Characterization of deposits on inside surfaces of lwr cladding

Samples of Zircaloy cladding from UO₂ fuel rods, irradiated in water-cooled power reactors under two regimes of operation, were examined in a scanning electron microscope to determine the nature of the deposits formed on the inner surfaces of the cladding. In one case, the fuel rods examined had operated under conditions that resulted in significant fission gas release and solid fission product deposition on the cladding walls. In the other case, the rod examined had operated at a low power to a large burnup but with almost no fission gas release. That cladding was remarkably free of solid deposits of fission product. The likely mechanism of transport of fission products from the fuel to the cladding was through the vapor. The volatilities of the fission product compounds were estimated on the basis of the compounds expected to be present in the light of the oxygen potential calculated for the system.     

A method for statistical steady state thermal analysis of reactor cores

This paper presents a method for performing a statistical steady state thermal analysis of a reactor core. The technique is only outlined here since detailed thermal equations are dependent on the core geometry. The method has been applied to a pressurised water reactor core and the results are presented for illustration purposes.Random hypothetical cores are generated using the Monte-Carlo method. The technique shows that by splitting the parameters into two types, denoted core-wise and in-core, the Monte Carlo method may be used inexpensively. The idea of using extremal statistics to characterise the low probability events (i.e. the tails of a distribution) is introduced together with a method of forming the final probability distribution. After establishing an acceptable probability of exceeding a thermal design criterion, the final probability distribution may be used to determine the corresponding thermal response value. If statistical and deterministic (i.e. conservative) thermal response values are compared, information on the degree of pessimism in the deterministic method of analysis may be inferred and the restrictive performance limitations imposed by this method relieved.     

Natural convection analyses and verification for LMFBR cores

Analytical model requirements for core natural convection analyses are reviewed. Then results from current modeling on intra-assembly flow and heat redistribution are compared with several sources of experimental data. Also, data are described on low flow rod bundle hydraulic characteristics. Numerous sensitivity studies are also presented which show the effect and importance of various parameters on core temperatures during natural circulation, including inter-assembly flow redistribution, pump flow coastdown, rod size and fuel type, control system scram worth and shutdown power level. A system of codes for making the natural circulation predictions is also described, i.e., a plant-wide dynamic code, a whole-core system dynamic code and a hot channel dynamic analysis code. The overall approach of verifying the core related codes is presented, along with the interaction and linkage between all the codes. Confirmation of this system of three codes will bee through prototypic data obtained from EBR-II and FFTF natural circulation experiments.     

Fuel performance analysis for PWR cores

Attainable discharge burnups for oxide and hydride fuels in PWR cores were investigated using the TRANSURANUS fuel performance code. Allowable average linear heat rates and coolant mass fluxes for a set of fuel designs with different fuel rod diameters and pitch-to-diameter ratios were obtained by VIPRE and adopted in the fuel code as boundary conditions. TRANSURANUS yielded the maximum rod discharge burnups of the several design combinations, under the condition that specific thermal-mechanical fuel rod constraints were not violated. The study shows that independent of the fuel form (oxide or hydride) rods with (a) small diameters and moderate P/Ds or (b) large diameters and small P/Ds give the highest permissible burnups limited by the rod thermal-mechanical constraints. TRANSURANUS predicts that burnups of ∼74 MWd/kg U and ∼163 MWd/kg U (or ∼65.2 MWd/kg U oxide-equivalent) could be achieved for UO₂ and UZrH_x fuels, respectively. Furthermore, for each fuel type, changing the enrichment has only a negligible effect on the permissible burnup. The oxide rod performance is limited by internal pressure due to fission gas release, while the hydride fuel can be limited by excessive clad deformation in tension due to fuel swelling, unless the fuel rods will be designed to have a wider liquid metal filled gap. The analysis also indicates that designs featuring a relatively large number of fuel rods of relatively small diameters can achieve maximum burnup and provide maximum core power density because they allow the fuel rods to operate at moderate to low linear heat rates.     

Thermal-hydraulic analysis for wire-wrapped PWR cores

This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH_1.6 (referred to as U-ZrH_1.6) or UO₂ fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA).The thermal-hydraulic performance of U-ZrH_1.6 and UO₂ were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.     

A numerical method for predicting seismic-induced impact between subassemblies in LMFBR cores

Structural problems that incorporate impact usually require small time steps for numerical integration and, consequently, solutions are very expensive. Since impact is one of the basic phenomena involved in any study of fast breeder reactor cores subjected to seismic disturbances, it is imperative that simplified models be introduced to make safety studies economically feasible. An approach is proposed in this paper whereby just the magnitude of each impulse and not the history of the impact force is determined for each impact. The consequence of the procedure is that the maximum time step is governed by system parameters and not by a detailed characteristic of the contact region such as an equivalent contact spring. The concepts are developed with the use of a single-degree-of-freedom model and the approach is applied to a three-assembly model of a reactor core. To obtain a solution, it is shown that the proposed direct approach may result in computer time that is less by an order of magnitude then the time required by the more conventional contact spring and gap element method.     

Applications of ANNs in flow and heat transfer problems in nuclear engineering: A review work

Artificial Neural Networks (ANNs) have been applied to deal with flow and heat transfer problems over the past two decades. In the present paper, recent work on the applications of ANNs for predicting the flow regime, pressure drop, void fraction, critical heat flux, onset of nucleate boiling, heat transfer coefficient and boiling curve has been reviewed, respectively. As can be noted in this review work, various types of ANNs can be employed as predictors with acceptable precisions. At the end of this review, methods to improve performance of ANNs and further applications of ANNs in flow and heat transfer problems were introduced.     

Environmentally assisted crack growth in low alloy steels — results and their relevance to lwr components

The paper describes the influence of corrosion on crack initiation and crack growth in low-alloy steels in high temperature water and the relevance of data determined by corrosion tests to the component behaviour. The test results, gained by laboratory experiments and the literature data available are analyzed and evaluated with respect to the in-service conditions of the components (e.g. RPV, feedwater-, main steam-line etc.). As a result of this evaluation, it can be stated that due to the current boundary conditions the RPV and other important components of the primary circuits in the water/steam cycle of LWR's are not endangered by stress corrosion cracking.     

Magnetic alloy cores for the HIRFL-CSRm compressor cavity

Four types of magnetic alloy cores,labeled as V1,V2,A1 and A2,were produced by Liyuan Corp.Ltd.,for the radio frequency compression cavity of HIRFL-CSRm.In this work,their permeability,quality factor (Q value) and shunt impedance were measured before installing them into the cavity.The results show that the V1,V2 and A2 have higher permeability and shunt impedance,and lower Q value,and are suitable to the radio frequency compression cavity.     

A method for statistical steady state thermal analysis of reactor cores: Application and assessment of confidence limits

In a previous publication [1] the author presented a method for undertaking statistical steady state thermal analyses of reactor cores. The present paper extends the technique to an assessment of confidence limits for the resulting probability functions which define the probability that a given thermal response value will be exceeded in a reactor core. Establishing such confidence limits is considered an integral part of any statistical thermal analysis and essential if such analyses are to be considered in any regulatory process. In certain applications the use of a best estimate probability function may be justifiable but it is recognised that a demonstrably conservative probability function is required for any regulatory considerations.     

Thermal-hydraulic analysis techniques for axisymmetric pebble bed reactor cores

A mathematical model for the analysis of coupled thermal-hydraulic problemsin steady-state pebble bed nuclear reactor cores is presented. The bed has been treated macroscopically as a generating, conducting porous medium. The model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, and new coefficients of the viscous and inertial loss terms are presented. The remaining equations in the model make use of continuity and thermal energy balances on the solid and fluid phases. None of the usual simplifying assumptions such as constant properties, constant velocity flow or negligible conduction and/or radiation are used. A computer program based on this model has been constructed; it has been validated by comparing predictions with measured values of previous experiments. Validation of the nonlinear fluid flow model is reported in a companion paper.     

An improved thermal-hydraulic analysis method for rod bundle cores

A new thermal-hydraulic computational procedure, which employs the perturbation method, is described for pressurized water reactor design and analysis. This THINC-IV computer program calculates the three-dimensional coolant and enthalpy behavior in an open lattice reactor core at steady-state flow and power conditions. Design evaluation of Westinghouse PWRs is carried out using this program with the correlations described herein. The THINC-IV calculation scheme differs from existing thermal-hydraulic programs in that lateral momentum equations, including both inertial and frictional effects are incorporated. A perturbation technique is employed to solve the momentum equations in the three coordinate directions. The perturbed axial velocity is obtained by solution of a field equation formed from the perturbed momentum equations. The crossflow velocities are then determined from the lateral momentum equations and are used with the complete continuity and energy equations to solve for the three-dimensional enthalpy distribution in the core. Calculations performed by this method compared well with measurements for five available experiments, giving confidence in its use for PWR analysis.     

A study on reactivity insertion controlled LMR cores with metallic fuel

An inherently safe core concept with metallic fuel for sodium cooled fast reactor is proposed that has a negative void reactivity at the loss of coolant events without scram as well as a small excess reactivity during the operation cycle. The relationship of sodium void reactivities and burn-up reactivities to different core configurations has been studied quantitatively to clarify the core concept for large metallic fuel reactors. It has shown that the sodium void reactivity is greatly dependent on the core shapes while the excess reactivity is on the fuel compositions. It has also indicated that the core configuration that enables to enhance the neutron streaming through the region above the active core at coolant voiding is most effective to decrease sodium void reactivity.

A 3000 MWt core composed of the flat inner core and annular outer core where the fuel volume fraction is relatively high and the sodium plenum is placed just above the active core region has been selected as a candidate core.

The maximum excess reactivity of the candidate core at UTOP is about 0.4 $ and it can be reduced to approximately zero by power or inlet temperature adjustment during the operation cycle, meanwhile the sodium void reactivity is as low as -1.3 $ in negative that is enough to prevent ULOF sequences.     

Heterogeneous cores for improved safety performance: A case study: The supercritical water fast reactor

Light water reactor (LWR) technology is nowadays the most successful commercial application of fission reactors for the production of electricity. However, in the next few years, nuclear industry will have to face new and demanding challenges: the need for sustainable and cheap sources of energy, the need for public acceptance, the need for even higher safety standards, the need to minimize the waste production are only a few examples. It is for these very reasons that a few next generation nuclear reactor concepts were selected for extensive research and development; super critical water reactors are among them. The use of a supercritical coolant would allow for higher thermal efficiencies and a more compact plant design, since steam generators, or steam separators and driers would not be needed, hence achieving a better economy. Moreover, because of the high heat capacity of supercritical water, relatively less coolant would be needed to refrigerate the reactor, therefore the feasibility to design a water cooled fast reactor: the supercritical water fast reactor (SCFR). This system presents unique features combining well-known fast and light water reactor characteristics in one design (e.g. a tendency to a positive void reactivity coefficient together with loss of coolant accident – LOCAs as a design basis accident). The core is in fact loaded with highly enriched MOX fuel (average plutonium content of 23%), and presents a peculiar and significant geometrical and material heterogeneity (use of radial and axial blankets, solid moderator layers, 12 different enrichment zones). The safety analysis of this very complex core layout, together with the optimization of the void reactivity effect through core design, is the main objective of this work.     

Nonlinear dynamic analysis of prismatic elements for high-temperature gas-cooled reactor cores

The high-temperature gas-cooled reactor (HTGR) core consists of several thousand prismatic graphite fuel elements arranged in columns within a prestressed concrete vessel. A major research and development effort was initiated in 1970 at General Atomic Company to study the dynamic response of the HTGR core arrangement to seismic excitation.This paper presents a discussion of the history and some of the results of this effort, with respect to advances made in the development of analytical methods. The computer programs developed to perform the analysis are described, along with certain techniques and the modeling required to utilize them. The purpose is to describe the nonlinear dynamic analysis techniques employed to analyze the HTGR core. Correlation of the codes is beyond the scope of the paper and will be discussed in subsequent publications.Each fuel column in the HTGR core is composed of stacked elements doweled together to ensure alignment of the coolant channels. Gaps exist between columns, allowing the elements to impact during a seismic disturbance. Analysis of this type of structure by standard structural dynamics techniques is not possible since both nonlinearities and discontinuities exist. One- and two-dimensional models of the three-dimensional core have been developed with explicit time integration methods. Various methods to treat the impact between elements are discussed.Three computer codes were developed. CRUNCH-1D models a one-dimensional horizontal strip through the core; CRUNCH-2D, a two-dimensional horizontal planar section; and MCOCO, a two-dimensional vertical planar section. The dynamic characteristics of these three representations of the full core structure are compared and the methods evaluated in the text. Plans for additional development and work to improve the techniques are also discussed.     

Thermal hydraulic analysis for grid supported pressurized water reactor cores

This paper presents the methodology and results for thermal hydraulic analysis of grid supported pressurized water reactor cores using U(45% wt)-ZrH_1.6 hydride fuel in square arrays. The same methodology is applied to the design of UO₂ oxide fueled cores to provide a fair comparison of the achievable power between the two fuel types. Steady-state and transient design limits are considered. Steady-state limits include: fuel bundle pressure drop, departure from nucleate boiling ratio, fuel temperature (average for UO₂ and centerline/peak for U-ZrH_1.6), and fuel rod vibrations and wear. Transient limits are derived from consideration of the loss of flow and loss of coolant accidents, and an overpower transient.In general, the thermal hydraulic performance of U-ZrH_1.6 and UO₂ fuels is very similar. Slight power differences exist between the two fuel types for designs limited by rod vibrations and wear, because these limits are fuel dependent. Large power increases are achievable for both fuels when compared to the reference core power output of 3800 MW_th. In general, these higher power designs have smaller rod diameters and larger pitch-to-diameter ratios than the reference core geometry. If the pressure drop across new core designs is limited to the pressure drop across the reference core, power increases of ∼400 MW_th may be realized. If the primary coolant pumps and core internals could be designed to accommodate a core pressure drop equal to twice the reference core pressure drop, power increases of ∼1000 MW_th may be feasible.