Azcatl-CRP: An ant colony-based system for searching full power control rod patterns in BWRs

We show a new system named AZCATL-CRP to design full power control rod patterns in BWRs. Azcatl-CRP uses an ant colony system and a reactor core simulator for this purpose. Transition and equilibrium cycles of Laguna Verde Nuclear Power Plant (LVNPP) reactor core in Mexico were used to test Azcatl-CRP. LVNPP has 109 control rods grouped in four sequences and currently uses control cell core (CCC) strategy in its fuel reload design. With CCC method only one sequence is employed for reactivity control at full power operation. Several operation scenarios are considered, including core water flow variation throughout the cycle, target different axial power distributions and Haling conditions. Azcatl-CRP designs control rod patterns (CRP) taking into account safety aspects such as k_eff core value and thermal limits. Axial power distributions are also adjusted to a predetermined power shape.     

Effects of hydrogen water chemistry on radiation field buildup in BWRs

A number of boiling water reactor (BWR) plants worldwide are currently operating under hydrogen water chemistry (HWC). In some reactors, when switching from normal water chemistry (NWC) to HWC, an increase in the recirculation piping dose rates has been observed. Understanding the key factors which affect the dose rate increase is the subject of our current investigation. Laboratory experiments have been conducted under control chemistry conditions to examine the rates of ⁶⁰Co deposition and the characteristic of oxide films formed on stainless steel surfaces. The activity buildup data obtained from two operating BWRs are carefully reviewed and discussed in this paper. Based on both laboratory and reactor data, a plausible mechanism of enhanced activity buildup under HWC conditions is hypothesized.     

Improvement of circuit oscillation generated by underwater high voltage pulse discharges based on pulse power thyristor

High voltage fracturing technology was widely used in the field of reservoir reconstruction due to its advantages of being clean, pollution-free, and high-efficiency. However, high-frequency circuit oscillation occurs during the underwater high voltage pulse discharge process, which brings security risks to the stability of the pulse fracturing system. In order to solve this problem, an underwater pulse power discharge system was established, the circuit oscillation generation conditions were an...     

Hydrogen water chemistry for BWRs: a status report on the EPRI development program

Many boiling water reactors (BWRs) have experienced extensive intergranular stress corrosion cracking (IGSCC) in their austenitic stainless steel reactor coolant system piping, resulting in serious adverse impacts on plant capacity factors, O&M costs, and personnel radiation exposures. A major research program to provide remedies for BWR pipe cracking was co-funded by EPRI, GE, and the BWR Owners Group for IGSCC Research between 1979 and 1988. Results from this program show that the likelihood of IGSCC depends on reactor water chemistry (particularly on the concentrations of ionic impurities and oxidizing radiolysis products) as well as on material condition and the level of tensile stress. Tests have demonstrated that the concentration of oxidizing radiolysis products in the recirculating reactor water of a BWR can be reduced substantially by injecting hydrogen into the feedwater. Recent plant data show that the use of hydrogen injection can reduce the rate of IGSCC to insignificant levels if the concentration of ionic impurities in the reactor water is kept sufficiently low. This approach to the control of BWR pipe cracking is called hydrogen water chemistry (HWC). This paper presents a review of the results of EPRI's HWC development program from 1980 to the present. In addition, plans for additional work to investigate the feasibility of adapting HWC to protect the BWR vessel and major internal components from potential stress corrosion cracking problems are summarized.     

Heat conduction analyses on rewetting front propagation during transients beyond anticipated operational occurrences for BWRs

Our previous study investigated the rewetting behavior of dryout fuel surface during transients beyond anticipated operational occurrences for BWRs, which indicated the rewetting velocity was significantly affected by the precursory cooling defined as cooling immediately before rewetting. This study further investigated the previous experiments by conducting additional experimental and numerical heat conduction analyses to characterize the precursory cooling. For the characterization, the precursory cooling was first defined quantitatively based on evaluated heat transfer rates; the rewetting velocity was investigated as a function of the cladding temperature immediately before the onset of the precursory cooling. The results indicated that the propagation velocity appeared to be limited by the maximum heat transfer rate near the rewetting front. This limitation was consistent with results of the heat conduction analysis using heat transfer models for the precursory cooling expressed as a function of distance from the rewetting front, the maximum wetting temperature, and the heat transfer coefficients in the wetted region. This paper also discusses uncertainties in the evaluation of transient heat flux from the measured surface temperature, and technical issues requiring further investigation.     

Inspection findings in austenitic RPV internals of German BWR plants and BWRs built in other countries and resulting measures for Isar 1 nuclear power station

As visual examinations carried out in autumn 1994 detected cracks in a German BWR plant due to intergranular stress corrosion cracking (IGSCC) in several core shroud components manufactured from 1.4550 steel, precautionary examinations and assessments were performed for all other plants. In accordance with these analyses, it can be stated for Isar 1 that the heat treatment to which the components in question were subjected in the course of manufacture cannot have caused sensitization of the material, and that crack formation due to the damage mechanism primarily identified in the reactor vessel internals at Würgassen Nuclear Power Station need not be feared. Although the material and corrosion–chemical assessments performed to date did not give any indications for the other crack formation mechanisms that are theoretically relevant for reactor vessel internals (IGSCC due to weld sensitization, IASCC (irradiation assisted stress corrosion cracking)), visual examinations with a limited scope will be carried out with the independant expert's agreement during the scheduled inservice inspections. The fluid-dynamic and structure-mechanical analyses showed that the individual components are subjected only to low loadings, even in the event of accidents, and that the safety objectives shutdown and residual heat removal can be fulfilled even in the case of large postulated cracks. The fracture-mechanics analyses indicated critical through-wall crack lengths which, however, can be promptly and reliably detected during random inservice inspections even when assuming stress corrosion cracking and irradiation-induced low-toughness material conditions. In addition, both the VGB and the Isar 1 plant are pursuing further prophylactic measures such as alternative water chemistry modes and an appropriate repair and replacement concept.     

Prospects for developing renewable energy sources for local power generation

Unconventional renewable energy sources can and must find practical applications in our country, first and foremost, in regions with autonomous power generation. In this niche, where high power generation costs are the norm, power plants using renewal energy sources are already attractive for economic and ecological reasons.     

Raising power levels at VVÉR-440 power stations

Translated from Atomnaya Énergiya, Vol. 69, No. 2, pp. 79–81, August, 1990.     

Remarks on the inadequacy of one-group heterogeneous theories for the interpretation of incore neutron noise in BWRs

By employing the frequency-dependent source-sink method of Feinberg and Galanin, it is shown that similarly to homogeneous diffusion theory, an explicit one-group heterogeneous theory with no slowing-down taken into account is not sufficient for the interpretation of incore neutron noise measurements in BWRs or in any other situation in which the distance between the perturbation and the detector is small.It is shown that, although in contrast to homogeneous one-group theory, in the heterogeneous approach there will always be a local component associated with attenuation effects in the pure moderator, if no slowing-down effects are taken into account, either the spatial relaxation length of the global component will be under-estimated, or the spatial relaxation length of the local component will be over-estimated.     

低速重离子在热等离子体中电子能量损失的Z1振荡效应

关于离子束和等离子体相互作用的实验和理论结果现已有大量报道【’一幻。这些研究的最重要的结果就是重离子在热等离子体中的能量损失较之在冷物质中有显著增强，而且离子的有效电荷也较在冷物质中大大增加。关于在冷物质中低速重离子电子阻止本领的电荷ZI振荡在实验上和理论上都有明确的结论＊司，而在等离子体中，低速重离子阻止本领随着＆的变化还没有得到较透彻的研究。我们应用量子散射理论具体计算了热等离子体靶对低速重离子的电子阻止本领。应用散射理论，带有电荷为ZI，速度为［）的重离子贯穿热等离子体时的电子阻止本领为（dE…     

Heat accumulators at nuclear power stations

Conclusions It has been shown to be technically possible and economically desirable to introduce heat accumulators into nuclear power stations and CHPS, particularly FWA, DHWA, and DSWA. The largest power-control range is attained with HA of DHWA, PTA, and DSWA types and is 120–140% for stations with HA of the water class in relation to the nominal reactor power or up to 160% for stations with PTA.The inclusion of HA in a nuclear station provides higher PUF and consequently maintains high rates of production for secondary nuclear fuel with nuclear stations working in variableload mode. The introduction of HA at nuclear stations will enable one to fit turbines of power 1.2–1.4 times larger than the reactor power with HA of the water class or by more than a factor 1.5 for PTA, which provides a reliable means of covering peak loads by means of nuclear stations.Translated from Atomnaya Énergiya, Vol. 56, No. 6, pp. 389–396, June, 1984.     

Nuclear power technologies at the stage of sustainable nuclear power development

It is not simple to solve the problem of competitiveness of nuclear power technologies in evolutionary upgrading the conventional nuclear power plants (NPP) such as light water reactors (LWR), which requires high expenditure for safety. Moreover, the existing LWRs cannot provide nuclear power (NP) for a long time (hundreds of years) because the efficiency of use of natural uranium is low and closing the nuclear fuel cycle (NFC) for those reactors is not expedient.The highlighted problem can be solved in the way of use of innovative nuclear power technology in which natural uranium power potential is used effectively and the intrinsic conflict between economic and safety requirements has been essentially mitigated.The technology that is most available and practically demonstrated is the use of reactors SVBR-100 — small power multi-purpose modular fast reactors (100 MWe) cooled by lead-bismuth coolant (LBC). This technology has been mastered for nuclear submarines’ reactors in Russia.High technical and economical parameters of the NPP based on RF SVBR-100 are determined from the fact that the potential energy stored in LBC per a volume unit is the lowest.The compactness of the reactor facility SVBR-100 that results from integral arrangement of the primary circuit equipment allows realizing renovation of power-units LWRs, the vessels’ lifetime of which has been expired. So due to this fact, high economical efficiency can be obtained.The paper also validates the economical advantage of launching the uranium-fueled fast reactors with further changeover to the closed NFC with use of plutonium extracted from the own spent nuclear fuel in comparison with launching fast reactors directly with on uranium-plutonium fuel on the basis of plutonium extraction from spent nuclear fuel of LWRs.     

Experimental reactor noise—A review on noise-analytic measurements of thermohydraulic parameters in operating BWRs and their interpretations

This article reviews the results and interpretations of measurements inferred from neutron-noise in BWRs with respect to their applicability for Code-Verification, Monitoring and Anomaly-Detection, which are the final goals of neutron-noise investigations in BWRs.It will be shown that for the time-being, none of these goals have been reached and furthermore, in the near future, rigorous code-verification by means of neutron-noise in operating BWRs (measurements of fluid velocity and voidfraction as well as flow-pattern identification), due to the complexity of the system (the measured values are unknown types of averages of the flows in four 8 × 8 bundles) seems to be rather difficult.     

Critical heat flux under flow oscillation of water at low-pressure, low-flow conditions

In this work, the effect of flow oscillations on critical heat flux (CHF) is investigated for water flow in vertical round tubes at low-pressure, low-flow (LPLF) conditions. An experimental study has been conducted to investigate the difference in CHF between forced and natural circulations, and between stable and oscillating flow conditions with three vertical round tube test sections (5.0 mm ID×0.6 m in length, 6.6 mm ID×0.5 m in length, and 9.8 mm ID×0.6 m in length) for mass fluxes below 400 kg m⁻² s⁻¹ under near atmospheric pressure. It is found that flow oscillations can drastically reduce the CHF, in particular for natural-circulation conditions. In addition to the experiments, CHF correction factors for flow oscillation effects are developed for forced and natural circulations, respectively, based on the experimental data of the present work and others.