Conference Agenda

In this study, a mass and energy (M/E) release analysis was performed on the postulated main steam line break (MSLB) accidents in the Advanced Power Reactor 1000 (APR1000). The M/E release rate was calculated using the SPACE-ME 1.0 code, developed by KEPCO Engineering & Construction Company, Inc. (KEPCO E&C), for various break areas ranging from an area fraction (AF) of 0.1 to 1.0, where AF 1.0 corresponds to the maximum double-ended guillotine break area. The initial core power was evaluated at 102%, 75%, 50%, 20%, and 0% of full power (%FP). To ensure conservative results, the break flow phase separation model and wall heat transfer multiplier were adopted. A simplified conservative model for the passive auxiliary feedwater system was used. The containment pressure and temperature responses were analyzed using the CAP 3.1 code with the calculated M/E release rates. A single failure of containment spray system was assumed. The highest containment peak pressure and temperature were found to be 0.7925 and 0.9531, respectively, which are normalized values with respect to the design values. The design margins of 20.75% for pressure and 4.69% for temperature during the most limiting MSLB accident indicate that the APR1000 containment can maintain its integrity well during the MSLB accidents. In conclusion, the new M/E release analysis methodology using SPACE-ME code is expected to be highly applicable to analyzing the postulated MSLB accidents in the APR1000.

This study presents the experimental results of large-scale containment thermal-hydraulics phenomena driven by the combined effects of steam injection and spray activation in a postulated Design Basis Accident (DBA) scenario, specifically a main steam line break. This experimental campaign, named P1A1_5 and P1A1_6, is part of the OECD/NEA PANDA project series. These experimental data could contribute to the assessment and validation of advanced computational tools for containment analysis.

The experiments were conducted in Vessel 1 of PANDA, a cylindrical confinement with 8 m in height and 4 m in diameter. A compartment representing the steam generator tower model was inside the Vessel 1. Initial conditions involved pressurizing with air and steam at 2.5 bar. There were defined with two different steam superheating (P1A1_5, P1A1_6), and the spray was activated using a single nozzle. The phases of the experiment were as follows: steam injection (phase 1), combined steam and spray injection (phase 2), and spray-only injection (phase 3 for P1A1_6).

Results showed that during containment depressurization, steam remains superheated above the spray nozzle. In contrast, below the spray nozzle, the fluid and saturation temperatures are approximately same value. Thus, the primary effects of steam injection and spray activation are the depressurization of Vessel 1 and the cooling of fluid and gas temperatures. Upon spray activation, the pressure and temperature gradients, especially below the spray nozzle, decrease more sharply over time compared to the phase 2. This is due to the enhanced momentum mixing and droplet behavior, particularly below the spray nozzle.

The NEA Committee on the Safety of Nuclear Installations (CSNI) has long supported international collaborations to enhance confidence in nuclear safety codes and experimental validation. One such initiative is the International Standard Problem (ISP), which began in the early 1970s and continues today.

A new ISP-52 was proposed upon the recommendations of the WGAMA/WGFS report: “Analyses of Design Extension Condition without Significant Fuel Degradation (DEC-A) for Operating Nuclear Power Plants” which highlighted the need for computer code validations for DEC-A conditions. For this purpose, the ETHARINUS project provided experimental data related to the PKL III J5.1 Run1 and Run2 tests. The latter addressed DEC-A scenario of Multiple Steam Generator Tube Rupture (MSGTR), which may occur following a severe earthquake, with limited safety system availability. In Run1 two double-ended guillotine breaks were considered in three out of four steam generators (SGs), while in Run2 the scenario was extended to all four SGs. Both test results were made available to the ISP-52 participants, but only Run 2 was selected for the “blind” and “open” analytical exercises.

This paper presents the main steps that have been followed to carry out the PKL III J5.1 Run1 and Run2 experiments and provide a description of the main events that took place during the course of the transient as well as the effectiveness of the operator actions (primary bleed, and manual activation of the ECCS) and the available ECCS to bring the system to safe shutdown conditions.

The TRACE code is used at Gösgen NPP to perform scoping simulations of plant behavior in response to extreme, very unlikely, external events. This allows plant personnel to investigate how the plant changes and refurbishments under study increase the safety margins in the event of Design Extension Condition (DEC), such as a Station Black-Out (SBO) with failure of the on-site emergency power supply. The safety of the unit is ensured thanks to the special emergency safety systems, bunkered and thus SBO-proven.

This study analyzes the benefits from the refurbished SEFW (Special Emergency Feedwater) tanks. In addition, this study investigates the automatic partial cooldown via Atmospheric Relief Valves (ARVs) (plant change not yet realized). Three simulations are presented in the paper. The full autarky time of 10 hours without operator actions is considered. In the first two cases the unit is kept at hot-shutdown conditions (no cooldown) and the secondary pressure is limited, respectively, by the cyclic opening of the Safety Relief Valves (SRVs) and the ARVs with automatic partial cooldown. The third case implements, based on the second case with ARV partial cooling, the manual cooldown procedure (slow gradient, 10 K/h) to reach cold shutdown.

The results of the simulations show that the automatic partial cooldown reduces the amount of primary coolant released into the containment by opening of the pressurizer safety valves. The increased safety margins of the plant in case of SBO are determined in 53 h (hot-shutdown state) and 42 h (cold-shutdown state) without external water injection.

During a Loss Of Coolant Accident (LOCA), in PWR’s, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from sumps in the lower part of the reactor building. A filtering system is implemented to collect debris, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the core. IRSN has launched an experimental R&D project investigating the clogging of sump filters by integral tests performed in the VIKTORIA loop, which was equipped successively with two types of 2 m² filters used in 900MWe NPP’s. The debris carried to the filter generate at 80°C (with chemistry) a very quick increase of the pressure drop across the filter (≈ 1 to 7 kPa according to the debris source term) that could be due to rapid chemical effects further to fibers corrosion. The two types of filters (rectangular pockets or planar types) behave very differently with rather low head losses for the second type. The recent experiments performed with less amount of fibers (by replacement of part of fibrous materials by RMI metallic insulation) led to significantly reduce the head loss without any consequences on the downstream behavior (debris transferred to the core). The increase of the duration of the corrosion of zinc in acidic conditions (as a sensitivity study) lead to increase head losses by a factor 4 the which indicates the formation of chemical precipitates.

In the first decades of the construction of the nuclear reactors, the attention in the field of nuclear safety analyses was focused on major accidents starting from full power (with maximum initial energy in the system and decay heat). Later on, however, accidents occurring during shutdown were found to be as important as those occurring at full power. Abnormal operational events, postulated accidents and design extension conditions occurring during shutdown operational modes represent a significant contribution to the NPP risk due to the fact, that both preventive and mitigatory capabilities of the plant are partially or fully unavailable. Deactivation of safety features, equipment under maintenance or repair, reduced amount of coolant in some regimes, some instrumentation and measurements switched off or non-functionable; open primary circuit (loss of one barrier); and open containment (loss of another barrier) are the causes of the specific risk of accidents in the shutdown mode.

The core of the paper concentrates on the deterministic thermal-hydraulic (TH) safety analyses of the events starting from the shutdown operating modes of VVER-1000. Number of the performed analyses are long-term analyses specifying time windows for the operator (in situation with reduced availability of safety systems and their automatic actuation). Specification of VVER-1000 shutdown modes accompanied, availability of safety systems, methodology basis for the safety analyses, acceptance criteria, computer codes and their validation, list of scenarios analyzed for the VVER-1000, examples of analyses results, and incorporation of new analyses into Safety Analysis Report (SAR) will be described step by step.