Reactor Coolant System Components Research Articles

Analysis of depressurization ability for IRIS containment during SBLOCAs using RELAP5 and CONTEMPT

As an integral and modular pressurized water reactor (PWR), the international reactor innovative and secure (IRIS) eliminates the threat of large break LOCAs (LBLOCAs) by integrating all major reactor coolant system components into the reactor vessel and provides compact steel containment with a suppression pool. Unlike large reactor containment, IRIS containment has modest volume and compact internal structure. The analysis of traditional dry containment cannot be applied to the IRIS reactor. Despite the disappearance of LBLOCAs (exceeding 4 in), small steel containment still faces the threat of high–frequency small break LOCAs (SBLOCAs) (e.g., the largest possible 4 in chemical and volume control system (CVCS) line break and the lowest elevation 2 in direct vessel injection (DVI) line break). To identify the impact of these characteristics on the integrity of IRIS reactor containment, a safety assessment framework that combines deterministic and probabilistic methods is proposed for the pressure response of containment under SBLOCAs of IRIS reactor. First in the modeling phase, primary and secondary loop models have been built considering the design of IRIS, and accident consequences under typical SBLOCAs were evaluated using RELAP5/MODE4. A simplified model of containment was presented herein based on the primary structure of IRIS containment. Second in the input parameters identification and uncertainty quantification stage, the responses of containment were simulated using CONTEMPT–LT. Thus, the main factors affecting the depressurization capacity of the containment were identified, and the reasonable probability density distributions were determined. Third, in the uncertainty propagation stage, Monte Carlo method was used to map the uncertainty of input parameters to the output variable. Fourth in the sensitivity analysis and uncertainty analysis phase, the contribution of each input parameter was measured using sensitivity analysis, and their interactions were investigated. The range of deviations of the containment pressure response was measured using uncertainty analysis, and finally, the calculated results were combined with the containment fragility curve to evaluate containment failure probability.

Data-Driven Maintenance Priority and Resilience Evaluation of Performance Loss in a Main Coolant System

The main coolant system (MCS) plays a vital role in the stability and reliability of a nuclear power plant. However, human errors and natural disasters may cause some reactor coolant system components to fail, resulting in severe consequences such as nuclear leakage. Therefore, it is crucial to perform a resilience analysis of the MCS, to effectively reduce and prevent losses. In this paper, a resilience importance measure (RIM) for performance loss is proposed to evaluate the performance of the MCS. Specifically, a loss importance measure (LIM) is first proposed to indicate the component maintenance priority of the MCS under different failure conditions. Based on the LIM, RIMs for single component failure and multiple component failures were developed to measure the recovery efficiency of the system performance. Finally, a case study was conducted to demonstrate the proposed resilience measure for system reliability. Results provide a valuable reference for increasing the system security of the MCS and choosing the appropriate total maintenance cost.

Methods for failure assessment of NPP pressure boundary under severe accident loading

During postulated high-pressure core melt accidents, high temperatures can be reached by the reactor coolant system (RCS) pressure boundary due to hot gases from the core melt process, which may lead to failure in terms of leak or break and subsequent depressurization. Location, time and size of the failure can have large implications on the further accident progression. Nevertheless, not for all possibly affected components and failure modes suitable integrity assessment methods, in-depth structure mechanical analyses or in some cases even simple models for the integration into severe accident codes are available. Investigations on the events in Fukushima Daiichi emphasize the importance of methods for RCS failure assessment within severe accident analysis.This work concentrates on the assessment of gross boundary failure, e.g. of the RCS piping, from a mechanical point of view. First, failure modes of RCS components as well as the deformation and fracture behavior of typical RCS materials in the relevant loading regime are discussed. Established assessment methods based on linear damage accumulation and the use of Finite Element (FE) codes as well as a newly developed method based on a zero-dimensional approach are presented. The methods are compared among each other and to a large-scale experiment with a DN700 pipe and international benchmark analyses within the OECD/NEA Components and Structures under Severe Accident Loading (COSSAL) activity. Results of the application to an exemplary severe accident scenario in a PWR and corresponding parametric studies are summarized. An outlook is given on the analysis of local boundary failure of components, potentially responsible for small early leaks.

A new direction in PWR simplification

A new approach to PWR simplification is presented, in which a compact Reactor Coolant System (RCS) configuration is introduced, particularly suited for a power level in the range of 600 MWe. Customary PWR primary system components are eliminated to achieve this RCS simplification. For example, RCS pressure control through a “self-pressurization” mode, with core exit at saturation temperature with less than 1% steam, allows elimination of a pressurizer. Also, mechanical control rods are replaced by reactivity control using negative moderator void and temperature coefficient together with variable speed primary pumps, and with an upgrade in the safety boration function. Decay heat removal in shutdown conditions is realized through the secondary side rather than through primary side equipment. The compact RCS can be installed in a small volume, high-pressure containment. The containment is divided into two leak-tight zones separated by a partition plate. Safety equipment installed in one of the two zones will be protected against adverse ambient conditions from leaks or breaks in the other zone. The partition facilitates management of coolant inventory within the RCS and the containment following RCS leaks or breaks. In particular, the safety injection system as commonly known, consisting of accumulators and multiple stages of injection pumps can be discarded and replaced by gravity-driven flooding tanks. Space available around major RCS components is adequate to avoid compromising accessibility during maintenance or in-service inspection operations. In addition, the two-zone, high-pressure containment provides extra margins in severe accident mitigation. Finally, the proposed containment has a much smaller size than customary large dry containments in PWR practice and it can be anticipated that Nuclear Island building size will similarly be reduced.

A methodology of risk assessment, management, and coping actions for nuclear power plant (NPP) hit by high-explosive warheads

A detailed Loss of Coolant Accident (LOCA) analysis in an AP1000 NPP was performed, followed by a definition of the vulnerability analysis principles, and analysis of blast loads and fragments impact created by a nearby explosion. The AP1000 NPP performs excellently to small-break LOCA due to in-structure shock, with the 10 CFR 50.46 Acceptance Criteria fully accomplished. Impulsive dynamic loads resulting from blast waves and fragments impact of GBU-28 (Guided Bomb Unit) were considered for a nearby explosion. We model the structure and the main reactor components using the MSC/Dytran code to obtain accurate internal acceleration levels at critical points. We account for the appropriate blast wave interaction with the soil and the soil interaction with the containment structure, rather than using empirical formulas. The model includes the shielding structure with its concrete base, the support structures for the reactor, the steam generators, and the pressurizer. The combined effect of bomb fragmentation and blast loading was also considered using a cylindrical fragmentation model and the blast model of Kingery-Bulmash, assuming a hemispherical charge. A comprehensive risk assessment methodology composed of four phases was developed. The methodology is comprised of: (I) System analysis, (II) Hazard analysis, (III) Damage assessment, and (IV) Risk analysis of the in-structure shock consequences. Using seismic fragility curves for analysis of the expected failure modes according to explosion events faced difficulties since no published data was found. Adjustments to these fragility curves were made using median acceleration limits on components designed to withstand airplane crash, together with standard deviations taken from the given earthquake fragility tables. The findings reveal that the probabilities of failure of the reactor coolant system components resulting from a GBU-28 nearby hit, namely the pressurizer, the cooling pumps, and valves are quite high (greater than 1∙10−4).

Extension of a Level 2 PSA Event Tree Based on Results of a Probabilistic Dynamic Safety Analysis of Induced Steam Generator Tube Rupture

This paper presents the approach of extending a classical generic event tree (ET) of a Level 2 Probabilistic Safety Analysis to the results of a probabilistic dynamic safety analysis. The example of creep-induced steam generator tube rupture has been chosen. The results of an Analysis of Thermal Hydraulics of Leaks and Transients with Core Degradation (ATHLET-CD)/Monte Carlo Dynamic Event Tree (MCDET) simulation analyzing the failure of reactor coolant system components by creeping in a scenario of a high-pressure core melt accident in a generic pressurized water reactor (PWR) have been implemented in the ET. From the results of these simulations, a set of parameters has been extracted and integrated into the ET along with their probability distributions. The effect of these parameters on both the progression of the severe accident sequence under consideration and the release categories of a generic German PWR plant, respectively, is discussed in this paper.

Integral PWR-Type Small Modular Reactor Developmental Status, Design Characteristics and Passive Features: A Review

In recent years, the trend in small modular reactor (SMR) technology development has been towards the water-cooled integral pressurized water reactor (iPWR) type. The innovative and unique characteristics of iPWR-type SMRs provide an enhanced safety margin, and thus offer the potential to expand the use of safe, clean, and reliable nuclear energy to a broad range of energy applications. Currently in the world, there are about eleven (11) iPWR-type SMRs concepts and designs that are in various phases of development: under construction, licensed or in the licensing review process, the development phase, and conceptual design phase. Lack of national and/or internatonal comparative framework for safety in SMR design, as well as the proprietary nature of designs introduces non-uniformity and uncertainties in regulatory review. That said, the major primary reactor coolant system components, such as the steam generator (SG), pressurizer (PRZ), and control rod drive mechanism (CRDM) are integrated within the reactor pressure vessel (RPV) to inherently eliminate or minimize potential accident initiators, such as LB-loss of coolant accidents (LOCAs). This paper presents the design status, innovative features and characteristics of iPWR-type SMRs. We delineate the common technology trends, and highlight the key features of each design. These reactor concepts exploit natural physical laws such as gravity to achieve the safety functions with high level of margin and reliability. In fact, many SMR designs employ passive safety systems (PSS) to meet the evolving stringent regulatory requirements, and the extended consideration for severe accidents. A generic classification of PSS is provided. We constrain our discussion to the decay heat removal system, safety injection system, reactor depressurization system, and containment system. A review and comparative assessment of these passive features in each iPWR-type SMR design is considered, and we underline how it maybe more advantageous to employ passive systems in SMRs in contrast to conventional reactor designs.

An analysis on the consequences of a severe accident initiated steam generator tube rupture

An analysis on the thermal hydraulic behavior, core damage progression and radiological releases of a steam generator tube rupture event initiated by a station black out event for APR1400 is provided in this paper. The analysis was performed by using MELCOR, where counter-current natural circulation, mixing in the steam generator inlet plenum, creep ruptures of reactor coolant system components and U-tube, and mitigation actions by the operator are modeled. A station black out accident with an opening of atmospheric steam dump valves (ADVs) at the time of severe accident management strategy entry condition with a failure of a supply of auxiliary feed water to the steam generators selected as a base case resulted in a creep rupture of U-tubes with a direct release of radioactive materials into the atmosphere. It was shown that a mitigation action by the operator for an aggressive cooldown of the reactor coolant system using ADVs and supplying the water to the steam generator even after the creep rupture of U-tubes was effective. It arrested core melt progression and significantly reduced the amount of radiological releases to the atmosphere.

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences. The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA. A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach. This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR, which is the representative case for high pressure sequences. MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident. In addition, a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty. The probability for steam generator tube creep rupture is estimated at 0.17.