[en] In the framework of the code assessment and maintenance program (CAMP) of the US NRC the reflood model of the RELAP5 code is being validated at Forschungszentrum Karlsruhe (FZK) using data from the test PKL-IIB.5. The primaerkreislauf-(PKL) test IIB.5 simulates a double-ended break of the cold leg of a German 1300 MWe PWR with emphasis on the reflood heat transfer phenomena. The PKL-facility includes all major primary coolant circuit components and some secondary components, and a containment. After transient initiation the coolant is dumped to the containment while the ECC-systems (accumulators and low pressure injection systems) inject a large amount of cold water leading to a progressive core rewetting. Under such conditions complex heat transfer mechanisms take place in the core region. A novel reflood model, developed at PSI, was implemented in the code version RELAP5/MOD3.2.2Gamma (322G). The heat transfer coefficient in the film and transition boiling flow regime is dependent on the distance from the quench front. The transition boiling heat transfer is predicted by the empirical Weisman correlation. The post-test calculation of the PKL-IIB.5 test showed that the PSI-reflood model is able to describe the reflooding process in an reasonable manner. But the empirical Weisman correlation tends to over-predict the transition boiling heat transfer. Hence the semi-mechanistic FZK-transition boiling model was implemented in RELAP5 instead of the Weisman correlation. The resulting code version was named RELAP5/MOD3.2.2G+FZK (322G+FZK). Based on the re-calculation of the PKL-test with the modified RELAP5-version it can be concluded that the rewetting temperature predicted by the FZK-transition boiling model is much closer to the experimental data than that obtained using the Weisman approach. Results of these investigations are presented and discussed in this report. (orig.)

[en] In the framework of the code assessment and maintenance program (CAMP) of the US nuclear regulatory commission (NRC) the RELAP5 code system is being validated at Forschungszentrum Karlsruhe (FZK). The validation work is focused on the assessment of the RELAP5-reflood model. The data obtained from the integral test LOFT-LP-LB-1 is used to validate the RELAP5-reflood model implemented in the version RELAP5/MOD3.2.2Gamma (322Gamma). The test LP-LB-1 simulates a double-ended, off-set shear of the cold leg of a pressurized water reactor (PWR) coincident with the loss of off-site power. A post-test calculation of the LOFT-test was performed using a model developed at Paul Scherrer institut (PSI) for RELAP5. Results of these investigations are presented and discussed in this report. Based on the predictions with the original version 322Gamma it can be stated that the overall system behaviour and the core thermal response are reasonable predicted by RELAP5. The reflooding process is qualitatively well predicted by this code version. The cladding temperature in several bundle elevation are closer to measured data compared to the ones of earlier versions. But PSI-model tends to under-predict the rewetting temperature due to the use of the empirical Weisman correlation to determine the transition boiling heat transfer. Therefore the FZK-transition boiling correlation was implemented in the original version instead of the Weisman model (code version 322Gamma+FZK). Recalculations of the LP-LB-1 test with the version 322Gamma+FZK showed that the rewetting temperature in all axial elevations better fits with experimental data now. (orig.)

[en] The invention has been aimed at an irradiation can for activations of irradiated materials in research, irradiation, and nuclear power plant reactors in order to enlarge the feasibilities of irradiation in the reactor core, to meet the requirements of a safe reactor operation and to ensure a higher irradiation capacity. The outer shape of the irradiation can is adequate to that of a fuel assembly (hexagonal, square or cylindrical). For that reason a joint insertion of irradiation cans and fuel assemblies into the reactor core is possible. The can is utilizable for long-time irradiations, too. The irradiated material has been positioned inside the irradiation can in sealed tubes

[en] The high performance light water reactor (HPLWR) project is being carried out within the 5. European Framework Programme (Contract N^o FIKI-CT-2000-00033). The main objective of this project is to evaluate the technological merits and economics of a high efficiency light water reactor (LWR) operating at thermodynamically supercritical conditions. One of the concepts developed by the Tokyo University, the high temperature supercritical light water reactor (SCLWR-H), was chosen as the reference plant for the HPLWR-project. A major activity of the HPLWR-project is the assessment of the appropriateness of RELAP5 - developed for light water reactors - to perform steady-state and safety analysis of the reference plant. The investigation of such reactor concepts whose operating conditions are far beyond the operation range of current light water reactors, is very challenging for codes like RELAP5. Since RELAP5 was not developed for supercritical water conditions and therefore not validated in this domain, the investigations to check the appropriateness of RELAP5 are focused on the following areas: Thermo-physical properties of water in the supercritical region. Heat transfer mechanisms for wall/supercritical water and corresponding correlations. Development of a simplified plant model for steady state and transient analysis of the reference plant. Exploratory analysis of selected postulated transients and accidents. Despite the preliminary character of this work, the investigations performed for the reference design have demonstrated that RELAP5 is capable to qualitatively predict the plant behavior under both normal operation and different accidental conditions. But there are still problems in the prediction of the thermo-physical properties around the critical point in case of depressurization transients. In this report, the results obtained in each area will be presented and discussed in detail. The additional work needed for both further code improvement and assessment of the final HPLWR-plant design is listed. (orig.)

[en] The First Ph.D. Student Workshop ''Nuclear Safety Research'' of the Helmholtz Association of National Research Centers (HGF)'' was jointly organized by the Research Center Karlsruhe GmbH and the Energie Baden-Wuerttemberg AG (EnBW) from Wednesday 9th to Friday 11th March 2005. The workshop was opened with welcome greetings by Dr. Peter Fritz, Forschungszentrum Karlsruhe. Subsequently Dr. Joachim U. Knebel explained the main goals and the content of the workshop. The young scientists reported in 28 high-level presentations about their research work which covered a wide spectrum from reactor safety, partitions and transmutation, and innovative reactor systems, to safety research for nuclear waste disposal. The junior researchs showed excellent professional competence and demonstrated presentation qualities at the highest level. The successful funding of two Virtual Institutes, namely: the ''Competence in Nuclear Technologies'' and ''Functional Characteristics of Aquatic Interfaces both co-ordinated by Forschungszentrum Karlsruhe'', by the President of the Helmholtz Association Prof. Walter Kroell was the motivation for the organization of this first Ph.D. Student Workshop. Thanks to these two Virtual Institutes, the Reseach Center Karlsruhe and Juelich together with several univer-sities i.e. RWTH Aachen, Heidelberg, Karlsruhe, Muenster, and Stuttgart, have successfully financed eight Ph.D. and two post-doctoral students. Moreover, young scientists of the European Institute for Transuranium Elements (ITU) and additional seven Ph.D. Students, who are sponsored by the German nuclear industry (Framatome ANP, RWE Power, EnBW) in the frame of the Alliance Competence in on Nuclear Technology, and who are trained at Forschungszentrum Karlsruhe, actively contributed to this workshop. The EnBW-Award was handed over by Dr. Hans-Josef Zimmer, member of the board of directors of the EnBW-Kraftwerksgesellschaft, to Mrs. Ayelet Walter from the University of Stuttgart for the best lecture entitled ''System modelling of a HTR-reactor with He-turbine under operational and accidental conditions''. This Ph.D. Student Workshop demonstrates once again that a solid nuclear and scientific-oriented engineering education offers excellent career opportunities in the energy sector in both Germany and Europe. In this report a compilation of the compacts and the respective slides of all presentations are documented. (Orig.)

[en] Summary –This paper describes the investigations performed by KIT in the frame of the German WASA-BOSS projected funded by the Federal Ministry of Economics and Technology to improve severe accident codes and to analyse options to prevent or delay the failure of the main safety barriers e.g. reactor pressure vessel of PWR and BWR. The KIT contribution was focused on the study of the potential accident management to delay or prevent the failure of the reactor pressure vessel (RPV) of a BWR using ATHLET-CD. For a generic BWR, different severe accident sequences without and with different SAM-measures were extensively investigated. The obtained results have shown that SAM-measures can be effective to mitigate the accident progression if they start before material relocation in the lower plenum. This paper will discuss the SAM-measures to avoid or delay the RPV-failure by performing reflooding of the core in certain time windows and under which conditions it will fail. The modelling issues, assumptions and results will be presented and discussed.

[en] For more than ten years, the Frederic Joliot/Otto Hahn Summer School has been organized alternately by the Karlsruhe Research Center in Germany and the French Commissariat a l'Energie Atomique (CEA), Cadarache, in France. This year, the Summer School was held at the Center for Advanced Training in Technology and the Environment of the Karlsruhe Research Center on August 29 to September 7. The overarching topic of the event was the sustainability of nuclear power, including topical issues of generation-IV reactor concepts, transmutation and actinide separation, and geologic final storage. Next year's Frederic Joliot/Otto Hahn Summer School will be organized by CEA at Aix-en-Provence together with the Nuclear Safety Research Program of the Karlsruhe Research Center. (orig.)

[en] The main purpose of the computational OECD/NEA pressurized water reactor main steam line break (PWR MSLB) Benchmark is the evaluation of the prediction capability of advanced code systems by means of a code-to-code comparison. The postulated MSLB-transient is characterized by a strong non-symmetrical core thermal behaviour due to the feedback between neutron kinetics and plant thermal hydraulics. The analysis of such transients with pronounced spatial power distortion represents a considerable challenge for advanced code systems. The transient is initiated by a double-ended break of one main steam line when the reactor TMI-1 is operated at nominal power. The high heat removal through the break leads to a strong cooldown of the primary coolant. Under such conditions a power increase and a re-criticality of the core despite scram can not be excluded due to the negative reactivity coefficients. The MSLB-Benchmark enfolds three exercises as follows: Exercise 1: integral plant simulation with best-estimate codes using point kinetics, Exercise 2: multidimensional simulation of the core for given initial and boundary conditions, and Exercise 3: integral plant simulation with coupled, best-estimate codes using 3D-neutron kinetics models. Das Forschungszentrum Karlsruhe (FZK) and Framatome Advanced Nuclear Power/Erlangen (former Siemens/KWU) participated on the MSLB-Benchmark with the code system RELAP5/MOD3.2 for the Exercise 1: In this report, the integral plant model developed for this Exercise 1 together with the calculated results will be presented and discussed. (orig.)

[en] This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-Ⅶ.1 and ENDF/B-Ⅷ.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-Ⅷ.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors

[en] The heat transfer model of the RELAP5/MOD3.1 (R5M3) code was extensively reviewed and assessed. The most important deficiency of the current version of the code was attributed to its treatment of the transition boiling heat transfer regime. The current transition boiling model significantly underpredicts the liquid heat transfer rate. Since at low quality conditions the liquid boiling component is a major fraction of the total heat transfer, the current model underpredicts the quench temperature and the quenching rate under most conditions relevant to LOCA and degraded core analysis. Therefore, a new model has been developed and implemented in the R5M3 code for predicting the transition boiling heat transfer. The new model is based on an extension of the phenomenological formulation suggested originally by Chen. It utilizes only local state variables calculated by the R5M3 code and does not require other history parameters, such as quench position or CHF and minimum film boiling temperatures, which are not available at each time step. A number of separate effect and bundle test experiments are analyzed with the modified code version. The predictions are compared with those obtained by the frozen code version and with available experimental data. Several variables, such as well temperatures, vapor and liquid velocities, void fraction etc., are examined in order to evaluate the general prediction capability of the code in modeling boil-off and reflood transients. In addition, the current and the modified stand-alone transition boiling models are tested against a large sample of the available data-base on steady-state post dryout heat transfer. In all cases, the predictions of the modified model fit the measured data better. The temperature curves are physically and conceptually more sound than those predicted by the frozen code version. This is achieved by introducing a more realistic modeling of the transition boiling heat transfer which affects only one subroutine of the R5M3 code. (orig.)