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Статті в журналах з теми "BWR1"
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.
Повний текст джерелаLU, 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.
Повний текст джерелаAltiok, 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.
Повний текст джерелаOlvera-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.
Повний текст джерелаLANGE, 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.
Повний текст джерелаCui, 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.
Повний текст джерелаMacdonald, 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.
Повний текст джерелаHarder, James W., Jing Ma, Pascale Alard, Kevin J. Sokoloski, Edith Mathiowitz, Stacia Furtado, Nejat K. Egilmez та 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, № 1 (10 січня 2023): 41–57. http://dx.doi.org/10.1093/jleuko/qiac002.
Повний текст джерелаHarder, James W., Jing Ma, Pascale Alard, Xiang Zhang, Fang Yuan та 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, № 1_Supplement (1 травня 2021): 105.04. http://dx.doi.org/10.4049/jimmunol.206.supp.105.04.
Повний текст джерелаTeguh 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.
Повний текст джерелаДисертації з теми "BWR1"
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.
Повний текст джерелаChun, John Hwan. "Modeling of BWR water chemistry." Thesis, Massachusetts Institute of Technology, 1990. http://hdl.handle.net/1721.1/13660.
Повний текст джерелаSoma, Kovács István. "Simplified Simulator for BWR Instabilities." Thesis, KTH, Fysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-210626.
Повний текст джерелаFerroni, 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.
Повний текст джерелаThesis (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.
Morra, Paolo. "Design of annular fuel for high power density BWRs." Thesis, Massachusetts Institute of Technology, 2004. http://hdl.handle.net/1721.1/34448.
Повний текст джерелаIncludes 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.
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.
Повний текст джерелаIncludes 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.
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.
Повний текст джерелаQC 20101126
Melara, San Román José. "PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR." Doctoral thesis, Universitat Politècnica de València, 2016. http://hdl.handle.net/10251/61307.
Повний текст джерела[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
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.
Повний текст джерелаIncludes 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.
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.
Повний текст джерелаКниги з теми "BWR1"
Jones, Bethan Wyn. Bwrw blwyddyn. Caernarfon: Gwasg Gwynedd, 1997.
Знайти повний текст джерелаEtherton, Roy. Bwrw lliwiau. Aberystwyth: Gwasg Cambria, 1991.
Знайти повний текст джерела1935-, Owen William, and Ysgol Carreg-lefn, eds. Bwrw cyfrif 'rôl canrif. Carreg-lefn: Corff Llywodraethol Ysgol Gymuned Carreg-lefn, 1999.
Знайти повний текст джерелаKikō, Genshiryoku Anzen Kiban. Teishiji reberu 2PSA no kentō (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2005.
Знайти повний текст джерелаTheofanous, 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.
Знайти повний текст джерелаKikō, Genshiryoku Anzen Kiban. Reberu 2 PSA shuhō no seibi (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2005.
Знайти повний текст джерелаKikō, Genshiryoku Anzen Kiban, ed. Jishinji reberu 2 PSA no kaiseki (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2008.
Знайти повний текст джерелаKikō, Genshiryoku Anzen Kiban, ed. Jishinji reberu 2 PSA no kaiseki (BWR). [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2008.
Знайти повний текст джерелаU.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.
Знайти повний текст джерелаU.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.
Знайти повний текст джерелаЧастини книг з теми "BWR1"
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.
Повний текст джерелаCattant, 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.
Повний текст джерелаPrince, 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.
Повний текст джерелаLaundy, 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.
Повний текст джерелаReynolds, 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.
Повний текст джерелаGuttridge, 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.
Повний текст джерелаKim, 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.
Повний текст джерелаKim, 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.
Повний текст джерелаLutz, 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.
Повний текст джерелаKim, 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.
Повний текст джерелаТези доповідей конференцій з теми "BWR1"
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.
Повний текст джерелаTentner, 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.
Повний текст джерелаYan, 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.
Повний текст джерелаRanganath, 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.
Повний текст джерелаMorita, 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.
Повний текст джерелаWehle, 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.
Повний текст джерелаReisch, 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.
Повний текст джерелаWidera, 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.
Повний текст джерелаPark, 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.
Повний текст джерелаSantamarina, 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.
Повний текст джерелаЗвіти організацій з теми "BWR1"
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.
Повний текст джерелаA. Alsaed. BWR AXIAL PROFILE. Office of Scientific and Technical Information (OSTI), July 2005. http://dx.doi.org/10.2172/862029.
Повний текст джерелаJ. Huffer. BWR AXIAL PROFILE. Office of Scientific and Technical Information (OSTI), September 2004. http://dx.doi.org/10.2172/862152.
Повний текст джерелаLawing, 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.
Повний текст джерелаTan, 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.
Повний текст джерелаYoon, Su Jong. High Fidelity BWR Fuel Simulations. Office of Scientific and Technical Information (OSTI), August 2016. http://dx.doi.org/10.2172/1364486.
Повний текст джерелаSutherland, 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.
Повний текст джерелаCheng, 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.
Повний текст джерелаOtt, L. J. (Boiling water reactor (BWR) CORA experiments). Office of Scientific and Technical Information (OSTI), October 1990. http://dx.doi.org/10.2172/6434331.
Повний текст джерелаD.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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