Published by De Gruyter

Volume 84 Issue 4

Issue of Kerntechnik

Contents
Journal Overview

Contents/Inhalt

August 27, 2019

Page range: 209-211

Editorial

August 27, 2019

Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018

S. Kliem

Page range: 212-213

Technical Contributions/Fachbeiträge

August 27, 2019

Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments

R. Ferrer, J. Hykes, J. Rhodes

Page range: 214-227

Abstract

Studsvik has recently extended the CASMO5 advanced lattice physics code for the analysis of VVER 1000 and 1200 reactors. These extensions form the basis of CASMO5-VVER, which is primarily intended to compute homogenized nodal data for SIMULATE5-VVER. CASMO5-VVER leverages the latest nuclear data and numerical methods to VVER analyses. The current CASMO5 data library, based on the ENDF/B-VII.1 nuclear data evaluation, features a 586 energy group structure and nuclear data for hundreds of unique nuclides. Resonance self-shielding, based on the Equivalence Theory and an Optimal Two-Term Rational (OTTR) method, has been extended to support hexagonal geometry. The solution to the two-dimensional transport equation is based on the new Linear Source (LS) approximation for the Method of Characteristics (MOC). The acceleration of the MOC solution is attained through the implementation of a Coarse-Mesh Nonlinear Diffusion (CMND) acceleration. Results from preliminary validation against various critical experiments are presented in this work.

August 27, 2019

C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant

I. Pós, Z. Kálya, T. Parkó, M. Horváth, S. P. Szabó

Page range: 228-241

Abstract

The C-PORCA/HELIOS models have been used at NPP Paks as basic core neutron physics calculation tools for many years. C-PORCA is a node-wise diffusion model for the purpose of 3D core analysis. HELIOS is a well-known neutron transport code. Its utilisation at Paks NPP has a dual use. This code is a basic tool for preparation of homogenised few-group neutron cross sections inside fuel nodes and areas without fuel and the flexibility of HELIOS allows using it for testing. During the last decade some new kind of fuel assemblies were utilised in Paks. In order to ensure the accuracy and performance requirements of the off-line core analysis and in-core monitoring, continuous development and testing of the codes have been performed. In this paper the main characteristics of the diffusion solver applied in the C-PORCA model are described. The accuracy of this solver is also demonstrated on the basis of comparisons with different international references available in hexagonal geometry. The C-PORCA results have been compared against benchmark data produced in the framework of the AER (Atomic Energy Research) community in recent decades. All presented comparisons illustrate that the accuracy of the C-PORCA diffusion solver is excellent.

August 27, 2019

Adaptation of the gas gap simplified model in DYN3D code to new types of fuel

M. Ieremenko, Iu. Ovdiienko

Page range: 242-245

Abstract

Currently, new types of fuel are being considered to be introduced or already in the introduction process at Ukrainian NPPs with WWER. By means of a new version of the TRANSURANUS code, new functions of the gas gap thickness in dependence on the burnup have been created and implemented into the gas gap model of the reactor dynamics code DYN3D. These new functions cover all actual and perspective fuel types for the Ukrainian NPPs with WWER.

August 27, 2019

A procedure for verification of Studsvik's spent nuclear fuel code SNF

T. Simeonov, C. Wemple

Page range: 246-251

Abstract

Studsvik's approach to spent nuclear fuel analyses combines isotopic concentrations and cross-sections, calculated by the lattice physics codes, with core irradiation history data from reactor core simulator and tabulated isotopic decay data. These data sources are used and processed by the code Spent Nuclear Fuel (SNF) to predict spent nuclear fuel characteristics. The procedure for verification and validation of SNF is outlined in this paper which includes verification of the decay data as well as the numerical methods where they are applied into. The paper presents the applicability of SNF to analyses of spent nuclear fuel in comparisons to well established and recognized code in the spent fuel analyses field, ORIGEN, SCALE v6.2.2.

August 27, 2019

Extension of nodal diffusion solver of Ants to hexagonal geometry

A. Rintala, V. Sahlberg

Page range: 252-261

Abstract

The development of a new computational framework for core multi-physics problems, called Kraken, has been started at VTT Technical Research Centre of Finland Ltd. The framework consists of modular neutronics, thermal hydraulics and thermal mechanics solvers, and is based on the use of continuous-energy Monte Carlo reactor physics program Serpent. Ants is a new reduced order nodal neutronics program developed as a part of Kraken. The published methodology and first results of Ants has previously been limited to rectangular geometry steady state multigroup diffusion solutions. This work describes the solution methodology of Ants extended to hexagonal geometry steady state diffusion solutions. The first results using various two-dimensional and three-dimensional hexagonal geometry numerical benchmarks are presented. These benchmarks include the AER-FCM-001 and AER-FCM-101 three-dimensional VVER-440 and VVER-1000 mathematical benchmarks. The obtained effective multiplication factors of all considered benchmarks are within 18 pcm and the RMS relative assembly power relative differences are within 0.4% of the reference solutions.

August 27, 2019

VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6

M. Lovecký, J. Závorka, J. Vimpel

Page range: 262-266

Abstract

Geometry models for Monte Carlo transport codes have been using standard constructive solid geometry (CSG). The standard approach is using analytical equations for defining surfaces from which spatial cells are constructed. However, this approach can be quite time consuming and possibly error prone for complex models. Monte Carlo transport codes are continuously developed, one of the paths is using CAD-based mesh geometry. MCNP6 features unstructured meshes (UM) created with Abaqus/CAE as geometry description. Attila4MC package for creation of UM geometry from CAD model can be used for MCNP6 models. VVER-1000 fuel assembly model in UM geometry was created for TVSA-T.mod.2 fuel type. Basic validation of the model was performed, initially for criticality calculations. In the future, the model will be used for criticality safety analyses, preparation of boundary conditions for diffusion codes and radiation shielding analyzes of spent fuel transport and storage facilities.

August 27, 2019

Fuel cycles with PK-3+ FAs for VVER-440 reactors

P. Mikoláš, J. Vimpel

Page range: 267-279

Abstract

In order to increase the efficiency of fuel utilization at Dukovany NPP, the design of FA was changed by shroud removal and replacement with a structure called “Karkas”. Optimization of PK-3+ type FAs with different average enrichments was performed in order to find out those enrichment profiles with minimized non-uniform energy generation in FA (during burn-up). In addition, it was assumed that such a radial enrichment profile in FA could be achieved by making a change in the location of the fuel pin with a Gd 2 O 3 burnable absorber – from the 2 nd row to the 3 rd row of pins from the edge of the fuel assembly on the fuel assembly diagonal. The aim of this study was to achieve a full quadruplicate cycle, every 15 months (approx. 450 days) at 1475 MW t nominal power. Preliminary results indicate that combination of PK-3+ and Gd-2M+ fuel assemblies does not show any unusual phenomena from the point of view of reactor physics. The proposed strategy is based on B1C33 cycle implemented at Dukovany NPP that is designed to be 395 FPDs. Already in the first “transient” cycle (34 th ) loaded with 60 fresh PK-3+ FAs and 12 Gd-2M++ CAs, the reached length at EOR is 424 FPDs, which means stretch-out 26 effective days. Averaged, the transition cycle stretch-out length is 23.5 effective days. For steady cycles, this average value is 19.2 effective days.

August 27, 2019

Prospects for implementation of VVER nuclear fuel enriched above 5%

A. V. Ugryumov, A. B. Dolgov, A. I. Shaulskaya, Yu. M. Semchenkov, E. K. Kosourov, A. I. Osadchiy, A. M. Pavlovichev

Page range: 280-284

Abstract

JSC TVEL has carried out a technical and economic study with the involvement of the National Research Centre “Kurchatov Institute” in the use of nuclear fuel enriched above the current limit of 5 wt% for VVER-1000/1200. The article presents neutronic characteristics of developed 18- and 24-month fuel cycles based on fuel enriched above 5 wt% and assessment of nuclear safety for fabrication and handling with high enriched fuel.

August 27, 2019

Core loading optimisation in Slovak VVER-440 reactors

R. Zajac, J. Majerčík, C. Strmenský

Page range: 285-289

Abstract

VVER-440 reactors have been utilized in Slovakia since 1978. So far, the vast majority of their core loadings were designed in VUJE institute. This paper presents a description of the methods and procedures, which have been used for this purpose in the last decade. Main attention is focused on the calculating tools for core refuelling scheme optimization.

August 27, 2019

Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data

M. Horváth, I. Pós, T. Parkó

Page range: 290-296

Abstract

After a preparation period with VERONA upgrade and lead test assembly program, a new fuel type was introduced at MVM Paks NPP Ltd. This 4.7% average uranium enriched assembly type, together with the former 4.2% uranium enriched fuels, allowed us to lengthen the operating cycles to 15 months. Both fuel types contain gadolinium burnable poison, in six and three pins respectively. All of the four units have been converted to the C15 cycles, and have been being operated without any problems in the last few years. In this paper the test results of core design code HELIOS/C-PORCA, which is the basic model of VERONA, are outlined. C15 cycles were entirely investigated with the comprehensive study of measured and predicted (calculated) values of different reactor states. At first step, in order to prove the capability of the reactivity calculation of the nodal diffusion model, critical boric acid concentrations of different burnup and start up states were calculated and compared with measured values. During the next step of the verification process local in-core parameters were investigated. Measured neutron flux distributions (SPND signals) and coolant outlet temperatures were examined. SPND and thermo couple (TC) signals were predicted as a part of the monitoring system. The results of statistical investigations (average differences and standard deviations) for the applied fuel types are also presented.

August 27, 2019

Optimized 18-months low-leakage core loadings for uprated VVER-1000

A. L. Egorov, A. M. Pavlovichev, M. A. Sumarokov, S. M. Zaritskiy, A. S. Morozov

Page range: 297-301

Abstract

The VVER-1000 life extension and power uprating are important areas of research and development activities of JSC “Rosenergoatom”. Recent research has shown that the core baffle resource is one of the major limiting factors for the life extension of the Balakovo NPP unit 3. The core loading optimization is necessary to decrease the core baffle radiation dose for NPP life extension at uprated power. The paper contains the results of the optimization of 18-month low-leakage core loadings for the uprated to 104% Balakovo NPP unit 3. Moreover, the results of the radiation dose calculations at the reactor pressure vessel and core baffle are presented in the work. It is shown that the optimization of the core loadings reduces peaks of radiation doses and flattens its distributions along the core baffle, as well as gives the opportunity to extend the NPP life up to 60 years at the uprated power.

August 27, 2019

Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code

M. V. Suslov, I. G. Petkevich, M. A. Uvakin

Page range: 302-308

Abstract

The paper deals with the results of the Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code. The test was held at the 1st unit with VVER 1200 reactor on the 29 th of March. During the test the steam generators were refilled with EFW system at first. This system is connected to the SG steam space. Therefore, injection of water leads to the secondary pressure decrease. This phenomenon is difficult for simulation due to complicated structure of the SG internals. Also, the test provides unique data for the reactor facility cooldown at natural circulation in the primary circuit. The important data for reactor head and pressurizer cooldown were obtained. Thus, the test data provides the validation of several processes such as: emergency feedwater injection into the steam generator; steam generator filling with “cold” water from AFWS; pressure regulation at filled pressurizer; cooldown of the reactor head and the coolant circulation beneath the head. The results of the test were used for the validation of KORSAR/GP code, unit nodalization and equipment models.

August 27, 2019

Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code

A. I. Sinegribova, M. A. Uvakin, M. A. Bykov

Page range: 309-315

Abstract

The purpose of this paper is the assessment of the FA pin-by-pin model available in the KORSAR/GP code. The paper presents the results of the simulation of an “Control rod ejection” accident. This accident is of interest due to the power redistribution in the FAs effected by the ejection. The safety margins are calculated with the use of the “hot channel” model for NPP safety analyses. Only the fuel rods of the core with the highest power are taking into account. The absence of the change of the pin power distribution for the FA during the transient is compensated by increased conservatism. In this paper, the maximum fuel rod power was obtained using the pin-by-pin model.

August 27, 2019

Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies

Y. Perin, S. Nikonov, R. Henry, I. Pasichnyk, K. Velkov

Page range: 316-321

Abstract

The OECD/NEA Benchmark on NPP Kalinin Unit 3 “Switching-off of one of the four operating main circulation pumps at nominal power” based on measurements is at its end phase. A new benchmark is starting based on measurements from the NPP Rostov Unit 2 “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core”. The Rostov-2 core is loaded with assemblies of a new modern type “TVS-2M” which differs in construction in comparison with the “TVSA”-type implemented in the Kalinin-3 core. The goal of the performed study is to determine the thermohydraulic differences, during steady-state operation, between the two cores based on loads with different assembly types. Results of steady-state simulations are compared for Kalinin-3 and Rostov-2 cores, taken from the respective Benchmarks specifications. To ensure that the observed differences are coming only from the fuel assembly construction, the same axial power density is used for all cases. Calculations are done with an under-development version of the GRS system code ATHLET, which has sub-channel modeling capabilities. Thus, in addition to the typical assembly-wise models, calculations with seven assemblies modeled at the sub-channel (pin-by-pin) level were also performed and analyzed. This work will help the Benchmarks' Rostov-2 participants to set up more accurate core thermohydraulic models.

August 27, 2019

Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320

J. Hádek, R. Meca

Page range: 322-333

Abstract

The paper gives a description of conservative analysis of initiating event associated with uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín NPP. This event is included in the group of beyond design basis accidents. The aim of analysis is to determine also the time interval which is necessary for interventions leading to the deterrence of fuel damage. The failure of operator intervention to isolate dilution routes at intervals shorter than 30 min is assumed. Since the response of the whole NPP system influences the course of safety important parameters of the reactor core, the calculations were made by an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic system code ATHLET. It is shown that, in addition to exceeding the DNBR limit of more than 99 min from the start of the transient, the remaining safety acceptance criteria will not be violated until the end of the calculation.

August 27, 2019

Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies

J. Kuopanportti, T. Lahtinen

Page range: 334-339

Abstract

In this paper, the optimization of the assignment of spent fuel assemblies into final disposal canisters is considered. This application is of essential importance as the final disposal canisters are expensive and, on the other hand, there exists a limit for the canister-wise total heat load, which must not be exceeded. The study utilizes mathematical optimization algorithms that have been developed by Ranta in his D.Sc. thesis (Tampere University of Technology, 2012). In the applied formulation, the target of the optimization is to minimize the maximum canister-wise decay heat load at the time of canister formation. The optimization algorithms were utilized for analysing a fictional final disposal scenario for present and expected future spent fuel assemblies of Loviisa NPP. The paper concludes that, despite a huge amount of degrees of freedom, the algorithms are capable of finding practically a global optimum for the considered problem. The implemented software tool can be utilized for further final disposal optimization analyses.

August 27, 2019

Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel

M. Chernykh, D. Amian, S. Tittelbach, A. Bannani, W. Cebula, T. Funke, R. Hüggenberg

Page range: 340-345

Abstract

The phase-out of nuclear energy in Germany has triggered the demand for a comprehensive solution to dispose of damaged fuel rods, normally remaining in the fuel pond until the final shutdown of the NPP. In order to establish a disposal concept for damaged fuel rods suitable for the needs of German utilities, GNS has developed a first of its kind solution, the Integrated Quiver System for Damaged Fuel (GNS IQ®), which can be loaded into the transport and storage casks of the CASTOR® family also by GNS. The GNS IQ® features a robust yet simple design with a high mechanical stability, a reliable leak tightness and large safety margins for future requirements on safety analysis. It can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific needs of the customer. The quiver is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The paper gives a general overview of the disposal concept and provides a description of the Integrated Quiver System for Damaged Fuel with the focus on criticality safety assessment.

August 27, 2019

Neutron balance in two-component nuclear energy system

V. Blandinskiy

Page range: 346-350

Abstract

Most nuclear reactors under operation are thermal reactors, which consume 235 U in once-through fuel cycle resulting in ineffective resource utilization and dramatic SNF volume growth. However, sustainable nuclear energy system (NES) should provide NFC closing for all hazardous radionuclides to minimize its life-time within NES and to make risk to be proportional to NES capacity, rather than total energy produced. These two basic principles require enough amount of neutrons for both energy generation and hazardous radionuclides transition to fission products. Therefore, taking into account politic, economic and technological risks and uncertainties, these issues can be solved in terms of two-component NES consisting of both thermal and fast reactors. In this work two methods to estimate neutron balance in NES are discussed. The fist method is based on the analysis of nuclear transformation chain due to radioactive decays and neutron induced reactions. The second one is the most complete one and relies on reaction rates comparison. Neutron balance estimation approach is demonstrated for two-component NES case study.

About this journal

Kerntechnik is an independent journal for nuclear engineering (including design, operation, safety and economics of nuclear power stations, research reactors and simulators), energy systems, radiation (ionizing radiation in industry, medicine and research) and radiological protection (biological effects of ionizing radiation, the system of protection for occupational, medical and public exposures, the assessment of doses, operational protection and safety programs, management of radioactive wastes, decommissioning and regulatory requirements). For more than 75 years Kerntechnik offers original scientific and technical contributions, review papers and conference reports.
All articles are subject to thorough, independent peer review.