Safety and Integrity of Nuclear Fuel
Development of Neutronics Model for the Use in the Analysis of Fuel Safety and Fuel Cladding Mechanical Integrity
Tech Area / Field
- FIR-ENG/Reactor Engineering and NPP/Fission Reactors
- FIR-FUE/Reactor Fuels and Fuel Engineering/Fission Reactors
- FIR-MOD/Modelling/Fission Reactors
- FIR-NSS/Nuclear Safety and Safeguarding/Fission Reactors
3 Approved without Funding
Nuclear and Radiation Safety Center of Armenian Nuclear Regulatory Authority, Armenia, Yerevan
- Brookhaven National Laboratory, USA, NY, Upton\nEuropean Commission / Joint Research Center / Institute for Transuranium Elements, Germany, Karlsruhe
Project summaryAs the international industry develops the issue of environment pollution becomes more topical. The accepted standard codes for operation of various ecologically dangerous industrial facilities allow reducing the environment pollution to minimum when operated in standard (design) conditions. When man-induced accidents occur, the ecologically hazardous releases into environment could exceed allowed values hundred times.
The Armenian Nuclear Power Plant (ANPP) operated on the area of the Republic of Armenia (RA) is one of the facilities in the Southern Caucasus where an accident could result in ecological disaster on a regional scale.
In regard with great significance of the ANPP for energy development in the region the RA Government made a decision on the ANPP operation up to 2016 provided the required safety level is ensured.
Ensuring the safety and integrity of nuclear fuel is a cornerstone in the ensuring the safety of the reactor as a whole, since nuclear fuel is one of the key barriers of in-depth protection of nuclear power plants.
Assessments of the safety of nuclear fuel in the reactors are performed indirectly - by using the system codes only thermal characteristics of the fuel and its cladding are obtained (the temperature of the fuel and cladding, fuel enthalpy, etc.). However, the detailed evaluation of thermo-mechanical impacts on the fuel (fuel cladding elongation, swelling and cracking of the fuel, gas pressure change under the cladding, fuel cladding oxidation, corrosive wear, increasing the diameter of the fuel cladding, etc.) at the level of inpidual fuel rods could be performed only with special thermo-mechanical programs, such as, TRANSURANUS [1-18].
TRANSUARNUS program was originally developed for the analysis of the fuel of western design reactors. In recent years, intensive research work [1-4, 6, 11-12] had been carried out in order to apply TRANSURANUS for WWER (water-moderated reactors of Russian design) fuel safety analysis.
One of the most important parameters for assessing the characteristics of fuel is the radial distribution of neutron flux, and major actinides and fission products in the fuel matrix, that essentially depends on the fuel lattice and the neutron spectrum.
In the TRANSURANUS, to take into account the radial distribution of main actinides and fission products approximate formula had been used, that was developed for western design reactors and didn’t take into account neutron-physical peculiarities of WWER fuel.
To extend the scope of above mentioned approximate formula for the lattice and the spectrum of WWER fuel NRS STC performed complex analysis [2-3, 19-20], using the program MONTEBURNS . As a result, the approximation formula used in TRANSURANUS was significantly improved, which allowed to seriously improve the accuracy of the calculation of thermo-mechanical parameters of the WWER fuel. In particular, the difference between the calculated and experimental values of fuel centerline temperature was reduced almost 2 times .
However, this approximate formula is not universal, since currently different types/modifications of nuclear fuels are used both in Russian and western design reactors and neutronics characteristics themselves depend on the parameters of the fuel in the reactor core and the history of fuel operation. In particular, on the coefficients in the approximate formula seriously affect the history of power, changing the concentration of boric acid, temperature of fuel and moderator, changes in the isotopic composition of fuel, etc. [2-3, 19-20]. Considering all of the above mentioned differences in approximation formula is virtually impossible that significantly influence on the accuracy of safety and integrity assessment of the nuclear fuel. In this connection there is strong need to develop a single universal neutronics model for TRANSURANUS program, that enables an adequate and comprehensive approach to addressing issues related with safety and integrity of nuclear fuel.
The main objectives of the project are a) adaptation of the program TRANSURANUS for the safety analysis of WWER nuclear fuel and b) development of neutronics module for the program TRANSURANUS to apply for WWER fuel.
To achieve above mentioned objectives the following main tasks is expected to address:
- Feasibility analysis of development and implementataion of neutronics module
- Development separate neutronics module for TRANSURANUS
- Integration of new neutronics module in TRANSURANUS
- Verification and validation of the neutronics module
- Verification and validation of TRANSURANUS program with new neutronics module
The project has one participant organization from Armenia - Nuclear and Radiation Safety Scientific Technical Center of State Committee for Nuclear Safety Regulation at Government of Armenia.
NRS STC is one of the leading scientific centers in Armenia in the field of nuclear power plants’ safety, modeling of neutronics, thermal hydraulics and thermal mechanical processes. In recent years, NRS STC performed the following analysis in the field of neutronics calculations:
- Calculation of the isotopic composition of nuclear fuel with the use point (ORIGEN, SCALE), two-dimensional (TRITON, SCALE) and three-dimensional (MONTEBURNS) model of nuclear fuel
- Analysis of the kinetics of VVER-440 reactor of ANPP. The full scope reactivity induced accidents had been analyzed (control assembly ejection, withdrawal, drop, boron dilution, improper loading of fuel assemblies in the core)
- Independent verification analysis of ANPP neutronics analysis.
- Criticality safety analysis of spent nuclear fuel storage using the program KENOVI (SCALE)
- Residual heat release analysis of spent nuclear fuel using the programs ORIGEN, TRITON (SCALE).
The Project will provide:
- Adaptation TRANSURANUS program for WWER fuel
- Determination of the real radial distribution of neutron flux, and major actinides and fission products in the WWER fuel
- Increasing the accuracy of the safety assessment of WWER nuclear fuel
The project will help scientists and professionals from Armenia, formerly occupied by military problems, apply their knowledge to solve peaceful problems.
During TRANSURANUS modification period contacts were established with the leading experts in the field of nuclear fuel safety and neutronics analysis, developers of TRANSURANSUS (Joint Research Centre, Institute for Transuranium Elements) that are collaborators of this project. The main role of Western experts in the project will be the participation of the implementation of neutronics module in the TRANSURANUS, and validation of TRANSURANUS with new module.
The project collaborators will continue to work with project participants in the following areas:
- Participation in advisory committee in each area, providing a certain contribution and leadership in the implementation of the Project;
- Exchange of data during Project;
- Choice of methodology and computational tools in the development of specific models;
- Conduct peer reviews of project results;
- Participation in the working meetings;
- Comments on technical reports submitted to the ISTC;
- Conducting joint seminars and symposia;
- Collaboration on scientific exchanges in various fields of research
- Dissemination of the approved results of the project
The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.
ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.