NEUTRONICS AND NUCLEAR FUEL

As part of deterministic safety analysis, neutronics analysis is applied in all stages of nuclear fuel cycle: transportation, fuel utilization in the reactor core, and storage. The main capabilities and focus areas of Neutronics Analysis and Fuel Group (NFG) are:power distribution, nuclear instrumentation response, fuel management, and analysis of reactivity effects by PARCS-RELAP/TRACE coupled code system:

• reactivity induced reactor transient and accident analysis by 3D kinetics PARCS-RELAP/TRACE coupled code system,
• analysis of reactor pressure vessel fluence by the DORT neutron transport code,
• criticality safety analysis, including the analysis of burnup credit approach for spent nuclear fuel storage and cooling pools by KENO-VI (SCALE 6) and STARBUCS (SCALE 6) codes,
• development of neutron-nucleus interaction cross-section libraries by the HELIOS, TRITON (SCALE 6) codes,
• analysis of nuclear fuel depletion by ORIGEN (1D), TRITON (2D, 3D), MCNP 6.1 (3D), ORIGEN-KENO-VI (3D), MONTEBURNS (3D) codes,
• analysis of nuclear fuel thermal-mechanical behavior by TRANSURANUS code,
• spent fuel decay heat analysis by ORIGEN (SCALE 6) code, and
• shielding analysis (e.g. spent fuel cask, NPP containment rooms, liquid waste drums) by MORSE, MONACO/MAVRIC (SCALE 6) codes.

This group was established in 2003, and since then it has been successfully involved in regulatory review and independent safety analysis of Armenian Nuclear Regulatory Authority’s decision-making.

Nuclear Reactor Core Performance Analysis

This group is able to review and verify the analysis of the ANPP routine fuel loadings. For this very reason, by using the PARCS-RELAP/TRACE Program, the NRSC has developed a coupled-neutronics and thermal-hydraulics models of the ANPP reactor core.

ANPP Reactor Core Neutronics Model

For the ANPP rector core reflector the NRSC developed a special model with a side-dependent ADF. In addition, by comparative analysis with the Russian BIPR-7A neutronics code, the NRSC also verified and validated the reactor core model against the ANPP operational data. During the process of the reactor core model development, the NRSC suggested to modify several aspects of the PARCS source code so that to accommodate the WWER-440 reactor core and nuclear fuel peculiarities.

Analysis of accidents initiated with reactivity change

The analysis of the reactor core’s response to the different mechanisms of reactivity insertion is one of the most important parts of deterministic accident analysis. Furthermore, to verify SAR and to license new fuel with 3.82% enrichment, the NRSC carried out reactivity-accident analysis for the HZP and the HFP conditions.

Reactor Core Power distribution during Control Assembly Ejection Accident

Within the development of the ANPP Comprehensive Modernization Program, the NRSC carried out the RIA analysis of the ANPP. In addition, the NRSC also developed the reactor core’s model by the PARCS code.

Analysis of Reactor Pressure Vessel Fluence

Because of the fast neutron fluence the embrittlement of the RPV tends to limit the lifetime of reactor pressure vessel, therefore, to verify the SAR and to license new fuel with 3.82% enrichment, the NRSC carried out neutron fluence analyses.

Reactor 2D Model

The NRSC made analysis by DORT program which, in turn, uses discrete ordinates and flux synthesis methods.

Criticality Safety Analysis

Nuclear criticality safety analysis is an important domain of neutronics analysis, which is meant to demonstrate how to prevent nuclear and radiation accidents resulting from an inadvertent, self-sustaining nuclear chain reaction. The Results of NRSC criticality analysis were used to verify the ANPP SAR, the ANPP spent fuel pools safety, and the safety of NUHUMS type of spent fuel storage.

3D Model of Spent Fuel Transport Cask

These models were developed by KENO-VI and MCNP6.1 programs.

Development of Neutron-Nucleus Interaction, Cross-Section Libraries

One of the key elements assuring the quality of neutronics analysis is the availability of an adequate cross-section library. Consequently, the cross-section libraries are used in reactor core performance and in RIA analysis.

ANPP Reactor Core Reflector Model

By using the HELIOS-2 code NRSC developed the cross-section libraries, the fuel model, and the control assemblies.

Analysis of Nuclear Fuel Depletion

The isotopic composition of spent fuel is one of the main inputs for further criticality, decay heat, severe accident and shielding analyses. The results of the composition analysis were used for various independent verification safety analyses in ANRA’s decision-making.

3D Model of the WWER-440 Fuel Assembly

Depending on the accuracy of the problem 1D, 2D and 3D, models of the WWER-440 fuel assemblies were developed by using ORIGEN, TRITON and MCNP6 codes, respectively.

Development of the National Regulation in the Field of Nuclear Safety

In the development of national regulation the NAF is involved in the following areas:

• design safety requirements,
• requirements of safety assessment report’s format and content,
• reactor site evaluation requirements, and
• procedure of accounting for and control of nuclear materials.

International Cooperation

In the field of deterministic safety analysis the NAF is also involved in international projects such as topical projects, working group activities, missions, trainings, and so on. In light of this cooperation, we would like to particularly highlight the following partners:

• The Reactor Physics Working Group of the WWER Regulatory Forum

The WWER Regulatory Forum established the Reactor Physics Working Group (WG) on February 2014. The mission of WG is to develop common approaches and methodologies in the field of reactor core and nuclear fuel safety analysis for countries that use WWER technology. The group is also going to setup benchmarks to verify different codes used for WWER reactors safety analysis.

• The University of Michigan

In collaboration with the University of Michigan, the PARCS code was modified to correctly model the WWER-440 reactor core and to control assemblies’ peculiarities.

• The Brookhaven National Laboratory

In collaboration with the BNL, various regulatory research programs were carried out in the following areas: neutron fluence analysis, criticality safety analysis, burnup credit analysis, and cross-section library development.