Relevant bibliographies by topics / BWR1

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Academic literature on the topic 'BWR1'

Author: Grafiati
Published: 6 September 2023

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Journal articles on the topic "BWR1"

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Nilsson, Tina, Anna Sjöblom, Maria G. Masucci, and Lars Rymo. "Viral and Cellular Factors Influence the Activity of the Epstein-Barr Virus BCR2 and BWR1 Promoters in Cells of Different Phenotype." Virology 193, no. 2 (April 1993): 774–85. http://dx.doi.org/10.1006/viro.1993.1186.

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LU, TIEN-FU. "MODELING FOR STOCKPILE OPERATIONS ASSOCIATED WITH BULK SOLID MATERIALS USING BUCKET WHEEL RECLAIMER." International Journal of Information Acquisition 07, no. 04 (December 2010): 357–73. http://dx.doi.org/10.1142/s0219878910002270.

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Bucket wheel reclaimer (BWR) is one of the main equipment which has been widely used for stacking/reclaiming bulk materials (i.e., iron ore and coal) in ports, iron-steel plants, coal storages, and power stations onto/from stockpiles by mining industry. Generally speaking, current BWRs are mostly manually operated, remotely operated, or automated to simply follow predefined trajectory patterns for stacking and reclaiming operations. BWRs are indeed very large in size, heavy in weight, expensive in price, and slow in motion. It is commonly agreed in the industry that the current stacking/reclaiming efficiency can be largely improved in several areas to obtain huge amount of savings in dollar terms. However, as stockpiles and BWRs are always heavily engaged in production and cannot be long spared or frequently interrupted for the required studies and developments for efficiency improvement, a close to real simulation environment including stockpiles, BWRs, and the associated environment would be highly valuable and greatly beneficial to carry out necessary studies, planning, preparations, and evaluations. This paper presents the progresses of the modeling work achieved so far for the simulation of stockpile operations associated with bulk solid materials using BWRs. The content covers the modeling of stockpiles, typical BWR, voxel-based reclaiming trajectory generation, and their implementation in a simplified stockyard. The result demonstrates a powerful simulation environment is being woven together and can be used as a tool for further investigations to improve relevant production efficiencies.
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Altiok, E., J. Minarovits, L. F. Hu, B. Contreras-Brodin, G. Klein, and I. Ernberg. "Host-cell-phenotype-dependent control of the BCR2/BWR1 promoter complex regulates the expression of Epstein-Barr virus nuclear antigens 2-6." Proceedings of the National Academy of Sciences 89, no. 3 (February 1, 1992): 905–9. http://dx.doi.org/10.1073/pnas.89.3.905.

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Olvera-Guerrero, Omar Alejandro, Alfonso Prieto-Guerrero, and Gilberto Espinosa-Paredes. "A Novel Nonlinear BWR Stability Indicator Based on the Sample Entropy." Science and Technology of Nuclear Installations 2018 (November 1, 2018): 1–13. http://dx.doi.org/10.1155/2018/9852925.

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BWRs are thus far the simplest energy systems to transform fission energy into electrical power. However, there are still many aspects in their operation that, under certain conditions, may induce BWR unstable behavior. The default indicator to study BWR unstable behavior is the Decay Ratio (DR). However, due to the fact that BWRs show very complex responses under instability and responses that may even be chaotic, the DR might not be a suitable choice to rely on to accommodate for such intricate behavior. In this work a novel methodology based on the Sample entropy (SampEn) and the noise-assisted multivariate empirical mode decomposition (NA-MEMD) is introduced. Such methodology was developed thinking for a real time-implementation of a stability monitor. The proposed methodology was tested with a set of signals that stem from several nuclear power plants in operation today that have experienced in the past unstable events, each one of a different nature.
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LANGE, CARSTEN, DIETER HENNIG, and ANTONIO HURTADO. "A NOVEL RESULT IN THE FIELD OF NONLINEAR STABILITY ANALYSIS OF BOILING WATER REACTORS." International Journal of Bifurcation and Chaos 22, no. 02 (February 2012): 1250041. http://dx.doi.org/10.1142/s0218127412500411.

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The nonlinear stability analysis of boiling water nuclear reactors (BWRs) is conducted with the aid of so-called advanced, well validated, system codes and an advanced reduced order model to build a detailed mathematical understanding of the BWR behavior in the practical relevant parameter space. In the last years, the existence of Hopf-bifurcation points was confirmed by some researchers. In the framework of this paper, a parameter region was analyzed in which the coexistence of different stability states is realized. As a novel result, we found a parameter region in which stable fixed points, unstable limit cycles and stable limit cycles coexist. This system behavior can be explained by a saddle-node bifurcation of cycles (turning point). The existence of this solution type in a BWR system indicates the possibility of large amplitude limit cycle oscillations in the linear stable region.
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Cui, Weihua, Bao Song, Chao Fu, and Hui Wang. "Effect of pitch on mechanical properties of braided wire rope under winding and traction condition." Journal of Physics: Conference Series 2355, no. 1 (October 1, 2022): 012080. http://dx.doi.org/10.1088/1742-6596/2355/1/012080.

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Abstract In the tension stringing construction process of power transmission and transformation engineering, the braided wire ropes (BWRs) are in the state of winding and stretching when passing through the friction drum. Pitch is an important structural parameter of BWRs, which directly influences the mechanical behaviors under these conditions. Based on the YS9-8×19 wire rope, this project studies the effect of the rope strand pitch on the curvature and winding from both qualitative and quantitative aspects. Qualitative analysis initially explores the effect of the pitch on the mechanical behaviors. Based on the established “rope-wheel” solid model and numerical simulation model, the stress distribution and the variation trend of the maximum equivalent stress corresponding to different pitches in the winding traction state are obtained through the quantitative analyses of the numerical simulations under the same loads with different pitches. The reasonable pitch range of the BWR subjected to traction and bending load is further concluded, to provide the data reference for the manufacture of related wire ropes.
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Macdonald, Digby D., and George R. Engelhardt. "A Critical Review of Radiolysis Issues in Water-Cooled Fission and Fusion Reactors: Part II, Prediction of Corrosion Damage in Operating Reactors." Corrosion and Materials Degradation 3, no. 4 (November 30, 2022): 694–758. http://dx.doi.org/10.3390/cmd3040038.

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The radiolysis of water is a significant cause of corrosion damage in the primary heat transport systems (PHTSs) of water-cooled, fission nuclear power reactors (BWRs, PWRs, and CANDUs) and is projected to be a significant factor in the evolution of corrosion damage in future fusion reactors (e.g., the ITER that is currently under development). In Part I of this two-part series, we reviewed the proposed mechanisms for the radiolysis of water and demonstrate that radiolysis leads to the formation of a myriad of oxidizing and reducing species. In this Part II, we review the role that the radiolysis species play in establishing the electrochemical corrosion potential (ECP) and the development of corrosion damage due to intergranular stress corrosion cracking (IGSCC) in reactor PHTSs. We demonstrate, that the radiolytic oxidizing radiolysis products, such as O2, H2O2, HO2−, and OH, when in molar excess over reducing species (H2, H, and O22−), some of which (H2) are preferentially stripped from the coolant upon boiling in a BWR PHTS, for example, renders the coolant in many BWRs oxidizing, thereby shifting the ECP in the positive direction to a value that is more positive than the critical potential (Ecrit = −0.23 Vshe at 288 °C) for IGSCC in sensitized austenitic stainless steel (e.g., Type 304 SS). This has led to many IGSCC incidents in operating BWRs over the past five decades that has exacted a great cost on the plant operators and electricity consumers, alike. In the case of PWRs, the primary circuits are pressurized with hydrogen to give a hydrogen concentration of 10 to 50 cm3/kgH2O (0.89 to 4.46 ppm), such that no sustained boiling occurs, and the hydrogen suppresses the radiolysis of water, thereby inhibiting the formation of oxidizing radiolysis products from water. Thus, the ECP is dominated by the hydrogen electrode reaction (HER), although important deviations from the HER equilibrium potential may occur, particularly at low [H2]. In any event, the ECP is displaced to approximately −0.85 Vshe, which is below the critical potential for IGSCC in sensitized stainless steels but is also more negative than the critical potential for the hydrogen-induced cracking (HIC) of mill-annealed Alloy 600. This has led to extensive cracking of steam generator tubing and other components (e.g., control rod drive tubes, pressurizer components) in PWRs that has also exacted a high cost on operators and power consumers. Although the ITER has yet to operate, the proposed chemistry protocol for the coolant places it close to a BWR operating on Normal Water Chemistry (NWC) without boiling or, if hydrogen is added to the IBED-PHTS, close to a BWR on Hydrogen Water Chemistry (HWC). In the current ITER technology, the concentration of H2 in the IBED-PHTS is specified to be 80 ppb, which is the concentration that will be experienced in both the Plasma Flux Area (PFA) and in the Out of Plasma Flux Area (OPFA). That corresponds to 0.90 cc(STP) H2/KgH2O, compared with 20–50 cc(STP) H2/KgH2O employed in a PWR primary coolant circuit and 5.5 to 22 cc(STP) H2/KgH2O in a BWR on hydrogen water chemistry (HWC). We predict that a hydrogen concentration of 80 ppb is sufficient to reduce the ECP in the OPFA to a level (−0.324 Vshe) that is sufficient to suppress the crack growth rate (CGR) below the practical, maximum level of 10−9 cm/s (0.315 mm/a) at which SCC is considered not to be a problem in a coolant circuit but, in the PFA, the ECP is predicted to be 0.380 Vshe, which gives a calculated standard CGR of 2.7 × 10−6 cm/s. This is more than three orders in magnitude greater that the desired maximum value of 10−9 cm/s. We recommend that the HWC issue in ITER be revisited to develop a protocol that is effective in suppressing both the ECP and the CGR in the PFA to levels that permit the operation of the IBED-PHTS in accordance with the experience gained in fission reactor technology.
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Harder, James W., Jing Ma, Pascale Alard, Kevin J. Sokoloski, Edith Mathiowitz, Stacia Furtado, Nejat K. Egilmez, and Michele M. Kosiewicz. "Male microbiota-associated metabolite restores macrophage efferocytosis in female lupus-prone mice via activation of PPARγ/LXR signaling pathways." Journal of Leukocyte Biology 113, no. 1 (January 10, 2023): 41–57. http://dx.doi.org/10.1093/jleuko/qiac002.

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Abstract Systemic lupus erythematosus development is influenced by both sex and the gut microbiota. Metabolite production is a major mechanism by which the gut microbiota influences the immune system, and we have previously found differences in the fecal metabolomic profiles of lupus-prone female and lupus-resistant male BWF1 mice. Here we determine how sex and microbiota metabolite production may interact to affect lupus. Transcriptomic analysis of female and male splenocytes showed genes that promote phagocytosis were upregulated in BWF1 male mice. Because patients with systemic lupus erythematosus exhibit defects in macrophage-mediated phagocytosis of apoptotic cells (efferocytosis), we compared splenic macrophage efferocytosis in vitro between female and male BWF1 mice. Macrophage efferocytosis was deficient in female compared to male BWF1 mice but could be restored by feeding male microbiota. Further transcriptomic analysis of the genes upregulated in male BWF1 mice revealed enrichment of genes stimulated by PPARγ and LXR signaling. Our previous fecal metabolomics analyses identified metabolites in male BWF1 mice that can activate PPARγ and LXR signaling and identified one in particular, phytanic acid, that is a very potent agonist. We show here that treatment of female BWF1 splenic macrophages with phytanic acid restores efferocytic activity via activation of the PPARγ and LXR signaling pathways. Furthermore, we found phytanic acid may restore female BWF1 macrophage efferocytosis through upregulation of the proefferocytic gene CD36. Taken together, our data indicate that metabolites produced by BWF1 male microbiota can enhance macrophage efferocytosis and, through this mechanism, could potentially influence lupus progression.
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Harder, James W., Jing Ma, Pascale Alard, Xiang Zhang, Fang Yuan, and Michele M. Kosiewicz. "Male microbiota-associated metabolites restore macrophage efferocytosis in female lupus-prone mice via PPARγ and LXR signaling pathways." Journal of Immunology 206, no. 1_Supplement (May 1, 2021): 105.04. http://dx.doi.org/10.4049/jimmunol.206.supp.105.04.

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Abstract Sex and gut microbiota both influence systemic lupus erythematosus (SLE) development. We have found that male and lupus-prone female NZBxNZW F1 (BWF1) mice exhibit differences in gut microbiota and metabolomic profiles, and transfer of male microbiota suppresses disease in female mice. Here we determine how sex and microbiota may interact to affect SLE development. Transcriptomic analysis of female and male BWF1 spleens found phagocytosis-promoting genes were upregulated in males. Since defects in macrophage-mediated phagocytosis of apoptotic cells (efferocytosis) are found in SLE patients, we compared efferocytosis in vitro between male and lupus-prone female BWF1 mice. Macrophage-mediated efferocytosis was decreased in female compared to male BWF1 mice, but could be restored in female mice by transfer of male microbiota in vivo. Further transcriptomic analysis of spleens found that genes regulated by PPARγ and LXR receptor signaling were increased in male BWF1 mice. Our previous metabolomics analyses have found that two metabolites which can activate PPARγ and LXR signaling, phytol and its derivative phytanic acid, are higher in male BWF1 mice. We have also found that feeding phytanic acid to female mice delayed lupus onset. We show here that defective female BWF1 macrophage efferocytosis can be restored by phytanic acid treatment, and this requires both PPARγ and LXR signaling pathways. Furthermore, we found that phytanic acid may restore female BWF1 macrophage efferocytosis by upregulating the pro-efferocytic receptor, CD36. Taken together, our data indicate that BWF1 male microbiota produce metabolites that can enhance macrophage efferocytosis and suppress lupus.
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Teguh Sasono, Tjatur Udjianto, and Taufik Rizal. "RANCANGAN MULTISTAGE HIGH RECOVERY BRACKISH WATER REVERSE OSMOSIS PADA PLTU CILACAP KAPASITAS 660 MW." Jurnal Teknik Energi 6, no. 2 (February 17, 2020): 541–46. http://dx.doi.org/10.35313/energi.v6i2.1719.

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Rancangan sistem BWRO dengan persentase air hasil pengolahan yang tinggi (high recovery) akan mengurangi biaya operasinya. Untuk mencapai high recovery, skema multistage diterapkan dalam rancangan BWRO ini. Dengan menerapkan skema multistage dan nilai recovery per elemen sebesar 14,53%, sistem BWRO ini memiliki keandalan (reliability) yang baik dengan investasi yang rendah. Air yang diolah pada BWRO merupakan air keluaran dari Seawater Reverse Osmosis (SWRO), air keluaran SWRO ini masih mengandung mineral di dalamnya. Kandungan mineral di dalam air disebut Total Dissolved Solids (TDS), TDS merupakan parameter yang harus dikurangi jumlahnya. Air keluaran SWRO merupakan brackish water dengan kandungan TDS 400 mg/L, temperature 25"C, dan nilai pH 8. Perancangan BWRO dimulai dari menentukan kebutuhan jumlah air dan mengetahui karakteristik air yang akan diolah. Kemudian, dilakukan pemilihan elemen membran dan pressurevessel, menentukan recoveryrate, dan menentukan jumlah stage. Selanjutnya, adalah menentukan tekanan input dan menghitung parameter performansi BWRO. Selain TDS, parameter yang menjadi pesyaratan dalam rancangan BWRO ini adalah Specific Membrane Permeability (SMP) standar BWRO yaitu 4,9-8,3 Lmh/bar. Dari hasil rancangan didapat feed flow 127,28 m3/jam, permeateflow 113 m3/jam, jumlah vessel stage 1 dan 2 masing-masing 12 buah dan 4 buah, TDS 7,72 mg/L, SMP stage 1 dan stage 2 masing - masing 6,52 Lmh/bar dan 6,87 Lmh/bar.
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Dissertations / Theses on the topic "BWR1"

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Raub, Sebastian. "Transient behaviour in a BWR with Hafnium Cladding : Feasibility study of using BWRs as Higher Actinide Burners at the Example of Ringhals I." Thesis, KTH, Skolan för teknikvetenskap (SCI), 2011. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-38189.

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Transmutation of transuranic elements is of interest to lower storage unit cost and long-term radiotoxicity. To make use of existing infrastructure, the deployment of Boiling Water Reactors (BWRs) with hafnium cladding and Mixed Oxide (MOX) fuel was proposed, resulting in a hardening of the neutron spectrum. This work tests varying spatial fuel configurations for maximal burn-up, using Serpent, and study their behaviour in common accident scenarios, simulated by a coupled TRACE/PARCS software suite. To this end, we provide a software solution, which serves to transfer Serpent output of a user defined system in a cross section parameter file, readable by TRACE/PARCS. The results of the transfer were tested for safety performance and, if they provided satisfactory steady states, subjected to a turbine-trip event without bypass, with or without control rod SCRAM. Building on the works by Suvdantstseg [12] and Wallenius & Westlen [7], we chose a Transuranium (TRU) content of 16.48% and a Hafnium-content of 5% with various Higher Actinides (HA) contents and z-axis distributions, intended to either maximize safety performance or minimize void worth and study the results. The chosen fuel loading allows a safe shut-down for both accident scenarios. Sharply rising pressure inside the reactor vessel causes a void collapse. The TRU-content lowers the positive reactivity contribution of increased moderator density, compared to the Uranium Oxide (UOX) baseline. Nonetheless, using a Hf-content of 5% in the cladding and MOX-fuel with 16.48 TRU and 2.06 HA, the void coefficient stays negative during a transitional period of the shutdown, lasting for approximately 200 seconds, before before changing it’s sign.
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Chun, John Hwan. "Modeling of BWR water chemistry." Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/1721.1/13660.

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Soma, Kovács István. "Simplified Simulator for BWR Instabilities." Thesis, KTH, Fysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-210626.

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Ferroni, Paolo Ph D. Massachusetts Institute of Technology. "Steady state thermal hydraulic analysis of hydride fueled BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41263.

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Includes bibliographical references (p. 205-208).
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2006.
(cont.) Since the results obtained in the main body of the analysis account only for thermal-hydraulic constraints, an estimate of the power reduction due to the application of neutronic constraints is also performed. This investigation, focused only on the "New Core" cases, is coupled with an increase of the thickness of the gap separating adjacent bundles from 2 to 5 mm. Under these more conservative conditions, the power gain percentages are lower, ranging between 24% and 43% (depending on the discharge burnup considered acceptable) for the upper pressure drop limit, and between 17% and 32% for the lower pressure drop limit.
(cont.) The benefits of the latter approach are evident since the space occupied by the bypass channel for cruciform control rod insertion becomes available for new fuel and a higher power can be achieved. The core power is constrained by applying thermal-hydraulic limits that, if exceeded, may induce failure mechanisms. These limits concern Minimum Critical Power Ratio (MCPR), core pressure drop, fuel average and centerline temperature, cladding outer temperature and flow-induced vibrations. To limit thermal-hydraulic instability phenomena, core power and coolant flow are constrained by fixing their ratio to a constant value. In particular, each BWR/5 core has been analyzed twice, each time with a different pressure drop limit: a lower limit corresponding to the pressure drop of the reference core and an upper limit 50% larger. It has been demonstrated that, in absence of neutronic constraints and with the maximum allowed pressure drop fixed at the upper limit, the implementation of the hydride fuel yields power gain percentages, with respect to oxide cores chosen as reference, of the order of 23% when its implementation is performed following the "Backfit" approach and even higher (50-70%) when greater design freedom is allowed in the core design, i.e. in the "New Core" approach. Should the maximum allowed pressure drop be fixed at the lower limit, the power gain percentage of the "Backfit" approach would decrease to 17%, while that of the "New Core" approach would remain unchanged, i.e. 50-70%.
This thesis contributes to the Hydride Fuel Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in Light Water Reactors (LWRs). Considerable work has already been accomplished on hydride fueled Pressurized Water Reactor (PWR) cores. This thesis extends the techniques used in the PWR analysis to examine the potential power benefits resulting from the implementation of the hydride fuel in Boiling Water Reactors (BWRs). This work is the first step towards the achievement of a complete understanding of the economic implications that may derive from the use of this new fuel in BWR applications. It is a whole core steady-state analysis aimed at comparing the power performance of hydride fueled BWR cores with those of typical oxide-fueled cores, when only thermal-hydraulic constraints are applied. The integration of these results with those deriving from a transient analysis and separate neutronic and fuel performance studies will provide the data required to build a complete economic model, able to identify geometries offering the lowest cost of electricity and thus to provide a fair basis for comparing the performance of hydride and oxide fuels. Core design is accomplished for two types of reactors: one smaller, a BWR/5, which is representative of existing reactors, and one larger, the ESBWR, which represents the future generation of BWRs. For both, the core design is accomplished in two ways: a "Backfit" approach, in which the ex-bundle core structure is identical to that of the two reference oxide cores, and a "New Core" approach, in which the control rods are inserted into the bundles in the form of control fingers and the gap between adjacent bundles is fixed optimistically at 2 mm.
by Paolo Ferroni.
S.M.
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Morra, Paolo. "Design of annular fuel for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/34448.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2005.
Includes bibliographical references (p. 94).
Enabling high power density in the core of Boiling Water Reactors (BWRs) is economically profitable for existing or new reactors. In this work, we examine the potential for increasing the power density in BWR plants by switching from the current solid fuel to annular fuel cooled both on its inside and outside surfaces. The GE 8x8 bundle dimensions and fuel to moderator ratio are preserved as a reference to enable applications in existing reactors. A methodology is developed and VIPRE code calculations are performed to select the best annular fuel bundle design on the basis of its Critical Power Ratio (CPR) performance. Within the limits applied to the reference solid fuel, the CPR margin in the 5x5 and 6x6 annular fuel bundles is traded for an increase in power density. It is found that the power density increase with annular fuel in BWRs may be limited to 23%. This is smaller than possible for PWRs due to the different mechanisms that control the critical thermal conditions of the two reactors. The annular fuel could still be a profitable alternative to the solid fuel due to neutronic and thermal advantages.
by Paolo Morra.
S.M.
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Karahan, Aydin. "An evolutionary fuel assembly design for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2006. http://hdl.handle.net/1721.1/41304.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2007.
Includes bibliographical references (p. 138-140).
An evolutionary BWR fuel assembly design was studied as a means to increase the power density of current and future BWR cores. The new assembly concept is based on replacing four traditional assemblies and large water gap regions with a single large assembly. The traditional BWR cylindrical UO2-fuelled Zr-clad fuel pin design is retained, but the pins are arranged on a 22x22 square lattice. There are 384 fuel pins with 9.6 mm diameter within a large assembly. Twenty-five water rods with 27 mm diameter maintain the moderating power and accommodate as many finger-type control rods. The total number and positions of the control rod drive mechanisms are not changed, so existing BWRs can be retrofitted with the new fuel assembly. The technical characteristics of the large fuel assembly were evaluated through a systematic comparison with a traditional 9x9 fuel assembly. The pressure, inlet subcooling and average exit quality of the new core were kept equal to the reference values. Thus the power uprate is accommodated by an increase of the core mass flow rate. The findings are as follows: - VIPRE subchannel analysis suggests that, due to its higher fuel to coolant heat transfer area and coolant flow area, the large assembly can operate at a power density 20% higher than the traditional assembly while maintaining the same margin to dryout. - CASMO 2D neutronic analysis indicates that the large assembly can sustain an 18-month irradiation cycle (at uprated power) with 3-batch refueling, <5wt% enrichment with <60 MWD/kg average discharge burnup. Also, the void and fuel temperature reactivity coefficients are both negative and close to those of the traditional BWR core. - The susceptibility of the large assembly core to thermalhydraulic/neutronic oscillations of the density-wave type was explored with an in-house code.
(cont.) It was found that, while well within regulatory limits, the flow oscillation decay ratio of the large assembly core is higher than that of the traditional assembly core. The higher core wide decay ratio of the large assembly core is due to its somewhat higher (more negative) void reactivity coefficient. The pressure drop in the uprated core is 17 %Vo higher than in the reference core, and the flow is 20% higher; therefore, larger pumps will be needed. FRAPCON analysis suggests that the thermo-mechanical performance (e.g., fuel temperature, fission gas release, hoop stress and strain, clad oxidation) of the fuel pins in the large assembly is similar to that of the reference assembly fuel pins. A conceptual mechanical design of the large fuel assembly and its supporting structure was developed. It was found that the water rods and lower tie plate can be used as the main structural element of the assembly, with horizontal support being provided by the top fuel guide plate and core plate assembly, and vertical support being provided by the fuel support duct, which also supports the finger-type control rods.
by Aydin Karahan.
S.M.
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Gajev, Ivan. "Sensitivity and Uncertainty Analysis of BWR Stability." Licentiate thesis, KTH, Kärnkraftsäkerhet, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-26387.

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Best Estimate codes are used for licensing, but with conservative assumptions. It is claimed that the uncertainties are covered by the conservatism of the calculation. As Nuclear Power Plants are applying for power up-rates and life extension, evaluation of the uncertainties could help improve the performance, while staying below the limit of the safety margins.   Given the problem of unstable behavior of Boiling Water Reactors (BWRs), which is known to occur during operation at certain power and flow conditions, it could cause SCRAM and decrease the economic performance of the plant. Performing an uncertainty analysis for BWR stability would give better understating of the phenomenon and it would help to verify and validate (V&V) the codes used to predict the NPP behavior.   This thesis reports an uncertainty study of the impact of Thermal-Hydraulic, Neutronic, and Numerical parameters on the prediction of the stability of the BWR within the framework of OECD Ringhals-1 stability benchmark. The time domain code TRACE/PARCS was used in the analysis. This thesis is divided in two parts: Sensitivity study on Numerical Discretization Parameters (Nodalization, Time Step, etc.) and Uncertainty part.   A Sensitivity study was done for the Numerical Parameters (Nodalization and Time step). This was done by refining all possible components until obtaining Space-Time Converged Solution, i.e. further refinement doesn’t change the solution. When the space-time converged solution was compared to the initial discretization, a much better solution has been obtained for both the stability measures (Decay Ratio and Frequency) with the space-time converged model.   Further on, important Neutronic and Thermal-Hydraulic Parameters were identified and the uncertainty calculation was performed using the Propagation of Input Errors (PIE) methodology. This methodology, also known as the GRS method, has been used because it has been tested and extensively verified by the industry, and because it allows identifying the most influential parameters using the Spearman Rank Correlation.
QC 20101126
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Melara, San Román José. "PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR." Doctoral thesis, Universitat Politècnica de València, 2016. http://hdl.handle.net/10251/61307.

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[EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. If the oscillation amplitude is large, before the scram occurs the fuel rods may experience periodic dry-out and rewetting, or if the oscillation is larger enough, extended dry-out. The Decay Ratio (DR) is the typical linear stability figure of merit. For analytical estimation of DR frequency domain codes are very useful. These types of codes are very fast and their results are very robust in comparison with time domain codes, whose results may be dependent on numeric scheme and nodalization. The only drawback of frequency domain is that you are limited to the linear domain; however, because of regulatory requirements imposed by GDC-12, reactors must remain stable and, thus, reactors always operate in the linear domain. LAPUR is a frequency domain stability code that contains a mathematical description of the core of a boiling water reactor. It solves the steady state governing equations for the coolant and fuel, and the dynamic equations for the coolant, fuel and the neutron field in the frequency domain. Several improvements have been performed to the current version of the code, LAPUR5, in order to upgrade it for use with new fuel design types. The channel geometry has been changed from constant area to variable area. The local losses due to the spacers and contractions along the flow path have been upgraded to use industry standard correlations. This new version is LAPUR 6. In this work, in order to check the correct implementation of these changes, a two-fold LAPUR 6 validation has been performed: First, an exhaustive validation of the models implemented has been performed, comparing single channels LAPUR 6 outputs against SIMULATE-3 results. Cofrentes NPP SIMULATE-3 thermal-hydraulic models have been independently validated against experimental data. Second, a Methodology for calculating Decay Ratios with LAPUR 6 has been developed, defining a validation matrix against analytical and plant measured decay ratios. Analysis of measured data from the Cofrentes NPP has shown that decay ratios have values lower than 0.3 confirming the large stability margin of Cofrentes NPP when proper operating procedures are followed, and the comparison with LAPUR shows deviations less than +/- 0.1. Past experience suggests that the uncertainty in low decay ratio ranges is usually larger than with higher decay ratio values. Finally a BWR noise generator has been used for estimating the uncertainty of the signal analyses methods used in this work for experimental estimation of decay ratio from the autocorrelation function of the APRM or LPRM power signals.
[ES] Las oscilaciones de potencia y caudal en un BWR no son deseables. Una de las principales preocupaciones es asegurar, durante oscilaciones de potencia, el cumplimiento de la GDC 10 y 12. GDC 10 requiere que el núcleo del reactor se haya diseñado con un margen adecuado para asegurar que los límites admisibles establecidos en el diseño del combustible no se excederán en cualquier condición de operación normal, incluyendo los efectos de los sucesos operacionales anticipados. GDC 12 requiere garantías de que las oscilaciones de potencia que pueden resultar en condiciones que excedan los límites admisibles establecidos de diseño del combustible, o bien no son posibles o puedan ser detectadas y suprimidas de forma pronta y segura. Si la amplitud de la oscilación es grande, antes de que se produzca el scram las varillas de combustible pueden experimentar secados y remojados periódicos, o si las oscilaciones son suficientemente grandes, un secado extendido. La tasa de amortiguamiento (DR) es la típica figura de mérito de la estabilidad lineal. Para la estimación analítica de la DR los códigos en el dominio de la frecuencia son muy usados. Este tipo de códigos son muy rápidos y sus resultados son muy robustos en comparación con los códigos en el domino temporal, cuyos resultados pueden depender del esquema numérico y la nodalización. El único inconveniente de los códigos en el dominio de la frecuencia es que está limitado al dominio lineal; sin embargo, como los requerimientos regulatorios impuestos por el GDC-12, los reactores deben permanecer estables y, por lo tanto, los reactores deben operar siempre en el dominio lineal. LAPUR es un código de estabilidad en el dominio de la frecuencia que contiene una descripción matemática del núcleo de un reactor de agua en ebullición. Resuelve las ecuaciones de conservación en estado estacionario para el refrigerante y el combustible, las ecuaciones dinámicas para el refrigerante, el combustible y el campo neutrónico en el dominio de la frecuencia. Se han realizado varias mejoras a la versión actual del código, LAPUR 5, con el fin de actualizarlo para su uso con los nuevos tipos de diseño de combustible. La geometría del canal se ha cambiado, el área ha pasado de ser constante a poder considerar área variable. El cálculo de las pérdidas locales debido a los espaciadores y contracciones a lo largo del camino que sigue el flujo se han actualizado, pasando a utilizar correlaciones estándar de la industria. Esta nueva versión del código se ha denominado LAPUR 6. En este trabajo, con el fin de verificar la correcta implementación de estos cambios, se ha realizado una doble validación del código LAPUR 6: En primer lugar se ha realizado una validación exhaustiva de los modelos implementados, comparando los valores de salida de LAPUR 6 para un canal con los resultados de SIMULATE-3. Los modelos termohidráulicos de la CN Cofrentes de SIMULATE-3 han sido validados de forma independiente con los datos experimentales. En segundo lugar se ha desarrollado una metodología para el cálculo de la tasa de amortiguamiento con LAPUR 6, definiendo una matriz de validación de los valores de tasa de amortiguamiento analíticos con valores medidos en la planta. Las tasas de amortiguamiento medidos en la Central Nuclear de Cofrentes tienen valores inferiores al 0.3, confirmando el gran margen de estabilidad de la Central Nuclear de Cofrentes cuando se siguen los procedimiento de operación adecuados, y la comparación con los resultados de LAPUR muestra desviaciones de menos de +/- 0.1. La experiencia acumulada sugiere que la incertidumbre para los rangos bajos de tasas de amortiguamiento es generalmente más grande que para los valores altos. Por último se ha utilizado un generador de señales BWR para la estimación de la incertidumbre de los métodos de análisis de señales utilizados en este trabajo para la estimación experimental de la DR, a partir de la funci
[CAT] Les oscil·lacions de potència i flux en un BWR són molt poc desitjades. Una de les majors preocupacions és assegurar-se, durant les oscil·lacions de potència, del compliment de GDC 10 i 12. GDC 10 requerix que el nucli del reactor estiga dissenyat amb un marge apropiat per a assegurar que els limits admissibles establerts en el disseny del combustible no siguen superats davall cap condició d'operació normal, incloent els incidents esperats d'operació. GDC 12 requerix assegurar que les oscil·lacions de potència que poden resultar en condicions on es superen els limits admissibles establerts en el disseny del combustible no siguen possibles o puguen ser detectades de manera segura e immediata i suprimides. Si l'amplitud de les oscil·lacions és gran, abans que el scram ocórrega les barres experimenten un assecat i remullat periòdic, o si l'oscil·lació és prou gran, un assecat estés. La taxa d'amortiment (DR) és la típica figura de mèrit de l'estabilitat lineal. Per a l'estimació analítica de la DR són molt usats els codis en el domini de la freqüència. Este tipus de codis són molt ràpids i els seus resultats són molt robustos en comparació amb els codis en el domini temporal, els resultats del qual són molt dependents de l'esquema numèric i la nodalizació. L'únic inconvenient del domini de la freqüència és que està limitat al domini lineal, no obstant això, com els requeriments reguladors imposats pel GDC-12, els reactors han de mantener-se estables i, per tant, els reactors han d'operar sempre en el domini lineal. LAPUR és un codi d'estabilitat en el domini de la freqüència que conté una descripció matemàtica del nucli d'un reactor d'aigua en ebullició. Resol les equacions de govern estacionàries del refrigerant i el combustible, les equacions dinàmiques del refrigerant, el combustible i el camp neutrònic en el domini de la freqüència. S'han realitzat diverses millores a la versió anterior del codi, LAPUR 5, amb l'objectiu d'actualitzar-ho per al seu ús amb nous tipus de disseny de combustibles. La geometria del canal s'ha canviat d'àrea constant a variable. Les pèrdues locals degudes als espaciadors i contraccions al llarg del camí del flux s'han actualitzat per a utilitzar correlacions estàndard de la indústria. Esta nova versió és LAPUR 6. En este treball, amb l'objectiu de comprovar la correcta implementació d'estos canvis, s'ha realitzat una doble validació del LAPUR 6: Primer, s'ha realitzat una validació exhaustiva dels models implementats, comparant els valors d'eixida per a un canal de LAPUR 6 amb els resultats de SIMULATE-3. Els models termohidraúlics per a SIMULATE-3 de la Central Nuclear de Cofrentes s'han validat independentment amb dades experimentals. Segon, s'ha desenrotllat una Metodologia per al càlcul de la Taxa d'Amortiment amb LAPUR 6, definint una matriu de validació amb valors de taxes d'amortiment analítics i mesurats en la planta. Anàlisis de les dades mesurades en la Central Nuclear de Cofrentes mostren valors de les taxes d'amortiment inferiors al 0.3, confirmant el gran marge d'estabilitat de la Central Nuclear de Cofrentes quan se seguix un adequat procediment d'operació, i la comparació amb LAPUR mostra desviacions inferiors al +/- 0.1. L'experiència acumulada mostra que la incertesa en el rang de taxes d'amortiment baixes és normalment major que per a valors alts de les taxes d'amortiment. Finalment s'ha utilitzat un generador de senyals per a estimar la incertesa dels mètodes d'anàlisi del senyal utilitzats en este treball per a l'estimació experimental de la taxa d'amortiment emprant la funció d'autocorrelació dels senyals de potència APRM o LPRM.
Melara San Román, J. (2016). PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/61307
TESIS
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Hu, Rui Ph D. Massachusetts Institute of Technology. "Stability analysis of natural circulation in BWRs at high pressure conditions." Thesis, Massachusetts Institute of Technology, 2007. http://hdl.handle.net/1721.1/46431.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2007.
Includes bibliographical references (leaves 112-115).
At rated conditions, a natural circulation boiling water reactor (NCBWR) depends completely on buoyancy to remove heat from the reactor core. This raises the issue of potential unstable flow. oscillations. The objective of this work is to assess the characteristics of stability in a NCBWR at rated conditions, and the sensitivity to design and operating conditions in comparison to previous BWRs. Two kinds of instabilities, namely Ledinegg flow excursion and Density Wave Oscillations (DWO), have been studied. The DWO analyses were conducted for three oscillation modes: Single Channel thermal-hydraulic stability, coupled neutronics region-wide out-of-phase stability and core-wide in-phase stability. Using frequency domain methods, the three types of DWO stability characteristics of the NCBWR and their sensitivity to the operating parameters and design features have been determined. The characteristic equations are constructed from linearized equations, which are derived for small deviations around steady operating conditions. The Economic Simplified Boiling Water Reactor (ESBWR) is used in our analysis as a reference NCBWR design. It is found that the ESBWR can be stable with a large margin around the operating conditions by proper choice of the core inlet orifice scheme, and for appropriate power to flow ratios. In single channel stability analysis, neutronic feedback is neglected. Design features of the ESBWR, including shorter fuel bundle and use of part-length rods in the assemblies, tend to improve the thermal-hydraulic stability performance. However, the thermal-hydraulic stability margin is still lower than that of a typical BWR at rated conditions. In neutronic-coupled out-of-phase as well as in-phase stability analysis, the perturbation decay ratios for ESBWR at our assumed conditions are higher than that of a typical BWR (Peach Bottom 2) at rated conditions, due to its lower thermal-hydraulic stability margin and higher neutronic feedback.
(cont.) Nevertheless, the stability criteria are satisfied. To evaluate the NCBWR stability performance, comparison with BWR/Peach Bottom 2 at both the rated condition and maximum natural circulation condition has been conducted. Sensitivity studies are performed on the effects of design features and operating parameters, including chimney length, inlet orifice coefficient, power, flow rate, and axial power distribution, reactivity coefficients, fuel pellet-clad gap conductance. It can be concluded that the NCBWR and BWR stabilities are similarly sensitive to operating parameters.
by Rui Hu.
S.M.
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Luszczek, Karol. "Validation and Benchmarking of Westinghouse BWR lattice physics methods." Thesis, KTH, Reaktorteknologi, 2015. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-180563.

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A lattice physics code is a vital tool, forming a base of reactor coreanalysis. It enables the neutronic properties of the fuel assembly to becalculated and generates a proper set of data to be used by a 3-D full coresimulator. Due to advancement and complexity of modern Boiling WaterReactor assembly designs, a new deterministic lattice physics codeis being developed at Westinghouse Sweden AB, namely PHOENIX5.Each time a new code is written, its methodology of solving the neutrontransport equation, has to be validated to make sure it providesreliable output. In a wake of preparation for PHOENIX5 release andconsecutive validation efforts, a set of reference Monte Carlo calculationswas prepared, using the code Serpent. A depletion calculation with achosen type of branch cases was conducted. Methods implemented inPHOENIX5 are based on the Current Coupling Collision Probabilitymethod used in older versions of the code HELIOS. Therefore, a comparisonbetween reference Monte Carlo simulations and HELIOS 1.8.1is made, in order to discover problems inherent to the said method ofsolving the neutron transport equation. A special care should be givenduring PHOENIX5 validation, to issues highlighted in this work.Discrepancies in results of Serpent and HELIOS are attributed mostlyto disparities in the basic nuclear data used by the codes, as well as arange of approximations and corrections adopted by the deterministiccode.Serpent and HELIOS showed a good agreement in a typical voidrange (up to 90 % void) and ‘less’ challenging branches (coolant void,fuel temperature and spacer grid branches). More significant discrepanciesappeared for extreme cases with a very high void and control rodpresence (k1 differences as high as 1000 pcm) and rather pronouncedconcentrations of the natural boron dissolved in coolant (absolute differencesroughly at a level of 900 pcm). The issues do not seem to stemsolely from discrepancies in the nuclear data libraries used by Serpentand HELIOS.Moreover, a coolant void bias was consistently found in the resultsof branch calculation at changing coolant void. This confirms the analogousphenomenon found in previous studies of the CCCP based deterministiccodes. It most probably stems from the assumptions used bythe method while tackling the neutron transport equation, such as theflat source approximation, the isotropic scattering assumption and thetransport correction. An alternative transport correction approximationis proposed to alleviate this issue.
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Books on the topic "BWR1"

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Jones, Bethan Wyn. Bwrw blwyddyn. Caernarfon: Gwasg Gwynedd, 1997.

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Etherton, Roy. Bwrw lliwiau. Aberystwyth: Gwasg Cambria, 1991.

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1935-, Owen William, and Ysgol Carreg-lefn, eds. Bwrw cyfrif 'rôl canrif. Carreg-lefn: Corff Llywodraethol Ysgol Gymuned Carreg-lefn, 1999.

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Kikō, Genshiryoku Anzen Kiban. Teishiji reberu 2PSA no kentō (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2005.

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Theofanous, T. G. Performance of the liquid reactivity control system in BWRs. Washington, DC: Division of Regulatory Applications, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

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Kikō, Genshiryoku Anzen Kiban. Reberu 2 PSA shuhō no seibi (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2005.

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Kikō, Genshiryoku Anzen Kiban, ed. Jishinji reberu 2 PSA no kaiseki (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2008.

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Kikō, Genshiryoku Anzen Kiban, ed. Jishinji reberu 2 PSA no kaiseki (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2008.

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U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Licensee Performance and Quality Evaluation., ed. BWR and PWR off-normal event descriptions. Washington, DC: Division of Licensee Performance and Quality Evaluation, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1987.

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U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Licensee Performance and Quality Evaluation., ed. BWR and PWR off-normal event descriptions. Washington, DC: Division of Licensee Performance and Quality Evaluation, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1987.

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Book chapters on the topic "BWR1"

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Radvany, R. M., J. van Munster, S. Nakata, A. A. Biegel, Y. K. Paik, D. Middleton, K. Tokunaga, et al. "Antigen Society #14 Report (B40 CREG, BW60, BW61, BW41, BW48, B13)." In Immunobiology of HLA, 209–14. New York, NY: Springer New York, 1989. http://dx.doi.org/10.1007/978-1-4612-3552-1_36.

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Cattant, François. "BWRs Cracking." In Materials Ageing in Light-Water Reactors, 1889–999. Cham: Springer International Publishing, 2022. http://dx.doi.org/10.1007/978-3-030-85600-7_23.

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Prince, Robert. "Radiological Aspects of BWR Systems." In Radiation Protection at Light Water Reactors, 39–56. Berlin, Heidelberg: Springer Berlin Heidelberg, 2012. http://dx.doi.org/10.1007/978-3-642-28388-8_3.

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Laundy, Godfrey J., Marilyn S. Pollack, Martin G. Guttridge, and Peter T. Klouda. "Antigen Society #8 Report (Bw70, Bw71, Bw72)." In Immunobiology of HLA, 151–53. New York, NY: Springer New York, 1989. http://dx.doi.org/10.1007/978-1-4612-3552-1_28.

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Reynolds, R. S. "A BWR Fuel Channel Tracking System." In Artificial Intelligence and Other Innovative Computer Applications in the Nuclear Industry, 617–23. Boston, MA: Springer US, 1988. http://dx.doi.org/10.1007/978-1-4613-1009-9_75.

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Guttridge, M. G., G. J. Laundy, and P. T. Klouda. "Biochemical Variants of the Bw70 Antigen (Bw71, Bw72)." In Immunobiology of HLA, 343–44. New York, NY: Springer New York, 1989. http://dx.doi.org/10.1007/978-1-4612-3552-1_71.

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Kim, Young-Jin, and Peter L. Andresen. "Protective Insulated Coating for SCC Mitigation in BWRs." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 2103–19. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_126.

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Kim, Young-Jin, and Peter L. Andresen. "Protective Insulated Coating for SCC Mitigation in BWRs." In 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, 2103–16. Hoboken, New Jersey, Canada: John Wiley & Sons, Inc., 2012. http://dx.doi.org/10.1002/9781118456835.ch218.

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Lutz, Dan, Yang-Pi Lin, Randy Dunavant, Rob Schneider, Hartney Yeager, Aylin Kucuk, Bo Cheng, and Jim Lemons. "Hydriding Induced Corrosion Failures in BWR Fuel." In Zirconium in the Nuclear Industry: 17th Volume, 1138–71. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2014. http://dx.doi.org/10.1520/stp154320120198.

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Kim, Young-Jin, Peter L. Andresen, and Samson Hettiarachchi. "Use of Noble Metal Nanopartice for SCC Mitigation in BWRs." In Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors, 1993–2003. Cham: Springer International Publishing, 2011. http://dx.doi.org/10.1007/978-3-319-48760-1_119.

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Conference papers on the topic "BWR1"

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Yin, Shengjun, Terry L. Dickson, Paul T. Williams, and B. Richard Bass. "Stress Intensity Factor Influence Coefficients for External Surface Flaws in Boiling Water Reactor Pressure Vessels." In ASME 2009 Pressure Vessels and Piping Conference. ASMEDC, 2009. http://dx.doi.org/10.1115/pvp2009-77143.

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Over the service life of a nuclear power plant, the Boiling Water Reactor (BWR) may undergo many cool-down and heat-up thermal-hydraulic transients associated with, for example, scheduled refueling outages and other normal operational transients, or even possible overcooling transients. These thermal-hydraulic events can act on postulated surface flaws in BWRs and therefore impose potential risk on the structure integrity of Reactor Pressure Vessels (RPVs). Internal surface flaws are of interest for the BWRs under overcooling accidental scenarios, while external surface flaws are more vulnerable when the BWRs are subjected to heat-up transients. Stress Intensity Factor Influence Coefficient (SIFIC) databases for application to linear elastic fracture mechanics analyses of BWR pressure vessels which typically have an internal radius to wall thickness ratio (Ri/t) between 15 and 20 were developed for external surface breaking flaws. This paper presents three types of surface flaws necessary in fracture analyses of BWRs: (1) finite-length external surface flaws with aspect ratio of 2, 6, and 10. (2) infinite-length axial external surface flaws; and (3) 360° circumferential external surface flaws. These influence coefficients have been implemented and validated in the FAVOR fracture mechanics code developed at Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Although these SIFIC databases were developed in application to RPVs subjected to thermal-hydraulic transients, they could also be applied to RPVs under other general loading conditions.
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Tentner, Adrian, Simon Lo, Andrey Ioilev, Vladimir Melnikov, Maskhud Samigulin, Vasily Ustinenko, and Valentin Kozlov. "Advances in Computational Fluid Dynamics Modeling of Two Phase Flow in a Boiling Water Reactor Fuel Assembly." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89158.

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A new code, CFD-BWR [1], is being developed for the simulation of two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. CFD-BWR is a specialized module built on the foundation of the commercial CFD code STAR-CD [2] which provides general two-phase flow modeling capabilities. New models describing the inter-phase mass, momentum, and energy transfer phenomena specific for BWRs have been developed and implemented in the CFD-BWR module. A set of experiments focused on two-phase flow and phase-change phenomena has been identified for the validation of the CFD-BWR code and results of two experiment analyses focused on the radial void distribution are presented. The close agreement between the computed results, the measured data and the correlation results provides confidence in the accuracy of the models.
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Yan, Jin, and Andrew Mallner. "Sensitivity Study of Lower Plenum Boron Injection in a BWR." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75056.

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Boiling water reactors (BWRs) are equipped with a standby liquid control system (SLCS). The SLCS is used to inject boron to shutdown the reactor from full power condition in the event that the control rods fail to insert. In order for the SLCS system to shut down the reactor, adequate mixing of the borated solution with the reactor coolant is necessary. In BWRs prior to BWR 5, the boron injection points are located in the lower plenum. The objective of this project is to evaluate the impact of the operating conditions on the boron injection based on the understanding of the behavior of multi-species flow in typical pre-BWR 5 reactors by using computational fluid dynamics (CFD). The project is divided into two phases. At the first phase, a CFD model based on the test configurations of GE 1/6 scale test program was established. The results were validated against measurements conducted by GE during the 1/6 scale test program performed in 1981. The validation shows that the CFD can give accurate predictions of the boron mixing. The technical approach employed in the CFD model was adequate to capture the boron mixing process in the BWR lower plenum. The second phase of the project is the sensitivity study based on the same technical approach developed in the first phase. However, a simplified BWR lower plenum model was used due to the time constrain. In the sensitivity study, the baseline case and four additional cases with different operating conditions were investigated using the same CFD approach.
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Ranganath, Sampath, Robert G. Carter, Rajeshwar Pathania, Stefan Ritter, and Hans-Peter Seifert. "Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment." In ASME 2018 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/pvp2018-84257.

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Low alloy steels (LAS) used in the fabrication of reactor pressure vessel (RPV) and nozzles have been resistant to stress corrosion cracking (SCC) in the Boiling Water Reactor (BWR) environment. The plate material is SA533 Grade B and the nozzle material is SA508 Class 2 for most operating BWRs. While BWR service field experience with the LAS materials has been very good for there have been a limited number of SCC incidents where cracking has been reported especially in Alloy 182 RPV attachment (dissimilar metal) welds. This paper describes the methodology for the assessment of SCC crack growth rate (CGR) of LAS RPV components in the BWR environment. Specifically, it describes the development of CGR disposition lines (also called reference crack growth rate curves) for normal water chemistry (NWC) and hydrogen water chemistry (HWC) in BWR environments. In addition, based on more recent data from tests on the effect of chloride transients in NWC environments are also proposed.
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Morita, Ryo, Yuta Uchiyama, Fumio Inada, and Shiro Takahashi. "Considerations in Steam Piping Design for Prevention of an Acoustic Resonance at a Closed Side Branch." In ASME 2017 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/pvp2017-65244.

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Flow-induced acoustic resonances in piping with closed side branches or T-junctions are one of the causes of severe structural vibrations, which sometimes cause fatigue damage to piping and components in a power plant and many engineering applications. In this paper, on the basis of the results of steam flow experiments and calculations, the effects of the liquid phase on the flow-induced acoustic resonance at closed side branches in the steam flow piping of BWRs are described, and some suggestions for the steam piping design of BWRs are also given. The liquid phase in a steam flow forms droplets or liquid film, which may affect the amplitude, frequency and critical Strouhal number of the resonance. From the results of wet steam experiments and CFD calculations, we have found that in some cases the wetness of the steam flow may decrease the resonant amplitude and change the frequency owing to the interaction of the vortex generation or damping by the existence of the liquid film and droplets. Therefore, for the wet steam piping design of BWR, some suggestions for taking these effects into consideration, under actual BWR steam conditions are described.
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Wehle, F., A. Schmidt, S. Opel, and R. Velten. "Non-Linear BWR Stability Analysis." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48395.

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Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants already during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge oN BWR instabilities. The appropriate representation of the physical processes in the non-linear regime requires typically time domain stability analysis. The objective of this paper is to present a physical model, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. The application of this model gives a deeper insight into the physical reasons for the prevention of the uncontrolled divergence of BWR oscillations. The mechanisms that have a stabilizing effect are demonstrated.
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Reisch, Frigyes, and Hernan Tinoco. "Concept of a High Pressure Boiling Water Reactor, HP-BWR." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75032.

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Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits: 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel without numerous control rod penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340°C (15 MPa) steam instead of ∼285°C (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Steam separators inside the reactor vessel, and steam dryers together with additional separators can be installed inside or outside the containment -Simple dry containment.
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Widera, Martin. "The BWR RPV Internals Management Program of the German NPP Gundremmingen, Units B and C: Results and Conclusions." In ASME 2002 Pressure Vessels and Piping Conference. ASMEDC, 2002. http://dx.doi.org/10.1115/pvp2002-1373.

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Due to the core shroud cracks reported from numerous BWRs including the German KWU type BWR Wuergassen, a RPV internals management program for the Gundremmingen NPP (KRB-II) has been initiated in 1994. Major steps and the main results of this program are presented. As an interim result, surface condition of the weld regions and controlled post weld heat treatment (PWHT) in order to reduce the residual stresses seem to play an important role for resistance to crack initiation and growth. To support these assumptions, results of related R&D projects of the German utilities (VGB) are presented.
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Park, Pilyeon, Mirna Urquidi-Macdonald, and Digby D. Macdonald. "Application of the PDM (Point Defect Model) to the Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49098.

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The PDM [Point Defect Model, D. D. Macdonald, Pure Appl. Chem., 71, 951 (1999)] describes the corrosion of passive metals in aqueous media in terms of the generation and annihilation of point defects at the passive film interfaces. In the current work, we have modified the PDM to provide a comprehensive, atomic scale description of the growth of bilayer passive films on zirconium to simulate the corrosion of Zircaloy fuel cladding in BWRs and PWRs under high burn-up conditions. Two models have been formulated; one comprising a hydride inner (barrier) layer and an oxide outer layer and other comprising an oxide inner layer and an oxide outer layer for PWR and BWR cladding, respectively. Since there are currently no experimental data for the kinetics of defect generation and annihilation at the passive film interfaces for Zircaloys under PWR/BWR conditions, of the type that are required for this analysis, this paper focuses only on exploring and predicting trends in the corrosion behavior of Zircaloy by using prototypical values for various electrochemical parameters. We derive equations for predicting the barrier layer thickness as a function of the applied voltage, pH, porosity, and temperature for both BWR and PWR primary water chemistry conditions.
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Santamarina, A., N. Hfaiedh, R. Letellier, V. Marotte, S. Misu, A. Sargeni, C. Vaglio, and I. Zmijarevic. "Advanced Neutronics Tools for BWR Design Calculations." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89493.

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This paper summarizes the developments implemented in the new APOLLO2.8 neutronics tool to meet the required target accuracy in LWR applications, particularly void effects and pin-by-pin power map in BWRs. The Method Of Characteristics was developed to allow efficient LWR assembly calculations in 2D-exact heterogeneous geometry; resonant reaction calculation was improved by the optimized SHEM-281 group mesh, which avoids resonance self-shielding approximation below 23eV, and the new space-dependent method for resonant mixture that accounts for resonance overlapping. Furthermore, a new library CEA2005, processed from JEFF3.1 evaluations involving feedback from Critical Experiments and LWR P.I.E, is used. The specific “2005–2007 BWR Plan” settled to demonstrate the validation/qualification of this neutronics tool is described. Some results from the validation process are presented: the comparison of APOLLO2.8 results to reference Monte Carlo TRIPOLI4 results on specific BWR benchmarks emphasizes the ability of the deterministic tool to calculate BWR assembly multiplication factor within 200 pcm accuracy for void fraction varying from 0 to 100%. The qualification process against the BASALA mock-up experiment stresses APOLLO2.8/CEA2005 performances: pin-by-pin power is always predicted within 2% accuracy, reactivity worth of B4C or Hf cruciform control blade, as well as Gd pins, is predicted within 1.2% accuracy.
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Reports on the topic "BWR1"

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Ridens, Simons, and Brun. PR-316-15606-Z01 Equations of State Comparison for Pipeline Compressor Applications. Chantilly, Virginia: Pipeline Research Council International, Inc. (PRCI), July 2016. http://dx.doi.org/10.55274/r0010873.

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In order to have an improved understanding of the applicability of standard EOS in pipeline applications, a set of gas physical property tests were undertaken with sweet and sour natural gas and CO2 mixtures at typical pipeline compositions and conditions, including new high pressure dense phase applications. Specific gas properties tested include gas density (?), specific heat at constant volume (cv), and speed of sound (c or SOS). These results were compared to several of the most commonly used EOS, including NIST, GERG, AGA8, PR, SRK and BWRS, which were also compared to each other.
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A. Alsaed. BWR AXIAL PROFILE. Office of Scientific and Technical Information (OSTI), July 2005. http://dx.doi.org/10.2172/862029.

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J. Huffer. BWR AXIAL PROFILE. Office of Scientific and Technical Information (OSTI), September 2004. http://dx.doi.org/10.2172/862152.

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Lawing, Chase, Scott Palmtag, and Mehdi Asgari. BWR Progression Problems. Office of Scientific and Technical Information (OSTI), February 2021. http://dx.doi.org/10.2172/1838995.

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Tan, C. P., and G. Bagchi. BWR steel containment corrosion. Office of Scientific and Technical Information (OSTI), April 1996. http://dx.doi.org/10.2172/219387.

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Yoon, Su Jong. High Fidelity BWR Fuel Simulations. Office of Scientific and Technical Information (OSTI), August 2016. http://dx.doi.org/10.2172/1364486.

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Sutherland, W., M. Alamgir, J. Findlay, and W. Hwang. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification. Office of Scientific and Technical Information (OSTI), October 1985. http://dx.doi.org/10.2172/6349740.

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Cheng, L. Y., D. Diamond, and Gilad Raitses, Arnold Aronson Arantxa Cuadra. Trace Assessment for BWR ATWS Analysis. Office of Scientific and Technical Information (OSTI), April 2010. http://dx.doi.org/10.2172/1013471.

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Ott, L. J. (Boiling water reactor (BWR) CORA experiments). Office of Scientific and Technical Information (OSTI), October 1990. http://dx.doi.org/10.2172/6434331.

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D.P. Henderson and D.A. Salmon. Disposal Critcality Analysis Methodology: BWR Benchmarks. Office of Scientific and Technical Information (OSTI), August 1999. http://dx.doi.org/10.2172/840675.

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