Emergency Response Guidance for Reactor Vessel Pressurized Thermal Shock Events

1986 ◽  
Vol 108 (3) ◽  
pp. 346-351
Author(s):  
W. T. Kaiser ◽  
B. S. Monty
Keyword(s):  
Thermal Shock ◽  
Temperature Limit ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Plant Safety ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Accident Conditions ◽  
Core Cooling ◽  
Emergency Operating

The operational concern of pressurized thermal shock (PTS) can be minimized by proper operator guidance. This paper presents a method for calculating a pressure temperature limit curve for reactor vessel integrity which can be used to identify an ongoing potential PTS event. This method has been developed for use and is applicable to all pressurized water reactors. The curve is used in emergency operating procedures developed to prioritize various plant safety concerns including PTS and core cooling to ensure proper operator action during accident conditions. This paper emphasizes the development of the pressure-temperature limit and how it is used within the emergency operating procedures.

Fast Neutron Reduction Techniques for Plant Lifetime Extension

2010 ◽  
Author(s):  
Choon Sung Yoo ◽  
Byoung Chul Kim ◽  
Tae Je Kwon
Keyword(s):  
Fast Neutron ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Neutron Transport ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Reduction Techniques ◽  
Pressurized Thermal Shock ◽  
Water Reactors ◽  
Inner Vessel

A Pressurized Thermal Shock (PTS) Event is an event or transient in pressurized water reactors (PWRs) causing severe overcooling (thermal shock) concurrent with or followed by significant pressure in the reactor vessel. A PTS concern arises if one of these transients acts in beltline region of a reactor vessel where a reduced fracture resistance exists due to neutron irradiation. Such an event may produce a flaw or cause the propagation of a flaw postulated to exist near the inner vessel wall surface, thereby potentially affecting the integrity of the vessel. In this paper fast neutron flux reduction techniques were implemented to reduce the potential risk of PTS due to the neutron irradiation on the pressure vessel beltline region. And the RTPTS value for the end of life of the plant was projected using the fast neutron fluence obtained by neutron transport calculations according to the various core loading pattern and reduction program possible for the future cycles.

Stochastic Simulation of PWR Vessel Integrity for Pressurized Thermal Shock Conditions

1984 ◽  
Vol 106 (3) ◽  
pp. 223-229 ◽  
Author(s):  
P. S. Jackson ◽  
D. S. Moelling
Keyword(s):  
Thermal Shock ◽  
Stochastic Simulation ◽  
Bayesian Method ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Sampling Errors ◽  
Pressurized Water ◽  
Simulation Methodology ◽  
Pressurized Thermal Shock ◽  
Vessel Integrity

A stochastic simulation methodology is presented for performing probabilistic analyses of Pressurized Water Reactor vessel integrity. Application of the methodology to vessel-specific integrity analyses is described in the context of Pressurized Thermal Shock (PTS) conditions. A Bayesian method is described for developing vessel-specific models of the density of undetected volumetric flaws from ultrasonic inservice inspection results. Uncertainty limits on the probabilistic results due to sampling errors are determined from the results of the stochastic simulation. An example is provided to illustrate the methodology.

Emergency Operating Procedures (EOPs) for the AP1000 Simulator

2006 ◽  
Author(s):  
Yuichi Hayashi ◽  
Gianfranco Saiu ◽  
Richard F. Wright
Keyword(s):  
Pressure Control ◽  
Passive Safety ◽  
Background Information ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Investment Protection ◽  
Plant Safety ◽  
Safety Features ◽  
Core Cooling ◽  
Emergency Operating

The AP1000 is two-loop 1100 MWe advanced pressurized water reactor (PWR) that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 30 years of operating PWR experience. The AP1000 final design certification was approved by the NRC in December, 2005. A total of 34 Emergency Operating Procedures (EOPs) for operation of the AP1000 simulator have been prepared based on the AP1000 Emergency Response Guidelines (ERGs), background information documents and detailed plant information. These include 28 EOPs at power and 6 EOPs during shutdown. The AP1000 ERGs were developed by using the generic ERGs for the low pressure reference PWR plant as a basis. The AP1000 design differences from the reference plant were reviewed and reflected in the process of developing operational steps in each ERG. The provisions of the AP1000 PRA were also reviewed and incorporated into the ERGs. Although the AP1000 design does not require operator actions for the first 72 hours after accidents, the operator actions with both safety-related and nonsafety-related equipment have an important role to mitigate the consequence of accidents. In the event of a steam generator tube rupture (SGTR), although the AP1000 is designed so that no operator actions are required to recover from the event, there are actions that can be taken by the operator to limit the release of radioactive effluents from the ruptured SG. These actions include isolation of the ruptured SG and depressurization of the reactor coolant system (RCS) to terminate primary-to-secondary leakage, restoring reactor coolant inventory to ensure adequate core cooling and plant pressure control. It is expected that these operator actions should be incorporated into the ERG to reduce the fission product release. To support the development of the AP1000 ERGs, several transient and accident analyses were performed. These include analyses for LOCA, post-LOCA cooldown and depressurization, passive safety system termination, SGTR and faulted SG isolation. These analyses results were incorporated into the ERG background information documents. In the event of SGTR, several cases were analyzed, including consideration of operator recovery actions. These cases were modeled using the best-estimate state-of-art RELAP5 code. The analyses results show that operator recovery actions are effective for SGTR to be placed under operator control.

scholarly journals Experimental investigation of PWR accident scenarios at the PKL test facility

2019 ◽  
Vol 137 ◽  
pp. 01016 ◽  
Author(s):  
Rafał Bryk ◽  
Lars Dennhardt ◽  
Simon Schollenberger
Keyword(s):  
Severe Accident ◽  
Test Facility ◽  
Pressurized Water ◽  
Water Reactors ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Core Cooling ◽  
Safety Systems ◽  
Reactor Pressure ◽  
Additional Loss

PKL is the only test facility in Europe that replicates the entire primary side and the most important parts of the secondary side of western-type Pressurized Water Reactors (PWR) in the scale of 1:1 in heights. It is also worldwide the only test facility with 4 identical reactor coolant loops arranged symmetrically around the Reactor Pressure Vessel (RPV) for simulation of nonsymmetrical boundary conditions between the reactor loops. Thermal-hydraulic phenomena observed in PWRs are simulated in the PKL test facility for over 40 years. The analyses carried out in these years encompass a large spectrum of accident scenario simulations and corresponding cool-down procedures. The overall goal of the PKL experiments is to show that under accident conditions - even for extreme and highly unlikely assumptions as additional loss of safety systems - the core cooling can be maintained or re-established by automatic or operator- performed procedures and that a severe accident e.g. a core melt-down can be avoided under all circumstances. Another goal of the tests performed in the PKL facility is the provision of data for validation of thermal-hydraulic system codes. This paper presents recent modifications of the PKL facility, applied in order to adapt the facility to the latest western-type designs currently built in the world. The paper discusses also important results obtained in the last years.

Optimum Channel Inclination for Gas Venting Under Countercurrent Flow Limitations

2006 ◽  
Author(s):  
Y. Liao ◽  
K. Vierow
Keyword(s):  
Cooling System ◽  
Superficial Velocity ◽  
Countercurrent Flow ◽  
Pressurized Water ◽  
Inclination Angles ◽  
Surge Line ◽  
Water Reactors ◽  
Optimum Channel ◽  
Accident Conditions ◽  
Core Cooling

Countercurrent flow limitation (CCFL) in the pressurizer surge line of future Pressurized Water Reactors (PWR) with passive safety systems is an important phenomenon in reactor safety analysis. The pressurizer surge line is typically comprised of several sections with various inclination angles. Under certain accident conditions, countercurrent flow takes place in the surge line with liquid flowing down from the pressurizer and steam flowing up from the hot leg. The steam venting rate as well as the liquid draining rate may affect the Emergency Core Cooling System (ECCS) actuation. The objective herein is to develop a physics-based model for evaluating the effect of inclination angle on CCFL. For a given liquid superficial velocity in the countercurrent flow system of the pressurizer surge line, the gas superficial velocity should be as large as possible at the onset of flooding, so that the steam can vent as fast as possible without inhibiting the pressurizer drain rate. Thus the system could depressurize in a timely manner to initiate the ECCS actuation. As indicated by CCFL experiments, for a given liquid superficial velocity, the gas superficial velocity attains a greatest value at a certain channel inclination, which is defined as the optimum channel inclination. In the present work, an analytical model is proposed to predict the optimum channel inclination under simplified conditions. The model predictions compare favorably with experimental data.

The Impact of an Improved Flaw Model on a Pressurized Thermal Shock Evaluation

2002 ◽  
Author(s):  
T. L. Dickson ◽  
F. A. Simonen
Keyword(s):  
Thermal Shock ◽  
Computational Models ◽  
The United States ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Evaluation Program ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
The Impact ◽  
Flaw Characterization

The current regulations for pressurized thermal shock (PTS) were derived from computational models that were developed in the early-mid 1980s. The computational models utilized in the 1980s conservatively postulated that all fabrication flaws in reactor pressure vessels (RPVs) were inner-surface breaking flaws. It was recognized at that time that flaw-related data had the greatest level of uncertainty of the inputs required for the probabilistic-based PTS evaluations. To reduce this uncertainty, the United States Nuclear Regulatory Commission (USNRC) has in the past few years supported research at Pacific Northwest National Laboratory (PNNL) to perform extensive nondestructive and destructive examination of actual RPV materials. Such measurements have been used to characterize the number, size, and location of flaws in various types of welds and the base metal used to fabricate RPVs. The USNRC initiated a comprehensive project in 1999 to re-evaluate the current PTS regulations. The objective of the PTS Re-evaluation program has been to incorporate advancements and refinements in relevant technologies (associated with the physics of PTS events) that have been developed since the current regulations were derived. There have been significant improvements in the computational models for thermal hydraulics, probabilistic risk assessment (PRA), human reliability analysis (HRA), materials embrittlement effects on fracture toughness, and fracture mechanics methodology. However, the single largest advancement has been the development of a technical basis for the characterization of fabrication-induced flaws. The USNRC PTS-Revaluation program is ongoing and is expected to be completed in 2002. As part of the PTS Re-evaluation program, the updated risk-informed computational methodology as implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, including the improved PNNL flaw characterization, was recently applied to a domestic commercial pressurized water reactor (PWR). The objective of this paper is to apply the same updated computational methodology to the same PWR, except utilizing the 1980s flaw model, to isolate the impact of the improved PNNL flaw characterization on the PTS analysis results. For this particular PWR, the improved PNNL flaw characterization significantly reduced the frequency of RPV failure, i.e., by between one and two orders of magnitude.

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Sign in / Sign up

Export Citation Format

Share Document