Engineering topics

Ferroukhi, H.

Publications and source records attributed to Ferroukhi, H..

15 records

DOE OSTI2016

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving either only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Full core LOCA analysis for BWR/6 - Methodology and first results

As part of collaboration between the Paul Scherrer Institute (PSI) and the Swiss regulator ENSI, a study aimed at core-wide estimates of the fuel rod behaviour during large-break loss-of-coolant-accidents (LB-LOCA) in BWR/6 plants was recently conducted on the basis of a Swiss reactor. To perform such full core LOCA analysis, a methodology consisting of a multi-physics sequence of interfaced best-estimate models and codes was established: CASMO-4E/SIMULATE-3 (C4/S3) for the core modelling, TRACE for the plant system analysis and FALCON for the fuel thermomechanical calculations during base irradiation (BI) as well as for the transient analyses (TA). As starting point, reference C4/S3 models validated for all operated cycles of the given plant were used to select as situation target, a recent cycle representative of modern BWR core loadings and fuel assembly designs. Based on the selected C4/S3 cycle model, PSI in-house computational modules were developed or updated in order to set up the TRACE full core model with regards to assembly geometries, fuel rods thermophysical properties and initial 3-D power distributions. Also, an interface to FALCON BI models was established for the transfer of the irradiation history, including primarily linear heat generation rates (LHGR), for an equivalent fuel pin. For each assembly, the power of the equivalent pin is defined with the maximum power of all pins at each axial position. However, in order to reduce the computational time of the fuel performance analyses, the whole core was subdivided into groups of assemblies, each represented by the most powerful assembly within the group. On this basis, the FALCON BI analyses were conducted with coupling to the PSI advanced model for gas release and swelling (GRWS-A) and an interface was developed to transfer the calculated fuel rod characteristics at end of irradiation to initialize the core fuel rods heat structures in the TRACE model. Subsequently, double-ended recirculation line rupture scenarios at hot-full-power (HFP) conditions were investigated with TRACE in order to determine a conservative transient case as function of cycle burnup, single failure criterion and availability of safety systems. For the most conservative case of this study, an interface was set-up to transfer the TRACE results as boundary conditions to the FALCON TA models, including thus primarily time and axial distributions of the clad outer temperatures. As final step, the FALCON TA were conducted for the representative assemblies in order to infer core-wide estimates of the fuel rod behaviour with specific focus on equivalent cladding reacted levels, hoop strains as well as margins to failure criteria. The objective of this paper is to present this methodology along with the obtained first results. Regarding the latter, it is found that principally, even for the most conservative LOCA scenario of this study, the clad temperature excursion remains mild enough at all core locations such that no significant rod ballooning and no cladding failure is predicted, something in line with corresponding analyses conducted earlier for BWR/4 plants. (authors)

42 ENGINEERING↗

DOE OSTI2015

Complik: A versatile tool for automatizing the representation of material data in MCNP models - 14412

At the Paul Scherrer Institut, the CASMO/SIMULATE code system is used as reference scheme for core analyses of the Swiss reactors. Besides, Monte Carlo (MC) based neutron transport codes are in parallel being established both for out-of-core applications and full core analyses, in order to provide higher fidelity solutions to problems beyond the modelling capabilities of the standard core modelling methods. In this perspective, a necessity is to be able to conduct the MC calculations at any operating conditions, requiring thus proper initialisation of the thermal-hydraulic feedback as well as at any burnup, prompting the need for a detailed modelling of all material zones in the irradiated fuel assemblies. To tackle the later challenge, the development of the COMPLINK tool, aimed at initialising large-scale MC models of burned core configurations, using as basis upstream 2-D assembly/ 3-D core analyses results, was initiated and is presented in this paper. A main feature is the capability to extend in an automatic manner, a user given MC model in order to represent material data with a higher resolution, keeping track of the location of such materials also within nested repeated structures. This capability prevents the user from the burden of developing specific computational framework for a target application. Since the same functionalities are required to analyse out-of-core spent fuel managements systems, a first verification of COMPLINK is illustrated on the basis of a criticality safety evaluation (CSE) with MCNP6.1 of a Swiss PWR wet storage pool loaded with fresh and burned fuel assemblies. Starting from CASMO-5 depletion calculations at each axial layer of the depleted fuel assemblies and reconstructed based on realistic irradiation histories, COMPLINK is then applied to load the isotopic compositions into the material zones of the corresponding MCNP6.1 assembly models. On this basis, the CSE analyses are conducted with focus on investigating axial and radial modelling effects. All in all, the obtained results are found to be well in line with the impact on CSE expected at the considered burnup levels, providing thus confidence in the correct implementation of the COMPLINK functionalities and confirming its effectiveness as well as efficiency for the initialization of all material zones in large-scale MC models of complex and strongly heterogeneous burned configurations. (authors)

42 ENGINEERING↗

DOE OSTI2015

Towards application of a neutron cross-section uncertainty propagation capability in the criticality safety methodology - 14434

At the Paul Scherrer Institute (PSI) the methodology for criticality safety evaluations (CSE) used for Swiss applications is focused, following common practice, on establishing an upper subcriticality limit (USL) for the effective neutron multiplication factor (k{sub eff}), based on a statistical evaluation of the obtained validation results in terms of the calculated-to-benchmark k{sub eff} values. Recently, a new computational tool referred to as NUSS (Nuclear data Uncertainty Stochastic Sampling) was developed in order to propagate neutron-nuclear data (ND) uncertainties in calculations with Monte Carlo transport codes in combination with general-purpose ND libraries in ACE format. Thereby, this tool should now allow for a quantification of ND data uncertainties in CSE and to introduce these into the CSE criteria definition. This paper concerns an application of the NUSS methodology for criticality validation studies. Calculated k{sub eff} uncertainty estimations were obtained for the current PSI validation suite including in total 149 critical benchmark configurations for which in-house MCNP(X) models were developed, covering both uranium and mixed uranium-plutonium oxide fuels. The MCNPX-2.7.0 code was used together with the ENDF/B-VII.1 ND library while the ND variance-covariance matrixes (VCM) were taken from both SCALE-6.0 44-group libraries and also from ENDF/B-VII.1 files. Based on the newly obtained information on the uncertainties, a re-assessment of the lower tolerance bound (LTB) for k{sub eff} using the total uncertainties as weighting factors is proposed as means for the present PSI CSE methodology upgrade and some possible directions of further methodology enhancements are discussed next. (authors)

42 ENGINEERING↗

DOE OSTI2015

Computational scheme for burnup credit applied to long term waste disposal - 14448

The application of burnup credit (BUC) to criticality safety evaluations (CSE) of geological repositories for long term disposal of spent nuclear fuel (SNF) is under consideration in Switzerland. Therefore, in collaboration with the national waste management organization, PSI is conducting research towards the development of a BUC methodology tailored for Swiss specific applications. As part of this research, a computational sequence from the in-core burnup phase, through the interim storage, until final disposal is being established. The tools selected for this methodology, referred to as BUCSS-R (Burnup Credit System for Swiss reactors and Repositories), are CASMO for the in-core fuel depletion, the decay module of SERPENT for the decay calculations, and MCNP for the criticality calculations. In connection to BUCSS-R, a computational interface to validated reference core models of the Swiss reactors is being developed. The purpose is to apply reconstructed depletion histories based on real core operation conditions with CASMO assembly calculations, to transfer the pin wise nuclide compositions at discharge to the subsequent SERPENT decay calculations, and afterwards to the final MCNP criticality calculations. An objective of this paper is to present the elements of the BUCSS-R methodology, and to show a first application to so-called bounding analyses of a preliminary Swiss disposal canister design loaded with fuel assembly types operated in a Swiss PWR reactor. The rationale for these bounding analyses is that before starting the integration of methodologies for bias and uncertainty quantifications, the span of the k-eff variation between fresh and burnt fuel loadings should first be quantified in a realistic manner. The obtained results are reported in this paper together with studies performed in order to evaluate the effects from burnup levels and nuclides credit at flooded conditions on the k-eff values. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Comparison of SERPENT and CASMO-5M for pressurized water reactors models

The objective of this work is to perform a preliminary assessment of the capability of SERPENT to generate cross sections for a PWR Beginning-of-Life (BOL) isothermal mini-core by comparing a SERPENT/PARCS solution with the results obtained using a CASMO-5M/PARCS approach. The PARCS code is used instead of the usual SIMULATE-3 to analyze the Swiss Reactors, because interfaces with PARCS already exist to obtain neutronic data from SERPENT. For the PWR configurations, the differences between CASMO-5M and SERPENT solutions are within 200 pcm at the assembly level and thus rather small when considering the deterministic transport method (energy/angular/space discretization) in CASMO-5M versus the stochastic treatment of SERPENT, the statistical uncertainties in the Monte-Carlo approach as well as the eventual differences in nuclear data used by both codes. At the 2D mini-core level, no major difference is observed when comparing PARCS run with CASMO-5M versus SERPENT cross sections. For the generation of kinetic parameters, non trivial differences are observed due both to the methods and the data used. For the relatively limited number of configurations considered, it is hard to make any definitive statement on the benefits of using Monte Carlo codes in terms of nuclear data generation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Application of power time-projection on the operator-splitting coupling scheme of the TRACE/S3K coupled code

This paper reports refinement studies on the temporal coupling scheme and time-stepping management of TRACE/S3K, a dynamically coupled code version of the thermal-hydraulics system code TRACE and the 3D core simulator Simulate-3K. The studies were carried out for two test cases, namely a PWR rod ejection accident and the Peach Bottom 2 Turbine Trip Test 2. The solution of the coupled calculation, especially the power peak, proves to be very sensitive to the time-step size with the currently employed conventional operator-splitting. Furthermore, a very small time-step size is necessary to achieve decent accuracy. This degrades the trade-off between accuracy and performance. A simple and computationally cheap implementation of time-projection of power has been shown to be able to improve the convergence of the coupled calculation. This scheme is able to achieve a prescribed accuracy with a larger time-step size. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

On BWR regional oscillations with rotational symmetry line using SIMULATE-3K

A new stability analysis methodology is being developed at the Paul Scherrer Institute (PSI) using the best-estimate coupled neutronic/thermal- hydraulics code, SIMULATE-3K (S3K). This methodology has so far been validated against Leibstadt NPP (KKL) stability tests of C10, C13 and C19, which all show global (in-phase) oscillations. However, the methodology has not yet been validated for regional instabilities and to that aim, a special KKL cycle 07 stability test was selected. Indeed, during this test, the core not only showed growing power oscillation amplitudes in an out-of-phase regime but also an oscillating and rotating symmetry line. Thereby, it was selected in order to verify the S3K capability to predict regional instabilities and on that basis, obtain more insights towards understanding the causes for the oscillatory and rotational behaviour of symmetry lines. The results obtained so far are presented in this paper. First, it is found that the S3K results are in good agreement with measurements both qualitatively and quantitatively, although the resonance frequency is slightly over-predicted. Secondly, the excitation of the out-of-phase mode with oscillations as well as rotation of the symmetry line is also well captured i.e. in accordance to the experimental observations. Related to this, an in-depth analysis of LPRM signals indicates that two out-of-phase oscillation modes associated to two azimuthal neutronic modes are simultaneously excited. Furthermore, it is found that a superposition of these two modes will trigger the symmetry line dynamics and that the behaviour will be guided by the dominance ratio between these two modes. More precisely, the oscillatory behaviour is due to the superposition of the two azimuthal modes but with one dominant mode. The rotational behaviour is however due to the superposition of the two modes with comparable strengths. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Convergence studies of deterministic methods for LWR explicit reflector methodology

The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on very different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Preparation of nuclear libraries with deterministic and stochastic methods for LWR reflectors

The explicit reflector methodology is used in the system of codes CASMO-5 / SIMULATE-3 to include the reflector around the active core into the computational region and avoid adopting any ad-hoc or experimental albedo coefficients as boundary conditions. However, to complete the core calculation, a set of cross sections and discontinuity factors is needed for the reflector nodes and the accuracy of these nuclear parameters influences the final results, in particular along the peripheral regions of the core. In this paper the explicit reflector methodology of CASMO-5 is adopted to evaluate the few-group cross sections and discontinuity factors of the different reflector cases, based on the design of Generation II and III LWR reactors. In addition, in the perspective of using Monte-Carlo codes as a complementary option for lattice calculations of reflector configurations, the stochastic SERPENT code is also included as part of this benchmark. With the latter, the impact of applying 2-D reflector models with homogenized materials instead of explicit representation of the real geometrical structures is moreover evaluated and shown to be limited. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Stochastic sampling method with MCNPX for nuclear data uncertainty propagation in criticality safety applications

In the domain of criticality safety, the efficient propagation of uncertainty in nuclear data to uncertainty in k{sub eff} is an important area of current research. In this paper, a method based on stochastic sampling is presented for uncertainty propagation in MCNPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed to have a multivariate normal distribution and simple random sampling is performed following this presumed probability distribution. A verification of the developed stochastic sampling procedure with MCNPX is then conducted using the {sup 239}Pu Jezebel experiment as well as the PB-2 BWR and TMI-1 PWR pin cell models from the Uncertainty Analysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developed stochastic sampling approach predicts similar k{sub eff} uncertainties compared to conventional sensitivity and uncertainty methods. For the UAM models, slightly lower uncertainties are obtained when comparing to existing preliminary results. Further details of these verification studies are discussed and directions for future work are outlined. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Towards a reference numerical scheme using MCNPX for PWR control rod tip fluence estimations

Recent occurrences of cracks and fissures on the cladding tubes of PWR control rod (CR) fingers employed in the Swiss reactors prompted the need to develop more reliable analytical methods for CR tip fluence estimations. To partly address this need, a deterministic methodology based on SIMULATE-3/CASMO-4 was in recent years developed at PSI. Although this methodology has already been applied for independent support to licensing issues related to CR lifetime, two main questions are currently being the center of attention for further enhancements. First, the methodology relies on several assumptions that have so far not been verified. Secondly, an assessment of the achieved accuracy has not been addressed. In an attempt to answer both these open questions, it was considered appropriate to develop an alternative computational scheme based on the stochastic MCNPX code with the objective to provide reference numerical solutions. This paper presents the first steps undertaken in that direction. To start, a methodology for a volumetric neutron source transfer to full core MCNPX models with detailed CR as well as axial reflector representations is established. On this basis, the assumptions of the deterministic methodology are studied for selected CR configurations for two Beginning-of-Life cores by comparing the spatial neutron flux distributions obtained with the two approaches for the entire spectrum. Finally, for the high-energy range (E> 1 MeV) and for a few CRs, the new MCNPX scheme is applied to estimate the accumulated fluence over one real operated cycle and the results are compared with the deterministic approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Reactor physics studies for assessment of tramp uranium methods

This paper presents calculation studies towards validation of a methodology for estimations of the tramp uranium mass from water chemistry measurements. Particular emphasis is given to verify, from a reactor physics point of view, the justification basis for the so-called 'Pu-based model' versus the 'U-based model' as a key assumption for the methodology. The computational studies are carried out for a typical BWR fuel assembly with CASMO-5M and MCNPX. By approximating the evolution of fissile nuclides and the fraction of {sup 235}U fissions to total fissions in different zones of a fuel rod, including tramp uranium on the clad surface, it is found that Pu gives the dominant contribution to fissions for tramp uranium after an irradiation on the outer clad surface of at least one cycle in a BWR. Thus, the use of the so-called Pu model for the determination of the tramp uranium mass (this means in particular using the yields for {sup 239}Pu fission) appears justified in the cases considered. On that basis, replacing the older U model by a Pu model is recommended. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Nuclear data uncertainty propagation in a lattice physics code using stochastic sampling

A methodology is presented for 'black box' nuclear data uncertainty propagation in a lattice physics code using stochastic sampling. The methodology has 4 components: i) processing nuclear data variance/covariance matrices including converting the native group structure to a group structure 'compatible' with the lattice physics code, ii) generating (relative) random samples of nuclear data, iii) perturbing the lattice physics code nuclear data according to the random samples, and iv) analyzing the distribution of outputs to estimate the uncertainty. The scheme is described as implemented at PSI, in a modified version of the lattice physics code CASMO-5M, including all relevant practical details. Uncertainty results are presented for a BWR pin-cell at hot zero power conditions and a PWR assembly at hot full power conditions with depletion. Results are presented for uncertainties in eigenvalue, 1-group microscopic cross sections, 2-group macroscopic cross sections, and isotopics. Interesting behavior is observed with burnup, including a minimum uncertainty due to the presence of fertile U-238 and a global effect described as 'synergy', observed when comparing the uncertainty resulting from simultaneous and one-at-a-time variations of nuclear data. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2012

Transition to CASMO-5M and SIMULATE-3K for stability analyses of the Swiss BWRs

For stability analyses of the Swiss operating Boiling-Water-Reactors (BWRs), the methodology employed and validated so far at the Paul Scherrer Inst. (PSI) was based on the RAMONA-3 code with a hybrid upstream static lattice/core analysis approach using CASMO-4 and PRESTO-2. More recently, steps were undertaken towards a new methodology based on the SIMULATE-3K (S3K) code for the dynamical analyses combined with the CMSYS system relying on the CASMO/SIMULATE-3 suite of codes and which was established at PSI to serve as framework for the development and validation of reference core models of all the Swiss reactors and operated cycles. This paper presents a first validation of the new methodology on the basis of a benchmark recently organised by a Swiss utility and including the participation of several international organisations with various codes/methods. Now in parallel, a transition from CASMO-4E (C4E) to CASMO-5M (C5M) as basis for the CMSYS core models was also recently initiated at PSI. Consequently, it was considered adequate to address the impact of this transition both for the steady-state core analyses as well as for the stability calculations and to achieve thereby, an integral approach for the validation of the new S3K methodology. Therefore, a comparative assessment of C4 versus C5M is also presented in this paper with particular emphasis on the void coefficients and their impact on the downstream stability analysis results. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗