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’.     

Flaw density examinations of a clad boiling water reactor pressure vessel segment

As part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel. This boiling water reactor vessel was never in service. One objective was to evaluate the approximate 0.7- by 3-m (2- by 10-ft) segment of the vessel provided using ultrasonic flaw detection methods performed with both ASME Code techniques and supplemental ultrasonic methods. A second objective was to evaluate the inner surface stainless steel cladding for cracks with a high sensitivity penetrant examination. Both objectives were successfully completed. Five Code-recordable indications were detected ultrasonically; however, all were found to be anomalies associated with the cladding. One flaw was detected by the supplemental ultrasonic tests, and it was analyzed destructively. This flaw was pipelike indication, about 20 mm (0.8 in.) long extending along the length of the longitudinal weld in which it was located and was about 20 mm below the cladding surface. The flaw had a through-wall dimension (or length) of about 6 mm (0.24 in.) for an approximate 3-mm (0.1-in.) distance along the 20-mm major length. No flaws were detected by the penetrant examination of the cladding surface.     

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.     

Seismic response of reactor vessel internals for Korean standard nuclear power plant

This paper presents the development of seismic design criteria for the reactor vessel internals as a part of the standardization programme for the nuclear power plant in Korea. The seismic design loads of the reactor vessel internals are calculated using the reference input motions of reactor vessels taken from Yonggwang nuclear power plant units 3 and 4 which are being constructed in Korea. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components to the reactor vessel motions is carefully investigated.     

Seismic analysis of the APR 1400 reactor vessel internals using the model reduction method

To secure reliability of the seismic design of the reactor vessel internals (RVIs) through the finite element analysis, it is important to develop the accurate analysis model which can represent the geometric complexity of the RVIs. However, the seismic analysis requires too large computation cost to solve the complex equations; thus, it needs to reduce the overall size of the analysis model. Here, we apply a model reduction method based on the fixed-interface component mode synthesis (CMS) method to practical RVIs to solve complex numerical problems efficiently. To verify the model reduction method, several cases of the RVIs with different conditions are analyzed for the static and dynamic problems. Finally, the seismic analysis was performed with the suggested reduced model with the CMS method. The time history analysis is performed to extract important seismic responses at the specified locations, and the stress analysis is also performed to identify that the RVIs satisfy the seismic design. In the last part of the paper, an example of the design modification is suggested to reduce the stress intensity at the support locations.     

Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste.The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology.The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s.International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

Optimal control of a boiling water nuclear reactor

Application of optimal control to a boiling water nuclear reactor is the theme of this paper. The optimal control problem of a linearized model of a reactor is treated as a regulator problem and feedback control laws are derived to drive the system to steady state in the presence of disturbances. The weighting matrices in the performance index of the regulator problem are suitably changed to yield acceptable closed-loop responses for specific disturbances. The disturbances considered are (i) impulse change in temperature of water at inlet to plenum chamber and (ii) step change in throttle valve area. Then the feedback control laws are implemented on the nonlinear model to illustrate their effectiveness both for large and small disturbances.     

Spectral shift rod for the boiling water reactor

A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained.     

Analyses of pressure perturbation events in boiling water reactor

In order to design more stable and safer core configurations, experimental and theoretical studies about BWR (Boiling Water Reactor) instability have been performed to characterize the phenomenon and to predict the conditions for its occurrence. The instabilities can be caused by interdependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In this work, the RELAP5-MOD3.3 thermal-hydraulic system code and the PARCS-2.4 3D neutron kinetic code were coupled to simulate BWR transients. The pressure perturbation is considered in order to study in detail this type of transient. Two different algorithms developed at the University of Pisa were used to calculate the Decay Ratio (DR) and the natural frequency (NF) from the power oscillation signals obtained from the transient calculations. The validation of a code model set up for the Peach Bottom-2 BWR plant is performed against Low-Flow Stability Tests (LFST). The four series of Stability Tests were performed at Peach Bottom Unit 2 in 1977 at the end of cycle 2 in order to measure the reactor core stability margins at the limiting conditions used in design and safety analysis.     

Monitoring of the VVER nuclear reactor internals

This paper presents a review of the published pressurized water reactor accidents caused by internal vibrations. These accidents are evaluated with respect to their impact on the safety and economy of nuclear power generation. Subsequently, criteria for a monitoring system which would allow the prevention of such accidents are proposed. Structures and some results obtained at NRI e in the course of such system development are presented as well. The results comprise experimental and computational analyses of the behaviour of the VVER-440/V-213 reactor internals and primary coolant.     

Recycling heterogeneous americium targets in a boiling water reactor

One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations that model multiple reload cycles and observe standard operational constraints. These simulations are possible via our collaboration with the Westinghouse Electric Co. which facilitates the use of industrial-caliber design tools such as the PHOENIX-4/POLCA-7 sequence and the Core Master 2^© GUI work environment for fuel management.     

Homogenized model for fluid-structure interaction of the pressurized water reactor core internals during blowdown

A method for the numerical simulation of the pressurized water reactor core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. In order to investigate the global core motion during the blowdown accident, the core model describes the coupled fluid-rod motion with Homogenization methods. The heterogeneous fluid-rod mixture thus is treated as a special continuum with anisotropic material properties. Furthermore, the core model considers elastical rod forces against bending and axial straining and the direct interaction of neighbouring fuel elements, which is a highly nonlinear process due to the finite gaps. Because this effect is very important, two simulation models have been developed and are compared. All these models have been implemented into the blowdown code FLUX-4. With the new code version FLUX-5 the PWR-blowdown is parametrically investigated.     

Study and application of boiling water reactor jet pump characteristic

RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model.In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible.In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K_d, the suction flow junction form loss coefficients, K_S, the diffuser form loss coefficient, K_e, and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve.     

Response of reactor internals to fluctuating pressure forces

The response of nuclear reactor internal components to the turbulent coolant flowing through the annular downcomer region is investigated analytically based on a simplified cylindrical shell model. Powell's acceptance approach is used to compute the r.m.s. response as a function of position on the surface of the shell; the virtual mass effect due to fluid loading is included in the analysis. The input forcing function and damping ratios were obtained by experimental measurement in which the r.m.s. displacement was also measured. The theoretical-experimental agreement is very good.     

Dynamic response of reactor internals to pipe breaks

This paper investigates the horizontal responses of the reactor internals to a 14 inch safety injection nozzle break, which is expected to cause the largest loads of the branch line pipe breaks defined for a nuclear power plant. It examines the effects of two forcing terms, reactor vessel motions and internal hydraulic loads, and suggests a new procedure which can be used for tributary pipe break analysis. The analysis result confirms the applicability of the suggested procedure to small size tributary pipe break analysis. Also, this paper calculates the horizontal responses of the reactor internals to a 3 inch pressurizer spray line nozzle break, which is the only one remaining on the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake to show that seismic loads with a conservative margin may be used for the pipe break loads in the preliminary design.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.