Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment. The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30 min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis. This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP) with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

MMonCa: An Object Kinetic Monte Carlo simulator for damage irradiation evolution and defect diffusion

In this work, we introduce the Object Kinetic Monte Carlo (OKMC) simulator MMonCa and simulate the defect evolution in three different materials. We start by explaining the theory of OKMC and showing some details of how such theory is implemented by creating generic structures and algorithms in the objects that we want to simulate. Then we successfully reproduce simulated results for defect evolution in iron, silicon and tungsten using our simulator and compare with available experimental data and similar simulations. The comparisons validate MMonCa showing that it is powerful and flexible enough to be customized and used to study the damage evolution of defects in a wide range of solid materials.

On the production of N-2(+) ions at the N 1s edge of the nitrogen molecule

The N+2 ion yield of the N2 molecule has been measured at the N 1s → Rydberg excitations. It displays Fano-type line shapes due to interference between direct outer-valence photoionization and participator decay of the core-excited Rydberg states. The N+2 ion yield is compared with the total intensity of the outer-valence photoelectron lines obtained recently with electron spectroscopy (Kivimäki et al 2012 Phys. Rev. A 86 012516). The increasing difference between the two curves at the higher core-to-Rydberg excitations is most likely due to soft x-ray emission processes that are followed by autoionization. The results also suggest that resonant Auger decay from the core–valence doubly excited states contributes to the N+2 ion yield at the photon energies that are located on both sides of the N 1s ionization limit.

Benchmarking of calculation schemes in Apollo2 and COBAYA3 for VVER lattices

This paper presents solutions of the NURISP VVER lattice benchmark using APOLLO2, TRIPOLI4 and COBAYA3 pin-by-pin. The main objective is to validate MOC based calculation schemes for pin-by-pin cross-section generation with APOLLO2 against TRIPOLI4 reference results. A specific objective is to test the APOLLO2 generated cross-sections and interface discontinuity factors in COBAYA3 pin-by-pin calculations with unstructured mesh. The VVER-1000 core consists of large hexagonal assemblies with 2mm inter-assembly water gaps which require the use of unstructured meshes in the pin-by-pin core simulators. The considered 2D benchmark problems include 19-pin clusters, fuel assemblies and 7-assembly clusters. APOLLO2 calculation schemes with the step characteristic method (MOC) and the higher-order Linear Surface MOC have been tested. The comparison of APOLLO2 vs.TRIPOLI4 results shows a very close agreement. The 3D lattice solver in COBAYA3 uses transport corrected multi-group diffusion approximation with interface discontinuity factors of GET or Black Box Homogenization type. The COBAYA3 pin-by-pin results in 2, 4 and 8 energy groups are close to the reference solutions when using side-dependent interface discontinuity factors.

Propagation of Cross-Section Uncertainties in Criticality Calculations in the Framework of UAM-Phase I Using MCNPX-2.7e and SCALE-6.1

In the framework of the OECD/NEA project on Benchmark for Uncertainty Analysis in Modeling (UAM) for Design, Operation, and Safety Analysis of LWRs, several approaches and codes are being used to deal with the exercises proposed in Phase I, “Specifications and Support Data for Neutronics Cases.” At UPM, our research group treats these exercises with sensitivity calculations and the “sandwich formula” to propagate cross-section uncertainties. Two different codes are employed to calculate the sensitivity coefficients of to cross sections in criticality calculations: MCNPX-2.7e and SCALE-6.1. The former uses the Differential Operator Technique and the latter uses the Adjoint-Weighted Technique. In this paper, the main results for exercise I-2 “Lattice Physics” are presented for the criticality calculations of PWR. These criticality calculations are done for a TMI fuel assembly at four different states: HZP-Unrodded, HZP-Rodded, HFP-Unrodded, and HFP-Rodded. The results of the two different codes above are presented and compared. The comparison proves a good agreement between SCALE-6.1 and MCNPX-2.7e in uncertainty that comes from the sensitivity coefficients calculated by both codes. Differences are found when the sensitivity profiles are analysed, but they do not lead to differences in the uncertainty.

Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronicsand activation studies of a preliminary reaction chamber based in the following technologies: unpro-tected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steelVacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function ofthe6Li enrichment. After a parametric study, we select for study both a ?thin? and ?thick? blanket, with?high? and ?low?6Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we com-pute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partialpressure, the203Hg and210Po total activity in the LiPb loop, and the Vacuum Vessel thickness requiredto guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in thesafety related issues and structural viability, but it operates at higher6Li enrichment. It is selected forfurther improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varyingbetween 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector,are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

UPM activities: Neutronic/Thermohydrulic (COBRA-TF) coupling and Inventory prediction using SCALE6.1

UPM activities: Neutronic/Thermohydrulic (COBRA-TF) coupling and Inventory prediction using SCALE6.1

Preparing Exercise I-3: Optimization of cross-section tables using sensitivity coefficients in COBAYA3

Preparing Exercise I-3: Optimization of cross-section tables using sensitivity coefficients in COBAYA3

GRS Results for the Burnup Pin-cell Benchmark Propagation of Cross-Section, Fission Yields and Decay Data Uncertainties

GRS Results for the Burnup Pin-cell Benchmark Propagation of Cross-Section, Fission Yields and Decay Data Uncertainties

UPM Activities on Sensitivity and Uncertainty Analysis of Assembly Depletion

UPM Activities on Sensitivity and Uncertainty Analysis of Assembly Depletion

Analysis of Neutron Thermal Scattering Data Uncertainties in PWRs

Analysis of Neutron Thermal Scattering Data Uncertainties in PWRs

Progress on spent fuel data compilations for PWRs

A “Collaborative Agreement” involving the collective participation of our students in their last year of our “Nuclear Engineering Master Degree Programme” for: “the review and capturing of selected spent fuel isotopic assay data sets to be included in the new SFCOMPO database"

Generation of Fission Yield covariance data and application to Fission Pulse Decay Heat calculations

Generation of Fission Yield covariance data and application to Fission Pulse Decay Heat calculations

Application of a Bayesian/Generalised Least-Squares method to generate correlations between independent neutron fission yield data

Fission product yields are fundamental parameters for several nuclear engineering calculations and in particular for burn-up/activation problems. The impact of their uncertainties was widely studied in the past and valuations were released, although still incomplete. Recently, the nuclear community expressed the need for full fission yield covariance matrices to produce inventory calculation results that take into account the complete uncertainty data. In this work, we studied and applied a Bayesian/generalised least-squares method for covariance generation, and compared the generated uncertainties to the original data stored in the JEFF-3.1.2 library. Then, we focused on the effect of fission yield covariance information on fission pulse decay heat results for thermal fission of 235U. Calculations were carried out using different codes (ACAB and ALEPH-2) after introducing the new covariance values. Results were compared with those obtained with the uncertainty data currently provided by the library. The uncertainty quantification was performed with the Monte Carlo sampling technique. Indeed, correlations between fission yields strongly affect the statistics of decay heat. Introduction Nowadays, any engineering calculation performed in the nuclear field should be accompanied by an uncertainty analysis. In such an analysis, different sources of uncertainties are taken into account. Works such as those performed under the UAM project (Ivanov, et al., 2013) treat nuclear data as a source of uncertainty, in particular cross-section data for which uncertainties given in the form of covariance matrices are already provided in the major nuclear data libraries. Meanwhile, fission yield uncertainties were often neglected or treated shallowly, because their effects were considered of second order compared to cross-sections (Garcia-Herranz, et al., 2010). However, the Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Processing and validation of JEFF3-3.1.2 Cross-section Library into Various Formats: ACE, PENDF, GENDF, MATXSR and BOXER.

Following the processing and validation of JEFF-3.1 performed in 2006 and presented in ND2007, and as a consequence of the latest updated of this library (JEFF-3.1.2) in February 2012, a new processing and validation of JEFF-3.1.2 cross section library is presented in this paper. The processed library in ACE format at ten different temperatures was generated with NJOY-99.364 nuclear data processing system. In addition, NJOY-99 inputs are provided to generate PENDF, GENDF, MATXSR and BOXER formats. The library has undergone strict QA procedures, being compared with other available libraries (e.g. ENDF/B-VII.1) and processing codes as PREPRO-2000 codes. A set of 119 criticality benchmark experiments taken from ICSBEP-2010 has been used for validation purposes.