Integral Test Facility PKL: Experimental PWR Accident Investigation

Umminger, K.; Dennhardt, Lars; Schollenberger, Simon; Schoen, B.

doi:10.1155/2012/891056

Cited by 27 publications

(12 citation statements)

References 2 publications

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“…This scenario formed a main target of the OECD/PKL-2 program [11] that began in 2008. In the frames of the program the PKL III G tests campaign was conducted which additionally covered the topics indicated in the figure 1 [5].…”

Section: Pkl Test Programsmentioning

confidence: 99%

Experimental investigation of PWR accident scenarios at the PKL test facility

2019

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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.

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Section: Pkl Test Programsmentioning

confidence: 99%

Experimental investigation of PWR accident scenarios at the PKL test facility

2019

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“…Thus, ROSA/LSTF is a scaled model of a Westinghouse and has Tsuruga-2 as reference plant and PKL is a Konvoi type and has as reference plant Phillisburg-2 (Nakamura et al, 2009;Umminger et al, 2011). Both designs present technological differences and also different scale ratios between each technology and their plant of reference, so a brief description of both installations is presented.…”

Section: Experimental Facilitiesmentioning

confidence: 99%

“…To achieve this objective, different experiments are performed in the Rig-of-Safety Large Scale Test Facility (ROSA/LSTF) located in Japan, which represents a PWR Westinghouse design. The second programme was mainly focused on investigating safety issues relevant and complex heat transfer mechanisms in current pressurized water reactor PWR plants as well as for new PWR design concepts (Nakamura et al, 2009;Umminger et al, 2011;Jonnet et al, 2013). In this programme, the experiments were undertaken at Primarkreislauf Versuchsanlage facility (PKL) in Germany, which is a scaled model of a Konvoi PWR design.…”

Section: Introductionmentioning

confidence: 99%

Post-test analysis of the ROSA/LSTF and PKL counterpart test

Alberola

Querol

Gallardo

et al. 2016

Nuclear Engineering and Design

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h i g h l i g h t s• TRACE modelization for PKL and ROSA/LSTF installations.• Secondary-side depressurization as accident management action.• CET vs PCT relation.• Analysis of differences in the vessel models. a b s t r a c t Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety.Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations.

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“…Some researchers [7,8] have analyzed loss of primary coolant and SBO events in PWRs employing severe accident computer codes, but under the condition modeling for the pressure vessel internals except the core and the downcomer was more simplified than that in the best-estimate computer codes. Meanwhile, there have been some experimental researches on the PWR SBO scenarios with AM measures by using such integral test facilities as the PSB for VVER in Russia [9] and the PKL for Vorkonvoi-type PWR in Germany [10]. The experimental data, however, would be insufficient to clarify the AM measures effectiveness due to such atypical features as small volume and low pressure in the primary system.…”

Section: Introductionmentioning

confidence: 99%

ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

Takeda

Ohtsu

2018

Science and Technology of Nuclear Installations

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Three tests were carried out with the ROSA/LSTF (rig of safety assessment/large-scale test facility), which simulated accident management (AM) measures during station blackout transient with loss of primary coolant under assumptions of nitrogen gas inflow and total failure of high-pressure injection system in a pressurized water reactor. As the AM measures, steam generator (SG) secondary-side depressurization was done by fully opening the relief valves in both SGs, and auxiliary feedwater was injected into the secondary-side of both SGs simultaneously. Conditions for the break size and the onset timing of the AM measures were different among the three LSTF tests. In the three LSTF tests, the primary pressure decreased to a certain low pressure of below 1 MPa with or without the primary depressurization by fully opening the relief valve in a pressurizer as an optional AM measure, while no core uncovery took place through the whole transient. Nonuniform flow behaviors were observed in the SG U-tubes under natural circulation (NC) with nitrogen gas depending probably on the gas accumulation rate in the two LSTF tests that the gas accumulated remarkably. The RELAP5/MOD3.3 code predicted most of the overall trends of the major thermal hydraulic responses observed in the three LSTF tests. The code, however, indicated remaining problems in the predictions of the primary pressure, the SG U-tube collapsed liquid levels, and the NC mass flow rate after the nitrogen gas ingress as well as the accumulator flow rate through the analyses for the two LSTF tests, where the remarkable gas accumulation occurred.

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Integral Test Facility PKL: Experimental PWR Accident Investigation

Cited by 27 publications

References 2 publications

Experimental investigation of PWR accident scenarios at the PKL test facility

Experimental investigation of PWR accident scenarios at the PKL test facility

Post-test analysis of the ROSA/LSTF and PKL counterpart test

ROSA/LSTF Tests and Posttest Analyses by RELAP5 Code for Accident Management Measures during PWR Station Blackout Transient with Loss of Primary Coolant and Gas Inflow

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