and fuel rod failure occurred from rupture of the cladding with lower threshold energy deposition for failure as the initial internal pressure was increased. The cladding rupture was governed mainly by the cladding temperature rise, not by the rod internal pressure rise during the transient. The relationships between cladding burst pressure and cladding burst temperature and between cladding strain and cladding temperature at cladding rupture obtained in the present study under an RIA condition agree with the results obtained from various in- and ex-reactor experiments under a LOCA condition, although the obtained time-averaged strain rate of the Zircaloy cladding was much greater than that in a LOCA condition.     

Reactivity of Monochlorobenzene and Its Derivatives toward Solvated Electrons in Lower Aliphatic Alcohols

Monochlorobenzene and its para- or meta-substituted derivatives were irradiated with -γ-rays in methanol, ethanol and n-propanol containing nitrous oxide to elucidate the reactivities of these organic solutes toward solvated electrons, and also the effect of the substituent for the aromatic nucleus on the reactivities of these solutes. The value of G (N₂) is seen to decrease with increasing concentration of the organic solute. Relative rate constants of these organic solutes toward solvated electrons have been determined from the competition with nitrous oxide by observing the G(N₂) in there systems. Based on the results, the rate constants of the reactions of these derivatives of monochlorobenzene with solvated electrons relative to monochlorobenzene can be related to Hammett's σ function. The ρ value has been observed to be 3.3, 2.1 and 1.8 for methanol, ethanol and n-propanol respectively. The present data, together with that by Sherman⁽¹⁾ for 2-propanol, indicates that the ρ value is correlated linearly to the dielectric constant of the solvent. Some discussions are presented on the results thus obtained.     

LWR Fuel Safety Research with Particular Emphasis on RIA/LOCA and Other Conditions

Fuel safety research at Japan Atomic Energy Research Institute (JAERI) is reviewed on the major subjects including studies on fuel behavior under postulated Reactivity Initiated Accident (RIA), postulated Loss of Coolant Accident (LOCA) and normal operating conditions. Nuclear Safety Research Reactor (NSRR) at JAERI has been utilized extensively for the studies of fuel behavior under RIA conditions. For the studies of fuel rod and cladding behavior under LOCA conditions, outpile experiments were conducted. The work on this subject has been concluded. Pellet Cladding Interaction (PCI) has been major subject on fuel integrity study during normal operating conditions. Irradiation experiments at Halden Boiling Water Reactor (HBWR) as well as code development are described.     

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO₂ fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO₂ to UO_2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO_2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.     

Diffusion of Cesium in Stainless Steel under Condition of Oxygen Potential Controlled by Mo/MoO2 or NbO2/Nb2O5 Oxygen Buffer

The kinetic behaviors of cesium migration in SUS-316 stainless steel for a cladding material of fast breeder reactor were studied using radioactive ¹³⁴Cs as a tracer. The oxygen potential in the atmosphere surrounding the specimen was controlled by Mo/MoO₂or NbO₂/Nb₂O₅ oxygen buffer, which corresponds approximately to the one in fuel-cladding gap during irradiation. The concentration profile of Cs in the specimen was analyzed on the basis of diffusion theory. The temperature dependences of diffusion coefficients of Cs in SUS-316 stainless steel were expressed in the range of 650~800°C by the following equations:

In the oxygen potential controlled by Mo/MoO₂ oxygen buffer

D = 0.15 exp (?63,500/RT) (cm²/s).

In the oxygen potential controlled by NbO₂/Nb₂O₅ oxygen buffer

D = 9.0×10-⁵exp(?50,300/RT) (cm²/s)

where the activation energy is expressed in terms of cal/mol.