Numerical Study of Shutdown Coolability without Residual Heat Removal System (RHRS) via OECD-ATLAS Project

Abstract

The Residual Heat Removal System (RHRS) has been a cornerstone of safety and efficiency in nuclear power plants since the early days of commercial nuclear energy development. Its history is intertwined with the evolution of nuclear technology, and its significance lies in its ability to remove decay heat generated by the reactor core, even during shutdown conditions. This function is critical for preventing overheating, fuel damage, and maintaining system stability.

Advancements in RHRS design and technology have been pivotal in improving the overall safety and reliability of nuclear power plants. One notable advancement is the incorporation of multiple redundant trains, such as the two-train configuration. Redundancy in these systems enhances reliability and availability, reducing the risk of system failures that could lead to catastrophic consequences. Safety standards and regulatory requirements have become increasingly stringent over the years, driving the development of more robust and fail-safe RHRS designs. These systems undergo rigorous testing and validation processes to ensure they meet these elevated safety standards, providing assurance to both operators and the public.

This thesis delves into a specific aspect of RHRS performance and safety, focusing on the planned utilization of the OECD-ATLAS phase 2 B4.2 test and RELAP5/MOD3.3 codes. The primary objective is to simulate the loss of the RHRS under Mid-loop operating conditions in a nuclear power plant. This scenario involves deliberately reducing the inventory or coolant level in the reactor coolant system (RCS) on the primary side. Such reductions are meticulously executed for planned activities like maintenance and inspections, ensuring that the coolant level falls below specific thresholds for safe access to critical components, such as the pressurizer manway and steam generator inspection hatch. The Mid-loop operations are a critical aspect of nuclear plant maintenance and safety protocols. The deliberate lowering of coolant levels, while carefully controlled, poses unique challenges, and requires a comprehensive understanding of thermal- hydraulic phenomena. The objective of this research is to gain insights into these phenomena and to validate the B4.2 test through code simulation. The deliverables of this research are to construct a steady-state RELAP5 input code tailored to the specifics of the ATLAS B4.2 experiment and to develop transient RELAP5 simulations for pre-test calculations. In addition, the research incorporates sensitivity studies addressing critical factors like counter-current flow limitation (CCFL) implementation, heat loss effects, and the integration of pressurizer manway collection tanks.

As for the outcome of this research, the RELAP5/MOD3.3 code is an essential tool for simulating and analyzing the behavior of nuclear power plants. However, it has been observed that this code may have limitations in accurately predicting certain thermohydraulic phenomena, specifically CCFL and reflux condensation, under specific operating conditions. These conditions typically involve low pressure and reduced Reactor Coolant System (RCS) levels within the nuclear power plant.

Pre-test calculations and sensitivity studies have revealed discrepancies between the code's predictions and actual plant behavior, highlighting the need for modifications to address these shortcomings. To ensure the safety and reliability of nuclear power plants, it is imperative to refine and enhance the RELAP5/MOD3.3 code to better capture these thermohydraulic phenomena, especially when operating conditions deviate from the standard parameters. These modifications will contribute to more accurate simulations and improved safety assessments for nuclear power plants.

Date of Award	17 Nov 2023
Original language	American English
Supervisor	Yacine Addad (Supervisor)