[en] The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B₄C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B₄C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B₄C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.)

[en] This report summarizes the results of the in-pile tests with electrically heated fuel rod simulators. These simulator experiments were performed as reference tests to the nuclear tests under identical boundary conditions and in the same test loop (DK loop in the FR2 reactor) that was used in the nuclear tests. The objective of the reference tests was to answer the question if the mechanisms of fuel rod failure were different for electrically heated simulators from those for nuclear fuel rods. Both type of tests were to simulate the 2nd heatup phase of a LOCA. A total of eight transient tests within the BSS series were performed. As parameter the rod internal pressure was varied between 20 and 120 bar. The burst temperatures and burst pressures lie in the range of the burst data obtained with the nuclear test rods. The same holds for the burst strains and deformation profiles of the simulators. Thus, the test program did not reveal an influence of the nuclear parameters on fuel rod failure. (orig.)

[en] In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO₂ specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO₂ interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO₂ pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM)

[en] This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.)

[en] The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program was to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations were tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The BWR-type bundles consisted of 18 fuel rod simulators (heated and unheated rods), an absorber blade of steel containing eleven absorber rods filled with boron carbide powder. The larger bundle CORA-18 contained the same number of absorber rods but was made up of 48 fuel rod simulators. All BWR bundles were surrounded by a zircaloy shroud and the absorber blades by a channel box wall on each side, also made of zircaloy. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence was simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. In all experiments bundle destruction started in the upper region (axially) with melting of the absorber blade and the absorber rod cladding at about 1250 C by interaction of boron carbide and steel. After destruction of the channel box walls this melt attacked the zircaloy fuel rod cladding and started to interact with the UO₂ pellets. The test bundles also resulted in severe oxidation of the following components made of zircaloy: shroud, cladding, and grid spacers at the central and upper positions. Relocated absorber melt formed extended blockages at lower elevations of the bundles. This distribution of absorber material could in the type of reactor accident described with subsequent flooding of the partially destroyed reactor core with unborated water lead to local recriticalities. There was no difference in the behavior of the large bundle CORA-18 compared to the BWR test bundles of regular size, i.e. CORA-16 and CORA-17. Quenching (flooding) of a degraded BWR-type bundle (CORA-17) exhibits identical behavior as observed in the PWR-type quenching experiments: The bundle results in locally enhanced zircaloy/steam reaction causing a renewed temperature rise, an additional meltdown of materials, and an additional strong hydrogen generation. (orig.)

[en] The CORA experiments carried out in an out-of-pile facility at the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, are part of the ''Severe Fuel Damage'' (SFD) program. The experimental program is to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in a few cases up to 2400 C. In the CORA experiments two different bundle configurations are tested: PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles. The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators and nine unheated rods (full-pellet and absorber rods). Bundle CORA-5 contained one Ag/In/Cd - steel absorber rod whereas two absorber rods were used in CORA-12, CORA-15, and CORA-9. The larger bundle CORA-7 contained 5 absorber rods. CORA-12 was terminated by quenching with water from the bottom. In CORA-15 the heated and unheated rods were pressurized to achieve pronounced clad ballooning. Bundle CORA-9 was tested with a system pressure of 1.0 MPa instead of 0.22 MPa. The test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of a LWR. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy (Zry)-steam reaction started at about 1100 C, leading the bundles to maximum temperatures of approximately 2000 C. Rod destruction started with the failure of the absorber rod cladding at about 1200 C, i.e. about 250 K below the melting regime of steel. Penetration of the steel cladding was presumably caused by a eutectic interaction between steel and the zircaloy guide tube. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO₂ interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. There was no difference in the behavior of the large bundle CORA-7 compared to the test bundles of regular size. (orig.)

[en] In the tests with Al₂O₃ pellets, simulating burnable poison rods (98.6 wt.% Al₂O₃ + 1.4 wt.% B₄C), early melt formation at about 1350⁰C was observed. The liquefaction increases distinctly at 1500⁰C. In the resolidified melts two metallic phases: α-Zr(O) and (Zr, Al) alloy and one porous ceramic (ZrO₂/Al₂O₃) eutectic can be distinguished. Large blockages form at the lower end of the bundle. In the tests with UO₂ pellets the melting starts at the elevation of the Inconel grid spacer. By eutectic melt formation in contact with the Zircaloy the liquefaction begins already below the melting point of the Zry and Inconel. Further interaction of this melt with the UO₂ results in partial dissolution of the pellets. Solidification of the melt led to blockage formation at the lower end of the bundle, but at higher elevations compared to the tests with alumina pellets. At some locations fragmentation of fuel pellets to fine powder took place during cooldown. (orig./HP)

[en] The QUENCH experiments are to investigate the hydrogen source term that results from the water injection into an uncovered core of a light-water reactor (LWR). The test bundle is made of 21 fuel rod simulators with a length of approximately 2.5 m. 20 fuel rod simulators are heated over a length of 1024 mm, the one unheated fuel rod simulator is located in the center of the test bundle. Heating is carried out electrically using 6-mm-diameter tungsten heating elements installed in the center of the rods and surrounded by annular ZrO₂ pellets. The rod cladding is identical to that used in LWRs: Zircaloy-4, 10.75 mm outside diameter, 0.725 mm wall thickness. The test bundle is instrumented with thermocouples attached to the cladding and the shroud at 17 different elevations with an axial distance between the thermocouples of 100 mm. During the entire test up to the cooldown phase, superheated steam together with the argon as carrier gas enters the test bundle at the bottom end and leaves the test section at the top together with the hydrogen that is produced in the zirconium-steam reaction. The hydrogen is analyzed by three different instruments: two mass spectrometers and a ''Caldos 7 G'' hydrogen measuring device (based on the principle of heat conductivity). Part I of this report describes the results of test QUENCH-04 performed in the QUENCH test facility at the Forschungszentrum Karlsruhe on June 30, 1999. The objective of the experiment QUENCH-04 was to investigate the reaction of the non-preoxidized rod cladding on cooldown by steam rather than quenching by water. Part II of the present report deals with the results of the SVECHA/QUENCH (S/Q) code application to the FZK QUENCH bundle tests. The adaptation of the S/Q code to such kind of calculations is described. The numerical procedure of the recalculation of the temperature test data, and the preparation for the S/Q code input is presented. In particular, the results of the QUENCH-01 and QUENCH-04 test simulations are discussed. (orig.)

[en] This report describes the results of the Test Series G1 within the in-pile experimental program for the investigation of LWR fuel rod behavior. The results were obtained with single rods of a PWR design in the DK loop of the FR2 reactor at the Kernforschungszentrum Karlsruhe (KfK). The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechansims of fuel rod failure were being performed with irradiated and unirradiated rods. The main parameter of the test program is the burnup ranging from 2500 to 35 000 MWd/t. The results of test series G1 (35 000 MWd/t) with respect to the burst data, i.e. burst temperature, burst pressure, and burst strain, do not indicate major differences from the in-pile tests with unirradiated test specimens. (orig.)

[en] The main objective of the FR2-LOCA tests is to provide information about the effects of a nuclear environment on the mechanisms of fuel rod failure in the second heatup phase of a LOCA (Loss-of-Coolant Accident). The test rods have a heated length of 50 cm, and their radial dimensions are identical with those of a commercial German PWR (Pressurized Water Reactor). The tests are performed with unirradiated as well as with previously irradiated rods. The main parameter of the FR2 In-pile Test Program is the burnup, ranging from 2500 to 35000 MWd/t. The test rods are filled with helium to internal pressures representing the range of PWR rod pressures. The C tests showed cracked fuel after the irradiation (prior to the transients) and a fragmentation of fuel pellets after the transient tests due to cladding ballooning. The post test results indicated a reduction of the pellet stack length of the test rods caused by a redistribution of the fuel in the sections of major deformation. The burst data of the C tests did not reveal any difference between tests with unirradiated fuel rods and the irradiated fuel rods of this test series. This report documents the nominal and actual data of the test rods, the data recorded during the transient and steady-state operation, and the results of the non-destructive post-test examinations. (orig./HP)