ASSESSMENT OF REACTOR VESSEL INTEGRITY (ARVI)

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1 ASSESSMENT OF REACTOR VESSEL INTEGRITY (ARVI) B.Raj. Sehgal a, O. Kymalainen b, J.M. Bonnet c, R. Sairanen d, S. Bhandari e, M. Buerger f, J. Dienstbier g, Z. Techy h, T. Theofanous i a. KTH, Sweden; b. Fortum N.S., Finland; c. CEA, France; d. VTT, Finland; e. FRAMATOME, France; f. USTUTT, Germany; g. NRI, Czech Republic; h. VEIKI, Hungary; i. UCSB, USA SUMMARY The ARVI (Assessment of Reactor Vessel Integrity) cost shared Project involving the participation of 9 organizations from Europe and USA started in January 2000 and concluded in December The main aim of the ARVI project was to resolve the safety issues that remain unresolved for the melt vessel interaction phase of a severe accident. The objectives were to determine, (1) the mode and location of vessel failure for different vessel steels, (2) the role of penetrations in the vessel failure, (3) the fraction of the melt discharged to the containment (4) the effectiveness of the gap cooling, (5) the effectiveness of water ingression cooling, (6) the critical heat flux for external cooling of the lower head, and (7) the effects of melt stratification. In addition, the aim was also to analyze the experiments and develop models for some of the processes and to apply the data and models for the design of in-vessel melt retention (IVMR) accident management scheme. The major experimental project was EC-FOREVER in which data was obtained on melt pool natural convection and lower head creep and rupture. Additional tests were made in which the melt pool was flooded with water to determine the potential for in-vessel coolability. Two other experimental projects were also conducted. One was the COPO experiment, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, on enhancement of CHF for the external cooling of the lower head. The modeling activities focused on supporting the EC-FOREVER experiments. These were exploited to validate the industrial structural codes. Other modeling activities were on the (1) gap cooling model for CHF, (2) simple models for system code. Finally, the methodology and data were applied to design of IVMR severe accident management scheme for VVER-440/213 plants. The major results obtained are described on the test and in the conclusions section of the paper. A. INTRODUCTION The MVI project resolved the issues related to melt jet impingment, melt pool convection, vessel hole ablation and critical heat flux on cooling of the external wall of the reactor pressure vessel (RPV). The ARVI Project is concerned with the later phase of the in-vessel accident progression during which the issues are (a) in-vessel coolability (b) invessel melt retention through cooling the external surface of the vessel and (c) vessel failure. The mechanisms for in-vessel coolability of the convecting melt pool are (i) the gap cooling which has been credited for reducing the temperatures in the hot zone during the TMI-2

2 accident and (b) water ingression in the melt through the creation of inter-connected porosity. In case the melt pool can not be cooled by water flooding from inside, water flooding of the external surface has been proposed as the accident management scheme for AP-600, AP-1000 and SBWR Efforts were made in the ARVI Project to enhance the CHF in the external cooling by directing the flow on the external surface. The location and mode of the creep rupture would determine the discharge of the melt from the vessel to the containment and the subsequent loading on the containment. In this context it would be necessary to determine the role of the penetrations on the failure characteristics. The ARVI Project research focussed on gaining the knowledge needed to resolve the issues listed above. In this context, we should clarify that by resolution, we mean that sufficient knowledge is, or will be gained that predictions and evaluations, for prototypic reactor accidents, can be made within tolerable uncertainties. B. WORK PROGRAMME The whole program was divided into five work packages. They were further distributed into tasks, which were performed by different partners. Major parts of the project are the experiments, but analytical developments are simultaneously followed. WP1 is related to EC-FOREVER experiments. In these experiments hemispherical vessels with heated oxide melt, maintained at high temperature, are ruptured/cooled. The major findings from these experiments are (1) effectiveness of the gap cooling, (2) a multiaxial creep data base for various vessel steels, (3) effect of penetrations, (4) timing, mode and location of lower head failure, and (5) the fraction of melt discharged to containment. These experiments are the first in the World in which data on coupled melt pool convection and creep rupture were obtained at prototypic accident conditions. WP 2 is related to melt pool convection studies in COPO and BALI facilities and the CHF studies in the ULPU facility. COPO is a ½ scale facility in which two stratified melt pool convection experiments were performed. This is to complement the stratified experiments performed in the smaller 1/8 th scale SIMECO facility and to confirm the heat transfer data obtained from that facility. BALI facility was used to study the focusing-effect phenomena using water as a simulant fluid. A 2-D rectangular (4 m long, various heights) test section was employed, which was heated from below to simulate the heat flux from oxidic melt, cooled from the sides and cooled from the top with a plastic heat exchanger to simulate the radiative heat transfer. ULPU is a full-scale slice, full height facility, representing the external two-phase natural circulation flow path of the containment for AP It was modified to channel the external coolant by means of a baffle that is equidistance from the heater surface. The main idea is to achieve higher CHF with an organized flow of coolant. WP3 is concerned with the development and validation of analytical methods to predict creep failure of a PWR vessel. This is achieved by using three different codes or approaches. A coupled thermal hydraulics and creep analysis is performed with the commercial code ANSYS Multiphysics. Failure time location and the creep deformation of the vessel are obtained. A finite element structural code, PASULA, was employed to model the penetrations in the EC-FOREVER 3 and 3B experiments and to model the creep laws for various RPV steels. SYSTUS+ code was validated against the displacement data from the EC-FOREVER experiments for different creep laws.

3 WP4 is related to modeling activities, namely, CHF modeling in gap cooling and application of the in-vessel data and models in the reactor system codes. In WP4.2, TOLBIAC code, which uses 3 fields system of equations in a cylindrical or rectangular geometry, was used to analyze the stratified pool convection tests performed in the COPO facility. In WP4.3, the development and application of simplified, fast running models for corium behavior in the core and lower plenum was pursued, which can be used in a system code like KESS/ATHELET-CD. WP5 is related to the design of the in-vessel melt retention scheme for VVER- 440/213 plant with two codes, namely, MVITA developed at RIT and VESSEL developed in Hungary. The objective is to synthesize the results and methods of MVI and the ARVI Projects to design in-vessel melt retention (IVMR) scheme for a VVER-440 plant. C. Main Achievements The focus of this section will be on the main achievements made after FISA-2001 described in Sehgal et al and 2003a. So, we would not be presenting the work presented earlier. C.1 EC-FOREVER Experiments We decided to simulate the prototypic severe accident scenario of melt pool convection with the accident management action of vessel de-pressurization, i.e., the vessel pressure is maintained at ~25 bar. The vessel was at 1/10 th scale i.e. its outer diameter was ~400 mm and wall thickness was ~15 mm. Some experiments were designed with eight 1/10 th scale Inconel-600 penetrations, which were welded in the lower head of the vessel spanning from from the bottom pole of the vessel. In this program, we decided to use the American reactor steel (previously used by Sandia National Laboratory(SNL) in their LHF and OLHF programs) and the French reactor steel (supplied by FRAMATOME) for manufacturing the lower head of the FOREVER experiments. The German Steel formed the upper cylindrical part of the FOREVER experiments; its strength at high temperature is substantially less than that of the American and French RPV steel. The EC-FOREVER experiments are part of the work in WP1. The first four experiments named as: EC-FOREVER-1, -2, -3, -3B, were reported in the FISA-2001 meeting. These experiments observed (1) time of vessel failure, (2) vessel creep displacements, (3) mode and location of vessel failure and (4) effects of penetrations upon these parameters for a lower head made of French reactor steel. During the period after 2001, the EC-FOREVER-4 vessel failure experiment was performed with a lower head made of American steel. A failure strain of ~13.5% was observed in this experiment which was slightly less than the measured failure strain for the French steel. The failure location for both steels was at ~73 from the bottom of the vessel. The American steel lower head failed after 1.7 hours (equivalent time) of pressurization, which was significantly lower than that observed for French steel. Further, much larger opening (27% of the circumference) was observed for the American steel experiment, which lead to greater melt discharged from the vessel than for the French reactor steel. At the failure site, vessel wall looked like an aerofoil, with thickness reduction of more than 60%, which was significantly higher than that observed using French steel. The failure site in the American steel looked similar to those observed in the LHF and OLHF tests at SNL. EC-FOREVER-5 and 6 experiments dealt with the gap cooling and water ingression cooling mechanism. These two experiments were performed with a heat input of 30 kw in

4 order to ensure that a continuous crust covered the melt spherical surface all the way to the top edge of the hemisphere. Other conditions of the experiments remained the same as in earlier experiments. Because of the lower heat input, average creep rate of ~0.5%/hr -1 was observed. Experiments, under creep, were continued until maximum total strain of 5% was achieved. After that, water was poured over the melt. In EC-FOREVER-5, the heater was kept on for 30 minutes during the water addition, while the heater in EC-FOREVER-6 was switched off during that time. From the measured temperature data and the video observations it was clear that no cooling occurred for the oxide crust and vessel wall. This was confirmed in post-test when the vessel was cut open since no gap was found between the oxide crust and the vessel wall. Cooling of the melt did occur from the upper layer. From the measured data it was deduced that the average heat flux from the upper layer was ~0.867 MW/m 2 during the first 100 sec after the water addition. However, the upward heat flux degrades to values of 50 to 100 kw/m 2 after about 300 sec. Post-test examination showed that top mm of the melt pool were quenched; the rest cooled very slowly (overnight) by conduction. The experiment was analyzed with RELAP-5 since choking was observed in the exit piping. C.2 Analysis of EC-FOREVER Vessel Failure Experiments This topic is covered by WP3. Metallographic tests were performed on the samples of failed vessel material to detect the experiment-induced micro-structural changes. These investigations reveal that creep pores are formed at the highly loaded positions. Creep pores were found on the outside of the vessel but almost none inside the vessel. The number of creep pores increases significantly with strain and they finally coalesce and create a small crack. Qualitative agreement between microscopically observed damage distribution and the calculated damage parameter is found to be quite satisfactory. In addition, no chemical interaction between vessel and oxide material was observed in the post-test specimen. Coupled thermal hydraulic and creep analyses of EC-FOREVER experiments were performed with ANSYS Multiphysics code. Details of the approach have been described in Sehgal et al. 2001, 2003a and in Willschutz et al., Further calculations on EC- FOREVER-3 and EC-FOREVER-3B were performed to investigate the role of the penetration. Since in the experiment, penetrations were not placed axi-symmetrically a 3-D calculation was made. It was found that since the penetrations were not located in the hot zone, they did not effect the lower head failure characteristics. The PASULA code developed in VTT performed both 2-D and 3-D analyses to simulate penetration in the EC-FOREVER-3 and 3B experiments. Its also concludes that the penetration was not the weakest zone for failure since it was not in the high temperature zone. PASULA code was also used for the analysis on EC-FOREVER-4 experiment employing thermal boundary condition obtained from the ANSYS calculation. A reasonable agreement with experimental results was obtained. Additionally, PASULA code also predicted the crack opening area and crack length of the vessel failure, which were found to be in good agreement with the measured data. FRAMATOME conducted post-test analysis of EC-FOREVER-3B using the SYSTUS+ code. Experimentally measured temperature was used for the calculation. Good agreement between the experimental results and the analysis was found, especially in the displacement history and the time of failure. C.3 Development and Validation of The TOLBIAC code The TOLBIAC code is devoted to the simulation of the behavior of corium pool with natural convection within a structure, which may be the pressure vessel or a core catcher.

5 Its main characteristics are the use of a 3-fields equation system, in a 2-D cylindrical or rectangular geometry. The 3-fields are the metal components, the oxide components and the gas. From a reference calculation, on COPO it was found that there were significant difference between the calculated results (i.e. both heat flux and pool temperature) and the data obtained in WP2 on the COPO facility. So the influence of some parameters on the results was studied. It was found that the paraffin oil solidification temperature and the paraffin oil thermal conductivity play significant roles in determining the heat flux and pool temperature. The correlations giving the best results are the one of Kelkar and Patankar (1993), or Mayinger et al.(1976) for the water layer, the correlation of Fishenden and Saunders (1950) for horizontal exchange in the paraffin layer and the correlation of Churchill and Chu (1975) at the lateral wall of the paraffin layer. However, even using this set of correlations did not produce agreement for all test regimes. C.4 Application of MVI and ARVI data in a system code The work in C.4 and C.5 is covered by WP4. The coolability of a particulate debris bed in the core and lower head region were investigated with WABE-2D code. This code models the coolant flows in the porous medium. It was found that multi-dimensional effects are important and enhance the potential for coolability greatly e.g. through flows which can cool dense regions in a porous bed. Effects of gap were evaluated and it was found that gap flows can spread horizontally and are effective in cooling the bed. C. 5 Preliminary Design of IVMR SAM Scheme for VVER 440 NPP In WP5 thermal loads (imposed by core melt) on the external vessel wall were estimated by VESSEL code (using simplified 2-D steady state conduction model) and MVITA code (developed by KTH) modeling transient melt pool convection. Core melt consists of two parts; one is heavier oxide bottom layer and other is metal layer. Thickness of the metal layer is varied in the calculations. The VESSEL code results suggest that for metal layer thickness (~0.36 m), heat flux of external vessel wall remains always below CHF limit. However, for a metal layer of 0.1 m thickness, heat flux exceeds the CHF. Since, the VESSEL code employ simple model, the same cases were studied with MVITA code. The MVITA code predictions for both 0-36 m and 0.1 m layer are below the CHF. D. DISSEMINATION AND EXPLOITATION OF RESULTS The lower head creep failure results achieved in the EC-FOREVER 1, -2, -3, and 3B experiments were disseminated in several scientific meetings, e.g. ICONE, SMIRT, NURETH to the nuclear scientific community. The results achieved in the EC-FOREVER 4 experiments employing the American steel were disseminated to the participants of the OECD Project OLHF, and presented at the 2002 ICONE meeting. The results obtained in the EC-5 and EC-6 experiments are very significant for the issue of in-vessel melt coolability. A paper on those results has been submitted to Nuclear Engineering and Design. E. CONCLUSIONS ARVI project was successfully concluded at the end of Dec The work performed is described in Sehgal et al. 2003b.The project met all its deliverables and achieved very significant results towards the resolution of important issues in the late phase of in-vessel accident progression. It was found that the bounding scenario of the lower head

6 full of convecting melt pool experienced creep rupture. The location of failure was found to be ~70 0 above the bottom pole of the vessel. The mode of failure was a crack which extended azimuthally from ~62 0 (in French Steel) to ~95 0 (in American steel). The opening area varied with different steels. Not all the melt discharged from the vessel, even at pressure of 25 bars. This implies that the containment loading will be substantially reduced and the issues e.g. ex-vessel steam explosion, jet induced concrete ablation etc. may have to be re-evaluated. The EC-FOREVER test data has been used to benchmark the creep laws for the French and American steels. The analyses performed with the codes ANSYS Multiphysics, PASULA and SYSTUS and their agreement with the measured data lends confidence to the predictions made with the commercial and the industry codes for the vessel failure time and location. It is still not clear why the area of failure is different for different steels. Perhaps it depends on the very small content materials in the vessel steels. The Issue of in-vessel coolability by flooding of the melt pool was investigated in the tests EC-5 and 6. The experiments were conducted to provide conditions favouring the establishment of a gap, before water was added. However, no gap cooling was observed. It was also found that the water ingression quenching is limited to a layer of 6-7 cm in the melt pool, which is not sufficient to declare that in-vessel coolability can be accomplished for the melt pool in the bounding scenario. The effects of melt pool stratification were found to be very significant in terms of the thermal loading and the focussing effect. We believe this issue is still open, since the stratification configuration, which depends on the melt pool chemistry is not fully understood yet. It is being investigated in the OECD, MASCA Project. The ULPU experiments performed in WP2 have found that by organising the flow around the external surface of the lower head the CHF could be enhanced by up to 20 %. This is a substantial improvement, which would help to provide the vessel external cooling accident management scheme for reactors of >1000 MWe power. The IVMR scheme for the VVER-440 plants seems to feasible except for greater assurance that margins are adequate to accommodate the greater focussing effect that may be present for thin metal layers. REFERENCES Sehgal, B.R. et al., 2001 Assessment of Reactor Vessel Integrity, FISA-2001, EU Reasearch in Reactor Safety. Sehgal, B.R. 2003b, Assessment of Reactor Vessel Integrity, Final Report. Sehgal, B.R. et al., 2003a Assessment of Reactor Vessel Integrity(ARVI), Nucl. Engg. Des., 2753, Willschutz, H.G., Alstadt, E., Weiss, F.P., Sehgal, B.R., 2001 Coupled thermal structural analysis of LWR vessel creep failure experiments, Nucl. Engg. Des., 208,