Structural integrity assessment of aging nuclear reactor pressure vessels

Irradiation embrittlement reduces both the cleavage fracture toughness and the ductile tearing toughness of reactor pressure vessel (RPV) steels. Extensive research programs have investigated the fracture behavior of heavy-section vessels containing flaws. Information obtained from that research has been used to develop regulatory guidance for evaluating the structural integrity of irradiated RPVs. Additional research programs have developed fracture analysis methods, and generated the data required for their implementation. Regulatory guidance employs fracture analysis technology to assure that adequate fracture-prevention margins for RPVs are maintained throughout the licensed operating period of nuclear power plants.     

Numerical computation of a reactor vessel melting under external cooling conditions

Numerical computations are performed for melting and natural convection in the liquefied region of a reactor vessel under external cooling to find more thermal margin for in-vessel retention. Existing typical experiment and calculations for gallium melting are used for the validation. The transient flow field in the liquefied region and the melt front movement analyzed are compared with those from finite-element and finite-volume methods. Reasonable agreements are achieved with respect to melt progression and flow configuration in the liquefied zone. A three-dimensional geometrical model for an azimuthally 3° angular section of the APR1400 pressurized water reactor vessel is prepared based on this verification, and a conservative heat flux profile from the corium inside with a concentrated heat flux from the metallic layer of 2.1 MW/m², which is greater than maximum critical heat flux, is applied to the vessel model assuming constant exterior temperatures of 400 and 1000 K. The results show that even though the vessel inside heat flux is much greater than the critical heat flux, this does not intensively melt a vessel due to combined effects of latent heat absorption during the melting and the remaining heat spreading through the entire vessel.     

CHF enhancement by vessel coating for external reactor vessel cooling

In-vessel retention (IVR) is a key severe accident management (SAM) strategy that has been adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). One viable means for IVR is the method of external reactor vessel cooling (ERVC) by flooding the reactor cavity during a severe accident. As part of a joint Korean–United States International Nuclear Energy Research Initiative (K-INERI), an experimental study has been conducted to investigate the viability of using an appropriate vessel coating to enhance the critical heat flux (CHF) limits during ERVC. Toward this end, transient quenching and steady-state boiling experiments were performed in the subscale boundary layer boiling (SBLB) facility at the Pennsylvania State University using test vessels with micro-porous aluminum coatings. Local boiling curves and CHF limits were obtained in these experiments. When compared to the corresponding data without coatings, substantial enhancement in the local CHF limits for the case with surface coatings was observed. Results of the steady-state boiling experiments showed that micro-porous aluminum coatings were very durable. Even after many cycles of steady-state boiling, the vessel coatings remained rather intact, with no apparent changes in color or structure. Moreover, the heat transfer performance of the coatings was found to be highly desirable with an appreciable CHF enhancement in all locations on the vessel outer surface but with very little effect of aging.     

严重事故条件下压力容器完整性评价的研究进展

堆芯熔融物堆内滞留(In-Vessel Retention,IVR)是以AP1000为代表的第三代轻水反应堆严重事故管理的重要策略之一,也是严重事故条件下保证压力容器完整性(Reactor Vessel Integrity,RVI)的典型方法之一.该文综述了国外在严重事故条件下压力容器完整性试验研究和理论分析的现状,总...     

Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal–hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.     

Experience of the integrity assessment of the highly embrittled reactor pressure vessel for the life extension

The reactor pressure vessel (RPV) is the key component of pressurized water reactor. It has to comply with various rules and regulatory guides to ensure sufficient safety and operating margins during the plant lifetime. Thus, it is crucial to assure the integrity of RPV for an effective plant lifetime management program. In this paper, the status and the experiences of various integrity issues of the highly embrittled RPV are introduced. A circumferential weld in the beltline region of the Kori Unit 1 RPV was projected to be unable to satisfy the minimum upper-shelf energy requirement and the reference temperature-pressurized thermal shock requirement before the end of 40-year design lifetime. The detailed integrity assessments had been performed to resolve both issues and the results summarized. In addition several actions have been taken as aging management programs to assure the integrity of Kori Unit 1 RPV during the extended operation. Details of the activities such as, redefining initial reference temperature-nil ductility transition temperature, installing ex-vessel dosimetry, and withdrawal and testing of additional surveillance capsule are explained. Finally, the applicability of these and other activities including thermal annealing to mitigate the effects of the irradiation embrittlement are evaluated.     

An assessment methodology for in-vessel corium retention by external reactor vessel cooling during severe accidents in PWRs

A consistent probabilistic approach is proposed to evaluate the feasibility of in-vessel retention of the molten corium through external reactor vessel cooling (IVR-ERVC) during severe accidents of pressurized water reactors (PWRs). By combining the results of Level-1 probabilistic safety assessment, a critical heat flux correlation, and wall heat flux distributions calculated by a severe accident code with appropriate adjustment, we can reasonably predict the overall success probability of the IVR-ERVC from the viewpoint of thermal failure. The practicability of the proposed approach is illustrated with a preliminary application to the Korean Standard Nuclear Power Plant. This paper also discusses future developmental needs for more reliable assessment.     

External cooling of a reactor vessel under severe accident conditions

The TMI-2 accident demonstrated that a significant quantity of molten core debris could drain into the lower plenum during a severe accident. For such conditions, the Individual Plant Examinations (IPEs) and severe accident management evaluations, consider the possibility that water could not be injected to the RCS. However, depending on the plant specific configuration and the accident sequence, water may be accumulated within the containment sufficient to submerge the lower head and part of the reactor vessel cylinder. This could provide external cooling of the RPV to prevent failure of the lower head and discharge of core debris into the containment.This paper evaluates the heat removal capabilities for external cooling of an insulated RPV in terms of (a) the water inflow through the insulation, (b) the two-phase heat removal in the gap between the insulation and the vessel and (c) the flow of steam through the insulation. These results show no significant limitation to heat removal from the bottom of the reactor vessel other than thermal conduction through the reactor vessel wall. Hence, external cooling is a possible means of preventing core debris from failing the reactor, which if successful, would eliminate the considerations of ex-vessel steam explosions, debris coolability, etc. and their uncertainties. Therefore, external cooling should be a major consideration in accident management evaluations and decision-making for current plants, as well as a possible design consideration for future plants.     

Flow analyses using RELAP5/MOD3 code for OPR1000 under the external reactor vessel cooling

External reactor vessel cooling (ERVC) is considered as one of the most promising severe accident management strategies for an in-vessel corium retention (IVR). Heat removal capacity and water availability at the vessel outer surface can be key factors determining the success of ERVC measures. In this study, for the investigations on the effect of water availability in case of ERVC, flow analyses using the RELAP5/MOD3 code were performed. The analyses were focused to examine the flow behavior inside the reactor pressure vessel (RPV) insulator of the OPR1000 (Optimized Power Reactor 1000 MWe) under a cavity flooding. The current flow analyses results show that for the accident scenarios of station black out (SBO) and 9.6 in. large break loss of coolant accident (LBLOCA) under the ERVC, steam could not ventilate through the insulator and the pressure inside the RPV insulator increased abruptly. This induced a water sweep out and steam domination in the flow path inside the insulator. These flow analyses results indicate that sufficient water ingression and steam venting through the insulator can be a key factor determining the success of the ERVC in the operating nuclear power plant, OPR1000. According to the results of the sensitivity studies for the venting area, in terms of an effective flow circulation inside the insulator, an optimal venting is to assign four holes having a diameter of 0.3 m at the upper exit (hot leg level) of the insulator. And the effect of the inlet flow area at the insulator bottom is rather minor when compared to that of the outlet flow area of a steam venting.     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

Reactor pressure vessel integrity program

The reactor vessel report provides a technical evaluation of the effects of aging on the reactor vessel shell, lower vessel head, closure head, nozzles (and safe ends, if provided), interior attachments, and all associated pressure-retaining bolting. The evaluation is performed in accordance with the requirements of the license renewal rule, 10 CFR Part 54. This paper is based upon the B and W Owners Group report BAW-2251 (Mackay, W.H., Gross, L.B., Rinckel, M.A., Starkey, R.L., 1996. Demonstration of the management of aging effects for the reactor vessel. Report BAW-2251, Framatome Technologies, Lynchburg, VA), which was submitted to the NRC for review and approval in June 1996. The reactor vessel report demonstrates that existing utility programs adequately monitor and manage the relevant effects of aging of the reactor vessel such as cracking, loss of material, reduction in the fracture toughness of the vessel beltline material, and loss of mechanical closure integrity. The aging management programs credited include ASME Section XI, ISI, compliance to 10 CFR Parts 50.60 and 50.61, Plant Technical Specifications, and commitments to generic Nuclear Regulatory Commission communications. The aging management techniques described in this report permit continued safe operation over the period of extended operation associated with license renewal. In addition, the report includes the evaluation of three generic time-limited aging analyses: (1) fatigue of metallic components; (2) analyses and calculations performed to show compliance with NRC regulations governing reduction of fracture toughness of reactor vessel beltline materials; and (3) intergranular separations in low-alloy steel heat-affected zones under austenitic stainless steel cladding.     

Assessment of reactor vessel integrity (ARVI)

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants.The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, 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, which provided data and correlations of CHF due to the external cooling of the lower head.     

On the prediction of the reactor vessel integrity under severe accident loadings (RPVSA)

In order to allow more reliable predictions on the lower head response under core melt-down conditions, the temperature distribution has been analysed including the natural convection in the corium pool. Furthermore, the mechanical models and the failure criteria have been improved based on the RUPTHER and FASTHER experiments where typical temperature gradients are simulated. Lower head local melting as well as corium crust development has been addressed in the CORVIS experiments studying the contact between an alumina/iron thermite and a thick steel plate. The upper head loading by corium impact due to a postulated in-vessel steam explosion has been investigated by the BERDA experiments. Similarity rules were considered such that the results can be directly converted to reactor conditions. Based on these investigations admissible steam explosion energy releases are determined which the upper head can carry. If these limits are not exceeded the reactor containment cannot be endangered by broken head fragments. To provide the necessary basic data, mechanical material tests have been performed.     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

Study of spray cooling of a pressure vessel head of a boiling water reactor

The present paper deals with a theoretical analysis of the spray cooling of a Reactor Pressure Vessel (RPV) head in a Boiling Water Reactor (BWR). To this end a detailed computational model has been developed. The model predicts the trajectories, diameters and temperatures of subcooled droplets moving in saturated vapor. The model has been validated through comparison with experimental data, in which droplet temperatures were measured as functions of the distance that they cover in saturated vapor from the moment they leave the sprinkler outlet to the moment they impact on the RPV head inner wall. The calculations are in very good agreement with measurements, confirming the model adequacy for the present study. The model has been used for a parametric study to investigate the influence of several parameters on the cooling efficiency of the spray system. Based on the study it has been shown that one of the main parameters that govern the temperature increase in a subcooled droplet is its initial diameter. Comparisons are also made between conclusions from the theoretical model and observations made through flow and temperature measurements in the plant (Forsmark 1 and 2). One of these observations is that the rate at which the RPV head temperature decreases on the way down from hot to cold standby is constant and independent of the sprinkling flow rate as long as the flow rate is above a certain minimum value. Accordingly, the theoretical model shows that if one assumes that the cooling of the RPV head is through a water film built on the inner wall due to sprinkling, the heat removal rate is only very weakly dependent on the sprinkling flow rate.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Reactor pressure vessel structural integrity research

Development continues on the technology used to assess the safety of irradiation embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack-tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil-ductility temperature (NDT) performs better than the reference temperature for nil-ductility transition (RT_NDT) as a normalizing parameter for shallow flaw fracture toughness data, (3) biaxial loading can reduce the shallow flaw fracture toughness, (4) stress based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow flaw fracture toughness because in-plane stresses at the crack-tip are not influenced by biaxial loading, and (5) an implicit strain based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation induced shift in Charpy V-notch vs. temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

Investigation on constraint effect of reactor pressure vessel under pressurized thermal shock

In recent years, the integrity of reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) accident has been treated as one of the most critical issues. Under PTS condition, the combination of thermal stress due to a steep temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the RPV wall. As a result, cracks on the inner surface of RPVs can experience elastic-plastic behavior that can be explained using the J-integral. In such a case, however, the J-integral may possibly lose its validity due to the constraint effect. The degree of constraint effect is influenced by the loading mode, the crack geometry and the material properties. In this paper, three-dimensional finite element analyses are performed for various surface cracks to investigate the effect of clad thickness and crack geometry on the constraint effect. A total of 36 crack geometries are analyzed and results are presented by the two-parameter characterization based on the J-integral and the Q-stress.     

The integrity of the pressure containment of the high-temperature reactor

The safety concept for ensuring the integrity of the pressure retaining containment is determined by the structural and system-specific inherent safety characteristics and features of the high-temperature reactor. The integrity of the pressure retaining containment, i.e. the elimination of a major failure, is achieved by a system of measures ensuring a high standard of quality and safety. The fundamental cornerstones of this safety concepts are the stringent requirements in the design and manufacture in view of an optimized production technology as well as specific structural solutions such as, e.g., the prestressed concrete reactor vessel. Additional safety measures such as the quality control performed independently of the manufacturer's works and the in-service inspection, have to be considered as redundant safety measures. The in-service inspection can be limited to the confirmation of safety-relevant data and analysis of deviations from these data. Recurrent non-destructive tests within the PCRV are not required, however, possible to a certain extent.