scholarly journals Qualification of TRACE V5.0 Code against Fast Cooldown Transient in the PKL-III Integral Test Facility

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
Eugenio Coscarelli ◽  
Alessandro Del Nevo ◽  
Francesco D'Auria
Keyword(s):  
Heat Transfer ◽  
Test Facility ◽  
Steam Generators ◽  
Integral Test ◽  
Pressurized Thermal Shock ◽  
Vessel Integrity ◽  
Main Steam Line ◽  
Main Steam ◽  
Secondary Side ◽  
Reactor Pressure

The present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry caused by an increase of the heat transfer from the primary to the secondary side that induces a fast cooldown transient on the primary side-affected loop. The asymmetric overcooling effect requires an assessment of the reactor pressure vessel integrity considering PTS (pressurized thermal shock) and an assessment of potential recriticality following entrainment of colder water into the core area. The aim of this work is the qualification of the heat transfer capabilities of the TRACE code from primary to secondary side in the intact and affected steam generators (SGs) during the rapid depressurization and the boiloff in the affected SG against experimental data.

scholarly journals Integral Test Facility PKL: Experimental PWR Accident Investigation

2012 ◽  
Vol 2012 ◽  
pp. 1-16 ◽  
Author(s):  
Klaus Umminger ◽  
Lars Dennhardt ◽  
Simon Schollenberger ◽  
Bernhard Schoen
Keyword(s):  
Systematic Investigation ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Integral Test ◽  
Pressurized Water ◽  
Height Scale ◽  
Water Reactors ◽  
Accident Conditions ◽  
Secondary Side ◽  
Reactor Pressure

Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR) at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break1

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multidimensional numerical analysis of the transient thermal-hydraulic response of a steam generator (SG) secondary side to a double-ended guillotine break of the main steam line attached to the SG at a pressurized water reactor (PWR) plant. A simplified analysis model is designed to include both the SG upper space, which the steam occupies and a part of the main steam line between the SG outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport (SST) turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break (MSLB) accident, a constant amount of steam is assumed to be generated from the bottom of the SG upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present computational fluid dynamics (CFD) model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2–8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

Effects of Cladding on Reactor Pressure Vessel Integrity Based on the Revised PTS Rule (10 CFR 50.61)

2009 ◽  
Author(s):  
Vikram Marthandam ◽  
Timothy J. Griesbach ◽  
Jack Spanner
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Historical Perspective ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Potential Benefits ◽  
Vessel Integrity ◽  
Technical Basis ◽  
Reactor Pressure

This paper provides a historical perspective of the effects of cladding and the analyses techniques used to evaluate the integrity of an RPV subjected to pressurized thermal shock (PTS) transients. A summary of the specific requirements of the draft revised PTS rule (10 CFR 50.61) and the role of cladding in the evaluation of the RPV integrity under the revised PTS Rule are discussed in detail. The technical basis for the revision of the PTS Rule is based on two main criteria: (1) NDE requirements and (2) Calculation of RTMAX-X and ΔT30. NDE requirements of the Rule include performing volumetric inspections using procedures, equipment and personnel qualified under ASME Section XI, Appendix VIII. The flaw density limits specified in the new Rule are more restrictive than those stipulated by Section XI of the ASME Code. The licensee is required to demonstrate by performing analysis based on the flaw size and density inputs that the through wall cracking frequency does not exceed 1E−6 per reactor year. Based on the understanding of the requirements of the revised PTS Rule, there may be an increase in the effort needed by the utility to meet these requirements. The potential benefits of the Rule for extending vessel life may be very large, but there are also some risks in using the Rule if flaws are detected in or near the cladding. This paper summarizes the potential impacts on operating plants that choose to request relief from existing PTS Rules by implementing the new PTS Rule.

Structural Assessment of Main Steam Line and Containment Building Under Steam Explosion Conditions

2015 ◽  
Author(s):  
Seung-Huyn Kim ◽  
Yoon-Suk Chang ◽  
Sungchu Song ◽  
Yong-Jin Cho
Keyword(s):  
Steam Explosion ◽  
Failure Criteria ◽  
Steam Line ◽  
Dynamic Loads ◽  
Steam Generators ◽  
Structural Assessment ◽  
Concrete Failure ◽  
Nuclear Component ◽  
Main Steam Line ◽  
Main Steam

The steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is quite intense and rapid. This phenomenon may threat integrity of nuclear components and structures. For instance, the dynamic loads on the reactor cavity and the reactor lower plenum could potentially lead to failure of the main steam lines connected to the steam generators. In addition, since the main steam line extends to the containment wall, failure of the containment building may occur. The object of present study is to examine characteristics of Main Steam Line (MSL) and containment building under the steam explosion conditions. In this context, the corresponding FE models were generated and previously determined displacements of penetration piping were used as loading conditions. Subsequent FE analyses were conducted for the main steam line and containment building to calculate stresses and crack evaluations. Finally, structural assessment of nuclear component and structure combined with concrete failure criteria was performed and their results were discussed.

Power Rising and Descending Transient for the OTSG of a Small PWR

2020 ◽  
Author(s):  
Baihui Jiang ◽  
Zhiwei Zhou ◽  
Zhaoyang Xia ◽  
Qian Sun
Keyword(s):  
Heat Transfer ◽  
Flow Rate ◽  
Two Phase Flow ◽  
Flow Instability ◽  
Flow Stability ◽  
Phase Flow ◽  
Steam Generators ◽  
Two Phase ◽  
Water Reactors ◽  
Secondary Side

Abstract As key heat transfer system in small and medium size pressurized water reactors, once-through steam generators are important parts of energy exchange between primary and secondary circuits, and are very important for the design and operation of reactors. However, two-phase flow and heat transfer in once-through steam generators are very complicated. When a reactor experience power rising and descending transient, the heat removal of once-through steam generator, the flow rate, the inlet fluid temperature and outlet steam temperature will all change accordingly. Especially when a reactor is running at a low power, the flow rate of the secondary side of OTSG is extremely small and the single-phase region of the secondary side of OTSGs is also too small. The two-phase flow instability may occur, which has a serious impact on reactor operation and safety. So, a reasonable power-up and power-down transient scheme is required to ensure operational stability when starting up and shutting down a reactor. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LCC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. Scholars all over the world have carried out a large number of analysis of two-phase flow stability using RELAP5, and the results are reliable. This paper takes once through steam generators with given structural parameters as the research object, and uses RELAP5 as the calculation tool. The influencing factors of flow instability are discussed in this paper, and the operating parameters of the fluid on the primary and secondary sides are designed to satisfy the flow stability under different powers. And a set of power-up and power-down schemes for stable operation is proposed.

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