Cross-section modelling effects on pressurised water reactor main steam line break analyses

Because of the strong asymmetric overcooling effects occurring during a PWR main steam line break (MSLB) event, an accurate analysis of this transient requires the use of 3-D kinetics methods. An assessment has been made of the relative performance of the two kinetics solvers currently employed at PSI for such analyses, viz. CORETRAN and SIMULATE-3 K. For the purpose, the simulation of a hypothetical MSLB in a real operated PWR MOX cycle has been considered, employing consistent 3-D core models with specified thermal-hydraulic boundary conditions at the lower and upper plenums. Although the employed cross-section library is in both codes based on the same set of homogenised 2-group cross-sections prepared with CASMO-4, significant differences are shown to occur due to the smaller moderator reactivity coefficient calculated in CORETRAN. It is found that this stems largely from differences in the cross-section formalism, i.e. the manner in which feedback dependencies are modelled and interpolated for the cross-section sets.In particular, the CORETRAN cross-section formalism induces an inadequate treatment of coupled feedback effects, principally between boron density and moderator temperature, which renders the MSLB dynamics predictions quite sensitive to the methodology employed during the cross-section preparation. As such, transient-specific cross-section libraries need to be produced for reliable MSLB analysis in this case. The cross-section model for SIMULATE-3 K, on the other hand, is shown to be adequate for accurately capturing the coupled reactivity effects occurring during an MSLB. In this case, the sensitivity of the results to other sources of uncertainties becomes more apparent, e.g. to those related to the neutron data and/or the thermal-hydraulic boundary conditions. Considering that many other state-of-the-art advanced kinetics solvers have cross-section formalisms similar to that of CORETRAN, effects of the type currently investigated need to be taken into account while developing methodologies for assessing neutronics-related uncertainties in best-estimate transient analysis.     

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5 ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57 bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.     

Predicted impact of power uprate on the water chemistry of Kuosheng boiling water reactor

A theoretical model was adapted to evaluate the impact of power uprate on the water chemistry of a commercial boiling water reactor (BWR) in this work. In principle, the power density of a nuclear reactor upon a power uprate would change immediately, followed by water chemistry variations due to enhanced radiolysis of water in the core and near-core regions. It is currently a common practice for commercial BWRs to adopt hydrogen water chemistry (HWC) for corrosion mitigation. The optimal feedwater hydrogen concentration may be different after a power uprate is implemented in a BWR. A computer code DEMACE was used in the current study to investigate the impact of various power uprate levels on major radiolytic species concentrations and electrochemical corrosion potential (ECP) behavior of components in the primary coolant circuit of a domestic BWR-6 type reactor operating under either normal water chemistry or HWC. Our analyses indicated that under a constant core flow rate the chemical species concentrations and the ECP did not vary monotonously with increases in reactor power level at a fixed feedwater hydrogen concentration. In particular, the upper plenum and the upper downcomer regions exhibited uniquely higher ECPs at 108% and 115% power levels than at the other evaluated power levels.     

Simulation of a main steam line break accident using a coupled “system thermal-hydraulics,three-dimensional reactor kinetics,and hot channel analysis” code

For a realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients.     

Integrity of feedwater and main steam piping in KWU light water reactor plants

New standard catalogs for piping, supports, and valves have been introduced by Kraftwerk Union (KWU) for the first time in its Convoy series of PWR plants. These catalogs, underlying regulatory codes, and newly developed KWU specifications are described. Feedwater and main steam piping systems within the containment, including pipe supports and valves, are used to demonstrate the high quality level of piping technology achieved in the Federal Republic of Germany. Such quality standards ensure the integrity of single components as well as of the entire system, so that, under certain conditions, pipe whip restraints against postulated breaks have become unnecessary. The quality aspects apply basically for both PWR and BWR plants of KWU.     

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.     

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.     

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.     

Observation of in-core instrument tube vibrations in a boiling water reactor by evaluating reactor noise data

Using reactor noise techniques, a vibrating in-core instrument tube in the Mühleberg Boiling Water Reactor could be detected. The normalized auto power spectral density (NPSD) of the current fluctuations of a mini in-core chamber, fixed within the instrument tube, has been analyzed and repair. There was a strong evidence of a peak at the resonance frequency of the vibration at 2.6 Hz in the NPSD. It vanished promptly after the bypass-streaming at the lower core plate, being the physical origin of the vibrating instrument tube, was re-arranged.     

Investigation on a corrosion product deposit layer on a boiling water reactor fuel cladding

Recent investigations on the complex corrosion product deposits on a boiling water reactor (BWR) fuel cladding have shown that the observed layer locally presents unexpected magnetic properties. The magnetic behaviour of this layer and its axial variation on BWR fuel cladding is of interest with respect to non-destructive cladding characterization. Consequently, a cladding from a BWR was cut at elevations of 810 mm, where the layer was observed to be magnetic, and of 1810 mm where it was less magnetic. The samples were subsequently analyzed using electron probe microanalysis (EPMA), magnetic analysis and X-ray techniques (μXRF, μXRD and μXAFS).Both EPMA and μXRF have shown that the observed corrosion deposit layer which is situated on the Zircaloy corrosion layer consists mostly of 3-d elements’ oxides (Fe, Zn, Ni and Mn). The distribution of these elements within the investigated layer is rather complex and not homogeneous. The main phases identified by 2D μXRD mapping inside the layer are hematite and spinel phases with the common formula M_xFe_y(M_(1?x)Fe_(2?y))O₄, where M = Zn, Ni, Mn. It has been shown that the solid solutions of these phases were obtained with rather large differences between the parameter cell of the known spinels (ZnFe₂O₄, NiFe₂O₄ and MnFe₂O₄) and the investigated material. The comparison of EPMA with μXRD analysis shows that the ratio of Fe₂O₃/MFe₂O₄ (M = Zn, Ni, Mn) phases in the lower sample equals ～1/2 and in the higher one ～1/1 within the analyzed volume of the samples. It has been shown that this ratio, together with the thickness of the corrosion product deposit layer, effect its magnetic properties.     

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.     

Fuel lattice design in a boiling water reactor using an ant-colony-based system

This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U²³⁵ enrichment and Gd₂O₃ concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 × 10 BWR fuel lattice with an initial average U²³⁵ enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U²³⁵ enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U²³⁵ enrichment; whereas, the k-infinity was inside the ±100 pcm’s range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of the fuel design.     

Predicted impact of core flow rate on the corrosion mitigation effectiveness of hydrogen water chemistry for Kuosheng boiling water reactor

It is currently a common practice that a boiling water reactor (BWR) adopts hydrogen water chemistry (HWC) for mitigating corrosion in structural components in its primary coolant circuit. When the core flow rate (CFR) in a BWR is changed, the coolant residence time in the primary coolant circuit would be different. The concentrations of major redox species (i.e. hydrogen, oxygen, and hydrogen peroxide) in the coolant may accordingly vary due to different durations of radiolysis in the core and other near-core regions. A theoretical model by the name of DEMACE was used in the current study to investigate the impact of various CFRs (from 100% to 80.0%) on the effectiveness of HWC in a domestic BWR. Our analyses indicated that the HWC effectiveness at some locations could be downgraded due to a decrease in CFR. However, a lower CFR was instead beneficial to the corrosion mitigation efficiency of HWC at other locations. The impact of CFR on the HWC effectiveness could vary from location to location in a BWR and eventually from plant to plant.     

Behaviour of a PWR vessel dissimilar metal weld during a steam line break transient

This paper presents the results of a Finite Element analysis conducted on the behaviour of a 15 mm deep fully circumferential crack on the outer surface of a bi-metallic weld of a 900 MWe Pressurized Water Reactor vessel, during a Steam Line Break transient. The typical weld considered connects the stainless-steel hot leg with the ferritic vessel.This work is part of a Benchmark exercise organised in the framework of an European Commission (DG Environment) project on Transition Welds. However, the results presented here concern only those obtained at Framatome using SYSTUS+ code and deal with the thermo-mechanical analysis, the evaluation of the crack driving energy (J-integral) and a discussion on crack stability. These results are compared with those obtained using a simplified method developed earlier and presented at one of the international conferences.     

Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design

In boiling water reactor (BWR) design, safety scenarios such as main steam line break need to be evaluated. After the main steam line break, the steam will fill the upper dry well of the containment. It will then enter the vertical vent and eventually flow into the suppression pool via horizontal vents. The steam will create large bubbles in the suppression pool and cause the pool to swell. The impact of the pool swell on the equipment inside the pool and containment structure needed to be evaluated for licensing. GE has conducted a series of one-third scale three-vent air tests in supporting the horizontal vent pressure suppression system used in Mark III containment design for General Electric BWR plants. During the test, the air-water interface locations were tracked by conductivity probes. The pressure was measured at many locations inside the test rig as well. The purpose of the test was to provide a basis for the pool swell load definition for the Mark III containment. In this paper, a transient three-dimensional CFD model to simulate the one-third scale Mark III suppression pool swell process is illustrated. The Volume of Fluid (VOF) multiphase model is used to explicitly track the interface between the water liquid and the air. The CFD results such as flow velocity, pressure, interface locations are compared to the data from the test. Through comparisons, a technical approach to numerically model the pool swell phenomenon is established and benchmarked.     

Numerical investigations on power ramping procedure for a natural circulation boiling water reactor

Startup of a natural circulation boiling water reactor (NCBWR) is studied numerically, using a thermal-hydraulic system code RELAP5. A number of numerical experiments are carried out using various power ramps, and a suitable heat-up rate is identified to pressurize the reactor to the desired operating conditions in a reasonable time without considerable void generation in the core. It is observed that the occurrence of flashing in the riser section is unavoidable. Although flashing helps in steam production, the amplitude of flow oscillations induced by flashing is the event of concern, as in the case of the pressure tube type NCBWR studied here. Therefore, the feasibility of a complete single-phase startup is also examined and found not attractive. A new startup procedure, which completely bypasses the unstable two-phase region, is conceptualized, and the method to take the system to the operating condition without encountering flow oscillations is numerically investigated.