Joerg Konheiser

Application of different nuclides in retrospective dosimetry

The activities of nuclides produced via the neutron irradiation of reactor pressure vessel (RPV) steel are used to validate respective fluence calculations. Niobium, nickel, and technetium isotopes from RPV trepans of the decommissioned NPP Greifswald (VVER-440) have been analyzed. The activities were determined by TRAMO (Monte-Carlo) fluence calculations, newly applying 640 neutron-energy groups and ENDF/B7 data. Relative to earlier results, fluences up to 20% higher have been computed, leading to somewhat better agreement between measurement and calculation, particularly in the case of Tc-99. (authors)

Investigation of the effects of a variation of fuel assembly position on the ex-core neutron flux detection in a PWR

ABSTRACT This work shows the impact of potential displacements of the fuel assembly positions in the reactor core on the signal values of the ex-core instrumentation of a pressurized water reactor in order to understand in detail the impact on the calibration factor of ex-core detectors. This was done with a range of Monte Carlo calculations that simulated the detailed geometrical effect by stepwise changing of the positions of fuel assemblies for selected, conservative scenarios. First, criticality calculations were carried out for the chosen core configurations, and corresponding surface sources on the core barrel were determined. In these calculations, the distances were varied between the fuel assemblies which were in the line of sight of the ex-core instrumentation. A maximal change of the fluxes on the surface of the core barrel of 4%/mm could be calculated under conservative assumptions for the combination of displaced fuel assemblies. In addition, a dependence of this effect as a function of cycle burn-up was analyzed. In a second step, transport calculations for the ionization chambers were performed using the surface sources. An increase of the reaction rate at the chambers of up to 3%/mm has been calculated.

Study of the uncertainties due to position change of the PWR aeroball measurement system

ABSTRACT The aeroball measurement system (AMS) is an important in-core instrumentation in German pressurized water reactors. Therefore, it is essential to know the possible uncertainties of this system. One is the lack of knowledge of the positions of balls in the guide tubes. The position changes can be up to 7 mm. Since the neutron flux distribution is not constant across the guide tubes, different reaction rates can result from the displacements. Both fuel assembly and full core calculations were carried out with the Monte Carlo code MCNP5. Differences in the reaction rates of up to 2% could be determined. In most cases, differences are only up to 0.5%. The results were hardly influenced by burnup and boron concentration in the water moderator. For fuel assemblies containing gadolinium as a burnable poison, a more pronounced reduction could be observed in the direction towards the gadolinium fuel rods. Overall, it was found that the AMS measurement values are very robust with regard to possible variations of ball positions.

NEUTRON AND GAMMA FLUENCE AND RADIATION DAMAGE PARAMETERS OF EX-CORE COMPONENTS OF RUSSIAN AND GERMAN LIGHT WATER REACTORS

Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. This was the clear outcome of the surveillance specimen story at HFIR [1]. Depending on reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, affect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The paper presents some results of work which has been done in this respect. Absolute neutron and gamma flux spectra had been calculated for two core loading variants of the Russian PWR type VVER-1000, for a German 1300 MW PWR and for a German 900 MW BWR. The present investigations extend a similar work done before for VVER440 and VVER-1000 reactors [2]. Based on the calculated spectra, several fluence integrals and radiation damage parameters were derived for the region around the azimuthal flux maximum in the mid-plane at different radial positions between core and biological shield. The relative contributions of gamma radiation to the sum of gamma and neutron contributions are of special interest. As damage parameters the displacements per atom of iron are given separately for neutrons and gammas as well as estimations of the numbers of freely migrating defects (FMD). To get some notion about the uncertainties of the obtained dpa, the calculations were performed using different dpa cross section evaluations. Additionally, gamma produced dpa were calculated by means of the Monte Carlo code EGS4. Another parameter of practical interest for pressure vessel dosimetry and which depends on the gamma flux spectra is the contribution of photo-fissions to the activation of fission detector materials. Therefore, neutron and photo-fission rates were calculated for the fission detector reactions Np(n,f) and U(n,f). Most of the calculations were performed using a 3D synthesis of 2D/1D-flux distributions. To increase the reliability of the evaluations some of the calculations were repeated by another laboratory. To verify the accuracy of the results additional calculations with the continuous energy Monte Carlo code MCNP were performed for two reactors.

Deterministic and Monte Carlo Neutron Transport Calculation for Greifswald-1 and Comparison with Ex-vessel Measured Data

The results of a study of neutron and gamma field functionals derived by deterministic Sn and Monte Carlo calculation methods and by neutron activation measurements in application to the ex-vessel cavity of the VVER-440 reactor Greifswald-1 are presented. A good agreement of deterministic and stochastic calculation results with each other as well as with measurement results was found for neutron threshold detector reaction rates at ex-vessel positions. The influence of different numbers of cross-sectional groups on the calculation results is demonstrated.

Verification of Monte Carlo Calculations by Means of Neutron and Gamma Fluence Spectra Measurements behind and inside of Iron-Water Configurations

Neutron and gamma spectra were measured behind and inside of modules consisting of variable iron and water slabs that were installed in radial beams of the zero‐power training and research reactors AKR of the Technical University Dresden and ZLFR of the University of Applied Sciences Zittau/Goerlitz. The applied NE‐213 scintillation spectrometer did allow the measurement of gamma and neutron fluence spectra in the energy regions 0.3–10 MeV for photons and 1.0–20 MeV for neutrons. The paper describes the experiments and presents important results of the measurements. They are compared with the results of Monte Carlo transport calculations made by means of the codes MCNP and TRAMO on an absolute scale of fluences.

Testing of Neutron Data Libraries in Application to Reactor Pressure Vessel Dosimetry

Different neutron data libraries frequently used in reactor pressure vessel dosimetry and related applications were compared and tested using activation measurement data from the international interlaboratory ex-vessel experiment at the Russian type pressurized water reactor Balakovo-3. Besides different data for the neutron transport calculations also different dosimetry cross section files were investigated. At simple testing models the impact of frequently used group approximations on neutron and gamma fluence parameters was studied by comparison of DORT and MCNP calculation results. The fast neutron results relevant for pressure vessel embrittlement agree reasonably for different libraries and calculation methods but the results for low energy neutrons as well as for deep penetrations differ strongly in many cases.