THERMAL HYDRAULIC ANALYSIS AND MULTIPHYSICS SIMULATION

Analyzing the thermal hydraulic behavior of NPPs during transients and accident conditions is one of the major fields of evaluation of nuclear safety. The Thermal Hydraulic Analysis and Multiphysics Simulation Group (THG) at NRSC has an advanced and comprehensive set of computational tools, and is quite experienced to performing thermal hydraulic analyses and meeting the requirements of nuclear industry.To model and analyze thermal hydraulic processes NRSC uses specific analytical tools, namely:

• RELAP5 and TRACE codes are used for the analysis of system behavior for the entire spectrum of transients and accidents.
• MELCOR code is used for the analysis of beyond design-basis accidents, severe accidents that cause core melting, and for the transport of radionuclides in the reactor coolant system and reactor site.
• COCOSYS code is used for the simulation of processes in hermetically closed plant areas (containment).

This group is also well experienced in analyzing the radiological consequences during incidents and severe accidents. And the main tools used for radiological consequence analysis are RADTRAD and RASCAL codes.
Reactor safety analysis of both PWRs and BWRs relies on system codes, where the primary and secondary flows are modeled with a rather coarse nodalization. However, some safety issues had been identified, consequently, a better resolution had been required for the simulation tools. These issues are often related to situations when the 3D aspects of the flow and the geometrical effects have a significant influence on the safety criterion. Therefore, the NRSC started to extend its analytical capabilities by using the advanced CFD code namely ANSYS CFX code, which allows to model and analyse the 3D aspects of the flow.

Thermal Hydraulic Analysis using RELAP5 Code

Thermal hydraulic analyses are mainly performed for deterministic safety assessment of the NPP that is followed by the traditional concepts of engineering analysis. RELAP5 code is used to address plant behavior under specific pre-determined operational states and accident conditions and to assess design adequacy.
NRSC developed VVER-440 reactor’s model using RELAP5 code, which was then validated based on an actual transient at the Armenian Nuclear Power Plant (ANPP).

In the framework of different projects and by the usage of the developed model, NRSC performed numerous design basis accidents analyses, e.g.

• core cooling with low pressure pumps during the Loss of Coolant Accident (LOCA),
• full spectrum of a loss of coolant design-based accident,
• independent verification of justifications for replacement of the PRZ and the SG safety valves,
• primary to secondary leaks,
• accidents which can potentially lead to RPV cold overpressure,
• justification analysis of the ANPP EOPs,
• accident analysis performed within the framework of Developing a Comprehensive Modernization Program for the ANPP.

To assess all phenomena during an accident state, NRSC continuously updates its developed models. As for analyzing transient and steady-state neutronic thermal-hydraulic behavior in light of light water reactors, the NRSC has also started development of ANPP model based on the TRACE code.

Severe Accident Analysis

For the analysis of potential consequences of severe accidents, and for the possible counter measures under conditions as realistic as possible the following actions are needed: simulation of severe accident propagation in reactor coolant system, reactor cavity, containment, and in confinement buildings. NRSC uses MELCOR code to assess the reactor systems and their response to off-normal or accident conditions, namely:

• thermal hydraulic response of the reactor coolant system, the reactor cavity, the containment, and the confinement buildings to severe accidents,
• core uncovery (loss of coolant), fuel heat-up, cladding oxidation, fuel degradation (loss of rod geometry), and core melting and relocation,
• heat-up of RPV lower head the thermal and mechanical loading and failure of the vessel lower head, and the transfer of core materials to the reactor vessel cavity,
• core-concrete interaction and ensuing aerosol generation,
• in-vessel and ex-vessel hydrogen production, transport, and combustion,
• fission product release (aerosol and vapor), transport and deposition,
• behavior of radioactive aerosols in the reactor containment building, (including the scrubbing in water pools), and aerosol mechanics in the containment atmosphere such as particle agglomeration and gravitational settling,
• impact of engineered safety features on thermal hydraulic and radionuclide behavior.

By the use of the developed models the following analyses were performed:

• severe accidents with LOCA
• station blackout analysis
• assessment of spray system’s modification options.

Severe accident analyses were also used for developing emergency procedures for the Armenian Nuclear Regulatory Authority. The developed procedures allow the assessment and prognosis of ANPP core state, the release path, and the expected doses of the population.

Confinement Analysis

The simulation of severe accident propagation in reactor containments is required for the analysis of potential consequences, and for possible counter measures. NRSC uses COCSYS to develop a model for the assessment of the ANPP containment’s safety issues.

The following analyses were performed:

• analysis of the ANPP confinement response to the design basis accidents and the beyond design basis accidents taking into consideration the hydrogen safety,
• investigation of the ANPP confinement response to new DBA LOCA,
• investigation of aerosol and fission product behavior in the ANPP containment during severe accidents,
• investigation of the ANPP containment response to LOCA after the installation of Jet Vortex Condenser,
• investigation of sump clogging phenomena,
• investigation of feasible modification of the spray system.

Radiological Consequence Analysis

Radiological consequence analyses for incidents and accidents represent another field of expertise at NRSC. This field is about the analysis of possible scenarios during which radioactive materials are released into the environment. Once the conditions of the release (the type and the amount of the radioactive materials) are determined, the dispersion of these materials in the atmosphere is calculated by using the models developed by RADTRAD or RASCAL codes.

To calculate the radiological consequences of various transients, incidents and accidents, that are significant in terms of radiological releases, is the same as to verify the systems’ proper design and operation. It also aims to show that the discharge of radioactive products outside of the plant remains within the limits set out by radiation protection norms. Radiological consequence analyses are also performed to define the limits of emergency arrangement zones.

During the implementation of different projects, radiological consequence assessment was performed for the following design based accidents, beyond design based accidents and severe accidents:

• LOCA with equivalent diameter of 150mm
• steam generator tube rupture
• pressurizer safety valve opening and sticking in open position LOCA
• main steam line break
• double ended break of pressurizer’s surge line with equivalent diameter of 200mm
• double ended break of main circulation pipeline severe accident.

Computational Fluid Dynamics Simulations

Computational Fluid Dynamics (CFD) is a relatively new field that solves problems with fluid flow by using numerical methods computationally. In recent years, this powerful tool has been used in predicting fluid flow and heat transfer characteristics during normal and emergency operation of the NPPs. Because the areas of application have not been verified and approved yet, the opportunities are very limited.
To stay faithful to its tradition – always be up to date with current trends in nuclear analysis – NRSC started implementing the CFD capabilities over the last few years. Consequently, the NRSC purchased one of the most widespread and validated codes used for the CFD – the ANSYS CFX code. And as for the code’s usage, USAID provided with guidance to the thermal hydraulic specialists. The ANSYS CFX code is particularly useful in predicting 3D details of turbulence and multiphase fluid flow, which, in fact, would have not been predictable with traditional 1D RELAP and MELCOR lump parameter codes.
The NRSC has so far performed several analyses of flow phenomena at the ANPP Unit 2 VVER-440 reactor. And the cold coolant mixing phenomena during emergency core cooling injection and the hydrogen distribution inside the containment in case of LOCA were just the two of the performed analyses.

The ANPP unit 2’s temperature profile on coolant vessel interface after emergency core cooling system injection

Hydrogen density distribution in the containment in case of LOCA

Training

The THG provides training in the field of thermal hydraulic analysis, severe accident analysis, radiological consequence analysis, and a basic course on the ANSYS CFX code usage.