While these samples are representative of the content of Science.gov,

they are not comprehensive nor are they the most current set.

We encourage you to perform a real-time search of Science.gov

to obtain the most current and comprehensive results.

Last update: August 15, 2014.

1

Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations

This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.

Wagner, J.C.; DeHart, M.D.

2000-03-01

2

Detailed Burnup Calculations for Testing Nuclear Data

NASA Astrophysics Data System (ADS)

A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burnup calculations, using some of the main available evaluated nuclear data files (ENDF-B-VI-Rel.8, JEFF-3.0, JENDL-3.3), on an isotope-by-isotope basis as much as possible. The selected experimental burnup benchmarks are reference cases for LWR and HWR reactors, with analysis of isotopic composition as a function of burnup. For LWR (H2O-moderated uranium oxide lattices) four benchmarks are included: ATM-104 NEA Burnup credit criticality benchmark; Yankee-Rowe Core V; H.B.Robinson Unit 2 and Turkey Point Unit 3. For HWR (D2O-moderated uranium oxide cluster lattices), three benchmarks were selected: NPD-19-rod Fuel Clusters; Pickering-28-rod Fuel Clusters; and Bruce-37-rod Fuel Clusters. The isotopes with experimental concentration data included in these benchmarks are: Se-79, Sr90, Tc99, Ru106, Sn126, Sb125,1129, Cs133-137, Nd143, 145, Sm149-150, 152, Eul53-155, U234-235, 238, Np237, Pu238-242, Am241-243, and Cm242-248. Results and analysis of differences between calculated and measured absolute and/or relative concentrations of these isotopes for the seven benchmarks are included in this work.

Leszczynski, F.

2005-05-01

3

NASA Astrophysics Data System (ADS)

A rock-like oxide (ROX) fuel - light water reactor (LWR) burning system has been studied for efficient plutonium transmutation. For the improvement of small negative reactivity coefficients and severe transient behaviors of ROX fueled LWRs, a partial loading core of ROX fuel assemblies with conventional UO 2 assemblies was considered. As a result, although the reactivity coefficients could be improved, the power peaking tends to be large in this heterogeneous core configuration. The reactivity initiated accident (RIA) and loss of coolant accident (LOCA) behaviors were not sufficiently improved. In order to reduce the power peaking, the fuel composition and the assembly design of the ROX fuel were modified. Firstly, erbium burnable poison was added as Er 2O 3 in the ROX fuel to reduce the burnup reactivity swing. Then pin-by-pin Pu enrichment and Er content distributions within the ROX fuel assembly were considered. In addition, the Er content distribution was also considered in the axial direction of the ROX fuel pin. With these modifications, a power peaking factor even lower than the one in a conventional UO 2 fueled core can be obtained. The RIA and LOCA analyses of the modified core have also shown the comparable transient behaviors of ROX partial loading core to those of the UO 2 core.

Akie, H.; Sugo, Y.; Okawa, R.

2003-06-01

4

Power excursion analysis for high burnup cores

A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report.

Diamond, D.J.; Neymotin, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)

1996-02-01

5

Triton burnup measurements and calculations on TFTR

NASA Astrophysics Data System (ADS)

Measurements of the burnup of fusion product tritons in TFTR are presented. Interpretation of triton burnup experiments requires three accurate components: the measurement of the 2.5 MeV neutron emission, the measurement of the 14 MeV neutron emission and a calculation of the expected burnup ratio from the measured plasma parameters. The absolute calibration for the 14 MeV neutron measurements is provided by an NE213 proton recoil spectrometer. Time dependent burnup measurements for three plasma conditions selected for optimum detector operation are shown. Measurements of the time integrated triton burnup from copper activation foils (cross-calibrated to the NE213 measurements) are presented. Descriptions are provided of the neutron detectors and the plasma diagnostics whose data are used as input to the calculation of the expected burnup. All these measurements find that the triton burnup on TFTR is 1/2 +/- 1/4 the classical expectations for a wide variety of discharges. The burnup decreases for relatively longer triton slowing down times, implying possible fast ion diffusion coefficients of ~0.1 m2/s. Alternatively, burnup appears to decrease with increasing major radius of the triton source and edge safety factor qcyl, implying that ripple losses may be playing a role. Triton burnup is a very sensitive measure of anomalous fast ion transport; similar levels of diffusive transport in an ignited reactor would have minimal impact on the alpha particles.

Barnes, C. W.; Bosch, H.-S.; Hendel, H. W.; Huibers, A. G. A.; Jassby, D. L.; Motley, R. W.; Nieschmidt, E. B.; Saito, T.; Strachan, J. D.; Bitter, M.; Budny, R. V.; Hill, K. W.; Mansfield, D. K.; McCune, D. C.; Nazikian, R.; Park, H. K.; Ramsey, A. T.; Scott, S. D.; Taylor, G.; Zarnstorff, M. C.

1998-04-01

6

Detailed Burnup Calculations for Testing Nuclear Data

A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full

F. Leszczynski

2005-01-01

7

Triton burnup measurements and calculations on TFTR

Measurements of the burnup of fusion product tritons in TFTR are presented. Interpretation of triton burnup experiments requires three accurate components: the measurement of the 2.5 MeV neutron emission, the measurement of the 14 MeV neutron emission and a calculation of the expected burnup ratio from the measured plasma parameters. The absolute calibration for the 14 MeV neutron measurements is

C. W. Barnes; H.-S. Bosch; H. W. Hendel; A. G. A. Huibers; D. L. Jassby; R. W. Motley; E. B. Nieschmidt; T. Saito; J. D. Strachan; M. Bitter; R. V. Budny; K. W. Hill; D. K. Mansfield; D. C. McCune; R. Nazikian; H. K. Park; A. T. Ramsey; S. D. Scott; G. Taylor; M. C. Zarnstorff

1998-01-01

8

Sensitivity Study of Fuel Cost in Extended Burnup BWR Core

A sensitivity study on the fuel cost of an extended burnup BWR core has been carried out on the basis of a realistic model of discharge burnup extension. Full power operating length in months in a refueling cycle and the number of refueling batches are chosen as independent variables in the model to describe extended burnup cores of various types.

Yasuhiro KOBAYASHI; Kikuo UMEGAKI

1984-01-01

9

Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor

The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations

M. Q. Huda; S. I. Bhuiyan; T. Obara

2008-01-01

10

Burnup calculation methodology in the serpent 2 Monte Carlo code

This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)

Leppaenen, J. [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Isotalo, A. [Aalto Univ., Dept. of Applied Physics, P.O.Box 14100, FI-00076 AALTO (Finland)

2012-07-01

11

TRIGA fuel burn-up calculations and its confirmation

The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and

R. Khan; S. Karimzadeh; H. Böck

2010-01-01

12

Fast reactor 3D core and burnup analysis using VESTA

Burnup analyses using the VESTA code have been performed on a MOX-fuelled fast reactor model as specified by an IAEA computational benchmark. VESTA is a relatively new code that has been used for burnup credit calculations and thermal reactor models, but not typically for fast reactor applications. The detailed input and results of the IAEA benchmark provides an opportunity to gauge the use of VESTA in a fast reactor application. VESTA employs an ultra-fine multi-group binning approach that accelerates Monte Carlo burnup calculations. Using VESTA to compute the end of cycle (EOC) power fractions by enrichment zone showed agreement with the published values within 5%. When comparing the ultra-fine multi-group binning approach to the tally-based approach, EOC isotopic masses also agree within 5%. Using the ultra-fine multi-group binning approach, we obtain a wall-time speedup factor of 35 when compared to the tally-based approach for computing a k{sub eff} eigenvalue with burnup problem. The authors conclude the use of VESTA's ultra-fine multi-group binning approach with Monte Carlo transport performs accurate depletion calculations for this fast reactor benchmark. (authors)

Luciano, N.; Shamblin, J.; Maldonado, I. [Nuclear Engineering Dept., Univ. of Tennessee, Knoxville, TN 37996-2300 (United States)

2012-07-01

13

MONTE-CARLO BURNUP CALCULATION UNCERTAINTY QUANTIFICATION AND PROPAGATION DETERMINATION

MONTEBURNS is a Monte-Carlo depletion routine utilizing MCNP and ORIGEN 2.2. Uncertainties exist in the MCNP transport calculation, but this information is not passed to the depletion calculation in ORIGEN or saved. To quantify this transport uncertainty and determine how it propagates between burnup steps, a statistical analysis of a multiple repeated depletion runs is performed. The reactor model chosen is the Oak Ridge Research Reactor (ORR) in a single assembly, infinite lattice configuration. This model was burned for a 25.5 day cycle broken down into three steps. The output isotopics as well as effective multiplication factor (k-effective) were tabulated and histograms were created at each burnup step using the Scott Method to determine the bin width. It was expected that the gram quantities and k-effective histograms would produce normally distributed results since they were produced from a Monte-Carlo routine, but some of results do not. The standard deviation at each burnup step was consistent between fission product isotopes as expected, while the uranium isotopes created some unique results. The variation in the quantity of uranium was small enough that, from the reaction rate MCNP tally, round off error occurred producing a set of repeated results with slight variation. Statistical analyses were performed using the {chi}{sup 2} test against a normal distribution for several isotopes and the k-effective results. While the isotopes failed to reject the null hypothesis of being normally distributed, the {chi}{sup 2} statistic grew through the steps in the k-effective test. The null hypothesis was rejected in the later steps. These results suggest, for a high accuracy solution, MCNP cell material quantities less than 100 grams and greater kcode parameters are needed to minimize uncertainty propagation and minimize round off effects.

Nichols, T.; Sternat, M.; Charlton, W.

2011-05-08

14

Nuclear reactor fuel cell burnup calculations; an efficient method

A method of cell calculation for thermal neutrons has been developed for the simulation of a light water reactor fuel rod burnup. This method is based on the analytical solution of diffusion equation in the clad and moderator regions and numerical solution of integral transport equation in the fuel region. The accuracy of this technique has been verified by considering one energy group and two region fuel unit cell. The results of thermal disadvantage factor and neutron flux profile are obtained for various test problems. These results are then compared with results obtained from several other techniques of varying sophistication. Also included are the results of one energy group, three region unit cell. The present method provides satisfactory results, not only for the disadvantage factor, but also for the flux. This technique has been then implemented in the one dimensional burnup computer code MLASER, as well as the original LASER code. The new versions DTMLASER and DTLASER developed in this research have been applied to the sample problem supplied with the original LASER code. The new codes lead to a more efficient calculation resulting in saving of more than two-thirds of computational time for accuracy comparable to that of the previous codes.

Abdullah, K.M.S.

1989-01-01

15

Burnup concept for a long-life fast reactor core using MCNPX.

This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

2013-02-01

16

Accident source terms for boiling water reactors with high burnup cores.

The primary objective of this report is to provide the technical basis for development of recommendations for updates to the NUREG-1465 Source Term for BWRs that will extend its applicability to accidents involving high burnup (HBU) cores. However, a secondary objective is to re-examine the fundamental characteristics of the prescription for fission product release to containment described by NUREG-1465. This secondary objective is motivated by an interest to understand the extent to which research into the release and behaviors of radionuclides under accident conditions has altered best-estimate calculations of the integral response of BWRs to severe core damage sequences and the resulting radiological source terms to containment. This report, therefore, documents specific results of fission product source term analyses that will form the basis for the HBU supplement to NUREG-1465. However, commentary is also provided on observed differences between the composite results of the source term calculations performed here and those reflected NUREG-1465 itself.

Gauntt, Randall O.; Powers, Dana Auburn; Leonard, Mark Thomas

2007-11-01

17

Calculations on fission gas behaviour in the high burnup structure

The behaviour of fission gas in high burnup fuel during steady-state and transient conditions is of special interest for safety reasons. Despite this, mechanistic models that reflect the fission gas transport processes and reliably predict the evolution of the remaining fission gas in the high burnup structure (HBS) are largely missing today. We start to address this problem by developing

P. Blair; A. Romano; Ch. Hellwig; R. Chawla

2006-01-01

18

Using Laguerre polynomials to compute the matrix exponential in burnup calculations

An essential part of burnup analysis is to solve the burnup equations. The burnup equations can be regarded as a first-order linear system and solved by means of matrix exponential methods. Because of its large spectrum, it is difficult to compute the exponential of the burnup matrix. Conventional methods of computing the matrix exponential, such as the truncated Taylor expansion and the Pade approximation, are not applicable to burnup calculations. Recently the Chebyshev Rational Approximation Method (CRAM) has been applied to solve burnup matrix exponential and shown to be robust and accurate. However, the main defect of CRAM is that its coefficients are not easy to obtain. In this paper, an orthogonal polynomial expansion method, called Laguerre Polynomial Approximation Method (LPAM), is proposed to compute the matrix exponential in burnup calculations. The polynomial sequence of LPAM can be easily computed in any order and thus LPAM is quite convenient to be utilized into burnup codes. Two typical test cases with the decay and cross-section data taken from the standard ORIGEN 2.1 libraries are calculated for validation, against the reference results provided by CRAM of 14 order. Numerical results show that, LPAM is sufficiently accurate for burnup calculations. The influences of the parameters on the convergence of LPAM are also discussed. (authors)

She, D.; Zhu, A.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)

2012-07-01

19

The Large Core Code Evaluation Working Group Benchmark Problem Four was specified to provide a stringent test of the current methods which are used in the nuclear design and analyses process. The benchmark specifications provided a base for performing detailed burnup calculations over the first two irradiation cycles for a large heterogeneous fast reactor. Particular emphasis was placed on the

C. L. Cowan; R. Protsik; J. W. Lewellen

1984-01-01

20

Development and verification of fuel burn-up calculation model in a reduced reactor geometry

A fuel burn-up model in a reduced reactor geometry (2-D) is successfully developed and implemented in the Batan in-core fuel management code, Batan-FUEL. Considering the bank mode operation of the control rods, several interpolation functions are investigated which best approximate the 3-D fuel assembly radial power distributions across the core as function of insertion depth of the control rods. Concerning

Tagor Malem Sembiring; Peng Hong Liem

2008-01-01

21

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa [Korea Atomic Energy Research Institute, 150 Dukjin-dong, Yusung-gu, Taejon (Korea, Republic of)

2005-05-24

22

Group Constants Generation of the Pseudo Fission Products for Fast Reactor Burnup Calculations

NASA Astrophysics Data System (ADS)

The pseudo fission products for the burnup calculations of the liquid metal fast reactor were generated. The cross-section data and fission product yield data of ENDF/B-VI were used for the pseudo fission product data of U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242. The pseudo fission product data can be used with the KAFAX-F22 or -E66, which are the MATXS-format libraries for analyses of the liquid metal fast reactor at KAERI and were distributed through the OECD/NEA. The 80-group MATXS-format libraries of the 172 fission products were generated and the burnup chains for generation of the pseudo fission products were prepared.

Gil, Choong-Sup; Kim, Do Heon; Chang, Jonghwa

2005-05-01

23

Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B.

Richard G. Ambrosek; Gray S. Chang; Debbie J. Utterbeck

2005-02-01

24

NASA Astrophysics Data System (ADS)

For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

2014-06-01

25

Standard reactor simulation codes WIMS-D\\/4 and CITATION were employed to analyze the proliferation resistance potential and burnup characteristics of a novel natural uranium fueled nuclear research reactor [Annals of Nuclear Energy 31(12), 1331–1356]. It was found that the proposed core, which provides twice the flux per unit core power compared to similar natural uranium fueled, light water cooled, heavy water

Mohammad Javed Khan; Aslam; Nasir Ahmad

2005-01-01

26

Accuracy considerations for Chebyshev rational approximation method (CRAM) in Burnup calculations

The burnup equations can in principle be solved by computing the exponential of the burnup matrix. However, due to the difficult numerical characteristics of burnup matrices, the problem is extremely stiff and the matrix exponential solution has previously been considered infeasible for an entire burnup system containing over a thousand nuclides. It was recently discovered by the author that the eigenvalues of burnup matrices are generally located near the negative real axis, which prompted introducing the Chebyshev rational approximation method (CRAM) for solving the burnup equations. CRAM can be characterized as the best rational approximation on the negative real axis and it has been shown to be capable of simultaneously solving an entire burnup system both accurately and efficiently. In this paper, the accuracy of CRAM is further studied in the context of burnup equations. The approximation error is analyzed based on the eigenvalue decomposition of the burnup matrix. It is deduced that the relative accuracy of CRAM may be compromised if a nuclide concentration diminishes significantly during the considered time step. Numerical results are presented for two test cases, the first one representing a small burnup system with 36 nuclides and the second one a full a decay system with 1531 nuclides. (authors)

Pusa, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)] [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)

2013-07-01

27

Technique for Sensitivity Analysis of Space- and Energy-Dependent Burn-Up Calculations.

National Technical Information Service (NTIS)

A practical method is presented for sensitivity analysis of the very complex, space-energy dependent burn-up equations, in which the neutron and nuclide fields are coupled nonlinearly. The adjoint burn-up equations that are given are in a form which can b...

M. L. Williams J. R. White

1979-01-01

28

Propagation of Nuclear Data Uncertainties for ELECTRA Burn-up Calculations

NASA Astrophysics Data System (ADS)

The European Lead-Cooled Training Reactor (ELECTRA) has been proposed as a training reactor for fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by pure liquid lead. In this work, we propagate the uncertainties in 239Pu transport data to uncertainties in the fuel inventory of ELECTRA during the reactor lifetime using the Total Monte Carlo approach (TMC). Within the TENDL project, nuclear models input parameters were randomized within their uncertainties and 740 239Pu nuclear data libraries were generated. These libraries are used as inputs to reactor codes, in our case SERPENT, to perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory determines uncertainties in: the long-term radio-toxicity, the decay heat, the evolution of reactivity parameters, gas pressure and volatile fission product content. In this work, a methodology called fast TMC is utilized, which reduces the overall calculation time. The uncertainty of some minor actinides were observed to be rather large and therefore their impact on multiple recycling should be investigated further. It was also found that, criticality benchmarks can be used to reduce inventory uncertainties due to nuclear data. Further studies are needed to include fission yield uncertainties, more isotopes, and a larger set of benchmarks.

Sjöstrand, H.; Alhassan, E.; Duan, J.; Gustavsson, C.; Koning, A. J.; Pomp, S.; Rochman, D.; Österlund, M.

2014-04-01

29

Language in calculation: a core mechanism?

Although there is evidence that exact calculation recruits left hemisphere perisylvian language systems, recent work has shown that exact calculation can be retained despite severe damage to these networks. In this study, we sought to identify a "core" network for calculation and hence to determine the extent to which left hemisphere language areas are part of this network. We examined performance on addition and subtraction problems in two modalities: one using conventional two-digit problems that can be easily encoded into language; the other using novel shape representations. With regard to numerical problems, our results revealed increased left fronto-temporal activity in addition, and increased parietal activity in subtraction, potentially reflecting retrieval of linguistically encoded information during addition. The shape problems elicited activations of occipital, parietal and dorsal temporal regions, reflecting visual reasoning processes. A core activation common to both calculation types involved the superior parietal lobule bilaterally, right temporal sub-gyral area, and left lateralized activations in inferior parietal (BA 40), frontal (BA 6/8/32) and occipital (BA 18) regions. The large bilateral parietal activation could be attributed to visuo-spatial processing in calculation. The inferior parietal region, and particularly the left angular gyrus, was part of the core calculation network. However, given its activation in both shape and number tasks, its role is unlikely to reflect linguistic processing per se. A possibility is that it serves to integrate right hemisphere visuo-spatial and left hemisphere linguistic and executive processing in calculation. PMID:22079204

Benn, Yael; Zheng, Ying; Wilkinson, Iain D; Siegal, Michael; Varley, Rosemary

2012-01-01

30

Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

Holly R. Trellue

1998-12-01

31

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the produced neutrons and photons but the current version of MCNPX doesn't support depletion/burnup calculation of the subcritical system with the generated neutron source from the target. MCB can perform neutron transport and burnup calculation for subcritical system using external neutron source, however it cannot perform electron transport calculations. To solve this problem, a hybrid procedure is developed by coupling these two computer codes. The user tally subroutine of MCNPX is developed and utilized to record the information of the each generated neutron from the photonuclear reactions resulted from the electron beam interactions. MCB reads the recorded information of each generated neutron thorough the user source subroutine. In this way, the neutron source generated by electron reactions could be utilized in MCB calculations, without the need for MCB to transport the electrons. Using the source subroutines, MCB could get the external neutron source, which is prepared by MCNPX, and perform depletion calculation for the driven subcritical facility.

Gohar, Y.; Zhong, Z.; Talamo, A.; Nuclear Engineering Division

2009-06-09

32

Benchmarking TMI2 core and canister calculations

For the application of the Monte Carlo code KENO to the calculations of k\\/sub eff\\/ for Three Mile Island Unit 2 (TMI-2) fuel in various arrangements, values of code bias were required. The presumed core configuration lay outside the region for which the code has been validated. There is an abundance of critical experiment data relevant to power reactor fuel

R. L. Murray; D. S. Williams; J. C. Rommel

1986-01-01

33

An automated procedure for determining the optimal core loading pattern for a pressurized water reactor which maximizes end-of-cycle k/sub eff/ while satisfying constraints on power peaking and discharge burnup has been developed. The optimization algorithm combines a two energy group, two-dimensional coarse-mesh finite difference diffusion theory neutronics model to simulate core conditions, a perturbation theory approach to determine reactivity, flux, power and burnup changes as a function of assembly shuffling, and Monte Carlo integer programming to select the optimal loading pattern solution. The core examined was a typical Cycle 2 reload with no burnable poisons. Results indicate that the core loading pattern that maximizes end-of-cycle k/sub eff/ results in a 5.4% decrease in fuel cycle costs compared with the core loading pattern that minimizes the maximum relative radial power peak.

Hobson, G.H.

1985-01-01

34

TMI2 isotopic inventory calculations

Point isotopic depletion methods are used to develop spatially dependent fission product and heavy metal inventories for the TMI-2 core. Burnup data from 1239 fuel nodes (177 elements, 7 axial nodes per element) are utilized to preserve the core axial and radial power distributions. A full-core inventory is calculated utilizing 12 fuel groups (four burnup ranges for each of three

B. G. Schnitzler; J. B. Briggs

1985-01-01

35

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

NASA Astrophysics Data System (ADS)

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Damian, F.; Brun, E.

2014-06-01

36

Benchmarking TMI-2 core and canister calculations

For the application of the Monte Carlo code KENO to the calculations of k/sub eff/ for Three Mile Island Unit 2 (TMI-2) fuel in various arrangements, values of code bias were required. The presumed core configuration lay outside the region for which the code has been validated. There is an abundance of critical experiment data relevant to power reactor fuel assemblies and highly enriched systems. However, there are few experiments involving high boron content and close packing of fuel as expected in the damaged core configuration. Such fuel would exhibit a harder neutron spectrum than that of most thermal power reactors. There also are few experiments involving small containers of low-enrichment UO/sub 2/ with solid internal poisons, and thus exhibiting the relatively large neutron losses due to leakage and absorption by B/sub 4/C that characterize the canisters to be used for storage of fuel removed from the reactor. The procedures for the benchmarking effort and resulting values of biases to be added to the k/sub eff/ calculated by KENO are discussed.

Murray, R.L.; Williams, D.S.; Rommel, J.C.

1986-01-01

37

A chemical isotopic analysis of the actinides and fission products of a high-burnup PWR-UO2 fuel with an average burnup of 60.2 MWd\\/kgHM was carried out to accumulate extensive nuclide composition data. Furthermore, computational analysis was performed using the integrated burnup calculation code SWAT. The differences between the amounts obtained by the chemical isotopic analysis and SWAT calculation using JENDL-3.2, JENDL-3.3,

Akihiro SASAHARA; Tetsuo MATSUMURA; Giorgos NICOLAOU; Yoshiaki KIYANAGI

2008-01-01

38

NASA Astrophysics Data System (ADS)

The development of tools for nuclear data uncertainty propagation in lattice calculations are presented. The Total Monte Carlo method and the Generalized Perturbation Theory method are used with the code DRAGON to allow propagation of nuclear data uncertainties in transport calculations. Both methods begin the propagation of uncertainties at the most elementary level of the transport calculation - the Evaluated Nuclear Data File. The developed tools are applied to provide estimates for response uncertainties of a PWR cell as a function of burnup.

Sabouri, P.; Bidaud, A.; Dabiran, S.; Lecarpentier, D.; Ferragut, F.

2014-04-01

39

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01

40

Nuclear Calculation of the JMTR Core, (2). Revision of the Neutron Flux Calculation Method.

National Technical Information Service (NTIS)

The method of nuclear calculation for JMTR core configuration is well established. Due to advance in the calculation techniques, however, the method has been partly revised, as follows. In calculating the fast neutron group constants, space dependent spec...

H. Ando H. Iida Y. Nagaoka R. Oyamada

1976-01-01

41

Dose Rate Calculations for Rotary Mode Core Sampling Exhauster

This document provides the calculated estimated dose rates for three external locations on the Rotary Mode Core Sampling (RMCS) exhauster HEPA filter housing, per the request of Characterization Field Engineering.

FOUST, D.J.

2000-10-26

42

Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

2006-07-01

43

Perturbation and sensitivity theory for burnup analysis

Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonliner systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron\\/nuclide fields (in

1979-01-01

44

Spatial Kinetics Calculations of MOX Fueled Core: Variant 22

This work is part of a Joint US\\/Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactors and presents the results of spatial kinetics calculational benchmarks. The examinations were carried out with the following purposes: to verify one of spatial neutronic kinetics model elaborated in KI, to understand sensibility of the model to neutronics difference of UOX and MOX cores, to

Pavlovichev

2001-01-01

45

Improved calculation of core loss with nonsinusoidal waveforms

An extension to the Steinmetz equation is proposed, to enable estimation of hysteresis losses in magnetic core materials with nonsinusoidal flux waveforms. The new formulation is shown to avoid anomalies present in previous modified-Steinmetz-equation calculations of loss with nonsinusoidal waveforms. Comparison with experimental measurements in MnZn ferrite shows improved accuracy. The result may be optionally formulated in terms of an

Jieli Li; Tarek Abdallah; Charles R. Sullivan

2001-01-01

46

Calculation models of AEGIS/SCOPE2, a core calculation system of next generation

This paper provides calculation models of AEGIS/SCOPE2, a core calculation system of next generation. This paper especially focuses on the resonance calculation method of the AEGIS code, which has not been published yet. In the AEGIS code, effective cross sections in resonance region are generated from ultra-fine-group calculations utilizing tabulated collision probabilities for square pin-cells. In addition, the SPH method in the energy domain is used to preserve reaction rates in the energy-collapsed multi-group calculations. The validity of the resonance calculation method of the AEGIS code is verified through the comparison with the continuous energy Monte-Carlo calculation in pin-cell geometry. (authors)

Sugimura, N.; Ushio, T. [Nuclear Engineering, Ltd., 1-3-7 Tosabori, Nishi-ku, Osaka, 550-0001 (Japan); Yamamoto, A. [Nagoya Univ., Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan); Tatsumi, M. [Nuclear Fuel Industies, Ltd., 1-950 Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka, 590-0481 (Japan)

2006-07-01

47

FRAPCON-3 is a FORTRAN IV computer code that calculates the steady-state response of light water reactor fuel rods during long-term burnup. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include (1) heat conduction through the fuel and cladding, (2) cladding elastic and plastic deformation, (3) fuel-cladding mechanical interaction, (4) fission gas release, (5) fuel rod internal gas pressure, (6) heat transfer between fuel and cladding, (7) cladding oxidation, and (8) heat transfer from cladding to coolant. The code contains necessary material properties, water properties, and heat-transfer correlations. The codes` integral predictions of mechanical behavior have not been assessed against a data base, e.g., cladding strain or failure data. Therefore, it is recommended that the code not be used for analyses of cladding stress or strain. FRAPCON-3 is programmed for use on both mainframe computers and UNIX-based workstations such as DEC 5000 or SUN Sparcstation 10. It is also programmed for personal computers with FORTRAN compiler software and at least 8 to 10 megabytes of random access memory (RAM). The FRAPCON-3 code is designed to generate initial conditions for transient fuel rod analysis by the FRAPTRAN computer code (formerly named FRAP-T6).

Berna, G.A.; Beyer, G.A.; Davis, K.L.; Lanning, D.D.

1997-12-01

48

This report gives the procedure for determining the economical efficiency of the utilization of nuclear fuel in a reactor on the basis of calculated costs. The expression obtained for the fuet constituent of the costs of production of electrical energy enables one to make deductions regarding the aggregate effect of the cost of the fuel charge on the efficiency of

Yu. I. Koryakin; V. V. Batov; V. G. Smirnov

1964-01-01

49

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

DOE

1997-04-01

50

In this study, risk-significant pressurized-water reactor severe accident sequences are examined using MELCOR 1.8.5 to explore the range of fission product releases to the reactor containment building. Advances in the understanding of fission product release and transport behavior and severe accident progression are used to render best estimate analyses of selected accident sequences. Particular emphasis is placed on estimating the

Randall O. Gauntt; Dana Auburn Powers; Scott G. Ashbaugh; Mark Thomas Leonard; Pamela Longmire

2010-01-01

51

NASA Astrophysics Data System (ADS)

Potential curves and spectroscopic constants for electronic states of the Li 2 molecule dissociating into 2 s + 2 s, 2 s + 2 p, 2 s + 3 s, 2 p + 2 p, 2 s + 3 p, and 2 s + 3 d atomic configurations (49 states) are obtained from CI calculations with electronic core potentials, including core polarization effects.

Poteau, R.; Spiegelmann, F.

1995-06-01

52

Potential curves and spectroscopic constants for electronic states of the Li2 molecule dissociating into 2s + 2s, 2s + 2p, 2s + 3s, 2p + 2p, 2s + 3p, and 2s + 3d atomic configurations (49 states) are obtained from CI calculations with electronic core potentials, including core polarization effects.

R. Poteau; F. Spiegelmann

1995-01-01

53

Perturbation and sensitivity theory for reactor burnup analysis

Perturbation theory is developed for the nonlinear burnup equations describing the time-dependent behavior of the neutron and nuclide fields in a reactor core. General aspects of adjoint equations for nonlinear systems are first discussed and then various approximations to the burnup equations are rigorously derived and their areas for application presented. In particular, the concept of coupled neutron\\/nuclide fields (in

1979-01-01

54

Calculation methods for core distortions and mechanical behavior

This paper describes ABADAN, a general purpose, nonlinear, multi-dimensional finite element structural analyses computer code developed for the express purpose of solving large nonlinear problems as typified by the Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System design problem. All of the structural modeling features inherent in a general purpose finite element code and required to adequately model an LMFBR core restraint system are demonstrated. Typical results for a radial row and a sixty degree sector model of FFTF are presented. The sixty degree sector results are interpreted in terms of the design criteria that the core restraint system must satisfy. Extensions and adaptations of these modeling techniques to different core restraint design concepts can be readily achieved. 27 figures.

Sutherland, W.H.

1984-09-01

55

This is the final report of the High Burnup Effects Program (HBEP). It has been prepared to present a summary, with conclusions, of the HBEP. The HBEP was an international, group-sponsored research program managed by Battelle, Pacific Northwest Laboratories (BNW). The principal objective of the HBEP was to obtain well-characterized data related to fission gas release (FGR) for light water reactor (LWR) fuel irradiated to high burnup levels. The HBEP was organized into three tasks as follows: Task 1 -- high burnup effects evaluations; Task 2 -- fission gas sampling; and Task 3 -- parameter effects study. During the course of the HBEP, a program that extended over 10 years, 82 fuel rods from a variety of sources were characterized, irradiated, and then examined in detail after irradiation. The study of fission gas release at high burnup levels was the principal objective of the program and it may be concluded that no significant enhancement of fission gas release at high burnup levels was observed for the examined rods. The rim effect, an as yet unquantified contributor to athermal fission gas release, was concluded to be the one truly high-burnup effect. Though burnup enhancement of fission gas release was observed to be low, a full understanding of the rim region and rim effect has not yet emerged and this may be a potential area of further research. 25 refs., 23 figs., 4 tabs.

Barner, J.O.; Cunningham, M.E.; Freshley, M.D.; Lanning, D.D.

1990-04-01

56

This paper reports on a data base of the TRIGAP code that is generated for the 3-MW TRIGA MARK-II research reactor in Bangladesh. The library is created using the WIMS-D\\/4 code. Cross sections are calculated from zero burnup to 37% of initial Â˛ÂłâµU in 20 burnup steps. The created TRIGAP library is tested through practical calculations and is compared with

S. I. Bhuiyan; A. R. Khan; M. M. Sarker; M. Rahman; Z. G. Ara; M. A. Mannan; M. Musa; I. Mele

1992-01-01

57

Development of a core follow calculational system for research reactors

Over the last few years a comprehensive PWR and MTR core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments.

Mueller, E.Z.; Ball, G.; Joubert, W.R. [and others

1994-12-31

58

Hartree-Fock Calculations with Realistic Hard-Core Potential

The effective matrix elements of the two-nucleon Yale potential have been used in doing Hartree-Fock (HF) calculations in N=Z even nuclei (8<=A<=40). The ground-state energy and single-particle energies and wave functions have been calculated as a function of two deformation parameters. The calculated equilibrium shapes and binding energy per nucleon are found to be reasonably good. The difficulties in the

M. K. Pal; A. P. Stamp

1967-01-01

59

In-core thermal-hydraulic and fission product calculations for severe fuel damage analyses

Best-estimate calculations of realistic fission product source terms are presented for the Severe Fuel Damage (SFD) tests conducted in the Power Burst Facility (PBF), utilizing the Advanced Reactor Severe Accident Program (ARSAP) bulk mass transfer correlation. Computer codes were written to perform the thermal-hydraulic and fission product calculations for the SFD tests. Fewer and slower releases are predicted with the ARSAP mass transfer correlation, in good agreement with the test results. The ARSAP mass transfer model correlates the inverse fuel temperature with the product of release rate and grain size considering the fuel/cladding interaction. The empirical coefficients were developed from Oak Ridge National Laboratory (ORNL) high-burnup fuel data in the 770 to 2,275 K temperature range. The ORNL test data indicate that the fuel/cladding interaction takes effect above 2,000 K.

Suh, K.Y.; Sharon, A.; Hammersley, R.J. (Fauske Associates, Inc., Burr Ridge, IL (USA))

1989-11-01

60

Approximate Calculation Method for Second Order Sensitivity Coefficient

A simple method has been developed for calculating the second order sensitivity coefficient of static and burnup-dependent core performance parameters. The method is applied to a small and a large fast breeder reactors. Changes in core performance parameters due to 10% cross section changes are compared with that predicted by the first and the second order sensitivity analyses. Numerical results

Kazuhisa MATSUMOTO; Toshikazu TAKEDA; Tomoaki MASUDA

1994-01-01

61

Isotopic validation for PWR actinide-only burnup credit using Yankee Rowe data

Safety analyses of criticality control systems for transportation packages include an assumption that the spent nuclear fuel (SNF) loaded into the package is fresh or unirradiated. In other words, the spent fuel is assumed to have its original, as-manufactured U-235 isotopic content. The ``fresh fuel`` assumption is very conservative since the potential reactivity of the nuclear fuel is substantially reduced after being irradiated in the reactor core. The concept of taking credit for this reduction in nuclear fuel reactivity due to burnup of the fuel, instead of using the fresh fuel assumption in the criticality safety analysis, is referred to as ``Burnup Credit.`` Burnup credit uses the actual physical composition of the fuel and accounts for the net reduction of fissile material and the buildup of neutron absorbers in the fuel as it is irradiated. Neutron absorbers include actinides and other isotopes generated as a result of the fission process. Using only the change in actinide isotopes in the burnup credit criticality analysis is referred to as ``Actinide-Only Burnup Credit.`` The use of burnup credit in the design of criticality control systems enables more spent fuel to be placed in a package. Increased package capacity results in a reduced number of storage, shipping and disposal containers for a given number of SNF assemblies. Fewer shipments result in a lower risk of accidents associated with the handling and transportation of spent fuel, thus reducing both radiological and nonradiological risk to the public. This paper describes the modeling and the results of comparison between measured and calculated isotopic inventories for a selected number of samples taken from a Yankee Rowe spent fuel assembly.

NONE

1997-11-01

62

NASA Astrophysics Data System (ADS)

When representing the behavior of commercial spent nuclear fuel (SNF), credit is sought for the reduced reactivity associated with the net depletion of fissile isotopes and the creation of neutron-absorbing isotopes, a process that begins when a commercial nuclear reactor is first operated at power. Burnup credit accounts for the reduced reactivity potential of a fuel assembly and varies with the fuel burnup, cooling time, and the initial enrichment of fissile material in the fuel. With regard to long-term SNF disposal and transportation, tremendous benefits, such as increased capacity, flexibility of design and system operations, and reduced overall costs, provide an incentive to seek burnup credit for criticality safety evaluations. The Nuclear Regulatory Commission issued Interim Staff Guidance 8, Revision 2 in 2002, endorsing burnup credit of actinide composition changes only; credit due to actinides encompasses approximately 30% of exiting pressurized water reactor SNF inventory and could potentially be increased to 90% if fission product credit were accepted. However, one significant issue for utilizing full burnup credit, compensating for actinide and fission product composition changes, is establishing a set of depletion parameters that produce an adequately conservative representation of the fuel's isotopic inventory. Depletion parameters can have a significant effect on the isotopic inventory of the fuel, and thus the residual reactivity. This research seeks to quantify the reactivity impact on a system from dominant depletion parameters (i.e., fuel temperature, moderator density, burnable poison rod, burnable poison rod history, and soluble boron concentration). Bounding depletion parameters were developed by statistical evaluation of a database containing reactor operating histories. The database was generated from summary reports of commercial reactor criticality data. Through depletion calculations, utilizing the SCALE 6 code package, several light water reactor assembly designs and in-core locations are analyzed in establishing a combination of depletion parameters that conservatively represent the fuel's isotopic inventory as an initiative to take credit for fuel burnup in criticality safety evaluations for transportation and storage of SNF.

Sloma, Tanya Noel

63

Nuclide analysis in high burnup fuel samples irradiated in Vandellós 2

NASA Astrophysics Data System (ADS)

In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% 235U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellós 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values. The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column. Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records. Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement between measured and calculated values for most of the analysed isotopes, similar to those reported previously for lower burnup ranges. Thus, it could be concluded, that SAS2H results for high burnup samples are not subject to higher uncertainty and/or different biases than for lower burnup samples, and that the different isotopic experimental measurement methods provide accurate results with acceptable precision.

Zwicky, H. U.; Low, J.; Granfors, M.; Alejano, C.; Conde, J. M.; Casado, C.; Sabater, J.; Lloret, M.; Quecedo, M.; Gago, J. A.

2010-07-01

64

Nuclear Criticality Calculations on a Spherically Symmetric Gaseous-Core Reactor.

National Technical Information Service (NTIS)

The nuclear criticality calculations are performed on a two-region, spherically symmetric gaseous-core reactor. The core is fueled with U235 vapor and is externally moderated by various thicknesses of graphite at 4000K. A one-group (thermal) transport ana...

J. R. Rec

1964-01-01

65

/sup 18/O as a core plus two valence neutrons: A three-body Faddeev calculation

The nucleus /sup 18/O is studied assuming a three-body model: two neutrons outside an inert core of /sup 16/O: and solving the Faddeev equations. The calculated spectrum is in good agreement with experiment.

Ueta, K.; Miyake, H.; Mizukami, A.

1983-01-01

66

The paper shows the nuclear design calculation for the Japan-made ; reactor No. 1 (JRR-3), a 10 Mw natural uranium heavy-water research reactor. At ; the present stage of the nuclear design calculation an accurate and simple method ; is not available, so a few lattice parameters must be set from the preliminary ; calculation and then a more accurate

H. Ishikawa; T. Asaoka; Y. Sasakura

1959-01-01

67

NASA Astrophysics Data System (ADS)

A detailed investigation of the atomic structure and radiative parameters involving the lowest states within the 6p4, 6p36d, 6p37s, 6p37p and 6p37d configurations of neutral polonium is reported in the present paper. Using different physical models based on the pseudo-relativistic Hartree-Fock approach, the influence of intravalence, core-valence and core-core electron correlation on the atomic parameters is discussed in detail. This work allowed us to fix the spectroscopic designation of some experimental level energy values and to provide for the first time a set of reliable oscillator strengths corresponding to 31 Po I spectral lines in the wavelength region from 175 to 987 nm.

Quinet, Pascal

2014-09-01

68

Molecular Evolution of A First Core in 3 Dimensional Hydrodynamic Calculations

NASA Astrophysics Data System (ADS)

It is well established that stars are formed by gravitational collapse of molecular cloud cores. Collapsing cores initially undergo isothermal collapse. The isothermal condition breaks down at the density of 10-13 g cm-3, and the temperature starts rising. Increasing gas pressure decelerates the contraction, and the cores come to hydrostatic equilibrium with a radius of a few AU and a mass of 0.01 M?, which is called the first cores (e.g. Larson 1969). Observation of the first cores is important but challenging, since their lifetime is short ( 1000 yr). The mechanical property of the first cores have been studied by multi-dimensional hydrodynamic calculations considering interstellar magnetic fields and radiative transfer (e.g. Tomisaka 2002; Machida et al.2008; Tomida et al. 2010). In contrast, their chemical property is yet to be understood. It is important to reveal their chemical property in terms of which lines we should use to observe the first cores. In addition, the first cores evolve to protoplanetary disks (Saigo et al. 2008; Machida et al. 2010), hence the compositions of the first cores restrict the initial compositions of disks. We investigate molecular evolution of star forming cores that are initially rotating molecular cloud cores and collapse to form the first cores. The results of three dimensional hydrodynamic calculations (Matsumoto & Hanawa 2003) are adopted as physical models of the core. We trace trajectories of test particles in the hydrodynamic calculations, and molecular evolution is solved using low temperature chemical network (Garrod & Herbst 2006) at T < 100 K and high temperature network (Harada et al. 2010) at T > 100 K along the trajectories. We also consider three body reactions and collisional dissociations (Willacy et al. 1998). Trace particles fall into the first core almost spherically, and rotate in the first core where the spiral arms transports angular momentum. In our model with barotropic approximation, we find that in outer regions (R > 5 AU), the composition is similar to the low temperature chemistry. In intermediate regions (R 3 AU), hot-core like species, such as HCOOCH_3 and CH_3OCH_3 are generated. In central regions (R < 1 AU), complex molecules, such as HC_7N, HC_9N and NH_2CN, are formed in the gas phase.

Furuya, K.; Aikawa, Y.; Matsumoto, T.; Tomida, K.; Saigo, K.; Tomisaka, K.; Hersant, F.; Wakelam, V.

2011-05-01

69

NASA Astrophysics Data System (ADS)

The zero-core-contribution method is extended to study the photodetachment of heteronuclear diatomic molecules. Total photodetachment cross sections are calculated for the ions OH-, SH-, and SeH-. The anisotropy factor ? is also calculated for OH-. The agreement with measured values of these quantities is good.

Clodius, W. B.; Stehman, R. M.; Woo, S. B.

1983-08-01

70

RMC - A Monte Carlo Code for Reactor Core Analysis

NASA Astrophysics Data System (ADS)

A new Monte Carlo transport code RMC has been being developed by Department of Engineering Physics, Tsinghua University, Beijing as a tool for reactor core analysis on high-performance computing platforms. To meet the requirements of reactor analysis, RMC now has such functions as criticality calculation, fixed-source calculation, burnup calculation and kinetics simulations. Some techniques for geometry treatment, new burnup algorithm, source convergence acceleration, massive tally and parallel calculation, and temperature dependent cross sections processing are researched and implemented in RMC to improve the effciency. Validation results of criticality calculation, burnup calculation, source convergence acceleration, tallies performance and parallel performance shown in this paper prove the capabilities of RMC in dealing with reactor analysis problems with good performances.

Wang, Kan; Li, Zeguang; She, Ding; Liang, Jin'gang; Xu, Qi; Qiu, Yishu; Yu, Jiankai; Sun, Jialong; Fan, Xiao; Yu, Ganglin

2014-06-01

71

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

Karim, Julia Abdul [Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

2008-05-20

72

Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

NASA Astrophysics Data System (ADS)

The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

Karim, Julia Abdul

2008-05-01

73

Accounting for strong localized heterogeneities and local transport effect in core calculations

Two methods based on the variational nodal transport method have been developed to account for localized heterogeneities and local transport effects in full core calculations. A local mesh refinement technique relies on using the projected partial ingoing surface currents produced during coarse-mesh iterations as boundary conditions for fine-mesh calculations embedded within the coarse-mesh calculations. The outgoing fine-mesh partial currents are

J. M. Ruggieri; J. Y. Doriath; P. J. Finck; R. Boyer

1996-01-01

74

NASA Astrophysics Data System (ADS)

Benchmark calculations for reflector effects in fast cores were performed to validate the reliability of scattering data of structural materials in the major evaluated nuclear data libraries, JENDL-4.0, ENDF/B-VII.1 and JEFF-3.1.2. The criticalities of two FCA and two ZPR cores were analyzed by using a continuous energy Monte Carlo calculation code. The ratios of calculation to experimental values were compared between these cores and the sensitivity analyses were performed. From the results, the replacement reactivity from blanket to SS and Na reflector is better evaluated by JENDL-4.0 than by ENDF/B-VII.1 mainly due to the ?bar values of Na and 52Cr.

Fukushima, M.; Ishikawa, M.; Numata, K.; Jin, T.; Kugo, T.

2014-04-01

75

Interaction of loading pattern and nuclear data uncertainties in reactor core calculations

Along with best-estimate calculations for design and safety analysis, understanding uncertainties is important to determine appropriate design margins. In this framework, nuclear data uncertainties and their propagation to full core calculations are a critical issue. To deal with this task, different error propagation techniques, deterministic and stochastic are currently developed to evaluate the uncertainties in the output quantities. Among these is the sampling based uncertainty and sensitivity software XSUSA which is able to quantify the influence of nuclear data covariance on reactor core calculations. In the present work, this software is used to investigate systematically the uncertainties in the power distributions of two PWR core loadings specified in the OECD UAM-Benchmark suite. With help of a statistical sensitivity analysis, the main contributors to the uncertainty are determined. Using this information a method is studied with which loading patterns of reactor cores can be optimized with regard to minimizing power distribution uncertainties. It is shown that this technique is able to halve the calculation uncertainties of a MOX/UOX core configuration. (authors)

Klein, M.; Gallner, L.; Krzykacz-Hausmann, B.; Pautz, A.; Velkov, K.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS MbH, Boltzmannstr. 14, D- 85748 Garching b. Muenchen (Germany)

2012-07-01

76

Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a

J. W. Sterbentz

1999-01-01

77

Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core

The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements,

Zuying Gao; Lei Shi

2002-01-01

78

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4 - Revised Report

The Task Force on Reactor-Based Plutonium Disposition (TFRPD) was formed by the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) to study reactor physics, fuel performance, and fuel cycle issues related to the disposition of weapons-grade (WG) plutonium as mixed-oxide (MOX) reactor fuel. To advance the goals of the TFRPD, 10 countries and 12 institutions participated in a major TFRPD activity: a blind benchmark study to compare code calculations to experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At Oak Ridge National Laboratory, the HELIOS-1.4 code system was used to perform the comprehensive study of pin-cell and MOX core calculations for the VENUS-2 MOX core benchmark study.

Ellis, RJ

2001-06-01

79

Westinghouse high-burnup experience

To enhance nuclear plant fuel economy, utilities are requiring the fuel to operate under increasingly demanding conditions. Upratings, longer fuel cycles, and more aggressive fuel management have all led to higher burnups. Maximum batch discharge burnups have increased by >20% in the past few years, from the low 40 GWd/tonne U range in the early 1990s to >50 GWd/ tonne U at the present time. Although no fuel integrity issues have been observed in this time frame that have been directly related to increased fuel burnup, two burnup-related performance issues being actively investigated are incomplete rod cluster control assembly (RCCA) insertion (IRI) and fuel rod cladding corrosion.

Wilson, H.W.; Esposito, V.J.; Sabol, G.P. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1997-12-01

80

Calculation of the Thermal and Hydraulic States in Rod Cluster Cores of Light-Water Reactors.

National Technical Information Service (NTIS)

For calculating the three-dimensional steady distribution of the thermal and hydraulic states in rod cluster cores of light-water reactors, the subchannel analysis programs COLA 1 and COLA 2 have been developed. Both programs contain a multitude of compet...

H. Teichel

1977-01-01

81

The No Core Gamow Shell Model for ab-initio Nuclear Structure Calculations

NASA Astrophysics Data System (ADS)

We apply the Berggren basis in a No-Core Shell Model framework to calculate ground state (g.s.) energies of 3H, 4He and 5He. In our studies we use the Argonne ?18 and the chiral N3LO potentials, both of which are renormalized via a Vlow-k process.

Papadimitriou, G.; Barrett, B. R.; Rotureau, J.; Michel, N.; P?oszajczak, M.

2014-03-01

82

Benchmark calculation of no-core Monte Carlo shell model in light nuclei

The Monte Carlo shell model is firstly applied to the calculation of the no-core shell model in light nuclei. The results are compared with those of the full configuration interaction. The agreements between them are within a few % at most.

Abe, T.; Shimizu, N. [Department of Physics, the University of Tokyo, Hongo, Tokyo 113-0033 (Japan); Maris, P.; Vary, J. P. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Otsuka, T. [Department of Physics, University of Tokyo, Hongo, Tokyo 113-0033 (Japan); CNS, University of Tokyo, Hongo, Tokyo 113-0033 (Japan); NSCL, Michigan State University, East Lansing, Michigan 48824 (United States); Utsuno, Y. [ASRC, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

2011-05-06

83

Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

J. W. Sterbentz

1999-08-01

84

Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd\\/t

High-resolution gamma spectroscopy has been employed for the measurement of ÂąÂłâ´Cs\\/ÂąÂłâ·Cs, Âąâµâ´Eu\\/ÂąÂłâ·Cs and ÂąÂłâ´Cs\\/Âąâµâ´Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UOâ pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd\\/t have been experimentally characterised. Additionally, pin cell depletion calculations have been

S. Caruso; M. Murphy; F. Jatuff; R. Chawla

2006-01-01

85

Calculation of ex-core physical quantities using the 3D importance functions

NASA Astrophysics Data System (ADS)

Diverse physical quantities are calculated in engineering studies with penalizing hypotheses to assure the required operation margins for each reactor. Today, these physical quantities are obtained by direct calculations from deterministic or Monte Carlo codes. The related states are critical or sub-critical. The current physical quantities are for example: the SRD counting rates (source range detector) in the sub-critical state, the IRD (intermediary range detector) and PRD (power range detector) counting rates (neutron particles only), the deposited energy in the reflector (neutron + photon particles), the fluence or the DPA (displacement per atom) in the reactor vessel (neutron particles only). The reliability of the proposed methodology is tested in the EPR reactor. The main advantage of the new methodology is the simplicity to obtain the physical quantities by an easy matrix calculation importance linked to nuclear power sources for all the cycles of the reactor. This method also allows to by-pass the direct calculations of the physical quantity of irradiated cores by Monte Carlo Codes, these calculations being impossible today (too many isotopic concentrations / MCNP5 limit). This paper presents the first feasibility study for the physical quantities calculation outside of the core by the importance method instead of the direct calculations used currently by AREVA.

Trakas, Christos; De Laubiere, Xavier

2014-06-01

86

3D Neutron Transport PWR Full-core Calculation with RMC code

NASA Astrophysics Data System (ADS)

Nowadays, there are more and more interests in the use of Monte Carlo codes to calculate the detailed power density distributions in full-core reactors. With the Inspur TS1000 HPC Server of Tsinghua University, several calculations have been done based on the EDF 3D Neutron Transport PWR Full-core benchmark through large-scale parallelism. To investigate and compare the results of the deterministic method and Monte Carlo method, EDF R&D and Department of Engineering Physics of Tsinghua University are having a collaboration to make code to code verification. So in this paper, two codes are used. One is the code COCAGNE developed by the EDF R&D, a deterministic core code, and the other is the Monte Carlo code RMC developed by Department of Engineering Physics in Tsinghua University. First, the full-core model is described and a 26-group calculation was performed by these two codes using the same 26-group cross-section library provided by EDF R&D. Then the parallel and tally performance of RMC is discussed. RMC employs a novel algorithm which can cut down most of the communications. It can be seen clearly that the speedup ratio almost linearly increases with the nodes. Furthermore the cell-mapping method applied by RMC consumes little time to tally even millions of cells. The results of the codes COCAGNE and RMC are compared in three ways. The results of these two codes agree well with each other. It can be concluded that both COCAGNE and RMC are able to provide 3D-transport solutions associated with detailed power density distributions calculation in PWR full-core reactors. Finally, to investigate how many histories are needed to obtain a given standard deviation for a full 3D solution, the non-symmetrized condensed 2-group fluxes of RMC are discussed.

Qiu, Yishu; She, Ding; Fan, Xiao; Wang, Kan; Li, Zeguang; Liang, Jingang; Leroyer, Hadrien

2014-06-01

87

An improved resonance self-shielding method has been developed to accurately estimate the effective multigroup cross sections for heterogeneous fast reactor assembly and core calculations. In the method, the heterogeneity effect is considered by the use of isotopic escape cross sections while the resonance interference effect is accounted for through the narrow resonance approximation or slowing-down calculations for specific compositions. The isotopic escape cross sections are calculated by solving fixed-source transport equations with the method of characteristics for the whole problem domain. This method requires no pre-calculated resonance integral tables or parameters that are typically necessary in the subgroup method. Preliminary results for multi pin-cell fast reactor problems show that the escape cross sections estimated from the explicit-geometry fixed source calculations produce more accurate eigenvalue and self-shielded effective cross sections than those from conventional one-dimensional geometry models. (authors)

Lee, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States)] [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439-4842 (United States); Yang, W. S. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)] [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907-2017 (United States)

2013-07-01

88

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR.

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data. PMID:16604689

Kurosawa, Masahiko

2005-01-01

89

NASA Astrophysics Data System (ADS)

The up-to-date development of the armored vehicles conditions complication of armor constructions and increased slope of shell armored plates. Combined strikers (C/S) can be used to destroy armored vehicles. We can increase total weight of the core part to increase the striker's power. However, the increase of core part diameter is limited by body dimensions. Thus, we can increase core part weight by increasing its length. Because of C/S interaction with the barriers at large deviation angles, C/S's mechanical trajectory sparks in the barrier. This results in bending stress which occurs in the core part. Because of large deviation angles, the impact of the side surface of oblong core part against the cavity edge occurs. This increases the probability of core part destruction. The calculation technique for oblong core part penetration into different types of barriers is presented. The large number of factors can be calculated using this technique. It is assumed that the core part is destroyed when the tail part impacts against the cavity in the section where specific impact energy exceeds the critical value. Impact elasticity and destruction at bending stress were selected to be destruction criteria. The following core part destruction scenarios were investigated and calculated: (i) core head part is slightly destroyed but tail part of cylindrical shape penetrates deeper; (ii) core tail part is slightly destroyed but head part penetrates deeper, mass loss is taken into account; and (iii) after the impact, the core part is splitted up into two parts, then both of them penetrate into the barrier, one part is of ogival shape, the other is of cylindrical one. This calculation technique was applied to computational program, then critical angles at which core part side surface is still in contact with cavity surface, and the angles at which core part destruction occurs were calculated. Depths of core part penetration for different destruction scenarios were calculated.

Antsiferova, E. V.; Bogdanov, V. V.; Derebenko, E. V.; Lagutina, A. V.; Khmelnikov, E. A.

2006-08-01

90

NASA Astrophysics Data System (ADS)

Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established and supported radiation transport code, for further development of a Monte Carlo-linked depletion methodology which is essential to the future development of advanced reactor technologies that exceed the limitations of current deterministic based methods.

Fensin, Michael Lorne

91

VENUS-2 MOX Core Benchmark: Results of ORNL Calculations Using HELIOS-1.4

The Task Force on Reactor-Based Plutonium Disposition, now an Expert Group, was set up through the Organization for Economic Cooperation and Development/Nuclear Energy Agency to facilitate technical assessments of burning weapons-grade plutonium mixed-oxide (MOX) fuel in U.S. pressurized-water reactors and Russian VVER nuclear reactors. More than ten countries participated to advance the work of the Task Force in a major initiative, which was a blind benchmark study to compare code benchmark calculations against experimental data for the VENUS-2 MOX core at SCK-CEN in Mol, Belgium. At the Oak Ridge National Laboratory, the HELIOS-1.4 code was used to perform a comprehensive study of pin-cell and core calculations for the VENUS-2 benchmark.

Ellis, RJ

2001-02-02

92

Thermal-hydraulic calculations for the conversion to LEU of a research reactor core

The thermal-hydraulic analysis performed for the needs of the conversion of the open pool 5MW Greek Research Reactor (GRR-1) to a pure Low Enrichment (LEU) configuration is presented. The methodology was based on a complete set of neutronic calculations performed for the new core configuration, in compliance with pre-defined Operation Limiting Conditions. The hottest channel analysis approach was adopted, and

D. Grigoriadis; M. Varvayanni; N. Catsaros; E. Stakakis

2008-01-01

93

Effective-core-potential calculations of sulphur, selenium and tellurium dioxides and dihydrides

NASA Astrophysics Data System (ADS)

The ground-state electronic structures of SO 2, SeO 2, TeO 2, SH 2, SeH 2 and TeH 2 have been calculated with effective core potentials. Satisfactory agreement with experimental molecular geometries was achieved in the dioxides only after d-functions were included in the basis sets for S, Se and Te; however, these d-functions were not essential for the dihydrides. The importance of electron correlation to the determination of dissociation energy is also evident from these calculations.

Jŕnszky, J.; Bartram, R. H.; Rossi, A. R.; Corradi, G.

1986-02-01

94

Investigation of the Fundamental Constants Stability Based on the Reactor Oklo Burn-Up Analysis

NASA Astrophysics Data System (ADS)

The burn-up of few samples of the natural Oklo reactor zones 3, 5 was calculated using the modern Monte Carlo code. We reconstructed the neutron spectrum in the core by means of the isotope ratios: 147Sm/148Sm and 176Lu/175Lu. These ratios unambiguously determine the water content and core temperature. The isotope ratio of the 149Sm in the sample calculated using this spectrum was compared with experimental one. The disagreement between these two values allows one to limit a possible shift of the low lying resonance of 149Sm. Then, these limits were converted to the limits for the change of the fine structure constant ?. We have found out, that for the rate of ? change, the inequality ?rt? {? }/? ?rt<= 5× 10-18 is fulfilled, which is one order higher than our previous limit.

Onegin, M. S.; Yudkevich, M. S.; Gomin, E. A.

2012-12-01

95

Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.

Hanson, A.L.; Diamond, D.

2011-09-30

96

Designing Critical Experiments in Support of Full Burnup Credit

Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative estimate in criticality calculations. Because of the numerous economical benefits including all available negative reactivity (i.e., full burnup credit) could provide [2], it is advantageous to work toward full burnup credit. Unfortunately, comparatively little work has been done to include non-major actinides and other fission products (FP) in burnup credit analyses due in part to insufficient experimental data for validation of codes and nuclear data. The Burnup Credit Criticality Experiment (BUCCX) at Sandia National Laboratory was a set of experiments with {sup 103}Rh that have relevance for burnup credit [3]. This work uses TSUNAMI-3D to investigate and adjust a BUCCX model to match isotope-specific, energy-dependent k{sub eff} sensitivity profiles to those of a representative high-capacity cask model (GBC-32) [4] for each FP of interest. The isotopes considered are {sup 149}Sm, {sup 143}Nd, {sup 103}Rh, {sup 133}Cs, {sup 155}Gd, {sup 152}Sm, {sup 99}Tc, {sup 145}Nd, {sup 153}Eu, {sup 147}Sm, {sup 109}Ag, {sup 95}Mo, {sup 150}Sm, {sup 101}Ru, and {sup 151}Eu. The goal is to understand the biases and bias uncertainties inherent in nuclear data, and ultimately, to apply these in support of full burnup credit.

Mueller, Don [ORNL; Roberts, Jeremy A [ORNL

2008-01-01

97

Full Core 3-D Simulation of a Partial MOX LWR Core

A comparative analysis and comparison of results obtained between 2-D lattice calculations and 3-D full core nodal calculations, in the frame of MOX fuel design, was conducted. This study revealed a set of advantages and disadvantages, with respect to each method, which can be used to guide the level of accuracy desired for future fuel and fuel cycle calculations. For the purpose of isotopic generation for fuel cycle analyses, the approach of using a 2-D lattice code (i.e., fuel assembly in infinite lattice) gave reasonable predictions of uranium and plutonium isotope concentrations at the predicted 3-D core simulation batch average discharge burnup. However, it was found that the 2-D lattice calculation can under-predict the power of pins located along a shared edge between MOX and UO2 by as much as 20%. In this analysis, this error did not occur in the peak pin. However, this was a coincidence and does not rule out the possibility that the peak pin could occur in a lattice position with high calculation uncertainty in future un-optimized studies. Another important consideration in realistic fuel design is the prediction of the peak axial burnup and neutron fluence. The use of 3-D core simulation gave peak burnup conditions, at the pellet level, to be approximately 1.4 times greater than what can be predicted using back-of-the-envelope assumptions of average specific power and irradiation time.

S. Bays; W. Skerjanc; M. Pope

2009-05-01

98

Calculation of scattering characteristic of complex target on multi-core platform

NASA Astrophysics Data System (ADS)

The scattering characteristic of complex target from terrestrial and celestial background radiation has been widely used in such engineering fields as remote sensing, feature extraction, tracking and recognition of target thus having been an attractive field for many scientists for decades. In our method, the model of target is constructed using 3DMAX and the surface is divided into triangle facets firstly. Bidirectional Reflectance Distribution Function (BRDF) is introduced and MODTRAN is applied to calculate background radiation for a given time at a given place. Finally the scattering of each facet is added up to get the scattering of the target. As the background radiance comes in all directions and in a wide spectrum and the complex target always consists of thousands of facets, in general it takes hours to complete the calculation. Consequently this limits its use in the real time applications. Recent years have seen the continual development of multi-core CPU. As a result parallel programming on multi-cores has been more and more popular. In this paper, the openMP, Intel CILK ++, Intel Threading Building Blocks (TBB) are used separately to leverage the processing power of multi-cores processors. Our experiments are conducted on a DELL desktop based on an Intel I7- 2600K CPU running at 3.40 GHz with 8 cores and 16.0 GB RAM. The Intel Composer 2013 is employed to build the program. Also in OpenMP implementation, gcc is used. The results demonstrate that highest speedups for three parallel models are 5.06X, 5.02X, 5.15X respectively.

Guo, Xing; Wu, Zhensen; Linghu, Longxiang

2013-09-01

99

Sensitivity of ex-core neutron detectors to vibrations of PWR fuel assemblies

The response of an ex-core neutron detector to fuel assembly vibrations in an 1150-MWe Westinghouse pressurized-water reactor (PWR) was determined by performing space-dependent reactor-kinetics calculations. The effect on the detector response of reducing the soluble-boron concentration in the coolant and fuel burnup over the first fuel cycle was also determined. The results of the calculations indicate that the ex-core neutron

F. J. Sweeney; J. P. Renier

1983-01-01

100

The burnup dependence of light water reactor spent fuel oxidation

NASA Astrophysics Data System (ADS)

The air oxidation of fragments of Light Water Reactor (LWR) spent fuel with burnup in the range 16-42 MWd/kg M was studied using thermogravimetric analysis. Experiments were conducted in dry air over the temperature range 255-325sp°C. Mass increases were generally followed until the calculated oxygen-to-metal ratio reached 2.7. LWR spent fuel was shown to oxidize via the two step reaction UOsb2-> UOsb {2.4}-> U sb3 Osb8, where the UOsb{2.4} phase is similar to cubic Usb4Osb9. The transition of UOsb2 to UOsb{2.4} was shown to be dependent on the average grain size of the specimen, with smaller-grained fuels oxidizing faster. No correlation with other fuel parameters, such as burnup, was found. The Arrhenius activation energy was calculated as 109 ± 14 kJ molsp{-1}, in agreement with reported values for oxygen diffusion in UOsb{2+x}. The oxidation of UOsb{2.4} to Usb3Osb8 was found to be strongly dependent on both temperature and burnup. At low temperature or high burnup, the fuel fragments displayed a marked resistance to oxidation beyond UOsb{2.4}, in the form of a plateau with nearly constant mass that extended for as long as 3000 hours. Both the duration of the plateau and the time-rate-of-change in the O/M ratio beyond the plateau exhibited identical burnup dependencies within experimental errors. The coefficient for the burnup dependence of the activation energy was determined as 1.2 ± 0.2 kJ molsp{-1} per MWd/kg M. The activation energy extrapolated to zero burnup was shown to agree with the value of 146 ± 10 kJ molsp{-1} reported for the oxidation of unirradiated UOsb2. The correlation of burnup with the kinetics of oxidation, and the fact that burnup is a crude measure of the concentrations of fission products and higher actinides support the conclusion that substitution of fission products and higher actinides into the vacancies in the uranium sublattice of UOsb2 that result from fission stabilizes the fluorite structure with respect to oxidation beyond UOsb{2.4}. Simple estimates show that at least one half of the burnup-dependence of the activation energy may be accounted for by the increase in lattice energy from the contraction produced by substitution of fission products and heavier actinides for uranium vacancies.

Hanson, Brady Dean

101

Hybrid parallel code acceleration methods in full-core reactor physics calculations

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

2012-07-01

102

Sensitivity and Uncertainty Analysis to Burnup Estimates on ADS using the ACAB Code

Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled

O. Cabellos; J. Sanz; A. Rodríguez; E. González; M. Embid; F. Alvarez; S. Reyes

2005-01-01

103

SCF-X?-SW calculations are performed on the ground state and core 1s hole state for CH4, NH3, H2O and HF using the overlapping sphere model, Table 1 contains the molecular charge reorganization after the renormal of an 1s electron in these molecules. The screeming of the core is further discussed. (AIP)

John S. Tse

1980-01-01

104

SCF-Xalpha-SW calculations are performed on the ground state and core 1s hole state for CH4, NH3, H2O and HF using the overlapping sphere model, Table 1 contains the molecular charge reorganization after the renormal of an 1s electron in these molecules. The screeming of the core is further discussed. (AIP)

John S. Tse

1980-01-01

105

NASA Technical Reports Server (NTRS)

To determine the feasibility of coupling the output of a single-mode optical fiber into a single-mode rib waveguide in a temperature varying environment, a theoretical calculation of the coupling efficiency between the two was investigated. Due to the complex geometry of the rib guide, there is no analytical solution to the wave equation for the guided modes, thus, approximation and/or numerical techniques must be utilized to determine the field patterns of the guide. In this study, three solution methods were used for both the fiber and guide fields; the effective-index method (EIM), Marcatili's approximation, and a Fourier method. These methods were utilized independently to calculate the electric field profile of each component at two temperatures, 20 C and 300 C, representing a nominal and high temperature. Using the electric field profile calculated from each method, the theoretical coupling efficiency between an elliptical-core optical fiber and a rib waveguide was calculated using the overlap integral and the results were compared. It was determined that a high coupling efficiency can be achieved when the two components are aligned. The coupling efficiency was more sensitive to alignment offsets in the y direction than the x, due to the elliptical modal field profile of both components. Changes in the coupling efficiency over temperature were found to be minimal.

Tuma, Margaret L.; Weisshaar, Andreas; Li, Jian; Beheim, Glenn

1995-01-01

106

Strategies for Application of Isotopic Uncertainties in Burnup Credit

Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103}Rh have also been included.

Gauld, I.C.

2002-12-23

107

Ab initio calculation of core-valence-valence Auger spectra in closed shell systems

NASA Astrophysics Data System (ADS)

We propose an ab initio method to evaluate the core-valence-valence Auger spectrum of systems with filled valence bands. The method is based on the Cini-Sawatzky theory and aims at estimating the parameters by first-principles calculations in the framework of density-functional theory (DFT). Photoemission energies and the interaction energy for the two holes in the final state are evaluated by performing DFT simulations for the system with varied population of electronic levels. Transition matrix elements are taken from atomic results. The approach takes into account the nonsphericity of the density of states of the emitting atom, spin-orbit interaction in core and valence, and nonquadratic terms in the total-energy expansion with respect to fractional occupation numbers. It is tested on two benchmark systems, Zn and Cu metals, leading in both cases to L23M45M45 Auger peaks within 2 eV from the experimental ones. Detailed analysis is presented on the relative weight of the various contributions considered in our method, providing the basis for future development. Especially problematic is the evaluation of the hole-hole interaction for systems with broad valence bands: our method underestimates its value in Cu, while we obtain excellent results for this quantity in Zn.

Fratesi, G.; Trioni, M. I.; Brivio, G. P.; Ugenti, S.; Perfetto, E.; Cini, M.

2008-11-01

108

A unified approach for calculating the core melt frequency of a specific reactor caused by both internal and external accident initiators is demonstrated. Two classes of internal initiators are examined: transients, of which turbine trip is the chosen example; and loss-of-coolant events of various sizes. The concepts of hazard and fragility analysis first proposed for seismic risk analysis are linked to the frequencies of internal initiating events, and to the plant response as a function of the event intensity. Uncertainties are propagated using discrete probability distribution (DPD) arithmetic. Advantages of this approach include mathematical and conceptual consistency, and an improved uncertainty analysis, which are important considerations if risk studies are to be utilized in decision-making based on quantitative safety goals. 5 refs., 5 figs., 5 tabs.

Heising, C.D.; Lopes de Oliveira, V. [Iowa State Univ., Ames, IA (United States)

1995-02-01

109

Ab initio no-core full configuration calculations of light nuclei

We perform no-core full configuration calculations for a set of light nuclei including {sup 16}O with a realistic NN interaction, JISP16. We obtain ground-state energies and their uncertainties through exponential extrapolations that we demonstrate are reliable in {sup 2}H, {sup 3}H, and {sup 4}He test cases where fully converged results are obtained directly. We find that {sup 6}He, {sup 6}Li, and {sup 8}He are underbound by about 600 keV, 560 keV, and 1.7 MeV, respectively. {sup 12}C is overbound by about 1.7 MeV and {sup 16}O is overbound by about 16 MeV. The first excited 0{sup +} states in {sup 12}C and {sup 16}O are also evaluated but their uncertainties are significantly larger than the uncertainties for the ground states.

Maris, P.; Vary, J. P. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Shirokov, A. M. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa 50011 (United States); Skobeltsyn Institute of Nuclear Physics, Moscow State University, RU-119991 Moscow (Russian Federation)

2009-01-15

110

Applicability of 3D Monte Carlo simulations for local values calculations in a PWR core

NASA Astrophysics Data System (ADS)

As technical support of the French Nuclear Safety Authority, IRSN has been developing the MORET Monte Carlo code for many years in the framework of criticality safety assessment and is now working to extend its application to reactor physics. For that purpose, beside the validation for criticality safety (more than 2000 benchmarks from the ICSBEP Handbook have been modeled and analyzed), a complementary validation phase for reactor physics has been started, with benchmarks from IRPHEP Handbook and others. In particular, to evaluate the applicability of MORET and other Monte Carlo codes for local flux or power density calculations in large power reactors, it has been decided to contribute to the "Monte Carlo Performance Benchmark" (hosted by OECD/NEA). The aim of this benchmark is to monitor, in forthcoming decades, the performance progress of detailed Monte Carlo full core calculations. More precisely, it measures their advancement towards achieving high statistical accuracy in reasonable computation time for local power at fuel pellet level. A full PWR reactor core is modeled to compute local power densities for more than 6 million fuel regions. This paper presents results obtained at IRSN for this benchmark with MORET and comparisons with MCNP. The number of fuel elements is so large that source convergence as well as statistical convergence issues could cause large errors in local tallies, especially in peripheral zones. Various sampling or tracking methods have been implemented in MORET, and their operational effects on such a complex case have been studied. Beyond convergence issues, to compute local values in so many fuel regions could cause prohibitive slowing down of neutron tracking. To avoid this, energy grid unification and tallies preparation before tracking have been implemented, tested and proved to be successful. In this particular case, IRSN obtained promising results with MORET compared to MCNP, in terms of local power densities, standard deviations and computing times.

Bernard, Franck; Cochet, Bertrand; Jinaphanh, Alexis; Jacquet, Olivier

2014-06-01

111

PIUS core performance analysis

A detailed evaluation of the fuel-burnup dependent power distribution and the scram reactivity for the PIUS reactor design has been performed. The analyses were carried out using the CPM lattice physics and NODE-P2 core neutronics/thermal-hydraulics codes, and are based on the information provided in the PIUS Preliminary Safety Information Document. Cycle depletion calculations were performed for a set of nine representative initial core loadings and the three-dimensional core power distributions were determined. These calculations indicate that the PIUS radial F{sub {Delta}h} and total F{sub Q} power peaking is stronger than that indicated by the PIUS reference-design values. The scram reactivity resulting from the injection of highly borated pool water was calculated for a series of time-dependent linear ramp and square-wave pool flows. The three-dimensional distribution of the borated pool water throughout the core was modeled and the spatial reactivity effects of the distributed boron were determined. For pool flows that increase as a linear ramp, the spatial reactivity effects of the distributed boron were very small. In this case, a constant core-average boron reactivity coefficient can be used to model the PIUS scram reactivity.

Carew, J.F.; Aronson, A.; Cokinos, D.M.; Prince, A.; Selcow, E.C.

1996-03-01

112

A Initio Calculations of Metal-Metal Complexes Using Relativistic Effective Core Potentials

NASA Astrophysics Data System (ADS)

Metal-metal complexes containing quadruple bonds have attracted much interest in inorganic chemistry. The two types studied were molybdenum and rhenium complexes. Previous theoretical, mainly local density functional, and experimental studies have disagreed on photoelectron and ultraviolet spectroscopic assignments. This study calculated the first few peaks in both spectra for both complexes. In order to incorporate relativistic effects, which are significant for molecules containing heavy elements, as these complexes do, we use relativistic effective core potentials (RECP). Atomic basis sets were optimized for molybdenum, rhenium, and for the second row (Na-Ar) for use with these pseudopotentials. Hartree-Fock and MCSCF calculations for these complexes were performed on the ground state and several excited states. Spin-orbit splittings of these states were found using a double group spin-orbit configuration interaction (CI) program. The known metal-metal quadruple bond was investigated and, as previously shown, was not found at the Hartree -Fock level. Corresponding calculations at the MCSCF level on the ground state of these complexes showed this quadruple bond. The criteria chosen for the description of the bonds were the Mullikan populations, the molecular orbital and CI coefficients, natural orbital populations, and contour plots. The lowest excited state for the rhenium complexes was found to be a ^3A_1, delta-delta^* state with spin-orbit coupling of the order of 40 cm ^{-1}. Calculations have also been completed on two spectroscopically important excited states, the delta-delta^* singlet state and a ligand-metal charge transfer state. Monovalent complex cations were studied for several of the complexes and compared to the photoelectron spectra. For the dimolybdenum complexes, the effects of different strengths of pi-donating ligands on the metal -metal bond were studied and were found to have a profound effect on this bond--some complexes were better described as triple-bonded with a localized orbital on one of the metals.

Blaudeau, Jean-Philippe

113

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Jung, Y. S.; Joo, H. G. [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of)] [Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Yoon, J. I. [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)] [KEPCO Nuclear Fuel, 1047 Daedukdae-ro, Yuseong-gu, Daejeon (Korea, Republic of)

2013-07-01

114

Some E1 transitions in the francium isoelectronic sequence are computed in the 'Dirac-Fock + core-polarization' approximation, where core-valence electron correlation is treated in a semiclassical core-polarization picture. The obtained ionization energies and oscillator strengths are tested versus very accurate many-body perturbation treatment (MBPT) theoretical results published recently as well as versus available experimental data. The role of core-valence correlation (core

Jacek Migdalek; Agnieszka Glowacz-Proszkiewicz

2007-01-01

115

Non-Invasive Characterization of Burnup for PWR Spent Fuel Rods with Burnups > 80 GWd/t

High-resolution gamma spectroscopy has been employed for the measurement of {sup 134}Cs/{sup 137}Cs, {sup 154}Eu/{sup 137}Cs and {sup 134}Cs/{sup 154}Eu gamma intensity ratios from spent fuel with the purpose of deriving pin-averaged single-ratio burnup indicators for high and ultra-high burnups. Two UO{sub 2} pressurised water reactor (PWR) fuel rod segments with record burnup levels >80 GWd/t have been experimentally characterised. Additionally, pin cell depletion calculations have been performed for each sample with the deterministic code CASMO-4, using both its JEF2.2- and its ENDF/B-IV-based libraries, for three different descriptions of the fuel rod irradiation histories, in order to test the sensitivity of the results to neutron cross sections and to the depletion model employed. Measured and calculated ratios have then been compared. It is shown that the {sup 134}Cs/{sup 137}Cs ratio, frequently used as burnup monitor, is considerably less accurate for values exceeding 50 GWd/t; discrepancies of up to {approx}25% are found between measured and calculated values. The ratios built with the {sup 154}Eu concentration show much larger discrepancies, essentially because this isotope is rather poorly predicted as revealed by the use of different basic cross section data. (authors)

Caruso, S.; Murphy, M.; Jatuff, F.; Chawla, R. [Paul Scherrer Institute, Laboratory for Reactor Physics and Systems Behavior, CH-5232 Villigen PSI (Switzerland)

2006-07-01

116

FaCE: a tool for three body Faddeev calculations with core excitation

NASA Astrophysics Data System (ADS)

FaCE is a self contained program, with namelist input, that solves the three body Faddeev equations. It enables the inclusion of excitation of one of the three bodies, whilst the other two remain inert. It is particularly useful for obtaining the binding energies and bound state structure compositions of light exotic nuclei treated as three-body systems, given the three effective two body interactions. A large variety of forms for these interactions may be defined, and supersymmetric transformations of these potentials may be calculated whenever two body states need to be removed due to Pauli blocking. Program summaryTitle of program: FaCE (Faddeev with Core Excitation) Catalogue identifier: ADTW Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADTW Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Computers: The code is designed to run on any Unix/Linux workstation or PC. Operating systems: Linux or UNIX Program language used: Fortran-90 Numerical libraries used: Source code for 6 routines from the NAG and BLAS libraries is included to enable independent compilation. Memory required to execute with typical data: 9 Mbytes of RAM memory and 12 MB of hard disk space. No. of bits in a word: 32 or 64 No. of bytes in distributed program, including test data, etc.: 116 514 No. of lines in distributed program, including test data, etc.: 15 574 Distribution format: tar gzip file Nature of physical problem: The program calculates eigenenergies and eigenstates for the three body problem by solving the Faddeev equations. Method of solution: Given the two body effective potentials it performs the supersymmetric transformation in case where there are forbidden states to be removed. The three body wavefunction is expanded in hyperspherical coordinates, the hyper-angular part is a series of Jacobi polynomials and the hyper-radial part is written in terms of a Laguerre basis. Within this basis the three body matrix elements are calculated and the full three body Hamiltonian matrix is completed. The diagonalization process is performed after various reductions (isospin, orthonormal and Feshbach) to determine the energies. Finally the three body wavefunction is reconstructed and other bound state observables are calculated. Typical running time: 6 s on a 1.7 GHz Intel P4-processor machine.

Thompson, I. J.; Nunes, F. M.; Danilin, B. V.

2004-08-01

117

NASA Astrophysics Data System (ADS)

Use of U and U-Th fuels in CANDU type of reactors (CANDU-6 and ACR-700) on the once-through nuclear fuel cycle is investigated. Based on the unit-cell approximation with the homogeneous-bundle/core model, utilizing the MONTEBURNS code, burnup computations are performed; discharge burnups are determined and expressed as functions of 235U and Th fractions, when applicable. Natural Uranium Requirement (and Saving) and Nuclear Resource Utilization are calculated for varying fuel compositions. Results are analyzed to observe the effects of 235U and Th fractions, thus to reach conclusions about use of Th in CANDU-6 and ACR-700 on the once-through cycle.

Türkmen, Mehmet; Zabuno?lu, Okan H.

2012-10-01

118

Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements’ burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element’s reported burnup or provide a burnup estimate for an element with an unknown burnup.

Winston, Philip Lon; Sterbentz, James William

2001-04-01

119

Gross Gamma Dose Rate Measurements for TRIGA Spent Nuclear Fuel Burnup Validation

Gross gamma-ray dose rates from six spent TRIGA fuel elements were measured and compared to calculated values as a means to validate the reported element burnups. A newly installed and functional gamma-ray detection subsystem of the In-Cell Examination System was used to perform the measurements and is described in some detail. The analytical methodology used to calculate the corresponding dose rates is presented along with the calculated values. Comparison of the measured and calculated dose rates for the TRIGA fuel elements indicates good agreement (less than a factor of 2 difference). The intent of the subsystem is to measure the gross gamma dose rate and correlate the measurement to a calculated dose rate based on the element s known burnup and other pertinent spent fuel information. Although validation of the TRIGA elements' burnup is of primary concern in this paper, the measurement and calculational techniques can be used to either validate an element's reported burnup or provide a burnup estimate for an element with an unknown burnup. (authors)

Winston, P.L.; Sterbentz, J.W. [INEEL - Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID 83415 (United States)

2002-07-01

120

NASA Astrophysics Data System (ADS)

Leg 199 drilled a series of sites in the equatorial Pacific in order to investigate the paleoceanography of the Paleogene Pacific Ocean. The two deepest cored sites, (1218 and 1219) have provided continuous/near continuous spliced sedimentary sections and in situ wireline log data. Comparison of core to log data sets shows the familiar non-linear, increasing with depth, miss-match between the core (metres composite depth - mcd) and log (mbsf) depths and concomitant offset between core and log physical property data sets e.g. porosity, density, velocity. The depth miss-matches represent core expansion due to elastic rebound experienced by the sediments upon unloading i.e. removal of overburden stress, which is a function of the sediment void ratio and log of the effective in situ stress. The increasing depth offset observed between the 1218 core and log data is used to calculate an expansion index (Cr) for continuous discrete measurement intervals, down the core. The Cr values are used to re-compress the core (mcd) depth scale and as expected provide a good match with the log (mbsf) depths. The Cr values are also used to correct the core index property data, to in situ values. The quality of the corrected core index property data is good when compared with the in situ measured log data. Cr values are dependent upon the sediment composition (especially the quantity of clay) and core light absorption spectroscopy (LAS) data collected on Leg 199, provides a continuous down-core record of sediment composition, in terms of the percent clay, carbonate and opal. A relationship between the Cr values and the sediment LAS composition is established and is then applied to the Site 1219 core LAS data, allowing appropriate Cr values to be assigned to continuous, discrete core intervals. These composition based Cr values are then used to re-calculate the core (mcd) depths and correct the index property data to in situ values. The quality of the depth and index property corrections are checked by comparison with the in situ measured log data, and provide encouraging results.

Rea, B. R.

2002-12-01

121

A Simple and Efficient Control Rod Cusping Model for Three-Dimensional Pin-by-Pin Core Calculations

A new solution for the control rod cusping problem in the three-dimensional pin-by-pin core calculation is proposed in this paper. The current advanced nodal code resolves this issue by estimating the one-dimensional axial flux distribution in a partially rodded node. However, direct application of this approach to the three-dimensional pin-by-pin calculation is impractical since the leakage effect in the radial

Akio

2004-01-01

122

Analysis of TRAC-PF1 Calculated Core Heat Transfer for CCTF Test C1-5 (Run 14).

National Technical Information Service (NTIS)

A TRAC-PF1 post test calculation for CCTF test C1-5 (Run 14) was performed to assess the core thermal-hydraulic models in the TRAC-PF1 code during the reflood phase of a PWR LOCA. TRAC showed good agreement with data for heater rods turnaround temperature...

H. Akimoto

1985-01-01

123

Consideration of the depletion phenomena and isotopic uncertainties in burnup-credit criticality analysis places an increasing reliance on computational tools and significantly increases the overall complexity of the calculations. An automated analysis and data management capability is essential for practical implementation of large-scale burnup credit analyses that can be performed in a reasonable amount of time. STARBUCS is a new prototypic analysis sequence being developed for the SCALE code system to perform automated criticality calculations of spent fuel systems employing burnup credit. STARBUCS is designed to help analyze the dominant burnup credit phenomena including spatial burnup gradients and isotopic uncertainties. A search capability also allows STARBUCS to iterate to determine the spent fuel parameters (e.g., enrichment and burnup combinations) that result in a desired k{sub eff} for a storage configuration. Although STARBUCS was developed to address the analysis needs for spent fuel transport and storage systems, it provides sufficient flexibility to allow virtually any configuration of spent fuel to be analyzed, such as storage pools and reprocessing operations. STARBUCS has been used extensively at Oak Ridge National Laboratory (ORNL) to study burnup credit phenomena in support of the NRC Research program.

Gauld, I.C.

2001-07-20

124

Review of Technical Studies in the United States in Support of Burnup Credit Regulatory Guidance

Taking credit for the reduction in reactivity associated with fuel depletion can enable more cost-effective, higher-density storage, transport, disposal, and reprocessing of spent nuclear fuel (SNF) while maintaining sufficient subcritical margin to establish an adequate safety basis. Consequently, there continues to be considerable interest in the United States (U.S.), as well as internationally, in the increased use of burnup credit in SNF operations, particularly related to storage, transport, and disposal of commercial SNF. This interest has motivated numerous technical studies related to the application of burnup credit, both domestically and internationally, as well as the design of SNF storage, transport and disposal systems that rely on burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission (NRC) initiated a burnup credit research program in 1999, with support from the Oak Ridge National Laboratory (ORNL), to develop regulatory guidance and the supporting technical bases for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. Although this NRC research program has not been continuous since its inception, considerable progress has been achieved in many key areas in terms of increased understanding of relevant phenomena and issues, availability of relevant information and data, and subsequently updated regulatory guidance for expanded use of burnup credit. This paper reviews technical studies performed by ORNL for the U.S. NRC burnup credit research program. Examples of topics include reactivity effects associated with reactor operating characteristics, fuel assembly characteristics, burnable absorbers, control rods, spatial burnup distributions, cooling time, and assembly misloading; methods and data for validation of isotopic composition predictions; methods and data for validation of criticality calculations; and operational issues and data related to assembly burnup confirmation. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details, and provide a useful resource to others in the burnup credit community.

Wagner, John C [ORNL] [ORNL; Parks, Cecil V [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL

2010-01-01

125

A Simple and Efficient Control Rod Cusping Model for Three-Dimensional Pin-by-Pin Core Calculations

A new solution for the control rod cusping problem in the three-dimensional pin-by-pin core calculation is proposed in this paper. The current advanced nodal code resolves this issue by estimating the one-dimensional axial flux distribution in a partially rodded node. However, direct application of this approach to the three-dimensional pin-by-pin calculation is impractical since the leakage effect in the radial direction is significant and the one-dimensional model for axial flux distribution is no longer valid. This issue has been neither addressed nor resolved yet. In this paper, a new approach that utilizes the inverse of the spectral index obtained in the assembly calculation is used to estimate the flux distribution inside the partially rodded mesh. The proposed model was implemented in the SCOPE2 code, which is a three-dimensional pin-by-pin nodal-transport code for pressurized water reactor core calculations, and a verification calculation was carried out to confirm the validity of the proposed method. From the calculation results, oscillation in the differential worth of control rods (i.e., the cusping effect) is damped, and the proposed model can almost reproduce that obtained by the reference calculation. The additional computation time for the proposed model is negligible. Consequently, the proposed control rod cusping model is an attractive method in three-dimensional pin-by-pin calculations.

Yamamoto, Akio [Nuclear Fuel Industries, Ltd. (Japan)

2004-01-15

126

Light scattering calculations from Au and Au/SiO2 core/shell nanoparticles

NASA Astrophysics Data System (ADS)

Given the importance of the optical properties of Au and Au/SiO2 core/shell nanoparticles, in this article we focus our attention on the light scattering properties of such systems and on a relative comparison. In particular, we report theoretical results of angle-dependent light scattering intensity and scattering efficiency for Au and Au/SiO2 core/shell nanoparticles increasing the Au particle radius from 30 to 130 nm, and for Au/SiO2 core/shell particles changing the core-to-shell sizes ratio. Finally, a comparison between the scattering efficiency of the Au and Au/SiO2 core/shell nanoparticles is drawn. The results of this work can be used in the design of tunable efficiency light scattering devices (biological and molecular sensors, solar cells).

Ruffino, F.; Pugliara, A.; Carria, E.; Bongiorno, C.; Grimaldi, M. G.

2013-01-01

127

NASA Astrophysics Data System (ADS)

Two weak points in Performance Assessment (PA) exercises regarding the alteration of Spent Nuclear Fuel (SNF) are the contribution of the so-called Instant Release Fraction (IRF) and the effect of High Burn-Up Structure (HBS). This manuscript focuses on the effect of HBS in matrix (long term) and instant release of a Pressurised Water Reactor (PWR) SNF irradiated in a commercial reactor with a mean Burn-Up (BU) of 60 GWd/tU. In order to study the HBS contribution, two samples from different radial positions have been prepared. One from the centre of the SNF, labelled CORE, and one from the periphery, enriched with HBS and labelled OUT. Static leaching experiments have been carried out with two synthetic leaching solutions: bicarbonate (BIC) and Bentonitic Granitic Groundwater (BGW), and in all cases under oxidising conditions. IRF values have been calculated from the determined Fraction of Inventory in Aqueous Phase (FIAP). In all studied cases, some radionuclides (RN): Rb, Sr and Cs, have shown higher release rates than uranium, especially at the beginning of the experiment, and have been considered as IRF. Redox sensitive RN like Mo and Tc have been found to dissolve slightly faster than uranium and further studies might be needed to confirm if they can also be considered part of the IRF. Most of the remaining studied RN, mainly actinides and lanthanides, have been found to dissolve congruently with the uranium matrix. Finally, Zr, Ru and Rh presented lower release rates than the matrix. Higher matrix release has been determined for CORE than for OUT samples showing that the formation of HBS might have a protective effect against the oxidative corrosion of the SNF. On the contrary, no significant differences have been observed between the two studied leaching solutions (BIC and BGW). Two different IRF contributions have been determined. One corresponding to the fraction of inventory segregated in the external open grain boundaries, directly available to water and very significant during the first days of the experiment; and a second one corresponding to a less accessible, most probably located at the internal grain boundaries, one order of magnitude lower than the first one at equal given dissolution times but of much longer period of incidence. Unlike matrix release results, higher Cs IRF release was found for OUT than for CORE sample. This effect can be attributed to thermal migration of Cs to the periphery of the fuel during irradiation. In the case of Rb no clear differences were observed between CORE and OUT showing equilibrium between the opposing thermal migration and matrix effects. Finally, Sr CORE/OUT release ratio showed similar behaviour to matrix release, thus proving no significant thermal migration during irradiation.

Serrano-Purroy, D.; Clarens, F.; González-Robles, E.; Glatz, J. P.; Wegen, D. H.; de Pablo, J.; Casas, I.; Giménez, J.; Martínez-Esparza, A.

2012-08-01

128

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT

The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C{sub k}, approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C{sub k} > 0.95) were attained at or near the end of a reactor cycle. The C{sub k} values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K{sub eff} sensitivity profiles toward higher energies in the thermal region as compared to the K{sub eff} sensitivity profile of the generic cask. Parameters representing coverage of the application by the CRCs on an energy-dependent, nuclide-reaction specific level (i.e., effectiveness of the CRCs for validating the cross sections as used in the application) were also examined. Based on the CRCs with C{sub k} > 0.8 and an assumed relative standard deviation for uncovered covariance data of 25%, the relative standard deviation of K{sub eff} due to uncovered sensitivity data varies from 0.79% to 0.95% for cask burnups ranging from 10 to 60 GWd/MTU. As expected, this uncertainty in K{sub eff} is largely dominated by noncoverage of sensitivities from major actinides and hydrogen. The contributions from fission products and minor actinides are very small and comparable to statistical uncertainties in K{sub eff} results. These results (again, assuming a 25% uncertainty for uncovered covariance data) indicate that there could be approximately 1% uncertainty in the calculated application K{sub eff} due to incomplete neutronic testing (validation) of the software by the CRCs. However, this conclusion also assumes all other uncertainties in the complex CRC configurations (e.g., isotopic compositions of burned fuel, operation history, data) are well known. Thus, an evaluation of the uncertainties in the CRC configurations is needed prior to the use of CRCs for code validation (i.e., quantifying code bias and bias uncertainty).

Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Wagner, John C [ORNL

2009-01-01

129

Production of {sup 99}Mo at the annular core research reactor-recent calculative results

Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem-type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.

Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)

1997-12-01

130

Mo-99 production at the Annular Core Research Reactor - recent calculative results

Significant progress has been made over the past year in understanding the chemistry and processing challenges associated with {sup 99}Mo production using Cintichem type targets. Targets fabricated at Los Alamos National Laboratory have been successfully irradiated in fuel element locations at the Annular Core Research Reactor (ACRR) and processed at the Sandia Hot Cell Facility. The next goal for the project is to remove the central cavity experiment tube from the reactor core, allowing for the irradiation of up to 37 targets. After the in-core work is complete, the reactor will be capable of producing significant quantities of {sup 99}Mo.

Parma, E.J.

1997-11-01

131

NASA Astrophysics Data System (ADS)

We implemented the calculation of hyperfine tensors into such plane wave supercell code working with the projector augmentation wave method that incorporates hybrid density functional theory and the contribution of the spin polarization of the core states. We show that the combination of HSE06 hybrid density functional together with the contribution of the core spin polarization provides accurate results on prominent point defects in various semiconductors, where the latter effect may be enormously large, in contrast to previous expectations. We briefly discuss the relevance of our results in the light of realization of solid-state quantum bits by paramagnetic point defects.

Szász, Krisztián; Hornos, Tamás; Marsman, Martijn; Gali, Adam

2013-08-01

132

Radial Power Profile of MOX and LEU Fuel Pellet Versus Burnup

One of challenge to burn the WG-Pu in Mixed Oxide (MOX) fuel in light water reactors (LWR) is to demonstrate that the differences between WG-MOX, RG-MOX, and LWR LEU fuel are minimal, and therefore, the commercial MOX and LEU fuel experience base is applicable. The MCWO-calculated Radial Power Profile of LEU, Weapons Grade-MOX and Reactor Grade-MOX fuel pellets at various burnups are similar toward the end of life (50 GWd/t). Therefore, the LEU fuel performance evaluation code - FRAPCON-3 with modifications, such as, the detailed fission power profiles versus burnup, can be used in the MOX fuel pellet performance analysis. MCWO also calculated the {sup 240}Pu/Pu ratio in WG-MOX versus burnup, which reaches an average of 31.25% at discharged burnup of 50 GWd/t. It meets the spent fuel standard for WG-Pu disposition in LWR. (authors)

Chang, Gray S.; Pedersen, Robert C. [INEEL - Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID 83415 (United States)

2002-07-01

133

Investigation of resonant properties of metal core–shell nanoparticles using T-matrix calculations

Using T-matrix method, plasmon resonance properties of metal core–shell nanoparticles are systematically investigated. It is shown that dielectric\\/metal core–shell structure may be excited stronger at resonance than metal\\/dielectric and hetero-metal ones, but the resonance states are extremely sensible to the layers thickness. For three-layer nanospheres, resonance properties will be dominated by a sub-10nm silver outermost shell, while only weakened by

Lei Liu; Bin Wang; Xuewei Cao; Xiaoxuan Xu; Yufang Wang

2011-01-01

134

Considering the importance of thorium data and concerning about the accuracy of Th-232 cross section library, a series of experiments of thorium critical core carried out at KUCA facility of Kyoto Univ. Research Reactor Inst. have been analyzed. The core was composed of pure thorium plates and 93% enriched uranium plates, solid polyethylene moderator with hydro to U-235 ratio of 140 and Th-232 to U-235 ratio of 15.2. Calculations of the effective multiplication factor, control rod worth, reactivity worth of Th plates have been conducted by MVP code using JENDL-4.0 library [1]. At the experiment site, after achieving the critical state with 51 fuel rods inserted inside the reactor, the measurements of the reactivity worth of control rod and thorium sample are carried out. By comparing with the experimental data, the calculation overestimates the effective multiplication factor about 0.90%. Reactivity worth of the control rods evaluation using MVP is acceptable with the maximum discrepancy about the statistical error of the measured data. The calculated results agree to the measurement ones within the difference range of 3.1% for the reactivity worth of one Th plate. From this investigation, further experiments and research on Th-232 cross section library need to be conducted to provide more reliable data for thorium based fuel core design and safety calculation. (authors)

Mai, V. T.; Fujii, T.; Wada, K.; Kitada, T. [Osaka Univ., 2-1, Yamadaoka, Suita-shi, Osaka 565-0871 (Japan); Takaki, N.; Yamaguchi, A.; Watanabe, H. [Tokai Univ., 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa, 259-1292 (Japan); Unesaki, H. [Kyoto Univ. Research Reactor Inst., Asahiro-nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan)

2012-07-01

135

Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel.

Representative accident source terms patterned after the NUREG-1465 Source Term have been developed for high burnup fuel in BWRs and PWRs and for MOX fuel in a PWR with an ice-condenser containment. These source terms have been derived using nonparametric order statistics to develop distributions for the timing of radionuclide release during four accident phases and for release fractions of nine chemical classes of radionuclides as calculated with the MELCOR 1.8.5 accident analysis computer code. The accident phases are those defined in the NUREG-1465 Source Term - gap release, in-vessel release, ex-vessel release, and late in-vessel release. Important differences among the accident source terms derived here and the NUREG-1465 Source Term are not attributable to either fuel burnup or use of MOX fuel. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the NUREG-1465 Source Term. An additional radionuclide chemical class has been defined to account for release of cesium as cesium molybdate which enhances molybdenum release relative to other metallic fission products.

Salay, Michael (U.S. Nuclear Regulatory Commission, Washington, D.C.); Gauntt, Randall O.; Lee, Richard Y. (U.S. Nuclear Regulatory Commission, Washington, D.C.); Powers, Dana Auburn; Leonard, Mark Thomas

2011-01-01

136

Dependence of Fast Reactor Fuel Burnup Characteristics on Nuclear Data Libraries

In the framework of the development of burnup calculation method for commercial fast reactors, a sensitivity analysis was carried out to clarify the dependence of fuel burnup characteristics on nuclear data libraries (NDLs). The following NDLs were compared: JEF-2.2, ENDF\\/B-VI Release 5, JENDL-3.2 and JENDL-3.3. The NDL-dependence of material balance for main heavy metal nuclides (U, U, Pu, Pu and

Shigeo OHKI; Tomoyuki JIN

2005-01-01

137

Effect of Fuel Fraction on Small Modified CANDLE Burn-up Based Gas Cooled Fast Reactors

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE Burn-up has been performed. The objective of this research is to get optimal design parameters of such type reactors. The parameters of nuclear design including the critical condition, conversion ratio, and burn-up level were compared. These parameters are calculated by variation in the fuel fraction 47.5% up to

Menik Ariani; Zaki Su'Ud; Abdul Waris; Khairurrijal; Nur Asiah; M. Ali Shafii

2010-01-01

138

The REBUS Experimental Programme for Burn-Up Credit

An international programme called REBUS (Reactivity tests for a direct evaluation of the Burn-Up credit on Selected irradiated LWR fuel bundles) for the investigation of the burn-up credit has been initiated by the Belgian Nuclear Research Center SCK.CEN and Belgonucleaire with the support of USNRC, EdF from France, VGB, representing German nuclear utilities and NUPEC, representing the Japanese industry. The programme aims to establish a neutronic benchmark for reactor physics codes. This benchmark will qualify the codes to perform calculations of the burn-up credit. The benchmark exercise investigates the following fuel types with associated burn-up: - Reference 3.3% enriched UO{sub 2} fuel; - Fresh commercial PWR UO{sub 2} fuel; - Irradiated commercial PWR UO{sub 2} fuel (51 GWd/tM); - Fresh PWR MOX fuel; - Irradiated PWR MOX fuel (20 GWd/tM). Reactivity effects are measured in the critical facility VENUS. Fission rate and flux distributions in the experimental bundles will be determined. The accumulated burn-up of all rods is measured non-destructively in a relative way by gross gamma-scanning, while some rods are examined by gamma-spectrometry for an absolute determination of the burn-up. Some rods will be analyzed destructively with respect to accumulated burn-up, actinides content and TOP-19 fission products (i.e. those non-gaseous fission products that have most implications on the reactivity). Additionally some irradiated rods have undergone a profilometry and length determination. The experimental implementation of the programme has started in 2000 with major changes in the VENUS critical facility. Gamma scans, profilometry, length determination and gamma-spectrometry measurements on the MOX fuel have been performed. In the course of October 2001 the first fresh fuel configuration will be investigated. In the same period the commercial irradiated fuel will arrive at the SCK.CEN hot cells and will be re-fabricated into fuel rodlets of 1 meter length. (authors)

D'hondt, Pierre; Van der Meer, Klaas; Baeten, Peter [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Marloye, Daniel; Lance, Benoit; Basselier, Jacques [Belgonucleaire (Belgium)

2002-07-01

139

Burnup analysis of rock-like oxide fuel disks irradiated in the Japan Research Reactor No. 3

NASA Astrophysics Data System (ADS)

Burnup analysis of rock-like oxide (ROX) fuel disks has been carried out and the results have been compared with measured values. Two kinds of ROX disks: zirconia and thoria, were fabricated and irradiated in an irradiation hole of the Japan Research Reactor No. 3 (JRR-3). After irradiation, several post-irradiation examinations (PIE) were performed. Computer codes used for the calculations were the SRAC and the MVP-BURN codes. Firstly, the neutron spectrum in the irradiation hole was calculated using the SRAC code system. Fixed source problems were solved to obtain the neutron spectra and effective cross-sections of the disks and burnup calculations were performed. The calculated results of burnup, isotopic abundance of plutonium and production of americium and curium were compared with measurement values. Calculations overestimate the measured burnup by 7 15% and both codes largely underestimate the measured production of americium and curium isotopes. The calculated plutonium abundance agrees moderately well with the measured values.

Nakano, Y.; Akie, H.; Magara, M.; Takano, H.

1999-08-01

140

Ab initio shell-model calculation for ^{18}O in a restricted no-core model space

We perform an ab initio shell-model calculation for ^{18}O in a restricted no-core model space, microscopically deriving a two-body effective interaction and introducing a minimal refinement of one-body energies in the spsd or spsdpf model space. Low-lying energy levels, except for the experimental 0_{2}^{+} and 2_{3}^{+} states, are better described in the spsdpf space than in the spsd space. The

S. Fujii; B. R. Barrett

2009-01-01

141

Within the framework of density-functional theory, first-principles pseudopotential methods have been highly successful in modeling the valence-electron properties of solids in their ground states. In this paper, we introduce ``core-cancellation functions'' which are designed to improve the accuracy of the treatment of the exchange-correlation interaction. This formalism, expected to be especially effective for transition metals, is tested for bulk tungsten

N. A. W. Holzwarth; Y. Zeng

1994-01-01

142

Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses

This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.

Wagner, J.C.

2002-10-23

143

The sensitivity and uncertainty of various core burnup performance quantities (e.g., k[sub eff], burnup reactivity swing, local power density, etc.) to the heavy isotope fission spectrum parameters was investigated using depletion perturbation methods and ENDF\\/B-V covariance data. A brief description of the methods is followed by results of a 900-MW(thermal) fast reactor. The analysis here indicates that for a 900-MW(thermal)

T. J. Downar; J. Broda; J. Kritzer

1990-01-01

144

NASA Astrophysics Data System (ADS)

By combining density functional molecular dynamics simulations with a thermodynamic integration technique, we determine the free energy of metallic hydrogen and silica, SiO2, at megabar pressures and thousands of degrees Kelvin. Our ab initio solubility calculations show that silica dissolves into fluid hydrogen above 5000 K for pressures from 10 and 40 Mbars, which has implications for the evolution of rocky cores in giant gas planets like Jupiter, Saturn, and a substantial fraction of known extrasolar planets. Our findings underline the necessity of considering the erosion and redistribution of core materials in giant planet evolution models, but they also demonstrate that hot metallic hydrogen is a good solvent at megabar pressures, which has implications for high-pressure experiments.

González-Cataldo, F.; Wilson, Hugh F.; Militzer, B.

2014-05-01

145

Multidimensional transition-state theory calculations for nuclear dynamics of core-excited molecules

We have extended the transition-state theory to describe the dynamics of core hole excitation. This allows us to interpret the abnormally bland near-edge x-ray absorption fine structure spectrum of the gas phase benzonitrile molecule at the N 1s edge. We have brought to light different paths for the two most intensive resonances, going from the linear to the bent structure. The profile of each resonance consists of two different vibrational progressions corresponding to stretching modes and a broad continuum of bending excited states.

Carniato, Stephane; Gallet, Jean-Jacques [Laboratoire de Chimie Physique, Matiere et Rayonnement, UMR 7614, Universite Pierre et Marie Curie, 11, rue Pierre et Marie Curie, 75231 Paris Cedex 05 (France); Ilakovac, Vita [Laboratoire de Chimie Physique, Matiere et Rayonnement, UMR 7614, Universite Pierre et Marie Curie, 11, rue Pierre et Marie Curie, 75231 Paris Cedex 05 (France); Universite de Cergy Pontoise, 95031, Cergy Pontoise Cedex (France); Kukk, Edwin [Department of Physical Sciences, University of Oulu, P.O. Box 3000, FIN-90014, Oulu (Finland); Luo Yi [Theoretical Chemistry, Royal Institute of Technology, AlbaNova, S-106 91 Stockholm (Sweden)

2004-09-01

146

NASA Astrophysics Data System (ADS)

The photoabsorption spectra of highly excited Rydberg atoms in external fields are known to be related to closed classical orbits of the excited electron. For atoms in a pure magnetic field, scaled recurrence spectra have successfully been calculated from classical orbits using the semiclassical closed-orbit theory. Rydberg atoms in crossed fields are more complicated as regards their theoretical and numerical treatment, as the Hamiltonian is nonseparable in three degrees of freedom, in contrast to only two nonseparable degrees of freedom in a pure magnetic field. In this paper, we present an extension of the closed-orbit theory to three degrees of freedom, taking into account arbitrary quantum defects. Motivated by nonhydrogenic resonances discovered in experimental scaled recurrence spectra of rubidium atoms, we investigate the influence of the ionic core on the three-dimensional closed orbits, using a simple model potential. We find that the introduction of a core potential results in an extreme increase of the number of closed orbits as compared to hydrogen. The novel orbits appear to be composed of hydrogenic orbits and are created through scattering by the core potential. Investigating the classical deflection function of the model potential, general properties of the new orbits can be explained. With the closed-orbit theory extended to three nonseparable degrees of freedom, we are able to calculate scaled recurrence spectra for Rydberg atoms in crossed fields with arbitrary quantum defects. Our results are in good agreement with the experimental spectra. In particular, the nonhydrogenic resonances can be explained in terms of the new orbits created by classical core scattering.

Weibert, Kirsten; Main, Jörg; Wunner, Günter

1998-09-01

147

NASA Astrophysics Data System (ADS)

Herein, we report finite-element calculations of the effective (relative) permittivity of composite materials consisting of inclusions and inclusion arrays with a core-shell structure embedded in a surrounding host. The material making up the core of the two-dimensional structures, or cross sections of infinite three-dimensional objects (parallel, infinitely long, and identical cylinders) where the properties and characteristics are invariant along the perpendicular cross sectional plane, is assumed to have a negative real part of the permittivity, while the coating material (annular shell) is considered to be lossless. While strictly valid only in a dc situation, our analysis can be extended to treat electric fields that oscillate with time, provided that the wavelengths and attenuation lengths associated with the fields are much larger than the microstructure dimension in order that the homogeneous (effective-medium) representation of the composite structure makes sense. While one may identify features of the electrostatic resonance (ER) which are common to core-shell structures characterized by permittivities with real parts of opposite signs, it appears that the predicted ER positions are sensitive to the shell thickness and can be tuned through varying this geometric parameter. For example, we observe that the ER is broadened and shifted as the loss and the shell thickness are increased, respectively. We also argue that such core shell may also be valuable in controlling ER characteristics via polarization in an external electric field. In addition, by considering calculations of the electric field distribution, we find that the ER results in very strong and local-field enhancements into small parts of the shell perimeter. Our findings open up possibilities for the development of hybrid structures that could exploit the ER features for a particular application.

Mejdoubi, Abdelilah; Brosseau, Christian

2007-11-01

148

Some Nuclear Calculations of U-235-D2O Gaseous-Core Cavity Reactors

NASA Technical Reports Server (NTRS)

The results of a multigroup, diffusion theory study of spherical gaseous-core cavity reactors are presented in this report. The reactor cavity of gaseous U235 is enclosed by a region of hydrogen gas and is separated from an external D2O moderator-reflector by a zirconium structural shell. Some cylindrical reactors are also investigated. A parametric study of spherical reactors indicates that, for the range of variables studied, critical mass increases as: (1) Fuel region is compressed within the reactor cavity, (2) moderator thickness is decreased, (3) structural shell thickness is increased, and (4) moderator temperature is increased. A buckling analogy is used to estimate the critical mass of fully reflected cylindrical reactors from spherical results without fuel compression. For a reactor cavity of a 120-centimeter radius uniformly filled with fuel, no structural shell, a moderator temperature of 70 F, and a moderator thickness of 100 centimeters, the critical mass of a spherical reactor is 3.1 kilograms while that of a cylinder with a length-to-diameter ratio of 1.0 (L/D = 1) is approximately 3.8 kilograms and, with L/D = 2, 5.9 kilograms. For the range of variables considered for U235-D2O gaseous-core cavity reactors, the systems are characterized by 95 to 99 percent thermal absorptions, with the flux reaching a maximum in the moderator about 10 to 15 centimeters from the reactor cavity.

Ragsdale, Robert G.; Hyland, Robert E.

1961-01-01

149

Reactor physics calculations for {sup 99}Mo production at the annular core research reactor

The Isotope Production and Distribution Program at the U.S. Department of Energy has designated Sandia National Laboratories (SNL) as the most appropriate facility for the production of {sup 99}Mo, a radioisotope whose daughter, {sup 99m}Tc, is used in more than 36,000 medical procedures per day in the United States and is considered to be a vital medical diagnostic and treatment tool. The isotope would be produced at SNL using the annular core research reactor (ACRR) facility and collocated hot cell facility. The {sup 99}Mo would be produced using the fission process by irradiating {open_quotes}targets{close_quotes} coated with {sup 235}U in the form of highly enriched U{sub 3}O{sub 8}. After {approximately}7 days of continuous irradiation in the ACRR, a target would be re- moved from the reactor core for processing. The isotope would be extracted by chemically precipitating the molybdenum using the {open_quotes}Cintichem{close_quotes} process and would be shipped to the various pharmaceutical companies by commercial or chartered airline.

Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States)

1995-12-31

150

Isotope Analysis. Radiochemical Burn-Up Determination.

National Technical Information Service (NTIS)

The radiochemical analysis of heavy isotopes (uranium, plutonium, americium, and curium isotopes) and burn-up (Nd148 method) in samples of irradiated nuclear fuels is described. The analysis is based on isotope dilution technique in combination with mass ...

N. Rhod Larsen E. Larsen

1979-01-01

151

Comparison of Burnup Credit Uncertainty Quantification Methods.

National Technical Information Service (NTIS)

The recent revision of Interim Staff Guidance (ISG-8 revision 3), issued by the US Nuclear Regulatory Commission, on burnup credit for criticality safety analysis of pressurized water reactor spent fuel in transportation and storage casks expands the nucl...

G. Ilas I. C. Gauld K. S. Kim W. A. Wieselquist

2013-01-01

152

Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during

S. Tashakor; G. Jahanfarnia; M. Hashemi-Tilehnoee

2010-01-01

153

One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets

Don Mueller; Bradley T Rearden; Davis Allan Reed

2010-01-01

154

ERIC Educational Resources Information Center

Five activities are presented in this student workbook on using the electronic calculator. Following the directions for using the machine, problems are given on multiplying and dividing, finding percentages, calculating the area of assorted polygons, changing fractions to decimals, and finding squares and square roots. (JH)

Parma City School District, OH.

155

The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the

Federico Puente Espel

2010-01-01

156

NASA Astrophysics Data System (ADS)

Since the identification of f-orbital contribution to the bonding in PaO+, investigations into Pa cations have hoped to characterize as many of the electronic states possible.1 Electronic states of the Pan+ (n=0-4) ions have been investigated using multi-reference spin-orbit configuration interaction (MR-SOCI). Initial investigations using Dunning style correlation consistent double-{?} basis sets are re-examined with a larger triple-{?} basis, with the hope of supporting the order of electronic states. Calculations using Hartree-Fock and CI calculations on the neutral atom did not produce the known order of states. A case study was deemed necessary on similar electron configurations present in the low energy states of Pa2+ more specifically those generated from the 5f26d1 and 5f16d2 configurations. Comparison in the Pa2+ ion is complicated by the lack of experimental results, but the states are presumed to be similar sequence as those in the neutral atom, with the addition of two electrons in the 7s shell. In evaluating the impact of inclusion of the outer core, calculations including valence-outer core correlation were completed for the 5d, 6s, and 6p shells of the Pa2+ ion. The magnitude of these individual shell correlation calculations will allow for identification of the energy level shifts associated with even and odd configurations, better describing the energy order in both the Pa2+ ion case study and for the neutral Pa atom. Upon completion of this aspect of the Pa neutral atom study, the knowledge of the energy levels in the Pan+ (n=0-4) family of ions will be greatly expanded, and may yield a model for future studies of atomic actinide systems. Gibson {et al.} Organometallics 2007, 26, 3947-3956.

Mrozik, Michael K.; Pitzer, Russell M.; Bursten, Bruce E.

2010-06-01

157

Iterative Transport-Diffusion Methodology For LWR Core Analysis

NASA Astrophysics Data System (ADS)

This paper presents an update on the development of an advanced methodology for core calculations that uses local heterogeneous solutions for on-the-fly nodal cross-section generation. The Iterative Transport-Diffusion Method is an embedded transport approach that is expected to provide results with near 3D transport accuracy for a fraction of the time required by a full 3D transport method. In this methodology, the infinite environment used for homogenized nodal cross-section generation is replaced with a simulated 3D environment of the diffusion calculation. This update focuses on burnup methodology, axial leakage and 3D modeling.

Colameco, David; Ivanov, Boyan D.; Beacon, Daniel; Ivanov, Kostadin N.

2014-06-01

158

Nondestructive Assay of Nuclear Low-Enriched Uranium Spent Fuels for Burnup Credit Application

Criticality safety analysis devoted to spent-fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent-fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as burnup credit. To be used, burnup credit involves obtaining evidence of the reactivity loss with a burnup measurement.Many nondestructive assays (NDA) based on neutron as well as on gamma-ray emissions are devoted to spent-fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously, and the link to burnup is a power link. As a result, burnup determination with passive neutron measurement is extremely accurate.Some gamma emitters also have interesting properties in order to characterize spent fuels, but the convenience of the gamma spectrometric methods is very dependent on the characteristics of the spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels.Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail:1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several nuclear power plants in western Europe, gives the average burnup within a 5% uncertainty and also the extremity burnup.2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (active neutron interrogation, passive neutron counting, and gamma spectrometry)

Lebrun, Alain; Bignan, Gilles [Commissariat a l'Energie Atomique CEA (France)

2001-09-15

159

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually, we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J. P.; Maris, P.; Honkanen, H.; Li, J. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Shirokov, A. M. [Department of Physics and Astronomy, Iowa State University, Ames, Iowa, 50011 (United States); Skobeltsyn Institute of Nuclear Physics, Moscow State University, Moscow, 119991 (Russian Federation); Brodsky, S. J. [SLAC National Accelerator Laboratory, Stanford University, Menlo Park, California (United States); Harindranath, A. [Theory Group, Saha Institute of Nuclear Physics, 1/AF, Bidhannagar, Kolkata, 700064 (India); Teramond, G. F. de [Universidad de Costa Rica, San Jose (Costa Rica)

2009-12-17

160

Ab initio no core calculations of light nuclei and preludes to Hamiltonian quantum field theory

Recent advances in ab initio quantum many-body methods and growth in computer power now enable highly precise calculations of nuclear structure. The precision has attained a level sufficient to make clear statements on the nature of 3-body forces in nuclear physics. Total binding energies, spin-dependent structure effects, and electroweak properties of light nuclei play major roles in pinpointing properties of the underlying strong interaction. Eventually,we anticipate a theory bridge with immense predictive power from QCD through nuclear forces to nuclear structure and nuclear reactions. Light front Hamiltonian quantum field theory offers an attractive pathway and we outline key elements.

Vary, J.P.; Maris, P.; /Iowa State U.; Shirokov, A.M.; /Iowa State U. /SINP, Moscow; Honkanen, H.; li, J.; /Iowa State U.; Brodsky, S.J.; /SLAC; Harindranath, A.; /Saha Inst.; Teramond, G.F.de; /Costa Rica U.

2009-08-03

161

In 2010 life test of three LEU (19.7%) lead test assemblies (LTA) is expected in the existing WWR-K reactor core with regular WWR-C-type fuel assemblies and a smaller core with a beryllium insert. Preliminary analysis of test safety is to be carried out. It implies reconstruction of the reactor core history for last three years, including burnup calculation for each regular fuel assembly (FA), as well as calculation of characteristics of the test core. For the planned configuration of the test core a number of characteristics have been calculated. The obtained data will be used as input for calculations on LTA test core steady-state thermal hydraulics and on transient analysis.

Arinkin, F.; Chakrov, P.; Chekushina, L.; Gizatulin,, Sh.; Koltochnik, S.; Hanan, N.; Garner, P.; Nuclear Engineering Division; Kazakhstan Ministry of Energy and Mineral Resources

2010-03-01

162

NASA Astrophysics Data System (ADS)

Auger decay of the C2H2 double core-hole (DCH) states, including the single-site DCH (C1s-2), two-site DCH (C1s-1C1s-1), and satellite (C1s-2?-1?*+1) states, has been investigated experimentally using synchrotron radiation combined with multi-electron coincidence method, and theoretically with the assumption of the two-step sequential model for Auger decay of the DCH states. The theoretical calculations can reproduce the experimental two-dimensional Auger spectra of the C2H2 single-site DCH and satellite decays, and allow to assign the peaks appearing in the spectra in terms of sequential two-electron vacancy creations in the occupied valence orbitals. In case of the one-dimensional Auger spectrum of the C2H2 two-site DCH decay, the experimental and calculated results agree well, but assignment of peaks is difficult because the first and second Auger components overlap each other. The theoretical calculations on the Auger decay of the N2 single-site DCH state, approximately considering the effect of nuclear motion, suggest that the nuclear motion, together with the highly repulsive potential energy curves of the final states, makes an important effect on the energy distribution of the Auger electrons emitted in the second Auger decay.

Tashiro, Motomichi; Nakano, Motoyoshi; Ehara, Masahiro; Penent, Francis; Andric, Lidija; Palaudoux, Jérôme; Ito, Kenji; Hikosaka, Yasumasa; Kouchi, Noriyuki; Lablanquie, Pascal

2012-12-01

163

We report hybrid density functional theory-molecular mechanics (DFT/MM) calculations performed for glycine in water solution at different pH values. In this paper, we discuss several aspects of the quantum mechanics-molecular mechanics (QM/MM) simulations where the dynamics and spectral binding energy shifts are computed sequentially, and where the latter are evaluated over a set of configurations generated by molecular or Car-Parrinello dynamics simulations. In the used model, core ionization takes place in glycine as a quantum mechanical (QM) system modeled with DFT, and the solution is described with expedient force fields in a large molecular mechanical (MM) volume of water molecules. The contribution to the core electronic binding energy from all interactions within and between the two (DFT and MM) parts is accounted for, except charge transfer and dispersion. While the obtained results were found to be in qualitative agreement with experiment, their precision must be qualified with respect to the problem of counter ions, charge transfer and optimal division of QM and MM parts of the system. Results are compared to those of a recent study [Ottoson et al., J. Am. Chem. Soc., 2011, 133, 3120]. PMID:23160171

Niskanen, Johannes; Arul Murugan, N; Rinkevicius, Zilvinas; Vahtras, Olav; Li, Cui; Monti, Susanna; Carravetta, Vincenzo; Agren, Hans

2013-01-01

164

Two novel stilbene derivatives bearing anthracene core based on 1,3,4-oxadiazole were efficiently synthesized and characterized by (1)H-NMR, mass spectrometry and elemental analysis. The optical properties of the title compounds were investigated by UV-vis absorption and fluorescence emission spectra in different solvents. Chemical calculations were performed by density functional theory (DFT) at the (B3LYP)/6-31G* level. The results show the two compounds exhibit strong green fluorescence emission ranged from 489-493 nm, and the fluorescence quantum yield ranged from 0.78-0.92. Their HOMO and LUMO levels are (-5.44 eV, -2.25 eV) and (-5.45 eV, -2.28 eV), respectively. The influence of the solvent on the fluorescence intensities was also discussed. PMID:23666076

Li, Xinwei; Lu, Huixiong; He, Daohang; Luo, Chun; Huang, Jianjun

2013-09-01

165

From ab initio calculations to multiscale design of Si/C core-shell particles for Li-ion anodes.

The design of novel Si-enhanced nanocomposite electrodes that will successfully mitigate mechanical and chemical degradation is becoming increasingly important for next generation Li-ion batteries. Recently Si/C hollow core-shell nanoparticles were proposed as a promising anode architecture, which can successfully sustain thousands of cycles with high Coulombic efficiency. As the structural integrity and functionality of these heterogeneous Si materials depend on the strength and fracture energy of the active materials, an in-depth understanding of the interface and their intrinsic mechanical properties, such as fracture strength and debonding, becomes critical for the successful design of such and similar composites. Here, we first perform ab initio simulations to calculate these properties for lithiated a-Si/a-C interface structures and combine these results with linear elasticity expressions to model conditions that will avert fracture and debonding in these heterostructures. We find that the a-Si/a-C interface retains good adhesion even at high stages of lithiation. For average lithiated structures, we predict that the strong Si-C bonding averts fracture at the interface; instead, the structure ruptures within lithiated a-Si. From the calculated values and linear elastic fracture mechanics, we then construct a continuum level diagram, which outlines the safe regimes of operation in terms of the core and shell thickness and the state of charge. We believe that this multiscale approach can serve as a foundation for developing quantitative failure models and for subsequent development of flaw-tolerant anode architectures. PMID:24611810

Stournara, Maria E; Qi, Yue; Shenoy, Vivek B

2014-04-01

166

ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT

The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent {sup 235}U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU).

A.H. Wells

2004-11-17

167

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.

DeHart, M.D.

1996-05-01

168

The purpose of this work is to establish a method for evaluating the burn-up values of the rod-type TRIGA spent fuel by using gamma-ray spectrometry of the short-lived fission products 97Zr/97Nb, 132I, and 140La. Fuel irradiation history is not needed in this method. Short-lived fission-product activities were established by reirradiating the spent fuels in a nuclear reactor. Based on the measured activities, 235U burn-up values can be deduced by iterative calculations. The complication caused by 239Pu production and fission is also discussed in detail. The burn-up values obtained by this method are in good agreement with those deduced from the conventional method based on long-lived fission products 137Cs, 134Cs/137Cs ratio and 106Ru/137Cs ratio. PMID:10670930

Wang, T K; Peir, J J

2000-01-01

169

Sensitivity and Uncertainty Analysis to Burn-up Estimates on ADS Using ACAB Code

Within the scope of the Accelerator Driven System (ADS) concept for nuclear waste management applications, the burnup uncertainty estimates due to uncertainty in the activation cross sections (XSs) are important regarding both the safety and the efficiency of the waste burning process. We have applied both sensitivity analysis and Monte Carlo methodology to actinides burnup calculations in a lead-bismuth cooled subcritical ADS. The sensitivity analysis is used to identify the reaction XSs and the dominant chains that contribute most significantly to the uncertainty. The Monte Carlo methodology gives the burnup uncertainty estimates due to the synergetic/global effect of the complete set of XS uncertainties. These uncertainty estimates are valuable to assess the need of any experimental or systematic reevaluation of some uncertainty XSs for ADS.

Cabellos, O; Sanz, J; Rodriguez, A; Gonzalez, E; Embid, M; Alvarez, F; Reyes, S

2005-02-11

170

A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors

NASA Astrophysics Data System (ADS)

This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of ?-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design perturbations and then transformed through the Physical Programming (PP) paradigm to build an aggregate objective function without iteratively determining weights. Finally, the design is optimized with standard gradient-based methods. Through the power of ACE and the transparency of PP, the optimization system allows users to locate designs that best suit their multiobjective preferences with ease.

Touran, Nicholas W.

171

This study calculates the reduced uranium requirements and the economic incentives for increasing the burnup of current design LWR fuels from the current range of 25 to 35 MWD/Kg to a range of 45 to 55 MWD/Kg. The changes in fuel management strategies which may be required to accommodate these high burnup fuels and longer fuel cycles are discussed. The material behavior problems which may present obstacles to achieving high burnup or to license fuel are identified and discussed. These problems are presented in terms of integral fuel response and the informational needs for commercial and licensing acceptance. Research and development programs are outlined which are aimed at achieving a licensing position and commercial acceptance of high burnup fuels.

Stout, R.B.; Merckx, K.R.; Holm, J.S.

1981-01-01

172

NASA Astrophysics Data System (ADS)

A new hybrid functional for accurate descriptions of core and valence excitations, the core-valence Becke's three-parameter exchange (B3)+Lee-Yang-Paar (LYP) correlation functional (CV-B3LYP), is proposed. The construction of the new hybrid functional is based on the assessment that B3LYP performs well for properties concerning valence electrons and Becke's half-and-half exchange+LYP functional (BHHLYP), which includes 50% portion of Hartree-Fock exchange, performs well for core excitations. By using the appropriate portions of Hartree-Fock exchange for core and valence regions separately, CV-B3LYP overcomes the disadvantages of BHHLYP and B3LYP, which give inferior descriptions of valence and core excitations, respectively. Density functional theory (DFT) calculations with the CV-B3LYP functional reproduce core- and valence-orbital energies close to those of BHHLYP and B3LYP, respectively. Time-dependent DFT calculations with the CV-B3LYP functional yield both core- and valence-excitation energies with reasonable accuracy.

Nakata, Ayako; Imamura, Yutaka; Otsuka, Takao; Nakai, Hiromi

2006-03-01

173

NASA Astrophysics Data System (ADS)

Si/SiO2 core/shell quantum dots (QDs) have been shown with wavelength-tunable photoluminescence in addition to their inert, nontoxic, abundant, low-cost, biocompatible advantages. Due to their big size, here, we apply density-functional tight-binding (DFTB) method to perform calculations to study their structures and properties. We systematically investigate the effects of surface passivation, thickness of SiO2 shell, and Si/O ratio on the structures and properties of Si/SiO2 core/shell quantum dots. We find that hydroxyl passivated Si/SiO2 core/shell quantum dots are able to stabilize the quantum dots compared with hydrogen passivated Si/SiO2 core/shell quantum dots. By using DFTB method, we are able to study Si/SiO2 core/shell quantum dots of big size (3 nm) and we find that, in Si/SiO2 core/shell quantum dots, there are competing effects between quantum confinement (blueshift) and oxidation (redshift) with the decrease of the size of Si core. The transition point is when Si/SiO2 ratio is around 1:1. The effect of the thickness of SiO2 on energy gap is not as significant as the effect of the size of the Si core. Our study provides theoretical basis for designing Si quantum dots with tunable photoluminescence.

Dong, Huilong; Hou, Tingjun; Sun, Xiaotian; Li, Youyong; Lee, Shuit-Tong

2013-09-01

174

Uncertainty in the burnup reactivity swing of liquid-metal fast reactors

The uncertainty in the burnup reactivity swing Ďk{sub b} attributable to nuclear data uncertainties is analyzed using depletion-dependent sensitivity coefficients for single- and multicycle equilibrium depletion. Four systems are analyzed with design features that encompass many of the design options considered for current U.S. advanced liquid-metal reactor cores. These systems, while characterized by very different Ďk{sub b} values in the

T. J. Downar; H. Khalil

1991-01-01

175

NASA Astrophysics Data System (ADS)

We study the equilibrium textures of molecular orientation inside cylindrical fibers made of coaxial layers of bent-core smectics. We propose a free-energy model taking into account surface-like and bulk contributions—including layer-compression and electrostatic terms among others— with constant values of the material parameters. We follow the usual variational procedure of minimization of the free energy with respect to the tilt-angle profile ?(r) and obtain an Euler-Lagrange equation and its boundary condition. We solve the variational equations for the equilibrium configurations using a boundary-layer approximation and find multiple solutions. Since the equilibrium tilt profiles are found to be radially inhomogeneous, we select those with minimum distortions in order to find the lowest free-energy state. We minimize further the free energy of the system with respect to the fiber radius and find wider intervals of stability than those previously reported, depending on the balance of the material’s spontaneous polarization, elastic and electric divergence-of-polarization constants, and surface-tension coefficients. The bulk and surface-layer structures thus found could be used to calculate the allowed modes of propagation of electromagnetic waves inside the fiber.

Pérez-Ortiz, Román; Guzmán, Orlando; Reyes, J. Adrián

2011-07-01

176

We study the equilibrium textures of molecular orientation inside cylindrical fibers made of coaxial layers of bent-core smectics. We propose a free-energy model taking into account surface-like and bulk contributions--including layer-compression and electrostatic terms among others--with constant values of the material parameters. We follow the usual variational procedure of minimization of the free energy with respect to the tilt-angle profile ?(r) and obtain an Euler-Lagrange equation and its boundary condition. We solve the variational equations for the equilibrium configurations using a boundary-layer approximation and find multiple solutions. Since the equilibrium tilt profiles are found to be radially inhomogeneous, we select those with minimum distortions in order to find the lowest free-energy state. We minimize further the free energy of the system with respect to the fiber radius and find wider intervals of stability than those previously reported, depending on the balance of the material's spontaneous polarization, elastic and electric divergence-of-polarization constants, and surface-tension coefficients. The bulk and surface-layer structures thus found could be used to calculate the allowed modes of propagation of electromagnetic waves inside the fiber. PMID:21867190

Pérez-Ortiz, Román; Guzmán, Orlando; Reyes, J Adrián

2011-07-01

177

Three Dimensional Analysis of 3-Loop PWR RCCA Ejection Accident for High Burnup

The Rod Control Cluster Assembly (RCCA) ejection accident is a Condition IV design basis reactivity insertion event for Pressurized Water Reactors (PWR). The event is historically analyzed using a one-dimensional (1D) neutron kinetic code to meet the current licensing criteria for fuel rod burnup to 62,000 MWD/MTU. The Westinghouse USNRC-approved three-dimensional (3D) analysis methodology is based on the neutron kinetics version of the ANC code (SPNOVA) coupled with Westinghouse's version of the EPRI core thermal-hydraulic code VIPRE-01. The 3D methodology provides a more realistic yet conservative analysis approach to meet anticipated reduction in the licensing fuel enthalpy rise limit for high burnup fuel. A rod ejection analysis using the 3D methodology was recently performed for a Westinghouse 3-loop PWR at an up-rated core power of 3151 MWt with reload cores that allow large flexibility in assembly shuffling and a fuel hot rod burnup to 75,000 MWD/MTU. The analysis considered high enrichment fuel assemblies at the control rod locations as well as bounding rodded depletions in the end of life, zero power and full power conditions. The analysis results demonstrated that the peak fuel enthalpy rise is less than 100 cal/g for the transient initiated at the hot zero power condition. The maximum fuel enthalpy is less than 200 cal/g for the transient initiated from the full power condition. (authors)

Marciulescu, Cristian; Sung, Yixing; Beard, Charles L. [Westinghouse Electric Company, LLC (United States)

2006-07-01

178

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

179

DANDE: a linked code system for core neutronics/depletion analysis

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the cource of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is illustrated in this report by two sample problems. 25 refs.

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1986-01-01

180

DANDE: a linked code system for core neutronics/depletion analysis

This report describes DANDE - a modular neutronics, depletion code system for reactor analysis. It consists of nuclear data processing, core physics, and fuel depletion modules, and allows one to use diffusion and transport methods interchangeably in core neutronics calculations. This latter capability is especially important in the design of small modular cores. Additional unique features include the capability of updating the nuclear data file during a calculation; a detailed treatment of depletion, burnable poisons as well as fuel; and the ability to make geometric changes such as control rod repositioning and fuel relocation in the course of a calculation. The detailed treatment of reactor fuel burnup, fission-product creation and decay, as well as inventories of higher-order actinides is a necessity when predicting the behavior of reactor fuel under increased burn conditions. The operation of the code system is made clear in this report by following a sample problem.

LaBauve, R.J.; England, T.R.; George, D.C.; MacFarlane, R.E.; Wilson, W.B.

1985-06-01

181

The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

2011-07-01

182

Assessment of Reactivity Margins and Loading Curves for PWR Burnup Credit Cask Designs

This report presents studies to assess reactivity margins and loading curves for pressurized water reactor (PWR) burnup-credit criticality safety evaluations. The studies are based on a generic high-density 32-assembly cask and systematically vary individual calculational (depletion and criticality) assumptions to demonstrate the impact on the predicted effective neutron multiplication factor, k{sub eff}, and burnup-credit loading curves. The purpose of this report is to provide a greater understanding of the importance of input parameter variations and quantify the impact of calculational assumptions on the outcome of a burnup-credit evaluation. This study should provide guidance to regulators and industry on the technical areas where improved information will most enhance the estimation of accurate subcritical margins. Based on these studies, areas where future work may provide the most benefit are identified. The report also includes an evaluation of the degree of burnup credit needed for high-density casks to transport the current spent nuclear fuel inventory. By comparing PWR discharge data to actinide-only based loading curves and determining the number of assemblies that meet the loading criteria, this evaluation finds that additional negative reactivity (through either increased credit for fuel burnup or cask design/utilization modifications) is necessary to accommodate the majority of current spent fuel assemblies in high-capacity casks. Assemblies that are not acceptable for loading in the prototypic high-capacity cask may be stored or transported by other means (e.g., lower capacity casks that utilize flux traps and/or increased fixed poison concentrations or high-capacity casks with design/utilization modifications).

Wagner, J.C.

2002-12-17

183

Benefits of the delta K of depletion benchmarks for burnup credit validation

Pressurized Water Reactor (PWR) burnup credit validation is demonstrated using the benchmarks for quantifying fuel reactivity decrements, published as 'Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty,' EPRI Report 1022909 (August 2011). This demonstration uses the depletion module TRITON available in the SCALE 6.1 code system followed by criticality calculations using KENO-Va. The difference between the predicted depletion reactivity and the benchmark's depletion reactivity is a bias for the criticality calculations. The uncertainty in the benchmarks is the depletion reactivity uncertainty. This depletion bias and uncertainty is used with the bias and uncertainty from fresh UO{sub 2} critical experiments to determine the criticality safety limits on the neutron multiplication factor, k{sub eff}. The analysis shows that SCALE 6.1 with the ENDF/B-VII 238-group cross section library supports the use of a depletion bias of only 0.0015 in delta k if cooling is ignored and 0.0025 if cooling is credited. The uncertainty in the depletion bias is 0.0064. Reliance on the ENDF/B V cross section library produces much larger disagreement with the benchmarks. The analysis covers numerous combinations of depletion and criticality options. In all cases, the historical uncertainty of 5% of the delta k of depletion ('Kopp memo') was shown to be conservative for fuel with more than 30 GWD/MTU burnup. Since this historically assumed burnup uncertainty is not a function of burnup, the Kopp memo's recommended bias and uncertainty may be exceeded at low burnups, but its absolute magnitude is small. (authors)

Lancaster, D. [NuclearConsultants.com, 187 Faith Circle, Boalsburg, PA 16827 (United States); Machiels, A. [Electric Power Research Inst., Inc., 3420 Hillview Avenue, Palo Alto, CA 94304 (United States)

2012-07-01

184

NASA Technical Reports Server (NTRS)

The effective-index method and Marcatili's technique were utilized independently to calculate the electric field profile of a rib channel waveguide. Using the electric field profile calculated from each method, the theoretical coupling efficiency between a single-mode optical fiber and a rib waveguide was calculated using the overlap integral. Perfect alignment was assumed and the coupling efficiency calculated. The coupling efficiency calculation was then repeated for a range of transverse offsets.

Tuma, Margaret L.; Beheim, Glenn

1995-01-01

185

Post Irradiation Examination of High Burnup PWR Fuel

Post irradiation examination (PIE) of a high burnt lead fuel assembly, which was irradiated to demonstrate fuel integrity at high burnup, was performed before the start of the full batch loading of high burnup fuel of 48GWd\\/t maximum fuel assembly burnup.The lead fuel assembly was 17×17 B-type PWR fuel which was supplied by Nuclear Fuel Industries, Ltd. (NFI) and achieved

Shin INOUE; Kazuo MORI; Taro OKAMOTO; Akira OE

1994-01-01

186

S and 4 Reactor: Operating Lifetime and Estimates of Temperature and Burnup Reactivity Coefficients

The S and 4 reactor has a sectored, Mo-14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor is loaded with UN fuel, cooled with a He-Xe gas mixture at {approx}1200 K and operates at steady thermal power of 550 kW. Following a launch abort accident, the axial and radial BeO reflectors easily disassemble upon impact so that the bare reactor is subcriticial when submerged in wet sand or seawater and the core voids are filled with seawater. Spectral Shift Absorber (SSA) additives have been shown to increase the UN fuel enrichment and significantly reduce the total mass of the reactor. This paper investigates the effects of SSA additions on the temperature and burnup reactivity coefficients and the operational lifetime of the S and 4 reactor. SSAs slightly decrease the temperature reactivity feedback coefficient, but significantly increase the operating lifetime by decreasing the burnup reactivity coefficient. With no SSAs, fuel enrichment is only 58.5 wt% and the estimated operating lifetime is the shortest (7.6 years) with the highest temperature and burnup reactivity feedback coefficients (-0.2709 cent /K and -1.3470 $/atom%). With europium-151 and gadolinium-155 additions, the enrichment (91.5 and 94 wt%) and operating lifetime (9.9 and 9.8 years) of the S and 4 reactor are the highest while the temperature and burnup reactivity coefficients (-0.2382 and -0.2447 cent /K; -0.9073 and 0.8502 $/atom%) are the lowest.

King, Jeffrey C.; El-Genk, Mohamed S. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM 87131 (United States); Chemical and Nuclear Engineering Dept., University of New Mexico, Albuquerque, NM 87131 (United States)

2006-01-20

187

Microstructural modeling of thermal conductivity of high burn-up mixed oxide fuel

NASA Astrophysics Data System (ADS)

Predicting the thermal conductivity of oxide fuels as a function of burn-up and temperature is fundamental to the efficient and safe operation of nuclear reactors. However, modeling the thermal conductivity of fuel is greatly complicated by the radially inhomogeneous nature of irradiated fuel in both composition and microstructure. In this work, radially and temperature-dependent models for effective thermal conductivity were developed utilizing optical micrographs of high burn-up mixed oxide fuel. The micrographs were employed to create finite element meshes with the OOF2 software. The meshes were then used to calculate the effective thermal conductivity of the microstructures using the BISON [1] fuel performance code. The new thermal conductivity models were used to calculate thermal profiles at end of life for the fuel pellets. These results were compared to thermal conductivity models from the literature, and comparison between the new finite element-based thermal conductivity model and the Duriez-Lucuta model was favorable.

Teague, Melissa; Tonks, Michael; Novascone, Stephen; Hayes, Steven

2014-01-01

188

Development of HELIOS/CAPP code system for the analysis of block type VHTR cores

In this paper, the HELIOS/CAPP code system developed for the analysis of block type VHTR cores is presented and verified against several VHTR core configurations. Verification results shows that HELIOS code predicts less negative MTC and RTC than McCARD code does and thus HELIOS code overestimates the multiplication factors at the states with high moderator and reflector temperature especially when the B{sub 4}C BP is loaded. In the depletion calculation for the VHTR single cell fuel element, the error of HELIOS code increases as burnup does. It is ascribed to the fact that HELIOS code treats some fission product nuclides with large resonances as non-resonant nuclides. In the 2-D core depletion calculation, a relatively large reactivity error is observed in the case with BP loading while the reactivity error in the case without BP loading is less than 300 pcm. (authors)

Lee, H. C.; Han, T. Y.; Jo, C. K.; Noh, J. M. [Korea Atomic Energy Research Inst., 989-111, Daedeok-Daero, Yuseong-Gu, Daejeon (Korea, Republic of)

2012-07-01

189

NASA Astrophysics Data System (ADS)

We analyze the present-day data on the periods of free oscillations and amplitudes of the forced nutations of the Earth for evaluating the admissible range of the mass and moment of inertia for the liquid core. The initial model for this study is taken in the form of the model distribution of density and mechanical Q parameters of the mantle suggested in (Molodenskii, 2010; 2011a; 2011b). This model was constructed by the steepest descent method in the space of 64 parameters, which determine the distribution of density and parameters of mechanical Q in the mantle, liquid outer core, and solid inner core of the Earth. We assumed the Q parameter of the mantle and inner solid core to be constant and sought for the density variations for the simplest two-parameter model of the piecewise-linear functions with the jumps on the boundary between the liquid core and the mantle and at the olivine-spinel phase transition at a depth of 670 km in the mantle. After this, the computations were repeated for the other distributions of Q (which were also assumed to be unchanged) that correspond to their limiting admissible values. Using this approach, we managed to find the most probable values of the mass and moment of inertia of the liquid core and determine the admissible range of their values. According to our estimates, the ratios of the mass and moments of inertia of the liquid core to the mass and moment of inertia of the whole Earth fall in the intervals 0.317996 ± 0.00065 and 0.110319 ± 0.00022, respectively. These values are lower than the corresponding values for the PREM model (0.322757 and 0.112297) by (1.48 ± 0.30)% and (1.76 ± 0.35)%, respectively. The interpretation of these results requires the revision and thorough analysis of the data on the admissible temperature range of the liquid core and (or) its chemical composition.

Molodenskii, S. M.; Molodenskii, M. S.

2013-07-01

190

Analyzing the rod drop accident in a BWR with high burnup fuel

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 170 cal\\/g. Because high burnup fuel may fail at much lower fuel enthalpies, the best-estimate of the enthalpy and

D. J. Diamond; L. Neymotin

1997-01-01

191

One of the challenges associated with implementation of burnup credit is the validation of criticality calculations used in the safety evaluation; in particular the availability and use of applicable critical experiment data. The purpose of the validation is to quantify the relationship between reality and calculated results. Validation and determination of bias and bias uncertainty require the identification of sets of critical experiments that are similar to the criticality safety models. A principal challenge for crediting fission products (FP) in a burnup credit safety evaluation is the limited availability of relevant FP critical experiments for bias and bias uncertainty determination. This paper provides an evaluation of the available critical experiments that include FPs, along with bounding, burnup-dependent estimates of FP biases generated by combining energy dependent sensitivity data for a typical burnup credit application with the nuclear data uncertainty information distributed with SCALE 6. A method for determining separate bias and bias uncertainty values for individual FPs and illustrative results is presented. Finally, a FP bias calculation method based on data adjustment techniques and reactivity sensitivity coefficients calculated with the SCALE sensitivity/uncertainty tools and some typical results is presented. Using the methods described in this paper, the cross-section bias for a representative high-capacity spent fuel cask associated with the ENDF/B-VII nuclear data for 16 most important stable or near stable FPs is predicted to be no greater than 2% of the total worth of the 16 FPs, or less than 0.13 % k/k.

Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Reed, Davis Allan [ORNL

2010-01-01

192

Globally, there exists a long history in reprocessing in evaluation of the shipper/receiver difference (SRD) on spent nuclear fuel (SNF) received and processed. Typically, the declared shipper s values for uranium and plutonium in SNF (based on calculations involving the initial manufacturer s data and reactor operating history) are used as the input quantities to the head-end process of the facility. Problems have been encountered when comparing these values with measured results of the input accountability tank contents. A typical comparison yields a systematic bias indicated as a loss of 5 7 percent of the plutonium (Pu) and approximately 1 percent for the uranium (U). Studies suggest that such deviation can be attributed to the non-linear nature of the axial burnup values of the SNF. Oak Ridge National Laboratory and Texas A&M University are co-investigating the development of a new method, via Nondestructive Assay (NDA) techniques, to improve the accuracy in burnup and Pu content quantification. Two major components have been identified to achieve this objective. The first component calculates a measurement-based burnup profile along the axis of a fuel rod. Gamma-ray data is collected at numerous locations along the axis of the fuel rod using a High Purity Germanium (HPGe) detector designed for a wide range of gamma-ray energies. Using two fission products, 137Cs and 134Cs, the burnup is calculated at each measurement location and a profile created along the axis of the rod based on the individual measurement locations. The second component measures the U/Pu ratio using an HPGe detector configured for relatively low-energy gamma-rays including x-rays. Fluorescence x-rays from U and Pu are measured and compared to the U/Pu ratio determined from a destructive analysis of the sample. This will be used to establish a relationship between the measured and actual values. This relationship will be combined with the burnup analysis results to establish a relationship between fission product activity and Pu content. It is anticipated that this new method will allow receiving facilities to make a limited number of NDA, gamma-ray, measurements to confirm the shipper declared values for burnup and Pu content thereby improving the SRD.

Saavedra, Steven F [ORNL; Charlton, William S [Texas A& M University; Solodov, Alexander A [ORNL; Ehinger, Michael H [ORNL

2010-01-01

193

Monte Carlo burnup code linking MCNP and REBUS.

National Technical Information Service (NTIS)

The REBUS-3 burnup code, used in the ANL RERTR Program, is a very general code that uses diffusion theory (DIF3D) to obtain the fluxes required for reactor burnup analyses. Diffusion theory works well for most reactors. However, to include the effects of ...

N. A. Hanan

1998-01-01

194

Designing Critical Experiments in Support of Full Burnup Credit

Burnup credit is the process of accounting for the negative reactivity due to fuel burnup and generation of parasitic absorbers over fuel assembly lifetime. For years, the fresh fuel assumption was used as a simple bound in criticality work for used fuel storage and transportation. More recently, major actinides have been included [1]. However, even this yields a highly conservative

Don Mueller; Jeremy A Roberts

2008-01-01

195

Microstructural analysis of LWR spent fuels at high burnup

NASA Astrophysics Data System (ADS)

The microstructural changes in commercial light-water reactor (LWR) fuels irradiated to average burnups near 50 MWd/kgM were studied by analytical transmission electron microscopy and Auger electron spectrometry. Several poorly understood aspects of the fuel behavior were examined, including (a) precipitation of the fission gases in dense, highly pressurized inclusions, (b) apparent solution of Cs, Ba, Zr, and Te in the UO 2 matrix, and (c) the "rim effect" involving restructuring of the enhanced burnup region at the fuel outer edges. Initial observations of the high-burnup rim showed an extremely fine-grained structure formed by recrystallization of the original UO 2. The restructuring is a burnup-induced instability of the UO 2, possibly driven by the stored energy of fission products in solution, and is expected to extend across LWR fuel pellets irradiated to higher burnups.

Thomas, L. E.; Beyer, C. E.; Chariot, L. A.

1992-06-01

196

Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T. [Nuclear Energy System Safety Div., Japan Nuclear Energy Safety Organization, 4-1-28 Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

2012-07-01

197

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity

John M Scaglione; Don Mueller; John C Wagner

2011-01-01

198

The coarse mesh transport method (COMET) is a highly accurate and efficient computational tool which predicts whole-core neutronics behaviors for heterogeneous reactor cores via a pre-computed eigenvalue-dependent response coefficient (function) library. Recently, a high order perturbation method was developed to significantly improve the efficiency of the library generation method. In that work, the method's accuracy and efficiency was tested in a small PWR benchmark problem. This paper extends the application of the perturbation method to include problems typical of the other water reactor cores such as BWR and CANDU bundles. It is found that the response coefficients predicted by the perturbation method for typical BWR bundles agree very well with those directly computed by the Monte Carlo method. The average and maximum relative errors in the surface-to-surface response coefficients are 0.02%-0.05% and 0.06%-0.25%, respectively. For CANDU bundles, the corresponding quantities are 0.01%-0.05% and 0.04% -0.15%. It is concluded that the perturbation method is highly accurate and efficient with a wide range of applicability. (authors)

Zhang, D.; Rahnema, F. [Georgia Institute of Technology, 770 State Street NW, Atlanta, GA 30332-0745 (United States)] [Georgia Institute of Technology, 770 State Street NW, Atlanta, GA 30332-0745 (United States)

2013-07-01

199

Needs of reliable nuclear data and covariance matrices for Burnup Credit in JEFF-3 library

NASA Astrophysics Data System (ADS)

Burnup Credit (BUC) is the concept which consists in taking into account credit for the reduction of nuclear spent fuel reactivity due to its burnup. In the case of PWR-MOx spent fuel, studies pointed out that the contribution of the 15 most absorbing, stable and non-volatile fission products selected to the credit is as important as the one of the actinides. In order to get a "best estimate" value of the keff, biases of their inventory calculation and individual reactivity worth should be considered in criticality safety studies. This paper enhances the most penalizing bias towards criticality and highlights possible improvements of nuclear data for the 15 FPs of PWRMOx BUC. Concerning the fuel inventory, trends in function of the burnup can be derived from experimental validation of the DARWIN-2.3 package (using the JEFF-3.1.1/SHEM library). Thanks to the BUC oscillation programme of separated FPs in the MINERVE reactor and fully validated scheme PIMS, calculation over experiment ratios can be accurately transposed to tendencies on the FPs integral cross sections.

Chambon, A.; Santamarina, A.; Riffard, C.; Lavaud, F.; Lecarpentier, D.

2013-03-01

200

NASA Astrophysics Data System (ADS)

In this paper, we want to emphasize the fact that many experimental properties of ceramics can be explained by the existence of a core-shell structure of the grains, particularly at small sizes. In this framework, we have studied BaTiO3 (BT) ceramics constituted of core-shell nanoparticles, nanowires, or nanoplanes by using ab initio derived effective Hamiltonian calculations whose application range is for large values of shell thickness and low values of shell permittivity. Many differences and new features compared to the situation of nanodots are induced by the core-shell structure. For instance, phase sequences are different; there is also a coexistence of vortices found by Naumov, Bellaiche, and Fu [I. I. Naumov, L. Bellaiche, and H. Fu, Nature (London)10.1038/nature03107 432, 737 (2004)] in the case of isolated dots with a homogeneous polarization, a transition from cubic paraelectric phase towards nonpolar rhombohedral phase, anomalies in dielectric permittivity associated with the onset of toroidal moments, etc. Afterwards, we compare these results with those obtained by the Landau theory of core-shell ceramics we have recently published. However, the ab initio calculations fail to capture the physics at small shell thickness and/or high shell permittivity, whereas the Landau theory fails to predict the peculiar properties of the phases in which vortices exist. Therefore, in a tentative way to build a global theory, we have constructed a Landau potential using both the polarization and the toroidal moment as competing order parameters, which allows us to propose a phase diagram, whatever the thickness and permittivity of the shell are.

Anoufa, M.; Kiat, J. M.; Kornev, I.; Bogicevic, C.

2013-10-01

201

The different (U-Pu) fuel compositions are considered for next generation of sodium fast breeder reactors. The considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to the cores with UO{sub 2} fuel cause increasing uncertainty of generating the group nuclear constants in those reactor systems. The calculation results of BFS-62-3A critical assembly which is considered as full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. For those systems the MCNP 5 calculations demonstrate significant sensitivity of effective multiplication factor K{sub eff} and spectral indices to nuclear data libraries. For {sup 235}U, {sup 238}U, {sup 239}Pu the results of calculated radial fission rate distributions against the reconstructed ones are analyzed. Comparative analysis of spectral indices, neutron spectra and radial fission rate distributions are performed using the different versions of ENDF/B, JENDL-3.3, JENDL-4, JEFF-3.1.1 libraries and BROND-3 for Fe, Cr isotopes. For analyzing the fission rate sensitivity to the plutonium presence in the fuel {sup 239}Pu is substituted for {sup 235}U (enrichment 90%) in the FA areas containing the plutonium. For {sup 235}U, {sup 238}U, {sup 239}Pu radial fission rate distributions the explanation of pick values discrepancies is based on the group fission constants analyses and possible underestimation of some features at the experimental data recovery method (Westcott factors, temperature dependence). (authors)

Mitenkova, E. F.; Novikov, N. V. [Nuclear Safety Inst. of Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Blokhin, A. I. [State Scientific Center of Russian Federation, Inst. of Physics and Power Engineering Named after A.I. Leypunsky, Bondarenko Square 1, Obninsk, Kaluga Region, 249030 (Russian Federation)

2012-07-01

202

Methodology for the Weapons-Grade MOX Fuel Burnup Analysis in the Advanced Test Reactor

A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2, and is therefore called the MCWO. MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. MCWO is capable of handling a large number of fuel burnup and material loading specifications, Advanced Test Reactor (ATR) lobe powers, and irradiation time intervals. MCWO processes user input that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Calculated results from MCNP, ORIGEN-2, and data process module calculations are output in succession as MCWO executes. The principal function of MCWO is to transfer one-group cross-section and flux values from MCNP to ORIGEN-2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN-2 back to MCNP in a repeated, cyclic fashion. The basic requirements of MCWO are a working MCNP input file and some additional input parameters; all interaction with ORIGEN-2 as well as other calculations are performed by CMO. This paper presents the MCWO-calculated results for the Reduced Enrichment Research and Test Reactor (RERTR) experiments RERTR-1 and RERTR-2 as well as the Weapons-Grade Mixed Oxide (WG-MOX) fuel testing in ATR. Calculations performed for the WG-MOX test irradiation, which is managed by the Oak Ridge National Laboratory (ORNL), supports the DOE Fissile Materials Disposition Program (FMDP). The MCWO-calculated results are compared with measured data.

G. S. Chang

2005-08-01

203

Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis

2011-01-01

204

NASA Astrophysics Data System (ADS)

Generalized relativistic effective core potential (GRECP) calculations of spectroscopic constants of the HgH molecule ground and low excited states and the HgH+ cation ground state are carried out, with correlation included by the Fock-space relativistic coupled cluster (RCC) method. Basis set superposition errors (BSSE) are estimated and discussed. It is demonstrated that connected triple excitations of the 13 outermost electrons are necessary to obtain accurate results for mercury hydride. Spectroscopic constants derived from potential curves which include these terms are in very good agreement with experiment, with errors of a few mbohr in Re, tens of wave numbers in excitation energies and vibrational frequencies, and proportionately for other properties. Comparison with previous calculations is also presented.

Mosyagin, Nikolai S.; Titov, Anatoly V.; Eliav, Ephraim; Kaldor, Uzi

2001-08-01

205

NASA Astrophysics Data System (ADS)

We report the calculations of core-excitation energies of first-row atoms using the time-dependent density functional theory (DFT) and the long-range correction (LC) scheme for exchange-correlation functionals, including LC-BOP, Coulomb-attenuated method BLYP, and our recently developed LCgau-BOP method, which includes a flexible portion of short-range Hartree-Fock (HF) exchange through the inclusion of a Gaussian function in the LC scheme. We show that the LC scheme completely fails to improve the poor accuracy of conventional generalized gradient approximation functionals, while the LCgau scheme gives an accuracy which is an order of magnitude better than BLYP and significantly better than B3LYP. A reoptimization of the two parameters controlling the inclusion of short-range HF exchange in the LCgau method enables the errors to be reduced to the order of 0.1 eV which is competitive with the best DFT methods we are aware of. This reparametrization does not affect the LC scheme and therefore maintains the high accuracy of predicted reaction barrier heights. Moreover, while there is some loss in accuracy in thermochemical predictions compared to the previously optimized LCgau-BOP, rms errors in the atomization energies over the G2 test set are found to be comparable to B3LYP. Finally, we attempt to rationalize the success of the LC and LCgau schemes in terms of the well-known self-interaction error (SIE) of conventional functionals. To estimate the role of the SIE, we examine the total energy calculations for systems with a fractional number of electrons, not only in the highest occupied molecular orbital but also in the 1s-characterized core orbital. Our conclusion is that the inclusion of short-range HF exchange in LC-type functionals can significantly alleviate the problems of the SIE in the core region. In particular, we confirm that the absence of the SIE diagnostics in the core orbital energies correlates with the accurate prediction of core-excitation energies using the newly optimized LCgau approach.

Song, Jong-Won; Watson, Mark A.; Nakata, Ayako; Hirao, Kimihiko

2008-11-01

206

PWR cores with silicon carbide cladding

The feasibility of using silicon carbide rather than Zircaloy cladding, to reach higher power levels and higher discharge burnups in PWRs has been evaluated. A preliminary fuel design using fuel rods with the same dimensions as in the Westinghouse Robust Fuel Assembly but with fuel pellets having 10 vol% central void has been adopted to mitigate the higher fuel temperatures that occur due to the lower thermal conductivity of the silicon carbide and to the persistence of the open clad-pellet gap over most of the fuel life. With this modified fuel design, it is possible to achieve 18 month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients and shutdown margin, while allowing batch average discharge burnups up to 80 MWD/kgU and peak rod burnups up to 100 MWD/kgU. Power uprates of 10% and possibly 20% also appear feasible. For non-uprated cores, the silicon carbide-clad fuel has a clear advantage that increases with increasing discharge burnup. Even for comparable discharge burnups, there is a savings in enriched uranium. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% up-rate. Silicon carbide's ability to sustain higher burnups than Zircaloy also allows the design of a licensable two year cycle with only 96 fresh assemblies, avoiding the enriched uranium penalty incurred with use of larger batch sizes due to their excessive leakage. (authors)

Dobisesky, J. P.; Carpenter, D.; Pilat, E.; Kazimi, M. S. [Center for Advanced Nuclear Energy Systems, Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, 77 Massachusetts Avenue 24-215, Cambridge, MA 02139-4307 (United States)

2012-07-01

207

Angular-dependent core hole screening effects have been found in the cobalt K -edge x-ray absorption spectrum of LiCoO2 , using high-resolution data and parameter-free general gradient approximation plus U calculations. The Co1s core hole on the absorber causes strong local attraction. The core hole screening on the cobalt nearest-neighbors induces a 2 eV shift in the density of states with

Amélie Juhin; Frank de Groot; György Vankó; Matteo Calandra; Christian Brouder

2010-01-01

208

BURNCAL: A Nuclear Reactor Burnup Code Using MCNP Tallies.

National Technical Information Service (NTIS)

BURNCAL is a Fortran computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. The code uses output parameters generated by the Monte Carlo neutronics code MCNP to determine the isotopic inven...

E. J. Parma

2002-01-01

209

Modeling depletion simulations for a high-burnup, highly heterogeneous BWR fuel assembly with scale

Extensive SCALE isotopic validation studies have been performed for various PWR fuel assembly designs and operating conditions, and to a lesser extent for BWR fuel assembly designs. However, no SCALE validation work has been documented for newer, highly heterogeneous BWR fuel assembly designs at high burnup. Isotopic benchmark calculations of the earlier, more geometrically uniform BWR fuel assemblies are less sensitive to simplification of the operating history details and certain modeling assumptions than heterogeneous fuel assemblies, particularly at high burnup. This analysis shows the capability of SCALE to simulate a complex highly heterogeneous SVEA96 Optima fuel assembly and illustrates the importance of the need for the highest possible accuracy and precision in isotope measurements intended to be used as benchmark-quality results. In addition, this analysis quantifies the impact of various modeling assumptions on the results. The sample for which the simulation results are reported here achieved a burnup 62 GWd/MTU and was analyzed as part of the MALIBU Extension program. (authors)

Smith, H. J. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)

2012-07-01

210

This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MW{sub th} power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MW{sub th} critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinide fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations. (authors)

Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

2006-07-01

211

Transverse buckling effects on solitary burn-up waves

A three-dimensional one-group diffusion model with explicit effects of burnup and feedback is studied for a so-called “candle reactor”. By a perturbation method the problem is reduced to a one-dimensional one, for which a solitary wave solution was obtained by van Dam (2000) [Self-stabilizing criticality waves. Annals of Nuclear Energy 27, 1505]. Therefore, such a travelling burn-up wave exists as

Xue-Nong Chen; Werner Maschek

2005-01-01

212

Fuel-cladding chemical interaction in mixed-oxide fuel at high burnup

The HEDL FCCI correlation has been expanded to peak burnups up to 14.5 at. %. Increases in depth of interaction were found to be linear with burnup to approximately twice the FFTF goal burnup. Examinations of the peak burnup pin P-23A-18 indicates that it was very close to a true end-of-life breach, demonstrating a substantial burnup capability for FFTF reference

L. A. Lawrence; J. W. Jost

1980-01-01

213

Portable gamma-ray holdup and attributes measurements of high- and variable-burnup plutonium

High burnup-plutonium holdup has been assayed quantitatively by low resolution gamma-ray spectrometry. The assay was calibrated with four plutonium standards representing a range of fuel burnup and {sup 241}Am content. Selection of a calibration standard based on its qualitative spectral similarity to gamma-ray spectra of the process material is partially responsible for the success of these holdup measurements. The spectral analysis method is based on the determination of net counts in a single spectral region of interest (ROI). However, the low-resolution gamma-ray assay signal for the high-burnup plutonium includes unknown amounts of contamination from {sup 241}Am. For most needs, the range of calibration standards required for this selection procedure is not available. A new low-resolution gamma-ray spectral analysis procedure for assay of {sup 239}Pu has been developed. The procedure uses the calculated isotope activity ratios and the measured net counts in three spectral ROIs to evaluate and remove the {sup 241}Am contamination from the {sup 239}Pu assay signal on a spectrum-by-spectrum basis. The calibration for the new procedure requires only a single plutonium standard. The procedure also provides a measure of the burnup and age attributes of holdup deposits. The new procedure has been demonstrated using portable gamma-ray spectroscopy equipment for a wide range of plutonium standards and has also been applied to the assay of {sup 239}Pu holdup in a mixed oxide fuel fabrication facility. 10 refs., 5 figs., 3 tabs.

Wenz, T.R.; Russo, P.A.; Miller, M.C.; Menlove, H.O. (Los Alamos National Lab., NM (United States)); Takahashi, S.; Yamamoto, Y.; Aoki, I. (Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan))

1991-01-01

214

A new deterministic method has been developed for the neutronics analysis of a pebble-bed reactor (PBR). The method accounts for the flow of pebbles explicitly and couples the flow to the neutronics. The method allows modeling of once-through cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times. This new work is distinguished from older methods by the systematically semi-analytical approach it takes. In particular, whereas older methods use the finite-difference approach (or an equivalent one) for the discretization and the solution of the burnup equation, the present work integrates the relevant differential equation analytically in discrete and complementary sub-domains of the reactor. Like some of the finite-difference codes, the new method obtains the asymptotic fuel-loading pattern directly, without modeling any intermediate loading pattern. This is a significant advantage for the design and optimization of the asymptotic fuel-loading pattern. The new method is capable of modeling directly both the once-through-then-out fuel cycle and the pebble recirculating fuel cycle. Although it currently includes a finite-difference neutronics solver, the new method has been implemented into a modular code that incorporates the framework for the future coupling to an efficient solver such as a nodal method and to modern cross section preparation capabilities. In its current state, the deterministic method presented here is capable of quick and efficient design and optimization calculations for the in-core PBR fuel cycle. The method can also be used as a practical "scoping" tool. It could, for example, be applied to determine the potential of the PBR for resisting nuclear-weapons proliferation and to optimize proliferation-resistant features. However, the purpose of this paper is to show that the method itself is viable. Refinements to the code are under way, with the objective of producing a powerful reactor physics analysis tool for PBRs.

Terry, William Knox; Gougar, Hans D; Ougouag, Abderrafi Mohammed-El-Ami

2002-07-01

215

The behavior of 1 MeV tritons produced in the d(d,p)t reaction is important to predict the properties of D-T produced 3.5 MeV alphas because 1 MeV tritons and 3.5 MeV alphas have similar kinematic properties, such as Larmor radius and precession frequency. The confinement and slowing down of the fast tritons were investigated by measuring the 14 MeV and the 2.5 MeV neutron production rates. Here the time resolved triton burnup measurements have been performed using a new type 14 MeV neutron detector based on scintillating fibers, as part of a US-Japan tokamak collaboration. Loss of alpha particles due to toroidal ripple is one of the most important issues to be solved for a fusion reactor such as ITER. The authors investigated the toroidal ripple effect on the fast triton by analyzing the time history of the 14 MeV emission after NB turn-off.

Nishitani, T.; Hoek, M.; Isobe, M.; Tobita, K.; Kusama, Y. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Harano, H. [Univ. of Tokyo (Japan). Dept. of Engineering; Wurden, G.A.; Chrien, R.E. [Los Alamos National Lab., NM (United States)

1995-10-01

216

Burnup of Cadmium Decoupler Material in the Spallation Neutron Source Moderators

At the Spallation Neutron Source being constructed at Oak Ridge National Laboratory, power levels will be greater than at any other operating pulsed spallation neutron scattering facility. Some of the moderators at the facility will contain cadmium that will be used to tailor neutron time distributions by absorbing low-energy neutrons. Because of the higher operating power levels, indications are that there will be considerable burnup of this cadmium during the lifetime of the moderators. Cadmium burnup rates have been calculated for locations around the moderators. Assumed operating conditions for these calculations were a 2-mA beam of 1-GeV protons on the mercury target for 5,000 operating hours per year and a three-year lifetime for the moderators and inner-plug assembly. With the present proposed cadmium thickness in the moderator region (0.05 cm), Monte Carlo calculations indicate considerable depletion of the active cadmium isotope. In places, the calculations indicate complete depletion. An obvious solution to the problem would be to increase the cadmium thickness with a concomitant increase in heat load. Results from some cadmium heating calculations are also presented for a cadmium thickness of 0.05 cm.

Murphy, BD

2001-08-21

217

For several years, Monte Carlo burnup/depletion codes have appeared, which couple a Monte Carlo code to simulate the neutron transport to a deterministic method that computes the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3 dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the time-expensive Monte Carlo solver called at each time step. Therefore, great improvements in term of calculation time could be expected if one could get rid of Monte Carlo transport sequences. For example, it may seem interesting to run an initial Monte Carlo simulation only once, for the first time/burnup step, and then to use the concentration perturbation capability of the Monte Carlo code to replace the other time/burnup steps (the different burnup steps are seen like perturbations of the concentrations of the initial burnup step). This paper presents some advantages and limitations of this technique and preliminary results in terms of speed up and figure of merit. Finally, we will detail different possible calculation scheme based on that method. (authors)

Dieudonne, C.; Dumonteil, E.; Malvagi, F.; Diop, C. M. [Commissariat a l'Energie Atomique et aux Energies Alternatives CEA, Service d'Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)] [Commissariat a l'Energie Atomique et aux Energies Alternatives CEA, Service d'Etude des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA/LTSD, F91191 Gif-sur-Yvette cedex (France)

2013-07-01

218

Study of compact fast reactor core designs

A study is conducted to investigate conceptual liquid-metal reactor (LMR) core concepts, employing some unconventional design features for improved economics and safety. The unconventional design elements are used to supplement the conventional measures, which alone have apparently not led to an attractive LMR design for the 21st century. Better economics are obtained through simplicity and compactness of the core design. For simplicity, internal scattered blankets are omitted. Core compactness is achieved by maximum power flattening, resulting from axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core compactness, the in-core control rods are replaced by reflector controls. For improved safety, the general objective is to reduce both coolant-void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be minimize coolant-void reactivity and to manage the burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors, and inner moderating regions, it is not possible to overcome the fact that both coolant-void and burnup reactivities cannot be reduced simultaneously to desirably low levels. The only resolution of this dilemma appears to be to minimize coolant-void reactivity and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described.

Hamid, T.; Ott, K.O. (Purdue Univ., West Lafayette, IN (United States))

1993-02-01

219

Core Design and Operation Optimization Methods Based on Time-Dependent Perturbation Theory.

National Technical Information Service (NTIS)

A general approach for the optimization of nuclear reactor core design and operation is outlined; it is based on two cornerstones: a newly developed time-dependent (or burnup-dependent) perturbation theory for nonlinear problems and a succesive iteration ...

E. Greenspan

1983-01-01

220

Accurate theoretical energy level, lifetime, and transition probability calculations of core-excited Fe XVI were performed employing the relativistic Multireference Moller-Plesset perturbation theory. In these computations the term energies of the highly excited n {<=} 5 states arising from the configuration 1s {sup 2}2s{sup k} 2p{sup m} 3l {sup p} nl' {sup q}, where k + m + p + q = 9, l {<=} 3 and p + q {<=} 2 are considered, including those of the autoionizing levels with a hole-state in the L-shell. All even and odd parity states of sodium-like iron ion were included for a total of 1784 levels. Comparison of the calculated L-shell transition wavelengths with those from laboratory measurements shows excellent agreement. Therefore, our calculation may be used to predict the wavelengths of as of yet unobserved Fe XVI, such as the second strongest 2p-3d Fe XVI line, which has not been directly observed in the laboratory and which blends with one of the prominent Fe XVII lines.

Diaz, F.; Vilkas, M. J.; Ishikawa, Y. [Department of Chemistry and the Chemical Physics Program, University of Puerto Rico, P.O. Box 23346, San Juan, PR 00931-3346 (Puerto Rico); Beiersdorfer, P., E-mail: beiersdorfer1@llnl.gov [Physics Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States)

2013-07-01

221

NASA Astrophysics Data System (ADS)

Monte Carlo methods are increasingly being used for whole core reactor physics modelling. We describe a number of recent developments to the MONK nuclear criticality and reactor physics code to implement parallel processing, mesh-dependent burn-up and coupling to both thermal hydraulics and gamma transport codes. Results are presented which demonstrate the e_ects of gamma heating in a MONK calculation coupled to the MCBEND Monte Carlo shielding code. Experimental validation of the mesh-dependent tracking and gamma coupling methods is provided by comparison with the results of the NESSUS experiment. The gamma heating calculated by coupled MONK-MCBEND, and the neutron heating calculated by MONK, both compare well against measurement. Finally results are presented from a parallel MONK calculation of a highly detailed PWR benchmark model, which show encouraging speed-up factors on a small development cluster.

Richards, Simon D.; Davies, Nigel; Armishaw, Malcolm J.; Dobson, Geoff P.; Wright, George A.

2014-06-01

222

This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

1987-10-01

223

Design and analysis of a nuclear reactor core for innovative small light water reactors

NASA Astrophysics Data System (ADS)

In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

Soldatov, Alexey I.

224

Use of burnup credit for transportation and storage

Burnup credit is the application of the effects of fuel burnup to nuclear criticality design. When burnup credit is considered in the design of storage facilities and transportation casks for spent fuel, the objectives are to reduce the requirements for storage space and to increase the payload of casks with acceptable nuclear criticality safety margins. The spent-fuel carrying capacities of previous-generation transport casks have been limited primarily by requirements to remove heat and/or to provide shielding. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced significantly. Thus a considerable weight margin is available to the designer for increasing the payload capacity. One method to achieve an increase in capacity is to reduce fuel assembly spacing. The amount of reduction in assembly spacing is limited by criticality and fuel support structural concerns. The optimum fuel assembly spacing provides the maximum cask loading within a basket that has adequate criticality control and sufficient structural integrity for regulatory accident scenarios. The incorporation of burnup credit in cask designs could result in considerable benefits in the transport of spent fuel. The acceptance of burnup credit for the design of transport casks depends on the resolution of system safety issues and the uncertainties that affect the determination of criticality safety margins. The remainder of this report will examine these issues and the integrated approach under way to resolve them. 20 refs., 2 figs.

Sanders, T.L.; Ewing, R.I. (Sandia National Labs., Albuquerque, NM (USA)); Lake, W.H. (USDOE Office of Civilian Radioactive Waste Management, Washington, DC (USA))

1991-01-01

225

A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

NASA Astrophysics Data System (ADS)

Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4®; to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4®; in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23,000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selcted (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. In these conditions two main subjects will be discussed: the Monte Carlo variance calculation and the assessment of the diffusion operator with two energy groups for the core calculation.

Jaboulay, J.-C.; Damian, F.; Douce, S.; Lopez, F.; Guenaut, C.; Aggery, A.; Poinot-Salanon, C.

2014-06-01

226

Burnup increase and Power Up-rate - Operation history of KKL

The Leibstadt nuclear power plant in Switzerland? (KKL), a GE BWR/6 boiling water reactor with an up-rated thermal power of 3600 MW and a nominal net electrical output of 1145 W has been operated for more than 20 years. The core today consists of 648 modern 10x10 assemblies with part length rods which results in a power density of 32 kW/kg Uranium or 63 kW/dm{sup 3}. The plant is operated in a 12 month cycle with shut downs in August. During the last 15 years the transformation of the core was carefully monitored and different fuel assemblies and control rods have been evaluated for an optimized performance of the core. Experience has been gained on core design with control cell core operation and a number of operational issues like change of the isothermal temperature coefficient (ITC), water chemistry with zinc addition and operation with failed fuel. The fuel and fuel assembly behavior has been monitored with numerous fuel inspections on lead use assemblies and selected reload batch assemblies. They have established a good understanding of an optimal fuel performance up to high burnup and the inspection techniques applied in the spent fuel pool on site normally during the outage. (authors)

Ledergerber, G.; Kaufmann, W.; Ritter, A.; Greiner, D. [Kernkraftwerk Leibstadt AG, CH-5325 Leibstadt (Switzerland); Parmar, Y.; Jacot-Guillarmod, R.; Krouthen, J. [Nordostschweizerische Kraftwerke AG, CH-5400 Baden (Switzerland)

2007-07-01

227

NASA Astrophysics Data System (ADS)

We have performed athermal periodic plane-wave density functional calculations within the generalised gradient approximation on the bcc, fcc and hcp structures of Fe1-XNiX alloys (X=0, 0.0625, 0.125, 0.25, and 1) in order to obtain their relative stability and elastic properties at 360 GPa and 0 K. For the hcp structure, using ab initio molecular dynamics, we have also calculated the elastic properties and wave velocities for X=0, 0.0625, and 0.125, at 360 GPa and 5500 K, with further calculations for X=0, and 0.125 at 360 GPa and 2000 K. At 0 K, the hcp structure is the most stable for X=0, 0.0625, 0.125, and 0.25, with the fcc structure becoming the most stable above X0.45; the bcc structure is not the most stable phase for any composition. At 0 K, compressional and shear wave velocities are structure dependent; in the case of fcc the velocities are very similar to pure Fe, but for the hcp structure the addition of Ni strongly reduces VS. Ni also reduced velocities in fcc iron, but to a lesser extent. However, at 5500 K and 360 GPa, Ni has little effect on the wave velocities of the hcp structure, which remain similar to those of pure iron throughout the range of compositions studied and, in the case of VS, >30% greater than that from seismological models. The effect of temperature on Fe-Ni alloys is, therefore, very significant, indicating that conclusions based on the extrapolation of results obtained at much lower temperatures must be treated with great caution. The significance of temperature is confirmed by the additional simulation at 2000 K for X=0, and 0.125 which reveals a remarkably linear temperature dependence of the change in VS relative to that of pure iron. At 0 K, the maximum anisotropy in VP is found to be only very weakly dependent on nickel content, but dependent on structure, being 15% for fcc and 8% for hcp. For the hcp structure at 2000 and 5500 K, the maximum anisotropy in VP is also 8% and almost independent of the Ni content. We conclude that Ni can safely be ignored when considering its effect on the seismic properties of hcp-Fe under core pressures and temperatures and that the negligible effect of nickel on the physical properties of iron in the core arises not because of the chemical similarities between iron and nickel, but because of the high temperature of the system.

Martorell, Benjamí; Brodholt, John; Wood, Ian G.; Vo?adlo, Lidunka

2013-03-01

228

National Technical Information Service (NTIS)

Disrupted-core (transition-phase) behavior has been evaluated for a hypothetical, unprotected transient undercooling accident in an early version of the heterogeneous-core liquid-metal-cooled fast breeder reactor (LMFBR) developed for the Conceptual Desig...

L. B. Luck G. P. DeVault M. W. Asprey C. R. Bell

1982-01-01

229

BWR Fuel Bundle Extended Burnup Program. Technical Progress Report, January 1982-December 1982.

National Technical Information Service (NTIS)

At the start of this period there were four fuel bundles operating to extended burnup in the Monticello Nuclear Generating plant. All four of these bundles successfully completed their planned extended burnup operation and were discharged in September 198...

J. A. Baumgartner

1983-01-01

230

NASA Astrophysics Data System (ADS)

Angular-dependent core hole screening effects have been found in the cobalt K -edge x-ray absorption spectrum of LiCoO2 , using high-resolution data and parameter-free general gradient approximation plus U calculations. The Co1s core hole on the absorber causes strong local attraction. The core hole screening on the cobalt nearest-neighbors induces a 2 eV shift in the density of states with respect to the on-site 1s-3d transitions, as detected in the CoK pre-edge spectrum. Our density functional theory plus U calculations reveal that the off-site screening is different in the out-of-plane direction, where a 3 eV shift is visible in both calculations and experiment. The detailed analysis of the inclusion of the core hole potential and the Hubbard parameter U shows that the core hole is essential for the off-site screening while U improves the description of the angular-dependent screening effects. In the case of oxygen K edge, both the core hole potential and the Hubbard parameter improve the relative positions of the spectral features.

Juhin, Amélie; de Groot, Frank; Vankó, György; Calandra, Matteo; Brouder, Christian

2010-03-01

231

DRAGON/OPTEX predictions of channel power peaking factors and average exit burnup in CANDU-6

The lattice code DRAGON has been under development at Ecole Polytechnique for several years. Features of the code and its application to the analysis of Canada deuterium uranium (CANDU) reactor cells and supercells are discussed in companion papers. Different homogenization techniques are used to generate the two-group macroscopic cross sections required for reactor calculations. To test the significance of the various improvements or modifications to the lattice code or to its input microscopic cross-section library, simplified benchmark reactor calculations can be carried out with a diffusion code such as TRIVAC-2. The on-line refueling feature of the CANDU reactor introduces characteristic fuel management effects that influence the flux and power distributions in the reactor and determines the fuel performance. A simplified yet demanding benchmark procedure was introduced to calculate time-average power distributions, power peaking factors, device reactivity worth, and average exit burnup, which characterize a given fuel management strategy.

Rozon, D.; Varin, R.E. [Ecole Polytechnique, Montreal (Canada)

1995-12-31

232

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

B. Boer; A. M. Ougouag

2010-09-01

233

High Burnup Effects Program. A State-of-the-Technology Assessment.

National Technical Information Service (NTIS)

As a part of Task 1 of the High Burnup Effects Program, a report titled High Burnup Effects, A State-of-the-Technology Assessment was prepared to provide an updated evaluation of the literature as it pertains to high burnup effects in Zircaloy-clad UO sub...

K. H. Rising E. R. Bradley R. E. Williford

1982-01-01

234

The expanded use of burnup credit in the United States (U.S.) for storage and transport casks, particularly in the acceptance of credit for fission products, has been constrained by the availability of experimental fission product data to support code validation. The U.S. Nuclear Regulatory Commission (NRC) staff has noted that the rationale for restricting the Interim Staff Guidance on burnup credit for storage and transportation casks (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issues of burnup credit criticality validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the isotopic composition (depletion) validation approach and resulting observations and recommendations. Validation of the criticality calculations is addressed in a companion paper at this conference. For isotopic composition validation, the approach is to determine burnup-dependent bias and uncertainty in the effective neutron multiplication factor (keff) due to bias and uncertainty in isotopic predictions, via comparisons of isotopic composition predictions (calculated) and measured isotopic compositions from destructive radiochemical assay utilizing as much assay data as is available, and a best-estimate Monte Carlo based method. This paper (1) provides a detailed description of the burnup credit isotopic validation approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias and uncertainty results based on a quality-assurance-controlled prerelease version of the Scale 6.1 code package and the ENDF/B-VII nuclear cross section data.

Radulescu, Georgeta [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

235

REACTIVITY LIFETIME AND BURNUP IN NUCLEAR FUELS (thesis)

Analytical methods for the prediction of the reactivity lifetime and ; burnup of nuclear fuels are developed. The analysis applies to those nuclear ; fuels whose changes in composition with time are due solely to neutron-absorption ; processes, so that the composition of any fuel species is a function only of the ; integrated flux time of its irradiation exposure.

de Ladonchamps

1963-01-01

236

Issues related to criticality safety analysis for burnup credit applications

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality

M. D. DeHart; C. V. Parks

1995-01-01

237

Test of Calorimetry for High Burn-Up Plutonium.

National Technical Information Service (NTIS)

Calorimetric measurements have been performed on a set of high burn-up (25000 MWd/t) Pu samples, ranging in mass between 60 g and 2.5 kg Pu, distributed as PuO2 powder embedded in stainless steel containers. The powers produced by these containers ranged ...

C. Beets R. Carchon P. Fettweis M. Corbellini D. D'Adamo

1984-01-01

238

Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel assembly vibrations contribute significantly to the ex-core neutron noise at nearly all frequencies in this range, probably due to mechanical or acoustic coupling with other vibrating internal structures. Space-dependent kinetics calculations show that ex-core neutron noise induced by fixed-amplitude fuel assembly vibrations will increase over a fuel cycle because of soluble boron and fuel concentration changes associated with burnup. These reactivity effects can also lead to 180/sup 0/ phase shifts between cross-core detectors. We concluded that it may be difficult to separate the changes in neutron noise due to attenuation (shielding) effects of structural vibrations from changes due to reactivity effects of fuel assembly motion on the basis of neutron noise amplitude or phase information. Amplitudes of core support barrel vibrations inferred from ex-core neutron noise measurements using calculated scale factors are likely to have a high degree of uncertainty, since these scale factors usually do not account for neutron noise generated by fuel assembly vibrations. Modifications in fuel management or design may also lead to altered neutron noise signature behavior over a fuel cycle.

Sweeney, F.J.; March-Leuba, J.; Smith, C.M.

1984-01-01

239

Investigation of Irradiation Behavior of SiC-Coated Fuel Particle at Extended Burnup

In current high-temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In safety design of the HTGR fuels, it is important to retain fission products within particles so that their release to primary coolant does not exceed an acceptable level. From this point of view, the basic design criteria for the fuel are to minimize the failure fraction of as-fabricated fuel coating layers and to prevent significant additional fuel failures during operation. The maximum burnup of the first-loading fuel of the High Temperature Engineering Test Reactor (HTTR) is limited to 3.6%FIMA (% fission per initial metallic atom) to certify its integrity during the operation. In order to investigate fuel behavior under extended burnup condition, irradiation tests were performed. The irradiation was carried out as HRB-22 and 91F-1A capsule irradiation tests. The fuel for the irradiation tests was called extended burnup fuel, whose target burnup and fast neutron fluence were higher than those of the first-loading fuel of the HTTR. In order to keep fuel integrity up to over 5%FIMA, the thickness of buffer and SiC layers of fuel particle were increased. The fuel compacts were irradiated in the HRB-22 and the 91F-1A capsules at the High Flux Isotope Reactor of Oak Ridge National Laboratory and at the Japan Materials Testing Reactor of the Japan Atomic Energy Research Institute, respectively. The comparison of measured and calculated release rate-to-birth rate ratios showed that there were additional failures in both irradiation tests. A pressure vessel failure model analysis showed that no tensile stresses acted on the SiC layers even at the end of irradiation and no pressure vessel failure occurred in the intact particles even in a particle with thin buffer layer with failed OPyC layer. The presumed failure mechanisms are additional through-coatings failure of as-fabricated SiC-failed particles or an excessive increase of internal pressure by the accelerated irradiation. Further study is needed to clarify the failure mechanism.

Sawa, Kazuhiro; Tobita, Tsutomu [Japan Atomic Energy Research Institute (Japan)

2003-06-15

240

NASA Technical Reports Server (NTRS)

The basic magnetic properties under various operating conditions encountered in the state-of-the-art DC-AC/DC converters are examined. Using a novel core excitation circuit, the basic B-H and loss characteristics of various core materials may be observed as a function of circuit configuration, frequency of operation, input voltage, and pulse-width modulation conditions. From this empirical data, a mathematical loss characteristics equation is developed to analytically predict the specific core loss of several magnetic materials under various waveform excitation conditions.

Triner, J. E.

1979-01-01

241

Analysis of burnup credit in fuel storage with CASMO

Recent trends in nuclear power plant operation have tended toward longer cycles with reload fuel of high (> 3.5 wt% Â˛ÂłâµU) enrichments. At the same time, the need for greater spent-fuel pool capacity has reduced storage canister spacing to the point where maximum allowable fresh enrichments are lower than those necessary for longer cycles. As a result, burnup credit analysis

Napolitano

1987-01-01

242

Development of Erbia-bearing Super High Burnup Fuel

In this paper, concept and development plan of the Erbia (Er{sub 2}O{sub 3})-bearing super high burnup (Er-SHB) fuel for LWRs are described. In order to reduce the number of spent fuel assemblies, utilization of high burnup fuels with higher uranium enrichment is effective. However, the upper limitation of enrichment for LWR fuels is 5 wt% and current advanced fuel assemblies for LWRs are already reaching this limit. Though various efforts to overcome the 5 wt% enrichment limit have been undergoing, it will require considerable cost that may offset the economic benefit of high burnup fuels. We are proposing another pathway. By adding low content ({>=}0.2 wt%) of Erbia in all UO{sub 2} powder, reactivity of high enrichment (>5 wt%) fuel is suppressed under that of current fuel assemblies, i.e. we leverage the negative reactivity credit of Erbia. Since Erbia is mixed into UO{sub 2} powder just after the re-conversion, we can avoid most of the criticality safety issues appearing in the front-end stream. Namely, major improvements and re-licensing for equipments in transportation, storage and fabrication process will not be necessary. Therefore, the Er-SHB fuel will significantly contribute to reduction of fuel cycle cost. (authors)

Akio, Yamamoto [Nagoya University, Furo-cho, Chikusa-ku, Nagoya (Japan); Toshikazu, Takeda [Osaka University, 2-1 Yamada-oka, Suita-shi, Osaka (Japan); Hironobu, Unesaki [Kyoto University, 2-1010 Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka (Japan); Masaaki, Mori [Nuclear Engineering Ltd., 1-3-7 Tosabori, Nishi-ku, Osaka (Japan); Masatoshi Yamasaki [Nuclear Fuel Industries, Ltd., 1-950 Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka (Japan)

2006-07-01

243

A Feasibility Study to Determine Cooling Time and Burnup of ATR Fuel Using a Nondestructive Technique1 Rahmat Aryaeinejad, Jorge Navarro, and David W Nigg Idaho National Laboratory Abstract Effective and efficient Advanced Test Reactor (ATR) fuel management require state of the art core modeling tools. These new tools will need isotopic and burnup validation data before they are put into production. To create isotopic, burn up validation libraries and to determine the setup for permanent fuel scanner system a feasibility study was perform. The study consisted in measuring short and long cooling time fuel elements at the ATR canal. Three gamma spectroscopy detectors (HPGe, LaBr3, and HPXe) and two system configurations (above and under water) were used in the feasibility study. The first stage of the study was to investigate which detector and system configuration would be better suited for different scenarios. The second stage of the feasibility study was to create burnup and cooling time calibrations using experimental isotopic data collected and ORIGEN 2.2 burnup data. The results of the study establish that a better spectra resolution is achieve with an above the water configuration and that three detectors can be used in the permanent fuel scanner system for different situations. In addition it was conclude that a number of isotopic ratios and absolute measurements could be used to predict ATR fuel burnup and cooling times. 1This work was supported by the U.S. Depart¬ment of Energy (DOE) under Battelle Energy Alliance, LLC Contract No. DE-AC07-05ID14517.

Jorge Navarro; Rahmat Aryaeinejad,; David W. Nigg

2011-05-01

244

Spent LWR fuel dry storage in large transport and storage casks after extended burnup

NASA Astrophysics Data System (ADS)

Dry spent LWR fuel storage is licensed for single fuel assemblies with rod burnup to 65 GWd/tHM. This allows dry spent fuel storage of reloads with a batch average up to 55 GWd/tHM. The leading defect mechanism for spent fuel rods in dry storage is hoop strain. Fuel rod degradation can be prevented by limiting creep. Post-pile creep of fuel rod cladding can be described conservatively by the creep of unirradiated cladding. In order to extend the database, internally pressurized creep samples were investigated for time intervals up to 10 000 h. Test temperatures were between 250 and 400°C, and the hoop stresses applied ranged from 80 to 150 N/mm 2. The resulting data were described mathematically by an interpolation formula. Based on the fuel assemblies end-of-life data the maximum CASTOR V cask storage temperature was calculated to be between 348°C and 358°C at the beginning.

Spilker, Harry; Peehs, Martin; Dyck, Hans-Peter; Kaspar, Guenter; Nissen, Klaus

1997-11-01

245

National Technical Information Service (NTIS)

The Core-Quasiparticle Coupling Model (CQCM) for odd-mass nuclei, which is based on dynamical field theory and the Bardeen-Cooper-Schrieffer (BCS) method, has been applied to two problems. In the first, a study of Pauli exchange effects for the odd partic...

P. B. Semmes

1985-01-01

246

ORIGEN2 was used to develop a data base of pressurized water reactor isotopic concentrations at various times after discharge with core burnup, specific power, enrichment, and neutron spectrum as variables. Results were analyzed to determine source term sensitivity to core management. Fuel rod power history was found to have an important effect on the source term. Activity and decay power

J. K. Wheeler; A. Sesonske

1986-01-01

247

Core-core and core-valence correlation

NASA Technical Reports Server (NTRS)

The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipole moment and 1 Sigma + - 1 Pi transition dipole moment were studied. The results of the FCI calculations are compared to those obtained using approximate methods. In addition, the generation of atomic natural orbital (ANO) basis sets, as a method for contracting a primitive basis set for both valence and core correlation, is discussed. When both core-core and core-valence correlation are included in the calculation, no suitable truncated CI approach consistently reproduces the FCI, and contraction of the basis set is very difficult. If the (nearly constant) core-core correlation is eliminated, and only the core-valence correlation is included, CASSCF/MRCI approached reproduce the FCI results and basis set contraction is significantly easier.

Bauschlicher, Charles W., Jr.; Langhoff, Stephen R.; Taylor, Peter R.

1988-01-01

248

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power

Suzuki Motoe

2000-01-01

249

NASA Astrophysics Data System (ADS)

Further calculations were performed to confirm that the scaling procedure in the calculation of core-electron binding energies proposed by Chong et al. [1] performs well for larger molecules. The procedure was tested on fifty-two new test cases including molecules involving elements from the third period such as SF 4 and ClF 3. In all cases the scaled pVTZ basis performs almost as well as the much larger cc-pV5Z. The average absolute deviation between the results from the scaled pVTZ and estimated complete basis set limit is 0.07 eV.

Pulfer, Mark; Hu, Ching-Han; Chong, Defano P.

1997-03-01

250

Recent observations of broadened spectral lines suggest that the dynamics of molecular clouds (MCs) are dominated by supersonic, turbulent motion. Furthermore, one observes sheet-like and filamentary structures on all resolvable scales. There exists recent theoretical work concerning the evolution of turbulent MC's and their fragmentation into dense cloud cores, which are presumably the progenitors of new stars. These calcualtions allowed

Olaf Kessel-Deynet

1999-01-01

251

NASA Astrophysics Data System (ADS)

In the current international guidelines and standards with regard to human exposure to electromagnetic waves, the basic restriction is defined in terms of the whole-body average-specific absorption rate. The rationale for the guidelines is that the characteristic pattern of thermoregulatory response is observed for the whole-body average SAR above a certain level. However, the relationship between energy absorption and temperature elevation was not well quantified. In this study, we improved our thermal computation model for rabbits, which was developed for localized exposure on eye, in order to investigate the body-core temperature elevation due to whole-body exposure at 2.45 GHz. The effect of anesthesia on the body-core temperature elevation was also discussed in comparison with measured results. For the whole-body average SAR of 3.0 W kg-1, the body-core temperature in rabbits elevates with time, without becoming saturated. The administration of anesthesia suppressed body-core temperature elevation, which is attributed to the reduced basal metabolic rate.

Hirata, Akimasa; Sugiyama, Hironori; Kojima, Masami; Kawai, Hiroki; Yamashiro, Yoko; Fujiwara, Osamu; Watanabe, Soichi; Sasaki, Kazuyuki

2008-06-01

252

A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% {delta}k. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% {delta}k. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

Yang, W.S.; Kim, T.K.; Grandy, C. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne IL 60439 (United States)

2007-07-01

253

Attainable Burnup in a LIFE Engine Loaded with Depleted Uranium

The Laser Inertial Fusion-based Energy (LIFE) system uses a laser-based fusion source for electricity production. The (D,T) reaction, beside a pure fusion system, allows the option to drive a sub-critical fission blanket in order to increase the total energy gain. In a typical fusion-fission LIFE engine the fission blanket is a spherical shell around the fusion source, preceded by a beryllium shell for neutron multiplications by means of (n,2n) reactions. The fuel is in the form of TRISO particles dispersed in carbon pebbles, cooled by flibe. The optimal design features 80 cm thick blanket, 16 cm multiplier, and 20% TRISO packing factor. A blanket loaded with depleted uranium and depleted in a single batch with continuous mixing can achieve burnup as high as {approx}85% FIMA while generating 2,000 MW of total thermal power and producing enough tritium to be used for fusion. A multi-segment blanket with a central promotion shuffling scheme enhances burnup to {approx}90% FIMA, whereas a blanket that is operated with continuous refueling achieves only 82% FIMA under the same constraints of thermal power and tritium self-sufficiency. Both, multi-segment and continuous refueling eliminate the need for a fissile breeding phase.

Fratoni, M; Kramer, K J; Latkowski, J F

2009-11-30

254

Method and apparatus for measuring burn-up of nuclear fuel in a reactor

US Patent & Trademark Office Database

The burn-up of nuclear fuel in a reactor is measured by producing two measuring signals, each of which is a function of the flux of a different neutron energy or group, and comparing the two signals to compute the burn-up.

1977-05-17

255

Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. This project investigated the feasibility of using the passive neutron counting and active neutron\\/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions

Zhongxiang Zhao

2004-01-01

256

Burnup verification measurements on spent fuel assemblies at Arkansas Nuclear One

Burnup verification measurements have been performed using the Fork system at Arkansas Nuclear One, Units 1 and 2, operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an internal calibration for the system in the form of a

Ewing

1995-01-01

257

Burnup verification measurements at U.S. Nuclear Facilities using the Fork system

Burnup verification measurements have been performed using the Fork system at the Oconee Nuclear Station of Duke Power Company, and at Arkansas Nuclear One (Units 1 and 2), operated by Energy Operations, Inc. Passive neutron and gamma-ray measurements on individual spent fuel assemblies were correlated with the reactor records for burnup, cooling time, and initial enrichment. The correlation generates an

Ewing

1995-01-01

258

An analysis of burnup reactivity credit for reactor RA spent fuel storage

The need for increasing the spent fuel storage capacity has led to the development of validated methods for assessing the reactivity effects associated with fuel burnup. This paper gives an overview of the criticality safety analysis methodology used to investigate the sensitivity of storage system reactivities to changes in fuel burnup. Results representing the validation of the methods are also

M. J Miloševi?; M. P Peši?

1998-01-01

259

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize

M. D. Dehart

1996-01-01

260

NASA Astrophysics Data System (ADS)

We propose and assess an exchange-correlation functional, BmLBLYP, in order to describe both core and valence excited states with high accuracy in the time-dependent density functional theory (TDDFT) calculations. The BmLBLYP functional is designed to adopt the modified van Leeuwen-Baerends (mLB) and the Becke 88 (B88) exchange functionals, combined with the Lee-Yang-Parr (LYP) correlation functional. The combination of BmLBLYP is based on the analysis that the LB94 functional behaves better for the core excitations than the B88 functional, while the opposite is true for the valence excitations. Numerical assessment confirms the high accuracy and wide applicability of the BmLBLYP functional.

Imamura, Yutaka; Nakai, Hiromi

2006-02-01

261

Study of a new compact fast reactor core design

A study was conducted to investigate conceptual Liquid Metal Reactor (LMR) designs, employing some unconventional design features, for improved economics and safety. The unconventional design elements were used to supplement the conventional design measures, which alone did not lead to a truly competitive LMR design. Better economics was obtained through simplicity and compactness of core design. For simplicity of core design, internal blankets were omitted. Core compactness was achieved by maximum power flattening. This was done by employing axial and radial enrichment zones along with axial and radial (BeO) reflectors. To further enhance core compactness, the in-core control rods were replaced by reflector controls. For improved safety, the objective was to reduce both coolant void and burnup reactivities. However, even with the use of a wide spectrum of unconventional design features, such as burnable poisons, peripheral reflectors and inner moderating regions, it was not possible to overcome the classical known fact that both coolant void and burnup reactivities cannot be reduced simultaneously. The only resolution of this dilemma appeared to be to minimize coolant void reactivity, and to manage the burnup reactivity losses, such that an accidental insertion of significant amounts of reactivity is mechanically not possible. A conceptual design with these characteristics is described in this thesis.

Hamid, T.

1990-01-01

262

Burnup and feasibility study of low power density PWR's

Operational and safety problems of current Pressurized Water Reactors are often associated with the high power density level of the cores. An alternate use of current-design cores is proposed by reducing the power density.The effects should be improved safety, improved ore utilization, and improved operational characteristics. A scoping study is performed in order to define core parameters suitable for optimization

Molins-Bartra

1981-01-01

263

National Technical Information Service (NTIS)

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code an...

N. A. Hanan

1998-01-01

264

MCNPX Monte Carlo burnup simulations of the isotope correlation experiments in the NPP obrigheim.

This paper describes the simulation work of the Isotope Correlation Experiment (ICE) using the MCNPX Monte Carlo computer code package. The Monte Carlo simulation results are compared with the ICE-Experimental measurements for burnup up to 30 GWD/t. The comparison shows the good capabilities of the MCNPX computer code package for predicting the depletion of the uranium fuel and the buildup of the plutonium isotopes in a PWR thermal reactor. The Monte Carlo simulation results show also good agreements with the experimental data for calculating several long-lived and stable fission products. However, for the americium and curium actinides, it is difficult to judge the predication capabilities for these actinides due to the large uncertainties in the ICE-Experimental data. In the MCNPX numerical simulations, a pin cell model is utilized to simulate the fuel lattice of the nuclear power reactor. Temperature dependent libraries based on JEFF3.1 nuclear data files are utilized for the calculations. In addition, temperature dependent libraries based ENDF/B-VII nuclear data files are utilized and the obtained results are very close to the JEFF3.1 results, except for {approx}10% differences in the prediction of the minor actinide isotopes buildup.

Cao, Y.; Gohar, Y.; Broeders, C. (Nuclear Engineering Division); (Inst. for Neutron Physics and Reactor Technology)

2010-10-01

265

NASA Astrophysics Data System (ADS)

New rhenium(I) dicarbonyl complexes containing cis-{Re(CO)2}+ fragment with redox non-innocent NNS donor ligands L1/L2 (L1, 1-methyl-2-{(o-thiomethyl)phenylazo}imidazole and L2, 1-ethyl-2-{(o-thioethyl)phenylazo}imidazole) having general formula cis-[ReX(CO)3(L1/L2)] (1/2) (X = Cl (1) and Br (2)) have been synthesized and characterized by both experimental and theoretical studies. The structural confirmation has been carried out for 1b. The complexes show quasireversible ReI/ReII oxidation and ligand based reduction in the cyclic voltammetric studies. The electronic structure and the nature of Resbnd CO bonding has been explained by means of DFT and NBO calculations. The spin allowed singlet-singlet electronic transitions of 1b and 2b have been calculated with TDDFT method, and the experimental spectra of the complexes have been discussed on this basis.

Jana, Mahendra Sekhar; Pramanik, Ajoy Kumar; Sarkar, Deblina; Biswas, Sujan; Mondal, Tapan Kumar

2013-09-01

266

Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

NASA Astrophysics Data System (ADS)

Reduced order modeling (ROM) has been recognized as an indispensable approach when the engineering analysis requires many executions of high fidelity simulation codes. Examples of such engineering analyses in nuclear reactor core calculations, representing the focus of this dissertation, include the functionalization of the homogenized few-group cross-sections in terms of the various core conditions, e.g. burn-up, fuel enrichment, temperature, etc. This is done via assembly calculations which are executed many times to generate the required functionalization for use in the downstream core calculations. Other examples are sensitivity analysis used to determine important core attribute variations due to input parameter variations, and uncertainty quantification employed to estimate core attribute uncertainties originating from input parameter uncertainties. ROM constructs a surrogate model with quantifiable accuracy which can replace the original code for subsequent engineering analysis calculations. This is achieved by reducing the effective dimensionality of the input parameter, the state variable, or the output response spaces, by projection onto the so-called active subspaces. Confining the variations to the active subspace allows one to construct an ROM model of reduced complexity which can be solved more efficiently. This dissertation introduces a new algorithm to render reduction with the reduction errors bounded based on a user-defined error tolerance which represents the main challenge of existing ROM techniques. Bounding the error is the key to ensuring that the constructed ROM models are robust for all possible applications. Providing such error bounds represents one of the algorithmic contributions of this dissertation to the ROM state-of-the-art. Recognizing that ROM techniques have been developed to render reduction at different levels, e.g. the input parameter space, the state space, and the response space, this dissertation offers a set of novel hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

Bang, Youngsuk

267

existing safety concepts was made in the paper. The modeled case was a CANDU natural uranium, 37 elements fuel bundle with a reference burnup of 685 GJ\\/kgU (7928.24 MWd\\/tU). The radiotoxicity of the light nuclide inventories, actinide, and fission-products was calculated in the paper. The calculation was made using the ORIGEN-S from ORIGEN4.4a in conjunction with the activation-burnup library and

ALEXANDRU OCTAVIAN; DAN GABRIEL

268

Sensitivity of Candu-Scwr Reactor Physics Calculations to Nuclear Data Files

A simplified MCNP model of a CANDU-SCWR lattice was used to test the sensitivity of the calculated reactivity to various nuclear data files involving issues of current interest. These tests were performed for cooled and voided conditions, with and without axial neutron leakage, for a uniform lattice of mid-life fuel and a mixed lattice with high-burnup and low-burnup fuel in

K. S. Kozier; G. R. Dyck

2006-01-01

269

NASA Astrophysics Data System (ADS)

Several alternative approximations of neutron transport have been proposed in years to move around the known limitations imposed by neutron diffusion in the modeling of nuclear cores. However, only a few complied with the industrial requirements of fast numerical computation, concentrating more on physical accuracy. In this work, the AN transport methodology is discussed with particular interest in core performance calculations. The implementation of the methodology in full core codes is discussed with particular attention to numerical issues and to the integration within the entire simulation process. Finally, first results from core studies in AN transport are analyzed in detail and compared to standard results of neutron diffusion.

Barbarino, Andrea; Tomatis, Daniele

2014-06-01

270

NASA Astrophysics Data System (ADS)

Recent observations of broadened spectral lines suggest that the dynamics of molecular clouds (MCs) are dominated by supersonic, turbulent motion. Furthermore, one observes sheet-like and filamentary structures on all resolvable scales. There exists recent theoretical work concerning the evolution of turbulent MC's and their fragmentation into dense cloud cores, which are presumably the progenitors of new stars. These calcualtions allowed to study the dynamics in MCs under the influence of thermal pressure and self-gravity. However, in high mass SF regions, the massive stars start interacting with their parental cloud via stellar winds and ionizing radiation. These feedback processes disrupt the MC by heating and ionization, but could at the same time induce new star formation by compression of the MC. This work focused on the development of a new computational method which allows the inclusion of ionizing radiation in hydrodynamical simulations using Smoothed Particle Hydrodynamics (sph). In contrast to grid-based methods, sph is especially suited for investigating the fragmentation of turbulent media due to the independence from a fixed grid geometry. The implementation of ionizing radiation into sph is an important step toward the ability to model numerically the feedback of young stars on their molecular clouds. The first application is the ionization-driven implosion of density enhancements in MC's. http://www.mpia-hd.mpg.de/THEORY/preprints/kessel/1999/dissertation/head.html

Kessel-Deynet, Olaf

1999-12-01

271

The radial distribution of plutonium in high burnup UO 2 fuels

NASA Astrophysics Data System (ADS)

A new model (TUBRNP) is described which predicts the radial power density distribution as a function of burnup (and hence the radial burnup profile as a function of time) together with the radial profile of uranium and plutonium isotopes. Comparisons between measurements and the predictions of the TUBRNP model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnups between 21000 and 64000 MWd/t. It is shown to be in excellent agreement with experimental measurements and is a marked improvement on earlier versions.

Lassmann, K.; O'Carroll, C.; van de Laar, J.; Walker, C. T.

1994-02-01

272

Core-Core and Core-Valence Correlation.

National Technical Information Service (NTIS)

The effect of (1s) core correlation on properties and energy separations was analyzed using full configuration-interaction (FCI) calculations. The Be 1 S - 1 P, the C 3 P - 5 S and CH+ 1 Sigma + or - 1 Pi separations, and CH+ spectroscopic constants, dipo...

C. W. Bauschlicher S. R. Langhoff P. R. Taylor

1988-01-01

273

The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised. PMID:16381689

Osmera, B; Cvachovec, F; Kyncl, J; Smutný, V

2005-01-01

274

Development of base technology for high burnup PWR fuel improvement Volume 1 and 2.

National Technical Information Service (NTIS)

Development of base technology for high burnup nuclear fuel -Development of UO(sub 2) pellet manufacturing technology -Improvement of fuel rod performance code -Improvement of plenum spring design -Study on the mechanical characteristics of fuel cladding ...

Y. E. Kim S. H. Lee S. M. Bae J. G. Chung S. K. Chung

1995-01-01

275

Analysis of high-burnup fuel performance during load-follow operation

In Japan, an objective of the burnup extension of nuclear fuel is to raise the licensing limit of burnup from 39 to 48 GWd/t for pressurized water reactors (PWRs) in the near future. Because of an increasing ratio of nuclear power generation, the necessity of the load-follow operation, which responds flexibly to changing power demands, is more apparent. To evaluate accurately the mechanical integrity of PWR fuel at high burnup during a load-follow operation, the FEMAXI-III code, originally developed for analyses of fuel experiments, was modified, improving submodels to evaluate PWR fuel; the new code was named IRON. The results of verification work on the code using data on PWR fuel covering wide ranges of burnup and linear heat rate show that it has good predictability and, therefore, that the improvement was confirmed as effective.

Matsui, T.; Fukuya, K.; Kinoshita, M.

1987-01-01

276

Extension of the TRANSURANUS burnup model to heavy water reactor conditions

NASA Astrophysics Data System (ADS)

The extension of the light water reactor burnup equations of the TRANSURANUS code to heavy water reactor conditions is described. Existing models for the fission of 235U and the buildup of plutonium in a heavy water reactor are evaluated. In order to overcome the limitations of the frequently used RADAR model at high burnup, a new model is presented. After verification against data for the radial distributions of Xe, Cs, Nd and Pu from electron probe microanalysis, the model is used to analyse the formation of the high burnup structure in a heavy water reactor. The new model allows the analysis of light water reactor fuel rod designs at high burnup in the OECD Halden Heavy Water Reactor.

Lassmann, K.; Walker, C. T.; van de Laar, J.

1998-06-01

277

Development of Technical Basis for Burnup Credit Regulatory Guidance in the United States

In the United States (U.S.) there has been and continues to be considerable interest in the increased use of burnup credit as part of the safety basis for SNF systems and this interest has motivated numerous technical studies related to the application of burnup credit for maintaining subcriticality. Responding to industry requests and needs, the U.S. Nuclear Regulatory Commission initiated a burnup credit research program, with support from the Oak Ridge National Laboratory, to develop regulatory guidance and the supporting technical basis for allowing and expanding the use of burnup credit in pressurized-water reactor SNF storage and transport applications. The objective of this paper is to summarize the work and significant accomplishments, with references to the technical reports and publications for complete details.

Parks, Cecil V [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Gauld, Ian C [ORNL] [ORNL

2011-01-01

278

Review and Prioritization of Technical Issues Related to Burnup Credit for BWR Fuel.

National Technical Information Service (NTIS)

This report has been prepared to support technical discussion of and planning for future research supporting implementation of burnup credit for boiling-water reactor (BWR) spent fuel storage in spent fuel pools and storage and transport cask applications...

D. E. Mueller J. M. Scaglione S. M. Bowman W. J. Marshall

2013-01-01

279

National Technical Information Service (NTIS)

Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characteriz...

M. D. DeHart

1996-01-01

280

One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.

Scaglione, John M [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

2011-01-01

281

Analysis of fuel-cladding chemical interaction in mixed-oxide fuel at high burnup. [LMFBR

The character and extent of fuel-cladding chemical interaction (FCCI) were established for mixed uranium-plutonium oxide fuels at an initial oxygen-to-metal ratio of 1.984 irradiated in EBR-II to a peak burnup of approximately 90.0 MWd\\/kgM. Depths of interaction were correlated with fuel pin operating conditions, i.e., burnup and temperature, and results discussed in terms of a change in reaction kinetics.

L. A. Lawrence; J. W. Weber; J. L. Devary

1978-01-01

282

Irradiation performance of metallic driver fuel in Experimental Breeder Reactor II to high burnup

The Experimental Breeder Reactor II Mark-II metallic-driver-fuel element has been irradiated to high burnup to assess element lifetime and performance reliability. The purpose of this paper is to describe the irradiation performance of the Mark-II fuel element to 10 at.% burnup. Fission gas behavior, fuel deformation, fuel-cladding chemical interaction, fuel-cladding mechanical interaction, and cladding dilation are examined for their effect

R. E. Einziger; B. R. Seidel

1980-01-01

283

Analysis of fuel-cladding chemical interaction at high burnup. [LMFBR

The HEDL fuel-cladding chemical interaction (FCCI) wastage correlation has been revised to incorporate additional data from the HEDL P-23 high temperature test series at peak fuel pin burnups of approximately 90.0 MWd\\/kgM. The previous HEDL correlation was developed from a data base of fuel pins with peak burnups of approximately 50.0 MWd\\/kgM. Therefore, in order to predict the amount of

L. A. Lawrence; J. W. Weber; J. L. Devary

1977-01-01

284

High Burnup Effects Program A State-of-the-Technology Assessment

Various analytical models and empirical correlations describing the fission gas release phenomenon were examined. An evaluation was made of the current pertinent experimental data on the subject of high burnup fission gas release. Data reported by individual investigators were compared and evaluated in relation to their applicability to the content and scope of the High Burnup Effects Program. These evaluations then form the bases for defining the data needs, and the selection of variables to be studied in this program.

Rising, K. H.; Bradley, E. R.; Williford, R. E.; Freshley, M D.

1982-06-01

285

Creep assessment of Zry-4 cladded high burnup fuel under dry storage

Cladding creep rupture is thought to be the most likely and limiting failure mechanism of spent fuel in dry storage. In spite of being highly unlikely, the current trend towards high burnups is drawing further attention to the potential creep effect on cladding integrity of fuels burnt over 45 GWd\\/tU.This paper explores the burnup influence on cladding creep during dry storage

F. Feria; L. E. Herranz

2011-01-01

286

ACCURATE NUCLEAR FUEL BURNUP ANALYSES. First Quarterly Report December 1961February 1962

S> Activities in a program to develop mass spectrometric techniques for ; use in reactor fuel burnup analysis are reported. The program emphasis is on ; measurement of nonradioactive refractory fission products that can be related to ; burnup. A controlled irradiation program is being initiated to prepare foils of ; UÂ˛Âłâµ, PuÂ˛Âłâą, and UÂ˛ÂłÂł for use in development of

Rider

1962-01-01

287

A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

To evaluate the burnup potential of a fuel pin, a simplified fuel rod analysis code called FUEL-SIMplified model (FUELSIM) was developed using the general-purpose software VENSIM. FUELSIM is based on FRAPCON-3 models and validated against it. A sensitivity analysis was done using FUELSIM to determine the fuel parameters that have high importance in limiting the burnup potential of a fuel

Stella Maris Oggianu; Hee Cheon No; Mujid S. Kazimi

2004-01-01

288

Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

NASA Astrophysics Data System (ADS)

Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of ? rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that ? rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

Hadad, Kamal; Ayobian, Navid

289

Burnup simulations and spent fuel characteristics of ZrO 2 based inert matrix fuels

NASA Astrophysics Data System (ADS)

Reducing the inventory of long lived isotopes that are contained in spent nuclear fuel is essential for maximizing repository capacity and extending the lifetime of related storage. Because of their non-fertile matrices, inert matrix fuels (IMF's) could be an ideal vehicle for using light-water reactors to help decrease the inventory of plutonium and other transuranics (neptunium, americium, curium) that are contained within spent uranium oxide fuel (UOX). Quantifying the characteristics of spent IMF is therefore of fundamental importance to determining its effect on repository design and capacity. We consider six ZrO 2 based IMF formulations with different transuranic loadings in a 1-8 IMF to UOX pin-cell arrangement. Burnup calculations are performed using a collision probability model where transport of neutrons through space is modeled using fuel to moderator transport and escape probabilities. The lethargy dependent neutron flux is treated with a high resolution multigroup thermalization method. The results of the reactor physics model are compared to a benchmark case performed with Montebruns and indicate that the approach yields reliable results applicable to high-level analyses of spent fuel isotopics. The data generated show that a fourfold reduction in the radiological and integrated thermal output is achievable in single recycle using IMF, as compared to direct disposal of an energy equivalent spent UOX.

Schneider, E. A.; Deinert, M. R.; Herring, S. T.; Cady, K. B.

2007-03-01

290

Sensitivity analysis of hot channel calculation methods

In safety analysis, the fulfillment of acceptance criteria is usually evaluated by separate hot channel or\\/and hot assembly thermal hydraulic\\/fuel behavior calculations. The whole range of the relevant input parameters (e.g. power distributions, burnup, heat conduction data, inlet temperature, etc.) must be taken into account. Concerning these parameters, the most frequent conservative approach is to select the limiting values, partly

I. Panka; M. Telbisz

2007-01-01

291

Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of {sub 62}{sup 149}Sm and its dependence on the shift of a resonance position E{sub r} (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73{<=}{delta}E{sub r}{<=}62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant {alpha}. We obtain new, more accurate limits of -4x10{sup -17}{<=}{alpha}{center_dot}/{alpha}{<=}3x10{sup -17} yr{sup -1}. Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress.

Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G. [St. Petersburg Nuclear Physics Institute, Gatchina, RU-188-300, St. Petersburg (Russian Federation)

2006-12-15

292

NASA Astrophysics Data System (ADS)

The work presented in this thesis is a continuation of a master's thesis research project conducted by the author to gain insight into the applicability of inverse methods to developing adaptive simulation capabilities for core physics problems. Use of adaptive simulation is intended to improve the fidelity and robustness of important core attributes predictions such as core power distribution, thermal margins and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e. in-core instrumentations readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high fidelity and robust adapted core simulators models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e. reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement with measured observables while keeping core simulators models unadapted. At a first glance, devising such adaption for typical core simulators models would render the approach impractical. This follows, since core simulators are based on very demanding computational models, i.e. based on complex physics models with millions of input data and output observables. This would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulators models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulators input data presents a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. We demonstrate that the power of our proposed approach is mainly driven by taking advantage of this unfavorable situation. Our contribution begins with the realization that to obtain numerical solutions to demanding computational models, matrix methods are often employed to produce approximately equivalent discretized computational models that may be manipulated further by computers. The discretized models are described by matrix operators that are often rank-deficient, i.e. ill-posed. We introduce a novel set of matrix algorithms, denoted by Efficient Subspace Methods (ESM), intended to approximate the action of very large, dense, and numerically rank-deficient matrix operators. We demonstrate that significant reductions in both computational and storage burdens can be attained for a typical BWR core simulator adaption problem without compromising the quality of the adaption. We demonstrate robust and high fidelity adaption utilizing a virtual core, e.g. core simulator predicted observables with the virtual core either based upon a modified version of the core simulator whose input data are to be adjusted or an entirely different core simulator. Further, one specific application of ESM is demonstrated, that is being the determination of the uncertainties of important core attributes such as core reactivity and core power distribution due to the available ENDF/B cross-sections uncertainties. The use of ESM is however not limited to adaptive core simulation techniques only, but a wide range of engineering applications may easily benefit from the introduced algorithms, e.g. machine learning and information retrieval techniques highly depends on finding low rank approximations to large scale matrices. In the appendix, we present a stand-alone paper that presents a generalized framework for ESM, including the mathematical theory behind the algorithms and several demonstrative applications that are central to many engineering arenas---(a) sensitivity analysis, (b) parameter estimation, and (c) uncertainty analysis. We choose to do so to allow other engineers, applied mathematicians, and scientists from other scientific disciplines to take direct advantage of ESM without having to sail across the sea of reactor core calculations.

Abdel-Khalik, Hany Samy

293

Minor Actinides-Loaded FBR Core Concept Suitable for the Introductory Period in Japan

NASA Astrophysics Data System (ADS)

According to the Japan's Framework for Nuclear Energy Policy(1), a basic scenario for fast breeder reactors (FBRs) is that they will be introduced on a commercial basis starting around 2050 replacing light water reactors (LWRs). During the FBR introduction period, the Pu from LWR spent fuel is used for FBR startup. Howerver, the FBR core loaded with this Pu has a larger burnup reactivity due to its larger isotopic content of Pu-241 than a core loaded with Pu from an FBR multi-recycling core. The increased burnup reactivity may reduce the cycle length of an FBR. We investigated, an FBR transitional core concept to confront the issues of the FBR introductory period in Japan. Core specifications are based on the compact-type sodium-cooled mixed oxide (MOX)-fueled core designed from the Japanese FBR cycle feasibility studies, because lower Pu inventory should be better for the FBR introductory period in view of its flexibility for the required reprocessing amount of LWR spent fuel to start up FBRs. The reference specifications were selected as follows. Output of 1500MWe and average discharge fuel burnup of about 150GWd/t. Minor Actinides (MAs) recovered from LWR spent fuels which provide Pu to startup FBRs are loaded to the initial loading fuels and exchanged fuels during few cycles until equilibrium. We made the MA content of the initial loading fuel four kinds like 0%, 3%, 4%, 5%. The average of the initial loading fuel is assumed to be 3%, and that of the exchange fuel is set as 5%. This 5% maximum of the MA content is based on the irradiation results of the experimental fast reactor Joyo. We evaluated the core performances including burnup characteristics and the reactivity coefficient and confirmed that transitional core from initial loading until equilibrium cycle with loaded Pu from LWR spent fuel performs similary to an FBR multi-recycling core.

Fujimura, Koji; Sasahira, Akira; Yamashita, Junichi; Fukasawa, Tetsuo; Hoshino, Kuniyoshi

294

TRU transmutation in thorium-based heterogeneous PWR core

A thorium-based seed and blanket design concept for a conventional pressurized light water reactor (PWR) was proposed to enhance the proliferation resistance potential and fuel cycle economics. The KTF core was satisfied with neutronic and thermal-hydraulic design limit of conventional PWR, APR-1400. In order to evaluate transmutation capability of a thorium-based KTF core, U/Zr seed fuel mixed with 10% TRU which come from 1,000 MWe power reactor after 10 years decay was proposed and analyzed by transmutation indices such as D{sub j}, TEX and SR. KTF core showed an extended fuel cycle burnup; average burnup of seed was 79.5 MWd/kgHM and blanket was 94.6 MWd/kgHM. It means that residence time of TRU in the core could be long enough for transmutation when TRU is mixed in seed fuel. The amount of TRU production from conventional PWR could be transmuted in the KTF-TRU core, especially Am-241 isotope is remarkably transmuted by capture reaction. Even isotopes of curium were cumulated in the core during the burnup, however, KTF-TRU core could reduce the TRU in spent fuel by using well-thermalized neutron spectrum. Proliferation resistance potential of thorium based transmutation fuel is slightly increased. About 31% reduction of TRU amount was measured from reduced plutonium production from U-238. Total amount of Am-241 was reduced significantly, but total amount of minor actinide (MA) was reduced by 28% of its initial loading mass. (authors)

Bae, Kang-Mok; Lim, Jae-Yong; Kim, Myung-Hyun [Department of Nuclear Engineering, Kyung Hee University, YoungIn-shi, Gyeonggi-do, 449-701 (Korea, Republic of)

2004-07-01

295

Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO{sub 2}-Gd{sub 2}O{sub 3} poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

Santamarina, A.; Bernard, D.; Blaise, P.; Leconte, P.; Palau, J. M.; Roque, B.; Vaglio, C.; Vidal, J. F. [Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA, DEN, DER, SPRC, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)] [Commissariat a l'Energie Atomique et aux Energies Alternatives, CEA, DEN, DER, SPRC, Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

2013-07-01

296

NASA Astrophysics Data System (ADS)

An experimental LWR fuel with high 235U enrichment (8.6%) and large UO 2 grain size (15-20 ?m) was analysed by optical microscopy, SEM and EPMA. Although the high burn-up reached (69.8 GWd/tM), the porosity growth in the rim zone was found equivalent to a standard LWR fuel with barely 40 GWd/tM burn-up. Also, the grain subdivision associated with the rim structure was absent or scarcely visible around the pores of the outermost periphery, being evidenced that the three typical features of this structure, i.e. Xe depletion, pore formation and grain subdivision, did not appear simultaneously, but sequentially in this order as the local burn-up increased. According to the results, the use of higher 235U enrichments in LWR fuels may help to shift all these steps to higher local burn-ups, whereas the grain size of the UO 2 matrix may play only a secondary role in the initiation of these processes.

Spino, J.; Baron, D.; Coquerelle, M.; Stalios, A. D.

1998-08-01

297

Core Compactness of Progenitors

NASA Astrophysics Data System (ADS)

The compactness of the core of a pre-supernova star is one of the important unexplored issues in progenitor evolution. Recent studies have found the core compactness to be varying non-monotonically as a function of ZAMS mass. In this work we have calculated a large grid of 1D full stellar and naked C/O core models using the implicit hydrodynamic code KEPLER and the open source stellar evolution code MESA, in order to gain a better insight in core compactness' dependence on the stellar mass and convection physics. We find the complicated evolution during C, O and Si burning phases act as the main cause of the non-monotonic variation of compactness, and the whole compactness curve as a function of mass to be quite dependent on the treatment of semiconvection. We also conclude that the C/O core mass is the main discriminant of pre-supernova structure.

Sukhbold, Tuguldur; Woosley, S. E.; Paxton, B.; Heger, A.

2013-04-01

298

Core Compactness of Progenitors

NASA Astrophysics Data System (ADS)

The compactness of the core of a pre-supernova star is one of the important unexplored issues in progenitor evolution. Recent studies have found the core compactness to be varying non-monotonically as a function of ZAMS mass. In this work we have calculated a large grid of 1D full stellar and naked C/O core models using the implicit hydrodynamic code KEPLER and the open source stellar evolution code MESA, in order to gain a better insight in core compactness' dependence on the stellar mass and convection physics. We find the complicated evolution during C burning acts as the main cause of the non-monotonic variation of compactness, and the whole compactness curve as a function of mass to be quite dependent on the treatment of semiconvection. We also conclude that the C/O core mass is the main discriminant of pre-supernova structure.

Sukhbold, Tuguldur; Woosley, S. E.; Paxton, B.; Heger, A.

2013-01-01

299

The Bank of Canada uses core CPI inflation, the year-over-year rate of change of the consumer price index excluding food, energy, and the effects of changes in indirect taxes, as the operational guide for monetary policy. In this report we study the concept and measurement of core or underlying inflation more generally by examining several alternative measures of core inflation,

Seamus Hogan; Marianne Johnson; Thérčse Laflčche

2001-01-01

300

Fission gas release data are presented for five fuel rods irradiated at low fuel temperature (below 2700/sup 0/F) with burnups up to 90,000 MWD/MTM. Four of these rods contained ThO/sub 2/-UO/sub 2/ (33.6 weight percent UO/sub 2/) fuel pellets; the fifth rod contained ThO/sub 2/ pellets. These data supplement fission gas release information previously reported for 54 rods containing ThO/sub 2/-UO/sub 2/ and ThO/sub 2/ fuel, some of which experienced fuel temperaures up to 5000/sup 0/F and burnups to 56,000 MWD/MTM. These new data suggest that at burnups exceeding about 80,000 MWD/MTM a sharp increase in fission gas release occurs, possibly caused by microstructural changes in the fuel. This is similar to the behavior of UO/sub 2/ except that the increase occurs in UO/sub 2/ at lower burnup (approximately 40,000 MWD/MTM). The fission gas release calculational model previously reported has been modified to account for the observed increase in the low temperature component. The revised model provides a good best estimate of all the fission gas release data.

Giovengo, J.F.; Goldberg, I.; Sphar, C.D.

1982-05-01

301

Out-of-core nuclear fuel cycle economic optimization for nonequilibrium cycles

A methodology and associated computer code was developed to determine near-optimum out-of-core fuel management strategies. The code, named OCEON (Out-of-Core Economic OptimizationN), identified feed-region sizes and enrichments, and partially burned fuel-reload strategies for each cycle of a multi-cycle planning horizon, subject to cycle-energy requirements and constraints on feed enrichments, discharge burnups, and the moderator temperature coefficient. A zero-dimensional reactor physics

Comes

1987-01-01

302

NASA Astrophysics Data System (ADS)

Recently our combined SNL-LANL research team has succeeded in developing a global, seamless 3D tomographic P-velocity model (SALSA3D) that provides superior first P travel time predictions at both regional and teleseismic distances. However, given the variable data quality and uneven data sampling associated with this type of model, it is essential that there be a means to calculate high-quality estimates of the path-dependent variance and covariance associated with the predicted travel times of ray paths through the model. In this paper, we show a methodology for accomplishing this by exploiting the full model covariance matrix. Our model has on the order of 1/2 million nodes, so the challenge in calculating the covariance matrix is formidable: 0.9 TB storage for 1/2 of a symmetric matrix, necessitating an Out-Of-Core (OOC) blocked matrix solution technique. With our approach the tomography matrix (G which includes Tikhonov regularization terms) is multiplied by its transpose (GTG) and written in a blocked sub-matrix fashion. We employ a distributed parallel solution paradigm that solves for (GTG)-1 by assigning blocks to individual processing nodes for matrix decomposition update and scaling operations. We first find the Cholesky decomposition of GTG which is subsequently inverted. Next, we employ OOC matrix multiply methods to calculate the model covariance matrix from (GTG)-1 and an assumed data covariance matrix. Given the model covariance matrix we solve for the travel-time covariance associated with arbitrary ray-paths by integrating the model covariance along both ray paths. Setting the paths equal gives variance for that path. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

Hipp, J. R.; Encarnacao, A.; Ballard, S.; Young, C. J.; Phillips, W. S.; Begnaud, M. L.

2011-12-01

303

Development and preliminary verification of the 3D core neutronic code: COCO

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)

Lu, H.; Mo, K.; Li, W.; Bai, N.; Li, J. [Reactor Design and Fuel Management Research Center, China Nuclear Power Technology Research Inst., 47F/A Jiangsu Bldg., Yitian Road, Futian District, Shenzhen (China)

2012-07-01

304

Temperature and Burnup Correlated FCCI in U-10Zr Metallic Fuel

Metallic fuels are proposed for use in advanced sodium cooled fast reactors. The experience basis for metallic fuels is extensive and includes development and qualification of fuels for the Experimental Breeder Reactor I, the Experimental Breeder Reactor II, FERMI-I, and the Fast Flux Test Facility (FFTF) reactors. Metallic fuels provide a number of advantages over other fuel types in terms of fabricability, performance, recyclability, and safety. Key to the performance of all nuclear fuel systems is the resistance to “breach” and subsequent release of fission products and fuel constituents to the primary coolant system of the nuclear power plant. In metallic fuel, the experience is that significant fuel-cladding chemical (FCCI) interaction occurs and becomes prevalent at high power-high temperature operation and ultimately leads to fuel pin breach and failure. Empirical relationships for metallic fuel pin failure have been developed from a large body of in-pile and out of pile research, development, and experimentation. It has been found that significant in-pile acceleration of the FCCI rate is experienced over similar condition out-of-pile experiments. The study of FCCI in metallic fuels has led to the quantification of in-pile failure rates to establish an empirical time and temperature dependent failure limit for fuel elements. Up until now the understanding of FCCI layer formation has been limited to data generated in EBR-II experiments. This dissertation provides new FCCI data extracted from the MFF-series of metallic fuel irradiations performed in the FFTF. These fuel assemblies contain valuable information on the formation of FCCI in metallic fuels at a variety of temperature and burnup conditions and in fuel with axial fuel height three times longer than EBR-II experiments. The longer fuel column in the FFTF and the fuel pins examined have significantly different flux, power, temperature, and FCCI profiles than that found in similar tests conducted in the EBR-II and study of the differences between the two fuel systems is critical for design of large advanced sodium cooled fast reactor systems. Comparing FCCI layer formation data between FFTF and EBR-II indicates that the same diffusion model can be used to represent the two systems when considering time, temperature, burnup history, and axial temperature and power profiles. This dissertation shows that FCCI formation peaks further below the top of the fuel column in FFTF experiments than has been observed in EBR-II experiments. The work provided in this dissertation will help forward the design of advanced metallic fuel systems for advanced sodium cooled fast reactors by allowing the prediction of FCCI layer formation in full length reactor designs. This will allow the accurate lifetime prediction of fuel performance capability for new advanced sodium cooled fast reactors with extended core designs.

William J. Carmack

2012-05-01

305

NASA Astrophysics Data System (ADS)

In the course of upgrading the unit no. 5 reactor core of the Novovoronezh nuclear power plant, operational limits by local parameters, which limit the admissible linear power density and the relative power of fuel elements, were established. Due to applying modern computer technologies in systems of the in-core monitoring, the calculation of power density for all fuel elements in the real-time mode is implemented. To monitor the power density of fuel elements, the algorithm for determining the limiting linear power density is developed depending on the reactor core height and on the average nuclear fuel burnup. The admissible relative power of fuel elements is determined. In the course of the performed work, the excessive conservative limitations on nonuniformity of the reactor power density are excluded. The monitoring of power density by local parameters instead of indirect K q (fuel-assembly relative power) and K v (relative power of the fuel assembly section) made it possible to increase the fuel efficiency and to improve the economic parameters of fuel cycles of the unit no. 5 reactor core of the Novovoronezh nuclear power plant.

Prytkov, A. N.; Tereshchenko, A. B.; Kravchenko, Yu. N.; Boldyrev, N. V.; Pozychaniuk, I. V.; Lisitsyn, D. I.; Golubev, E. I.

2014-04-01

306

Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure {sup 134/137}Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. {sup 134/137}Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from {sup 134}Cs and the 662 keV peak from {sup 137}Cs/{sup 137m}Ba. Assembly No. 2 {sup 134/137}Cs ratio measured on 4-15-92 was 0.19. The measured {sup 134/137}Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of {sup 134}Cs and {sup 137}Cs. The measured {sup 134/137}Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of {sup 134}Cs and 8620--11646 curies of {sup 137}Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies {sup 134}Cs and 8973 curies {sup 137}Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured {sup 134}Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured {sup 137}Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material.

Haggard, D.L.; Tanner, J.E.

1997-06-01

307

A Simplified Approach for Evaluation of the Burnup Potential of Alternative Fuels

To evaluate the burnup potential of a fuel pin, a simplified fuel rod analysis code called FUEL-SIMplified model (FUELSIM) was developed using the general-purpose software VENSIM. FUELSIM is based on FRAPCON-3 models and validated against it. A sensitivity analysis was done using FUELSIM to determine the fuel parameters that have high importance in limiting the burnup potential of a fuel material. Among 16 parameters, 10 were identified as having high importance. For six fuel materials (uranium metal, UC, UN, Th/U metal, UO{sub 2}/ThO{sub 2} fuels, and UO{sub 2}), a simplified model for the pressure rise and volumetric changes inside the fuel is developed to estimate the operational index of each fuel; these models include only the variables with high importance. It was found that the highest burnup potential is that of the nitride fuel, followed by the UO{sub 2}/ThO{sub 2} fuel.

Oggianu, Stella Maris [Massachusetts Institute of Technology (United States); No, Hee Cheon [Korea Advanced Institute of Science and Technology (Korea, Republic of); Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)

2004-06-15

308

Calculation of BWR cycles with CASMO-3-MICROBURN-B and comparison with measurements

The complexity of modern fuel assembly designs, together with heterogeneous core loading patterns, high enrichments and discharge burnups and unconventional reactor operation concepts, challenge the accuracy of BWR steady-state core simulators. The Siemens program system CASMO-3-MICROBURN-B is based on modern concepts and detailed physical modelling for both neutronics and two-phase thermal hydraulics and as such is suited for the analysis

St. Misu; R. G. Grummer

1997-01-01

309

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and/or disposal and enable plants to operate with longer more economical cycle lengths and/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding material and have irradiated several test fuel rods through four cycles. The U.S. Department of Energy (DOE) joined FANP and the BWOG in supporting this project during its final phase of collecting and evaluating high burnup data through post irradiation examination (PIE).

John H. Strumpell

2004-12-31

310

NASA Astrophysics Data System (ADS)

Co K-edge x-ray absorption near edge structures (XANES) of LiCoO2 and CoO2 with a variety of structures, and their orientation dependence, have been evaluated by density functional theory calculations using supercells with a core hole. The spectrum of the layered rocksalt LiCoO2, which is available by experiment, is well reproduced. The effects of the stacking sequence of layered structures and those of Li/Co arrangement in ordered rocksalt polymorphs on XANES are systematically investigated. The spectral shape is sensitive to the stacking sequence in the layered LiCoO2 polymorphs, while it is insensitive in the corresponding CoO2. On the other hand, the ordered-rocksalt polymorphs of LiCoO2 with different Li/Co arrangements exhibit similar spectra, while strong dependence on the manner of the ordering is found in CoO2. Although XANES is rich in information, the use of theoretical fingerprints is desirable for reliable analysis of the oxidation states and identification of the stacking sequence and the manner of ordering.

Koyama, Yukinori; Arai, Hajime; Ogumi, Zempachi; Tanaka, Isao; Uchimoto, Yoshiharu

2012-02-01

311

The loss of gravitational energy on core formation is calculated for the case of simple unmixing of two components, whose equations of state are found from the present density distribution. Without allowance for thermal expansion, the mean energy available for heating is 600 eal\\/g; with an approximate allowance for thermal expansion, this is re- dueed to 400 eal\\/g, which is

Francis Birch

1965-01-01

312

Federal Register 2010, 2011, 2012, 2013

...Public Comment on Draft Test Plan for the High Burnup...Technologies, Office of Nuclear Energy, Department of...public comment on its draft test plan for the High Burnup...to the NRC on the CDP test plan at that time. The DOE's Office of Used Nuclear Fuel Disposition...

2013-11-12

313

National Technical Information Service (NTIS)

The burnup of natural U sub 3 O sub 8 that occurs by the action of thermal neutrons was determined, using the radioisotopes exp 144 Ce, exp 137 Cs, exp 103 Ru, exp 106 Ru and exp 95 Zr as monitors. The determination of the burnup was made using both destr...

I. I. L. Cunha

1979-01-01

314

Svalbard ice cores have not yet been fully exploited for studies of climate and environmental conditions. In one recently drilled ice core from Lomonosovfonna we have studied the methanesulphonic acid (MSA) records in relation to temperature and sea ice. During the present climatic conditions MSA appears to be negatively correlated with the sea ice conditions in the Barents Sea, and

Elisabeth Isaksson; Teija Kekonen; John Moore

315

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

2010-01-01

316

Core design studies for a 1000 MW{sub th} advanced burner reactor.

This paper describes the core design and performance characteristics of 1000 MW{sub th} Advanced Burner Reactor (ABR) core concepts with a wide range of TRU conversion ratio. Using ternary metal alloy and mixed oxide fuels, reference core designs of a medium TRU conversion ratio of {approx}0.7 were developed by trade-off between burnup reactivity loss and TRU conversion ratio. Based on these reference core concepts, TRU burner cores with low and high TRU conversion ratios were developed by changing the intra-assembly design parameters and core configurations. Reactor performance characteristics were evaluated in detail, including equilibrium cycle core performances, reactivity feedback coefficients, and shutdown margins. The results showed that by employing different assembly designs, a wide range of TRU conversion ratios from {approx}0.2 to break-even can be achieved within the same core without introducing significant performance and safety penalties.

Kim, T. K.; Yang, W. S.; Grandy, C.; Hill, R.; Nuclear Engineering Division

2009-04-01

317

The viscosity of the earth's core is probably the least well-known physical property of the earth. Miki [1952] gives an estimate, based on a theoretical calculation, that the dynamic viscosity lies between 10 - and 10 - poise. Malkus [1968] suggests the range 10 -' to 1 poise. Attenuation of S waves reflected from the core [Sato and Espinosa, 1967b;

Roger F. Gans

1972-01-01

318

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

Tomoyuki Uwaba; Masahiro Ito; Kozo Katsuyama; Bruce J. Makenas; David W. Wootan; Jon Carmack

2011-05-01

319

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39E26 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

Uwaba, Tomoyuki; Ito, Masahiro; Mizuno, Tomoyasu; Katsuyama, Kozo; Makenas, Bruce J.; Wootan, David W.; Carmack, Jon

2011-06-16

320

Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel assemblies, with LEU and HEU fuels, are provided to demonstrate the use of this method.

Hanan, N. A.

1998-10-14

321

The Chebyshev Rational Approximation Method (CRAM) has been recently introduced by the authors for solving the burnup equations with excellent results. This method has been shown to be capable of simultaneously solving an entire burnup system with thousands of nuclides both accurately and efficiently. The method was prompted by an analysis of the spectral properties of burnup matrices and it can be characterized as the best rational approximation on the negative real axis. The coefficients of the rational approximation are fixed and have been reported for various approximation orders. In addition to these coefficients, implementing the method only requires a linear solver. This paper describes an efficient method for solving the linear systems associated with the CRAM approximation. The introduced direct method is based on sparse Gaussian elimination where the sparsity pattern of the resulting upper triangular matrix is determined before the numerical elimination phase. The stability of the proposed Gaussian elimination method is discussed based on considering the numerical properties of burnup matrices. Suitable algorithms are presented for computing the symbolic factorization and numerical elimination in order to facilitate the implementation of CRAM and its adoption into routine use. The accuracy and efficiency of the described technique are demonstrated by computing the CRAM approximations for a large test case with over 1600 nuclides. (authors)

Pusa, M.; Leppaenen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)

2012-07-01

322

R and D of Oxide Dispersion Strengthening Steels for High Burn-up Fuel Claddings

Research and development of fuel clad materials for high burn-up operation of light water reactor and super critical water reactor (SCPWR) will be shown with focusing on the effort to overcome the requirements of material performance as the fuel clad. Oxide dispersion strengthening (ODS) steels are well known as a high temperature structural material. Recent irradiation experiments indicated that the

A. Kimura; H. S. Cho; J. S. Lee; R. Kasada; S. Ukai; M. Fujiwara

2004-01-01

323

Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 {mu}m in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307{degree}C rather than the normal 288{degree}C, a relatively thick (50 to 70 {mu}m) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs.

Chung, H.M. (Argonne National Lab., IL (USA))

1989-09-01

324

Development of hold-down spring for high burn-up fuel assembly.

National Technical Information Service (NTIS)

For the development of high burn-up fuel assembly, it is necessary to develop hold-down spring which guarantees more gap margin between fuel rods and end pieces, and also provides more hold-down force at EOL than current hold-down spring. To this end, the...

J. K. Lee

1991-01-01

325

National Technical Information Service (NTIS)

In-reactor densification of ThO sub 2 and ThO sub 2 --UO sub 2 fuel of low burnup and low power operation (hence low temperature) was assessed by measuring fuel pellet diameter changes. Pellet diameter changes ranged from nil in a large grain, low tempera...

G. L. Spahr

1978-01-01

326

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of 239Pu and 241Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can

Levent Özdemir; Banu Bulut Acar; Okan H. Zabuno?lu

2011-01-01

327

Evaluation of Cross Section Sensitivities in Computing Burnup Credit Fission Product Concentrations.

National Technical Information Service (NTIS)

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the u...

D. E. Mueller I. C. Gauld

2005-01-01

328

A validated methodology for evaluating burnup credit in spent fuel casks

The concept of allowing reactivity credit for the transmuted state of spent fuel offers both economic and risk incentives. This paper presents a general overview of the technical work being performed in support of the US Department of Energy (DOE) program to resolve issues related to the implementation of burnup credit. An analysis methodology is presented along with information representing

M. C. Brady; T. L. Sanders

1991-01-01

329

Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations

U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses

I. C. Gauld; D. E. Mueller

2005-01-01

330

The Effect of Pitch, Burnup, and Absorbers on a TRIGA Spent-Fuel Pool Criticality Safety

It has been shown that supercriticality might occur for some postulated accident conditions at the TRIGA spent-fuel pool. However, the effect of burnup was not accounted for in previous studies. In this work, the combined effect of fuel burnup, pitch among fuel elements, and number of uniformly mixed absorber rods for a square arrangement on the spent-fuel pool k{sub eff} is investigated.The Monte Carlo computer code MCNP4B with the ENDF-B/VI library and detailed three dimensional geometry was used. The WIMS-D code was used to model the isotopic composition of the standard TRIGA and FLIP fuel for 5, 10, 20 and 30% burnup level and 2- and 4-yr cooling time.The results show that out of the three studied effects, pitch from contact (3.75 cm) up to rack design pitch (8 cm), number of absorbers from zero to eight, and burnup up to 30%, the pitch has the greatest influence on the multiplication factor k{sub eff}. In the interval in which the pitch was changed, k{sub eff} decreased for up to {approx}0.4 for standard and {approx}0.3 for FLIP fuel. The number of absorber rods affects the multiplication factor much less. This effect is bigger for more compact arrangements, e.g., for contact of standard fuel elements with eight absorber rods among them, k{sub eff} values are smaller for {approx}0.2 ({approx}0.1 for FLIP) than for arrangements without absorber rods almost regardless of the burnup. The effect of burnup is the smallest. For standard fuel elements, it is {approx}0.1 for almost all pitches and numbers of absorbers. For FLIP fuel, it is smaller for a factor of 3, but increases with the burnup for compact arrangements. Cooling time of fuel has just a minor effect on the k{sub eff} of spent-fuel pool and can be neglected in spent-fuel pool design.

Logar, Marjan [University of Maribor (Slovenia); Jeraj, Robert [Jozef Stefan Institute (Slovenia); Glumac, Bogdan [Jozef Stefan Institute (Slovenia)

2003-02-15

331

Test plan for high-burnup fuel cladding behavior under loss-of- coolant accident conditions

Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding, and as result, mechanical properties of high-burnup fuels are degraded significantly. This may influence the current fuel cladding failure limits for loss-of- coolant-accident (LOCA) situations, which are based on fuel cladding behavior for zero burnup. To avoid cladding fragmentation and fuel dispersal during a LOCA, 10 CFR 50.46 requires that peak cladding temperature shall not exceed 1204 degrees C (2200 degrees F) and that total oxidation of the fuel cladding nowhere exceeds 0.17 times total cladding thickness before oxidation. Because of the concern, a new experimental program to investigate high-burnup fuel cladding behavior under LOCA situations has been initiated under the sponsorship of the U.S. Nuclear Regulatory Commission. A hot-cell test plan to investigate single-rod behavior under simulated LOCA conditions is described in this paper. In the meantime, industry fuel design and operating conditions are expected to undergo further changes as more advanced cladding materials are developed. Under these circumstances, mechanical properties of high-burnup fuel cladding require further investigation so that results from studies on LOCA, reactivity- initiated-accident (RIA), operational transient, and power-ramping situations, can be extrapolated to modified or advanced cladding materials and altered irradiation conditions without repeating major integral experiments in test reactors. To provide the applicable data base and mechanistic understanding, tests will be conducted to determine dynamic and static fracture toughness and tensile properties. Background and rationale for selecting the specific mechanical properties tests are also described.

Chung, H.M.; Neimark, L.A.; Kassner, T.F.

1996-10-01

332

Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

G. S. Chang

2006-07-01

333

Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at {approx}2.4, {approx}7 and {approx}11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of {approx}7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10{sup 15} n/cm{sup 2}/s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between {approx}410 deg. C and {approx}645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T. [Central Research Institute of Electric Power Industry - CRIEPI, 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Ougier, M.; Glatz, J.P. [European Commission Joint Research, Institute for Transuranium Elements - JRC-ITU, Postfach 2340, D-76125 Karlsruhe (Germany); Fontaine, B.; Breton, L. [Commissariat a l'Energie Atomique - CEA, Centrale Phenix, 30200 Bagnols Sur Ceze (France)

2007-07-01

334

XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8–54.9 GWd\\/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd\\/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold

M. Mogensen; J. H Pearce; C. T Walker

1999-01-01

335

RIA Limits Based On Commercial PWR Core Response To RIA

Reactivity insertion accident (RIA) limits have been under intense review by regulators since 1993 with respect to what should be the proper limit as a function of burnup. Some national regulators have imposed new lower limits while in the United States the limits are still under review. The data being evaluated with respect to RIA limits come from specialized test reactors. However, the use of test reactor data needs to be balanced against the response of a commercial PWR core in setting reasonable limits to insure the health and safety of the public without unnecessary restrictions on core design and operation. The energy deposition limits for a RIA were set in the 1970's based on testing in CDC (SPERT), TREAT, PBF and NSRR test reactors. The US limits given in radially averaged enthalpy are 170 cal/gm for fuel cladding failure and 280 cal/gm for coolability. Testing conducted in the 1990's in the CABRI, NSRR and IGR test reactors have demonstrated that the cladding failure threshold is reduced with burnup, with the primary impact due to hydrogen pickup for in-reactor corrosion. Based on a review of this data very low enthalpy limits have been proposed. In reviewing proposed limits from RIL-0401(1) it was observed that much of the data used to anchor the low allowable energy deposition levels was from recent NSRR tests which do not represent commercial PWR reactor conditions. The particular characteristics of the NSRR test compared to commercial PWR reactor characteristics are: - Short pulse width: 4.5 ms vs > 8 ms; - Low temperature conditions: < 100 deg. F vs 532 deg. F. - Low pressure environment: atmospheric vs {approx} 2200 psi. A review of the historical RIA database indicates that some of the key NSRR data used to support the RIL was atypical compared to the overall RIA database. Based on this detailed review of the RIA database and the response of commercial PWR core, the following view points are proposed. - The Failure limit should reflect local fuel enthalpy and the condition of the fuel, i.e. burnup and expected oxide levels as a function of enthalpy. Limits based on cladding oxide needs to take into account that in many core designs the highest oxide will generally be on high burnup rods on the core periphery which have low reactivity, and lower peaking factors. Otherwise excessively low limits based generally on oxide could restrict use of fuel from the spent fuel pool. (authors)

Beard, Charles L.; Mitchell, David B.; Slagle, William H. [Westinghouse Electric Company, 4350 Northern Pike, Monroeville, PA 15146-2886 and 5801 Bluff Road, Columbia, SC 29250 (United States)

2006-07-01

336

McCARD for Neutronics Design and Analysis of Research Reactor Cores

NASA Astrophysics Data System (ADS)

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

2014-06-01

337

Degraded core modeling in MELCOR

A package of phenomenological models has been developed for the MELCOR code system to calculate the thermal response of structures in the core and lower plenum of an LWR during a severe accident. This package treats all important modes of heat transfer within the core, as well as oxidation, debris formation, and relocation of core and structural materials during melting, candling, and slumping. Comparison of MELCOR and MARCON calculations for the Browns Ferry BWR primary system shows many areas of agreement during the early stages of core heatup and oxidation, but very large differences at later times. Many of these differences are attributed to the effects of candling predicted by MELCOR and the lack of any mechanistic candling or debris relocation models in MARCON. The melting and slumping behavior calculated by MELCOR is in qualitative agreement with our current understanding of the processes involved.

Summers, R.M.

1986-01-01

338

Calculation of Optimum Control Rod Operation Programme for Boiling Water Reactor.

National Technical Information Service (NTIS)

Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load dens...

L. Fehr

1978-01-01

339

Federal Register 2010, 2011, 2012, 2013

...major changes in the staff recommendations: (1) optional credit for fission product and minor actinide neutron absorbing isotopes in the SNF composition, and (2) misload analyses and additional administrative procedures in lieu of a burnup...

2012-10-03

340

NASA Astrophysics Data System (ADS)

Fuel service conditions proposed for the very high temperature reactor will be challenging. All major fuel-related design parameters (burnup, temperature, fast neutron fluence, power density, particle packing fraction) exceed the values that were qualified in the successful German UO 2 coated particle fuel development program in the 1980s. Of particular concern are the high burnup and high temperatures expected in the very high temperature reactor. In this paper, the challenges associated with high burnup and high temperature are evaluated quantitatively by examining the performance of the fuel in terms of different known failure mechanisms. Potential design solutions to ameliorate the negative effects of high burnup and high temperature are discussed. Also of concern are the effects of accelerated irradiation on coated fuel that often occur during irradiation testing. These effects are evaluated in this paper and recommendations concerning allowable levels of accelerations are presented.

Maki, John T.; Petti, David A.; Knudson, Darrell L.; Miller, Gregory K.

2007-09-01

341

A method for determination of linear energy release of a VVER fuel assembly near a rhodium self-powered neutron detector (SPND) is described. The dependence of SPND burnup on the charge passing through it is specified.

Kurchenkov, A. Yu., E-mail: s327@vver.kiae.ru [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15

342

The performance of MOX fuel irradiated in the advanced thermal reactor, FUGEN, to a burnup of 47.5GWd\\/t, was investigated by using a telescope, optical microscope, SEM and EPMA. Observations focused on elucidating the corrosion behavior of the cladding inner surface. A reaction layer was observed at burnups higher than about 35GWd\\/t. The relationship between the thickness of the reaction layer

Kosuke Tanaka; Koji Maeda; Shinji Sasaki; Yoshihisa Ikusawa; Tomoyuki Abe

2006-01-01

343

The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the

Michael A. Pope

2011-01-01

344

Fast-ion redistribution at sawtooth crashes as inferred from triton burn-up measurements

Based on a recent theory for fast-ion redistribution during sawtooth crashes, an approximate analytical model is given for describing the sawtooth induced changes in the emission profile of 14 MeV neutrons generated by triton burn-up. Comparison with measurements of sawtooth-induced neutron emission profile broadening on JET shows good agreement, which indicates that the theory for fast-ion redistribution is consistent with

D. Anderson; P. Batistoni; M. Lisak; F. Wising

1993-01-01

345

Post Irradiation Examination for Advanced Materials at Burnups Exceeding the Current Limit

Permitting fuel to be irradiated to higher burnups limits can reduce the amount of spent nuclear fuel (SNF) requiring storage and\\/or disposal and enable plants to operate with longer more economical cycle lengths and\\/or at higher power levels. Therefore, Framatome ANP (FANP) and the B&W Owner's Group (BWOG) have introduced a new fuel rod design with an advanced M5 cladding

John H. Strumpell

2004-01-01

346

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at

B. El Bakkari; T. El Bardouni; O. Merroun; Ch. El Younoussi; Y. Boulaich; E. Chakir

2009-01-01

347

Fuel\\/cladding chemical interaction in mixed-oxide fuel at high burnup

The character and extent of fuel\\/cladding chemical interaction (FCCI) have been established for mixed uranium-plutonium oxide, (U,Pu)Oâ, fuels irradiated in Experimental Breeder Reactor-II to peak fuel burnups to 14.5 at. % at beginning-of-life peak cladding temperatures to 730Â°C. The changes in character and the correlation of depth of FCCI were determined as functions of the initial as-fabricated fuel oxygen-tometal ratios

1984-01-01

348

A Genesis breakup and burnup analysis in off-nominal Earth return and atmospheric entry

NASA Technical Reports Server (NTRS)

The Genesis project conducted a detailed breakup/burnup analysis before the Earth return to determine if any spacecraft component could survive and reach the ground intact in case of an off-nominal entry. In addition, an independent JPL team was chartered with the responsibility of analyzing several definitive breakup scenarios to verify the official project analysis. This paper presents the analysis and results of this independent team.

Salama, Ahmed; Ling, Lisa; McRonald, Angus

2005-01-01

349

Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications

The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty,cycles (high burnup, boiling, aggressive chemistry) andto investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment.

Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim

2006-10-31

350

Feasibility and incentives for burnup credit in spent-fuel casks

The spent-fuel carrying capacities of previous-generation spent-fuel shipping casks have been primarily thermal and\\/or shielding limited. Shielding and heat transfer requirements for casks designed to transport older spent fuel with longer decay times are reduced considerably and cask capacities become criticality limited. Using burnup credit in the design of future casks can result in increased cask capacities as well as

T. L. Sanders; R. M. Westfall

1990-01-01

351

K-infinite trends with burnup, enrichment, and cooling time for BWR fuel assemblies

This report documents the work performed by ORNL for the Yucca Mountain project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k{sub inf} values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup\\/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a

Broadhead

1998-01-01

352

SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT

The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality

Georgeta Radulescu; Don Mueller; John C Wagner

2009-01-01

353

K-Effective Trends with Burnup, Enrichment, and Pooling Time for BWR Fuel Assemblies

This report documents the work performed by ORNL for the Yucca Mountain Project (YMP) M and O contractor, Framatome Cogema Fuels. The goal of this work was to obtain k{sub inf} values for infinite arrays of flooded boiling-water-reactor (BWR) fuel assemblies as a function of various burnup\\/enrichment and cooling-time combinations. These scenarios simulate expected limiting criticality loading conditions (for a

Broadhead

1998-01-01

354

SENSITIVITY COEFFICIENT GENERATION FOR A BURNUP CREDIT CASK MODEL USING TSUNAMI3D

The evolution of a complex criticality model for a burnup credit shipping cask to an accurate TSUNAMI-3D model for eigenvalue sensitivity coefficient generation is detailed in this paper. TSUNAMI-3D is a Monte Carlo-based eigenvalue sensitivity analysis sequence that was released with SCALE 5. In the criticality model, 32 fuel assemblies, each with 18 axial zones with differing depletion-dependent compositions, are

Donald E. Mueller; Bradley T. Rearden

355

Cladding metallurgy and fracture behavior during reactivity-initiated accidents at high burnup

High-burnup fuel failure during a reactivity-initiated accident has been the subject of safety-related concern. Because of wide variations in metallurgical and simulation test conditions, it has been difficult to understand the complex failure behavior from major tests in NSRR and CABRI reactors. In this paper, a failure model based on fracture toughness and microstructural characteristics is proposed in which fracture

H. M. Chung; T. F. Kassner

1996-01-01

356

Best-estimate computational methods are here used to analyse the thermo-mechanical behaviour of high-burnup UO2 fuel rods under postulated reactivity initiated accidents in light water reactors. The considered accident scenarios are the hot zero power rod ejection accident in pressurised water reactors and the cold zero power control rod drop accident in boiling water reactors. For these accidents, fuel enthalpy thresholds

Lars Olof JERNKVIST

2006-01-01

357

Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

NASA Astrophysics Data System (ADS)

The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

2014-06-01

358

Analysis of Burnup and Economic Potential of Alternative Fuel Materials in Thermal Reactors

A strategy is proposed for the assessment of nuclear fuel material economic potential use in future light water reactors (LWRs). In this methodology, both the required enrichment and the fuel performance limits are considered. In order to select the best fuel candidate, the optimal burnup that produces the lowest annual fuel cost within the burnup potential for a given fuel material and smear density ratio is determined.Several nuclear materials are presented as examples of the application of the methodology proposed in this paper. The alternative fuels considered include uranium dioxide (UO{sub 2}), uranium carbide (UC), uranium nitride (UN), metallic uranium (U-Zr alloy), combined thorium and uranium oxides (ThO{sub 2}/UO{sub 2}), and combined thorium and uranium metals (U/Th). For these examples, a typical LWR lattice geometry in a zirconium-based cladding was assumed. The uncertainties in the results presented are large due to the scarcity of experimental data regarding the behavior of the considered materials at high burnups. Also, chemical compatibility issues are to be considered separately.The same methodology can be applied in the future to evaluate the economic potential of other nuclear fuel materials including different cladding designs, dispersions of ceramics into ceramics, dispersions of ceramics into metals, and also for geometries other than the traditional circular fuel pin.

Oggianu, Stella Maris; No, Hee Cheon; Kazimi, Mujid S. [Massachusetts Institute of Technology (United States)

2003-09-15

359

Radiation-induced microstructural change in high burnup UO 2 fuel pellets

NASA Astrophysics Data System (ADS)

The formation mechanism of a unique microstructure, the rim structure, in high burnup UO 2 fuels has been elucidated by transmission electron microscopy (TEM). Specimens were prepared from the fuel peripheral region, using pellets which had been irradiated to a wide range of burnups (6-83 GWd/t; 10 GWd/t = 2.5 × 10 20 fissions/cm 3) in light water reactors. Dislocation density and volume fraction of intragranular bubbles increase with burnup. Low angle boundaries begin to form above 30-40 GWd/t. The TEM images and selected area electron diffraction (SAD) analyses of the rim structure observed in the 83 GWd/t fuel show: (1) sub-divided grains, 20-30 nm in size, with high angle boundaries due to the accumulation of an extremely high density of sub-boundaries; (2) recrystallized grains, 50-200 nm in size, adjacent to the sub-divided grain region, which are induced by the stored energy of the matrix; and (3) coarsened intragranular bubbles generated by radiation-induced excess vacancies.

Nogita, K.; Une, K.

1994-06-01

360

Effects of burnup on fission product release and implications for severe fuel damage events

Xe, Kr, and I fission-product release data from (a) Halden tests where release in intact rods was measured during irradiation at burnups to 18,000 MWd/t and fuel temperatures of 800 to 1800/sup 0/K, and (b) Power Burst Facility (PBF) tests where trace-irradiated fuel (approx. = 90 MWd/t) was driven to temperatures of >2400/sup 0/K and fuel liquefaction occurred are discussed and related to fuel morphology. Results from both indicate that the fission-product morphology and fuel restructuring govern release behavior. The Halden tests show low release at beginning of life with a 10-fold increase at burnups in excess of 10,000 MWd/t, due to the development of grain boundary interlinkage at higher burnups. Such dependence of release on morphology characteristics is consistent with findings from the PBF tests, where for trace-irradiated fuel, the absence of interlinkage accounts for the low release rates observed during initial fuel heatup, with subsequent enhanced Xe, Kr, and I release via liquefaction or quench-induced destruction of the grain structure. Morphology is also shown to influence the chemical release form of I and Cs fission products.

Appelhans, A.D.; Cronenberg, A.W.; Carboneau, M.L.

1984-01-01

361

Rim structure formation and high burnup fuel behavior of large-grained UO 2 fuels

NASA Astrophysics Data System (ADS)

Irradiation-induced fuel microstructural evolution of the sub-divided grain structure, or rim structure, of large-grained UO 2 pellets has been examined through detailed PIEs. Besides standard grain size pellets with a grain size range of 9-12 ?m, two types of undoped and alumino-silicate doped large-grained pellets with a range of 37-63 ?m were irradiated in the Halden heavy water reactor up to a cross-sectional pellet average burnup of 86 GWd/t. The effect of grain size on the rim structure formation was quantitatively evaluated in terms of the average Xe depression in the pellet outside region measured by EPMA, based on its lower sensitivity for Xe enclosed in the coarsened rim bubbles. The Xe depression in the high burnup pellets above 60 GWd/t was proportional to d-0.5- d-1.0 ( d: grain size), and the two types of large-grained pellets showed remarkable resistance to the rim structure formation. A high density of dislocations preferentially decorated the as-fabricated grain boundaries and the sub-divided grain structure was localized there. These observations were consistent with our proposed formation mechanism of rim structure, in which tangled dislocation networks are organized into the nuclei for recrystallized or sub-divided grains. In addition to higher resistance to the microstructure change, the large-grained pellets showed a smaller swelling rate at higher burnups and a lower fission gas release during base irradiation.

Une, K.; Hirai, M.; Nogita, K.; Hosokawa, T.; Suzawa, Y.; Shimizu, S.; Etoh, Y.

2000-01-01

362

Application of perturbation theory to lattice calculations based on method of cyclic characteristics

NASA Astrophysics Data System (ADS)

Perturbation theory is a technique used for the estimation of changes in performance functionals, such as linear reaction rate ratio and eigenvalue affected by small variations in reactor core compositions. Here the algorithm of perturbation theory is developed for the multigroup integral neutron transport problems in 2D fuel assemblies with isotropic scattering. The integral transport equation is used in the perturbative formulation because it represents the interconnecting neutronic systems of the lattice assemblies via the tracking lines. When the integral neutron transport equation is used in the formulation, one needs to solve the resulting integral transport equations for the flux importance and generalized flux importance functions. The relationship between the generalized flux importance and generalized source importance functions is defined in order to transform the generalized flux importance transport equations into the integro-differential equations for the generalized adjoints. Next we develop the adjoint and generalized adjoint transport solution algorithms based on the method of cyclic characteristics (MOCC) in DRAGON code. In the MOCC method, the adjoint characteristics equations associated with a cyclic tracking line are formulated in such a way that a closed form for the adjoint angular function can be obtained. The MOCC method then requires only one cycle of scanning over the cyclic tracking lines in each spatial iteration. We also show that the source importance function by CP method is mathematically equivalent to the adjoint function by MOCC method. In order to speed up the MOCC solution algorithm, a group-reduction and group-splitting techniques based on the structure of the adjoint scattering matrix are implemented. A combined forward flux/adjoint function iteration scheme, based on the group-splitting technique and the common use of a large number of variables storing tracking-line data and exponential values, is proposed to reduce the computing time when both direct and adjoint solutions are required. A problem that arises for the generalized adjoint problem is that the direct use of the negative external generalized adjoint sources in the adjoint solution algorithm results in negative generalized adjoint functions. A coupled flux biasing/decontamination scheme is applied to make the generalized adjoint functions positive using the adjoint functions in such a way that it can be used for the multigroup rebalance technique. Next we consider the application of the perturbation theory to the reactor problems. Since the coolant void reactivity (CVR) is a important factor in reactor safety analysis, we have decided to select this parameter for optimization studies. We consider the optimization and adjoint sensitivity techniques for the adjustments of CVR at beginning of burnup cycle (BOC) and k eff at end of burnup cycle (EOC) for a 2D Advanced CANDU Reactor (ACR) lattice. The sensitivity coefficients are evaluated using the perturbation theory based on the integral transport equations. Three sets of parameters for CVR-BOC and keff-EOC adjustments are studied: (1) Dysprosium density in the central pin with Uranium enrichment in the outer fuel rings, (2) Dysprosium density and Uranium enrichment both in the central pin, and (3) the same parameters as in the first case but the objective is to obtain a negative checkerboard CVR at beginning of cycle (CBCVR-BOC). To approximate the sensitivity coefficient at EOC, we perform constant-power burnup/depletion calculations for 600 full power days (FPD) using a slightly perturbed nuclear library and the unperturbed neutron fluxes to estimate the variation of nuclide densities at EOC. Sensitivity analyses of CVR and eigenvalue are included in the study. In addition the optimization and adjoint sensitivity techniques are applied to the CBCVR-BOC and keff-EOC adjustment of the ACR lattices with Gadolinium in the central pin. Finally we apply these techniques to the CVR-BOC, CVR-EOC and keff-EOC adjustment of a CANDU lattice of which the burnup period is extended f

Assawaroongruengchot, Monchai

363

NSDL National Science Digital Library

The Ethics CORE Digital Library, funded by the National Science Foundation, "brings together information on best practices in research, ethics instruction and responding to ethical problems that arise in research and professional life." It's a remarkable site where visitors can make their way through ethics resources for dozens of different professions and activities. The Resources by Discipline area is a great place to start. Here you will find materials related to the biological sciences, business, computer & information science, along with 14 additional disciplines. The Current News area is a great place to learn about the latest updates from the field. Of note, these pieces can easily be used in the classroom or shared with colleagues. The dynamism of the site can be found at the Interact with Ethics CORE area. Active learning exercises can be found here, along with instructional materials and visitors' own lessons learned.

364

NASA Astrophysics Data System (ADS)

The main aim of this research is the development and validation of computational schemes for advanced lattice codes. The advanced lattice code which forms the primary part of this research is "DRAGON Version4". The code has unique features like self shielding calculation with capabilities to represent distributed and mutual resonance shielding effects, leakage models with space-dependent isotropic or anisotropic streaming effect, availability of the method of characteristics (MOC), burnup calculation with reaction-detailed energy production etc. Qualified reactor physics codes are essential for the study of all existing and envisaged designs of nuclear reactors. Any new design would require a thorough analysis of all the safety parameters and burnup dependent behaviour. Any reactor physics calculation requires the estimation of neutron fluxes in various regions of the problem domain. The calculation goes through several levels before the desired solution is obtained. Each level of the lattice calculation has its own significance and any compromise at any step will lead to poor final result. The various levels include choice of nuclear data library and energy group boundaries into which the multigroup library is cast; self shielding of nuclear data depending on the heterogeneous geometry and composition; tracking of geometry, keeping error in volume and surface to an acceptable minimum; generation of regionwise and groupwise collision probabilities or MOC-related information and their subsequent normalization thereof, solution of transport equation using the previously generated groupwise information and obtaining the fluxes and reaction rates in various regions of the lattice; depletion of fuel and of other materials based on normalization with constant power or constant flux. Of the above mentioned levels, the present research will mainly focus on two aspects, namely self shielding and depletion. The behaviour of the system is determined by composition of resonant and predominantly scattering isotopes. When the concentration of resonant isotopes is small, its presence does not affect the flux shape which is smooth. But when the concentration becomes high, there will be dips in the flux where resonances of the isotopes occur. This will affect the reaction rate, which is a product of cross section and flux. The reaction rate will thus be lower than that when one does not consider the flux dip. This is the phenomenon of self shielding. Self shielding treatment is thus a very important aspect of reactor lattice analysis code. This needs to be correctly modelled to obtain a physically sound and acceptable solution. In this research we will be looking into behaviour of the advanced self shielding models that have been incorporated in the code DRAGON Version4. The self shielding models are primarily classified into two broad groups, which are based on "equivalence in dilution" and "subgroup approach". These self shielding models will be tested against a variety of lattices which include Canada Deuterium Uranium (CANDU-6), CANDU-New Generation (CANDU-NG), Light Water Reactor (LWR), and High Conversion Light Water Reactor (HCLWR). The fuel composition will vary from natural uranium oxide to enriched uranium oxide and plutonium-uranium mixed oxide (MOX). We will also consider the presence of strong neutron absorbers like gadolinium and dysprosium in the lattice. The coolant/moderator chosen for the analysis will be light water/heavy water or a combination. The lattice geometry will vary from square, hexagonal and annular. Thus a broad spectrum of lattices will be analysed to assess the behaviour of advanced self shielding models. The results obtained using DRAGON will be validated against that obtained using Monte Carlo code MCNP5. The reference solutions for all situations will be provided by MCNP5. The depletion behaviour of any lattice will depend on the power or flux normalization that is considered. In general the flux in various regions is estimated with reference to a single neutron absorbed a

Ramamoorthy, Karthikeyan

365

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

NASA Astrophysics Data System (ADS)

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Subhki, Muhamad Nurul; Ismail

2010-12-01

366

Modelling dislocation cores in Forsterite

NASA Astrophysics Data System (ADS)

Olivine (Mg,Fe)2SiO4 is considered as the main constituent of the Earth's upper mantle (down to 410 km deep). The rheology of, and convection in, the upper mantle is therefore controlled by the deformation mechanisms of this mineral. Numerous experimental studies have been undertaken leading to a good description of the deformation mechanisms and rheological properties of this mineral at ambient pressure. However, recent studies have show that [001] glide is enhanced over [100] glide when pressure increases or when trace amounts of water are dissolved in the crystals. These observations have a lot of implications on our understanding of the rheology of the upper mantle and call for a more detailed description of the dislocation cores and dynamics. The Peierls-Nabarro (PN) model including generalized stacking fault energies is a privileged tool to calculate core structures at a remarkably low cost. Moreover, the PN model, which is usually restricted to the description of planar cores, is very adapted to look for the most mobile core configurations. However, dislocation cores may exhibit distinct, low-energy, configurations that are not described by the PN model. We present here new calculations based on full atomistic calculations (using the THB1 potential) and a method coupling Peierls-Nabarro and element-free Galerkin methods. These techniques expand the possibilities of previously reported calculations, in particular in permitting modeling 3D dislocation cores. We show that, [100] dislocations may exhibit non collinear dissociation in the (010) plane following the reaction [100] = 1/6[3 0 1] +1/6[3 0 -1]. We also discuss several possible core structures for [001] screw dislocations, including non-planar core spreadings.

Cordier, P.; Metsue, A.; Carrez, P.; Walker, A. M.; Denoual, C.; Mainprice, D.

2008-12-01

367

Air core pulse transformer design

Cylindrical-air-core pulse transformers capable of passing high-voltage\\/high-energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. Special computer codes were written to evaluate their performance. The analysis includes calculation of the self and mutual inductances as determined by the dimensions and insulation

J. P. O'Loughlin; J. D. Sidler; Gerry J. Rohwein

1988-01-01

368

REACTOR PHYSICS CALCULATIONS FOR THE MSRE

A compilation is presented of results obtained to date from a number of ; reactor physics calculations for the molten salt reactor experiment (MSRE). ; Included are one-dimensional multigroup and two-dimensional twogroup calculations ; of critical mass, flux, and power density distributions; gamma heating in the ; core can, reactor vessel, and core support grid; drain tank criticality; and an

Nestor; C. W. Jr

1960-01-01

369

Influence of fuel vibration on PWR neutron noise associated with core barrel motion

Ex-core neutron detector noise has been utilized to monitor core support barrel (CSB) vibrations. In order to observe long-term changes, noise signals at Sequoyah-1 were monitored continuously during the whole first fuel cycle and part of the second cycle. Results suggest that neutron noise measurements performed infrequently may not provide adequate surveillance of the CSB because it may be difficult to separate noise amplitude changes due solely to CSB motion from changes caused by fuel motion and burnup. (DLC)

Sweeney, F.J.; March-Leuba, J.

1984-01-01

370

NASA Astrophysics Data System (ADS)

The performance of MOX fuel irradiated in the advanced thermal reactor, FUGEN, to a burnup of 47.5 GWd/t, was investigated by using a telescope, optical microscope, SEM and EPMA. Observations focused on elucidating the corrosion behavior of the cladding inner surface. A reaction layer was observed at burnups higher than about 35 GWd/t. The relationship between the thickness of the reaction layer and burnup was similar to that reported in the literature for conventional UO 2 fuel and other MOX fuels. The existence of a plutonium spot near the outer surface of the fuel pellet had no significant effect on the thickness of the reaction layer. A bonding layer was observed on the cladding inner surface. Its morphology and elemental distributions were not so different from those in BWR UO 2 fuel rods irradiated to high burnup, in which the fission gas release rate is high. In addition, the dependences of bonding layer formation on the burnup and linear heat rating were similar to results of UO 2 fuel rods. It was, thus, suggested that the bonding layer formation mechanism was similar in both UO 2 and MOX fuel rods.

Tanaka, Kosuke; Maeda, Koji; Sasaki, Shinji; Ikusawa, Yoshihisa; Abe, Tomoyuki

2006-10-01

371

Modeling of fuel rod steady state and transient behavior over the full range of burnup

The heightened recent attention given to fuel rod behavior at high burnup has largely been the result of a few reactivity initiation experiments conducted in France and Japan. Regardless of the merits of these tests, their outcome has underscored the need for improved analytical methods for both steady-state performance evaluation and transient safety analysis. At issue is the ability to reliably predict fuel rod behavior over the full range of burnup under all conditions. The overall process can be divided into two essential components: the collection of high-burnup material properties data and the development of predictive computer codes with essential special-effects models. Lessons learned from the Electric Power Research Institute`s (EPRI`s) recent efforts in understanding and properly interpreting the reactivity insertion accident test results indicate that a first-principles approach to fuel rod behavioral modeling within a robust analytical capability is vital for describing the complex interaction between the various phenomena involved. Thus, the heavy reliance on analysis-by-correlation and the excessive use of adjustable parameters is not adequate, especially in extrapolating the analytical results beyond the correlation database. Guided by this philosophy, EPRI has initiated a major fuel rod modeling effort that builds on two existing EPRI codes: the ESCORE code for steady-state analysis and the FREY code for transient analysis. The new combined code, FALCON, is not a mere merging of the one-dimensional ESCORE and the two-dimensional FREY but rather an innovative construct of robust numerics and relevant material models.

Yagnik, S.K.; Yang, R.L. [Electric Power Research Institute, Palo Alto, NC (United States); Rashid, Y.R.; Montgomery, R.O. [Anatech Research Corp., San Diego, CA (United States)

1997-12-01

372

The inert matrix fuels are a promising option to reduce-eliminate worldwide plutonium stockpiles by burning it in LWRs. These fuels, where plutonium is hosted in a U-free inert matrix phase, may reach high burning efficiency while preventing new plutonium build-up under irradiation. A specific investigation on CSZ and thoria inert matrices has been developed by ENEA since several years. In-pile testing on the ENEA-conceived innovative fuels is ongoing in the OECD Halden HBWR since June 2000 (IFA-652 experiment). The registered burnup at the end of 2005 is about 38 MWd.kgU{sub eq}{sup -1} vs. 45 MWd.kgU{sub eq}{sup -1} (40 MWd.kgUOX{sub eq}{sup -1}) target. Fuel pins are equipped with fuel temperature thermocouples, internal pressure transducers and fuel stack elongation sensors, with the task of studying thermal conductivity and its degradation with burnup, densification-swelling behaviour and the FGR. In this paper, the response at low burnup (< 7 MWd.kgU{sub eq}{sup -1}) of CSZ-based fuels loaded in IFA-652, is analysed by means of the TRANSURANUS code. To this purpose, a comprehensive modelling of the above mentioned un-irradiated fuels, mainly relying on the thermophysical characterisation performed at the JRC/ITU-Karlsruhe, has been implemented in a custom TRANSURANUS version (TU-IMF). A comparison of the code predictions vs. the experimental data, aimed at evaluating the early-stage under irradiation phenomena, particularly densification and relocation, has been performed. (authors)

Calabrese, R.; Vettraino, F. [ENEA, Nuclear Fission Division, via Martin di Monte Sole 4, 40129 Bologna (Italy); Tverberg, T. [OECD Halden Reactor Project, Institutt for energiteknikk, P.O. Box 175, N-1751 Halden (Norway)

2006-07-01

373

The high burnup BWR 9x9 lead use fuel assemblies, which have been designed for maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiations to confirm the reliability of the current safety evaluation methodology, and to accumulate data to judge the adequacy to apply it to the future higher burnup fuel. After 3 and 5 cycle irradiations, post irradiation examinations were performed for both 9x9 Type-A and Type-B fuel assemblies. Both Type LUAs utilize Zry-2 claddings, while there are deviation in the contents of impurity and alloying elements between Type-A and Type-B, especially in Fe and Si concentration. Measured oxide thicknesses of fuel rods showed no significant difference between after 3 and 5 cycle irradiation except for some rods at corner position in Type B LUA. The axial profile of hydrogen concentration and oxide thickness for the corner rods in Type B LUA after 5 cycle irradiation had peaks at the second lowest span from the bottom. The maximum oxide thickness is about 50 {mu}m on the surface facing the bundle outside at the second lowest span and dense hydrides layer (Hydride rim) is observed in peripheral region of cladding showing unexpected high hydrogen concentration. The results of calculated thermal-hydraulic conditions show that the thermal neutron flux at the corner position was higher than the other position. On the other hand, the void fraction and the mass flux were relatively lower at the corner position. The oxide thickness on spacer band and spacer cell of Zry-2 increases from 3 to 5 cycle irradiations. Spacer band of Zry-4 showed significantly thick oxide after 5 cycle irradiations but Hydrogen concentration was relatively small in contrast its obviously thick oxide in comparison with Zry-2 spacer bands. The large increase in hydrogen concentration was measured in Zry-2 spacers after 5 cycle irradiations and the evaluated hydrogen pick-up rate also increased remarkably. (authors)

Miyashita, Toshiyasu; Nakae, Nobuo; Ogata, Keizo; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization, 3-17-1 Toranomon, Minato-ku, Tokyo, 105-0001 (Japan); Matsumoto, Toshio [Grobal Nuclear Fuel - Japan, 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan); Kakiuchi, Kazuo [Nuclear Fuel Industries, Ltd., 3135-41 Muramatsu, Tokai-mura, Naka-gun, Ibaraki, 319-1196 (Japan)

2007-07-01

374

NASA Technical Reports Server (NTRS)

Fuel volume swelling and clad diametral creep strains were calculated for five fuel pins, clad with either T-111 (Ta-8W-2.4Hf) or PWC-11 (Nb-1Zr-0.1C). The fuel pins were irradiated to burnups between 2.7 and 4.6%. Clad temperatures were between 1750 and 2400 F (1228 and 1589 K). The maximum percentage difference between calculated and experimentally measured values of volumetric fuel swelling is 60%.

Davison, H. W.; Fiero, I. B.

1971-01-01

375

Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process.

Kobayashi, Yoko; Aiyoshi, Eitaro

2005-07-15

376

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01

377

Overview of core designs and requirements/criteria for core restraint systems

The requirements and lifetime criteria for the design of a Liquid Metal Fast Breeder Reactor (LMFBR) Core Restraint System are presented. A discussion of the three types of core restraint systems used in LMFBR core design is given. Details of the core restraint system selected for FFTF are presented and the reasons for this selection given. Structural analysis procedures being used to manage the FFTF assembly irradiations are discussed. Efforts that are ongoing to validate the calculational methods and lifetime criteria are presented.

Sutherland, W.H.

1984-09-01

378

National Technical Information Service (NTIS)

This work presents comparisons between measured fluxes obtained by activation of Manganese foils in the light water, enriched uranium research pool reactor RA-2 MTR (Materials Testing Reactors) fuel element and fluxes calculated by the finite element meth...

C. R. Calabrese C. R. Grant

1990-01-01

379

In accordance with the need to determine whether cracking of the ceramic core disks which will be constructed and used in the High Temperature Test Facility (HTTF) for heatup and cooldown experiments, a set of calculation were performed using Abaqus to investigate the thermal stresses levels and likelihood for cracking. The calculations showed that using the material properties provided for the Greencast 94F ceramic, cracking is predicted to occur. However, this modeling does not predict the size or length of the actual cracks. It is quite likely that cracks will be narrow with rough walls which would impede the flow of coolant gases entering the cracks. Based on data recorded at Oregon State University using Greencast 94F samples that were heated and cooled at prescribed rates, it was concluded that the likelihood that the cracks would be detrimental to the experimental objectives is small.

Brian D. Hawkes; Richard Schultz

2012-07-01

380

Thermal diffusivity of homogeneous SBR MOX fuel with a burn-up of 35 MWd/kgHM

NASA Astrophysics Data System (ADS)

The effect of burn-up on the thermal conductivity of homogeneous SBR MOX fuel is investigated and compared with standard UO 2 LWR fuel. New thermal diffusivity results obtained on SBR MOX fuel with a pellet burn-up of 35 MWd/kgHM are reported. The thermal diffusivity measurements were carried out at three radial positions using a shielded "laser-flash" device and show that the thermal diffusivity increases from the pellet periphery to the centre. The fuel thermal conductivity was found to be in the same range as for UO 2 of similar burn-up. The annealing behaviour was characterized in order to identify the degradation due to the out-of-pile auto-irradiation.

Cozzo, C.; Staicu, D.; Pagliosa, G.; Papaioannou, D.; Rondinella, V. V.; Konings, R. J. M.; Walker, C. T.; Barker, M. A.; Hervé, P.

2010-05-01

381

NASA Astrophysics Data System (ADS)

Structural analyses of dry casks containing high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). A test procedure was developed to simulate the effects of drying-storage temperature histories. Following drying-storage simulation, samples were subjected to ring-compression test (RCT) loading, which was used as a ductility screening test and to simulate pinch-type loading that may occur during cask transport. RCT samples with <2% offset strain prior to >50% wall cracking were assessed as brittle. Prior to testing high-burnup cladding, many tests were conducted with pre-hydrided Zircaloy-4 (Zry-4) and ZIRLO™ to determine target 400 °C hoop stresses for high-burnup rodlets. Zry-4 cladding segments, from a 67-GWd/MTU fuel rod, with 520-620 wppm hydrogen and ZIRLO™ cladding segments from a 70-GWd/MTU fuel rod, with 350-650 wppm hydrogen were defueled and tested. Following drying-storage simulation, the extent of radial-hydride precipitation was characterized by the radial-hydride continuity factor. It was found that the DBTT was dependent on: cladding material, irradiation conditions, and drying-storage histories (stress at maximum temperature). High-burnup ZIRLO™ exhibited higher susceptible to radial-hydride formation and embrittlement than high-burnup Zry-4. It was also observed that uniformly pre-hydrided, non-irradiated cladding was not a good surrogate for high-burnup cladding because of the high density of circumferential hydrides across the wall and the high metal-matrix ductility for pre-hydrided cladding.

Billone, M. C.; Burtseva, T. A.; Einziger, R. E.

2013-02-01

382

EBSD and TEM Characterization of High Burn-up Mixed Oxide Fuel

Understanding and studying the irradiation behavior of high burn-up oxide fuel is critical to licensing of future fast breeder reactors. Advancements in experimental techniques and equipment are allowing for new insights into previously irradiated samples. In this work dual column focused ion beam (FIB)/scanning electron microscope (SEM) was utilized to prepared transmission electron microscope samples from mixed oxide fuel with a burn-up of 6.7% FIMA. Utilizing the FIB/SEM for preparation resulted in samples with a dose rate of <0.5 mRem/h compared to approximately 1.1 R/h for a traditionally prepared TEM sample. The TEM analysis showed that the sample taken from the cooler rim region of the fuel pellet had approximately 2.5x higher dislocation density than that of the sample taken from the mid-radius due to the lower irradiation temperature of the rim. The dual column FIB/SEM was additionally used to prepared and serially slice approximately 25 um cubes. High quality electron back scatter diffraction (EBSD) were collected from the face at each step, showing, for the first time, the ability to obtain EBSD data from high activity irradiated fuel.

Melissa C Teague; Brian P. Gorman; Brandon D Miller; Jeffrey King

2014-01-01

383

Tritium Burn-up Depth and Tritium Break-Even Time

NASA Astrophysics Data System (ADS)

Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317 g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18 kg that is more than the least required amount of tritium storage to start up three of FEB-like fusion reactors.

Li, Cheng-Yue; Deng, Bai-Quan; Huang, Jin-Hua; Yan, Jian-Cheng

2006-08-01

384

R and D of Oxide Dispersion Strengthening Steels for High Burn-up Fuel Claddings

Research and development of fuel clad materials for high burn-up operation of light water reactor and super critical water reactor (SCPWR) will be shown with focusing on the effort to overcome the requirements of material performance as the fuel clad. Oxide dispersion strengthening (ODS) steels are well known as a high temperature structural material. Recent irradiation experiments indicated that the steels were quite highly resistant to neutron irradiation embrittlement, showing hardening without accompanying loss of ductility. High Cr ODS steels whose chromium concentration was in the range from 15 to 19 wt% showed high resistance to corrosion in supercritical pressurized water (SCPW). As for the susceptibility to hydrogen embrittlement of ODS steels, the critical hydrogen concentration required to hydrogen embrittlement is ranging 10{approx}12 wppm that is approximately one order of magnitude higher value than that of 9Cr reduced activation ferritic (RAF) steel. In the ODS steels, the fraction of helium desorption by bubble migration mechanism was smaller than that in the RAF steel, indicating that the ODS steels are also resistant to helium He bubble-induced embrittlement. Finally, it is demonstrated that the ODS steels are very promising for the fuel clad material for high burn-up operation of water-cooling reactors. (authors)

Kimura, A.; Cho, H.S.; Lee, J.S.; Kasada, R. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Ukai, S. [Japan Nuclear Cycle Development Institute, Tokai (Japan); Fujiwara, M. [Kobelco, Ltd, Takatsukadai, Nishi-ku, Kobe (Japan)

2004-07-01

385

TEM analysis of pellet-cladding bonding layer in high burnup BWR fuel

NASA Astrophysics Data System (ADS)

Detailed analysis of the pellet-cladding bonding layer in high burnup nuclear fuel has been done by transmission electron microscopy (TEM). A specimen was prepared from the fuel, which had been irradiated to the pellet average burnup of 49 GWd/tU (1.2×10 21 fissions/cm 3) in a boiling water reactor (BWR). A 20 ?m thick bonding layer which consisted of two regions was observed. In one region from the inner surface of the Zr liner cladding to 12-13 ?m away, the main species identified was ZrO 2 with a small amount of dissolved UO 2 also present. This ZrO 2 consisted of a mixture of cubic polycrystals of a few nanometers in size and an amorphous phase, but no monoclinic crystal, which is the stable ZrO 2 phase at temperatures below 1170°C, was found. In a second region from the pellet surface to about 7 ?m away, both a cubic solid solution of (U, Zr)O 2 and an amorphous phase existed, in which the concentrations of UO 2 and ZrO 2 changed continuously. The formation of substitutional solid solution progressed during irradiation due to a strong contact between cubic UO 2 and ZrO 2, which induced subsequent mutual diffusion of U and Zr. Phase transformation from monoclinic to cubic ZrO 2 and amorphization were discussed in connection with fission damage.

Nogita, K.; Une, K.; Korei, Y.

1996-08-01

386

Thermal conductivity of homogeneous and heterogeneous MOX fuel with up to 44 MWd/kgHM burn-up

NASA Astrophysics Data System (ADS)

New thermal diffusivity data for homogeneous SBR and heterogeneous MIMAS and OCOM MOX fuels are reported. No significant difference between the thermal diffusivity of the homogeneous and heterogeneous fuels was found at the burn-up up to 44 MWd/kgHM. These measurements, combined with previously published results or correlation functions for irradiated UO 2 and MOX were compared and it was found that separate correlations for these two fuels are not justified. A correlation for the thermal conductivity of irradiated UO 2 and MOX as a function of burn-up and irradiation temperature is proposed.

Staicu, D.; Cozzo, C.; Pagliosa, G.; Papaioannou, D.; Bremier, S.; Rondinella, V. V.; Walker, C. T.; Sasahara, A.

2011-05-01

387

National Institute of Standards and Technology Data Gateway

SRD 166 MEMS Calculator (Web, free access) This MEMS Calculator determines the following thin film properties from data taken with an optical interferometer or comparable instrument: a) residual strain from fixed-fixed beams, b) strain gradient from cantilevers, c) step heights or thicknesses from step-height test structures, and d) in-plane lengths or deflections. Then, residual stress and stress gradient calculations can be made after an optical vibrometer or comparable instrument is used to obtain Young's modulus from resonating cantilevers or fixed-fixed beams. In addition, wafer bond strength is determined from micro-chevron test structures using a material test machine.

388

Core formation in giant gaseous protoplanets

Sedimentation rates of silicate grains in gas giant protoplanets formed by disk instability are calculated for protoplanetary masses between 1 MSaturn to 10 MJupiter. Giant protoplanets with masses of 5 MJupiter or larger are found to be too hot for grain sedimentation to form a silicate core. Smaller protoplanets are cold enough to allow grain settling and core formation. Grain

Ravit Helled; Gerald Schubert

2008-01-01

389

Modelling Characteristics of Ferromagnetic Cores with the Influence of Temperature

NASA Astrophysics Data System (ADS)

The paper is devoted to modelling characteristics of ferromagnetic cores with the use of SPICE software. Some disadvantages of the selected literature models of such cores are discussed. A modified model of ferromagnetic cores taking into account the influence of temperature on the magnetizing characteristics and the core losses is proposed. The form of the elaborated model is presented and discussed. The correctness of this model is verified by comparing the calculated and the measured characteristics of the selected ferromagnetic cores.

Górecki, K.; Rogalska, M.; Zar?bski, J.; Detka, K.

2014-04-01

390

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

Clement, J. D.; Rust, J. H.

1978-01-01

391

NASA Technical Reports Server (NTRS)

This report gives an overall view of the CORE program at Goddard Space Flight Center (GSFC). It summarizes the different CORE sessions and gives information about the technical staff. The outlook summarizes the evolution of the different CORE programs.

Thomas, Cynthia; Vandenberg, Nancy

1999-01-01

392

This paper described the procedure of implementation and the possibilities of the BURNUP program. The purpose of the program is to predict the change in the nuclear composition of the materials of which a reactor is made in the course of its run and compute the radiation characteristics of the materials after their irradiation.

Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Centre Kurchatov Institute (Russian Federation)

2012-12-15

393

National Technical Information Service (NTIS)

The NATHALIE code intended for processing the data from the in-pile gamma -scanning measurements effected at Siloe and Melusine is discussed with the results obtained especially for spent fuel power and burnup determination, and when the migration of a gr...

A. Maman P. Chenebault J. P. Hairion F. Michel

1976-01-01

394

National Technical Information Service (NTIS)

The research and development program on the recycling of plutonium in light water reactors and the measurement campaign undertaken in the MINERVE reactor aim to provide the experimental data needed in support of the calculation methods for light water rea...

C. Golinelli

1982-01-01

395

Progress in the Research Programs to Elucidate Axial Cracking Fuel Failure at High Burnup

A fuel failure with an axial crack starting outside the cladding and penetrating inwards was experienced by high burnup BWR fuel rods in power ramp test. On the other hand, no fuel failure caused by power ramp test has been currently reported on PWR fuel rods at burnups higher than 50 GWd/t. Extensive research programs regarding hydrogen behaviors and mechanical performances on irradiated BWR and PWR fuel claddings have been carried out to clarify the mechanism of the axial cracking and to quantify the conditions to cause fuel failure. Hydrogen solid solubility measurement on irradiated Zircaloy-2 materials showed almost comparable results to those on unirradiated ones. Hydride re-distribution and re-orientation behaviors were tested by heating irradiated BWR claddings with Zr-liner under the conditions of applied radial heat flux (temperature gradient) and circumferential stress. Mechanical performances of BWR claddings were evaluated mainly by the internal pressurizing tests. Internal pressurization tests applying various pressurizing sequences, e.g. stepwise increase in pressure with holding intervals, were also conducted to simulate crack propagation behaviors. Some specimens demonstrated characteristic fracture surfaces similar to those observed on the failed fuel rods after the power ramp. Mechanical performances of irradiated PWR claddings were tested at temperatures of 573 to 723 K. Metallographic examination after tensile tests revealed a large number of incipient cracks within the region of cladding outer rim where a concentrated hydride layer (hydride rim) has been formed during irradiation. Crack propagation test using an expanding mandrel device demonstrated the crack propagation at 573 K but no propagation at 658 K. (authors)

Ogata, Keizo; Aomi, Masaki; Baba, Toshikazu; Kamimura, Katsuichiro [Japan Nuclear Energy Safety Organization, 3-17-1 Toranomon, Minato-ku, Tokyo 105-0001 (Japan); Etoh, Yoshinori [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Ito, Kunio [Grobal Nuclear Fuel - Japan Co., Ltd., 3-1 Uchikawa 2-chone, Yokosuka 239-0836 (Japan); Kido, Toshiya [Nuclear Development Corporation, 622-12 Funaishikawa, Tokai-mura, Ibaraki 319-1111 (Japan); Teshima, Hideyuki [Mitsubishi Heavy Industries, Ltd. 1-1, Wadasaki-cho 1-chome, Hyogo-ku, Kobe 652-8585 (Japan)

2007-07-01

396

One of my passions in life is to try and understand how we have developed the wonderful calculating ability we currently possess. Anyone with this hobby will undoubtedly start by looking back to see how the first PCs were developed, then progress back to older \\

Michael R. Williams

2004-01-01

397

NSDL National Science Digital Library

This web site, which is part of the NCTM Illuminations project, allows students to challenge themselves or opponents from anywhere in the world by playing games that are organized around content from the upper elementary and middle grades math curriculum. The games allow students to learn about fractions, factors, multiples, symmetry, as well as practice important skills like basic multiplication and calculating area.

2011-01-01

398

NSDL National Science Digital Library

This interactive calculator produced by Teachers' Domain helps you determine the mercury levels in various types of fish, and enables you to make more informed choices about which fish are safe to eat and which should be avoided or eaten infrequently.

Foundation, Wgbh E.

2010-12-23

399

NSDL National Science Digital Library

This simple inflation calculator uses the Consumer Price Index to adjust any given amount of money, from 1800 to 1998. Creator S. Morgan Friedman uses data from the Historical Statistics of the United States for statistics predating 1975 and the annual Statistics Abstracts of the United States for data from 1975 to 1998. Links to other online inflation information are also included.

Friedman, S. M.

400

Benchmark data for validating irradiated fuel compositions used in criticality calculations

To establish criticality safety margins utilizing burnup credit in the storage and transport of spent reactor fuels requires a knowledge of the uncertainty in the calculated fuel composition used in making the reactivity assessment. To provide data for validating such calculated burnup fuel compositions, radiochemical assays have been obtained as part of the United States Department of Energy From-Reactor Cask Development Program. Assay results and associated operating histories on the initial three samples analyzed in this effort are presented. The three samples were taken from different axial regions of a Pressurized Water Reactor fuel rod and represent radiation exposures of about 37, 27, and 44 GWd/MTU. The data are presented in a benchmark type format to facilitate identification/referencing and computer code input.

Bierman, S.R.; Talbert, R.J.

1994-10-01

401

Dislocation core-core interaction and Peierls stress in a model hexagonal lattice

A series of atomistic calculations is performed in order to explore dislocation core-core interactions and the Peierls stress in a model hexagonal lattice. The method of calculation is the lattice Green`s function method, using several pair potentials with variable parameters. We confirm that dislocation cores broaden as a pair of dislocations with opposite sign move closer to each other. Continuum theories are surprisingly accurate in describing the dislocation-dislocation interaction force even in the range of strong core-core overlap. However, our atomistic calculations show that while the relation between the Peierls stress and dislocation width is exponential as the Peierls-Nabarro model predicts, that model underestimates the Peierls stress by nearly a factor of 10{sup 4}.

Zhou, S.J.; Carlsson, A.E. [Department of Physics, Washington University, St. Louis, Missouri 63130 (United States)] [Department of Physics, Washington University, St. Louis, Missouri 63130 (United States); Thomson, R. [Materials Science and Engineering Laboratory, National Institute of Standards and Technology, Gaithersburg, Maryland 20899 (United States)] [Materials Science and Engineering Laboratory, National Institute of Standards and Technology, Gaithersburg, Maryland 20899 (United States)

1994-03-01

402

NSDL National Science Digital Library

This interactive applet helps students develop fluency and flexibility with numbers. At each of 6 difficulty levels the user is presented with 8 target numbers and a partial set of keys on a basic calculator (does not follow order of operations). The goal is to use the given keys to make as many of the target numbers as possible within the 3-minute time limit. Some levels include memory keys.

Barrow, Mandy

2008-01-01

403

Benchmark calculations of power distribution within assemblies.

National Technical Information Service (NTIS)

The main objective of this Benchmark is to compare different techniques for fine flux prediction based upon coarse mesh diffusion or transport calculations. We proposed 5 'core' configurations including different assembly types (17 x 17 pins, 'uranium', '...

C. Cavarec J. F. Perron D. Verwaerde J. P. West

1994-01-01

404

The influence of potassium on core and geodynamo evolution

NASA Astrophysics Data System (ADS)

We model the thermal evolution of the core and mantle using a parametrized convection scheme, and calculate the entropy available to drive the geodynamo as a function of time. The cooling of the core is controlled by the rate at which the mantle can remove heat. Rapid core cooling favours the operation of a geodynamo but creates an inner core that is too large; slower cooling reduces the inner core size but makes a geodynamo less likely to operate. Introducing potassium into the core retards inner core growth and provides an additional source of entropy. For our nominal model parameters, a core containing ~ 400 ppm potassium satisfies the criteria of present-day inner core size, surface heat flux, mantle temperature and cooling rate, and positive core entropy production. We have identified three possibilities that may allow the criteria to be satisfied without potassium in the core. (1) The core thermal conductivity is less than half the generally accepted value of 50 W m-1 K-1. (2) The core solidus and adiabat are significantly colder and shallower than results from shock experiments and ab initio simulations indicate. (3) The core heat flux has varied by no more than a factor of 2 over Earth history. All models we examined with the correct present-day inner core radius have an inner core age of <1.5 Gyr; prior to this time the geodynamo was sustained by cooling and radioactive heat production within a completely liquid core.

Nimmo, F.; Price, G. D.; Brodholt, J.; Gubbins, D.

2004-02-01

405

Parametric vibrations of cylindrical shells with a viscoelastic core

The method of calculating the axisymmetric and nonaxisymmetric parametric vibrations of a cylindrical shell bonded to an elastic core [2] is extended to the case of hollow and solid viscoelastic cores by substituting for the material moduli in the equations of motion of the core integral operators with kernels in the form of an exponential and a sum of exponentials.

A. E. Bogdanovich

1975-01-01

406

NSDL National Science Digital Library

This site contains many chemistry applets created by Jonathan Goodman and his group at Cambridge University. An example of an applet available is the Molecular Weight Calculation; whereby entering in a molecular formula, users are able to discover the HRMS weight, the molecular weight, the element percents, and the Molecular Ion Isotope Pattern. Interactive graphs are also available to assist chemistry students with concepts such as boiling points, pressure, and Consecutive First Step Reversible Reactions. Educators and students will also find many three dimensional depictions of the molecules including fused rings, aromatic rings, and Fullerenes.

Goodman, Jonathan

407

Pressure Vessel Calculations for VVER-440 Reactors

NASA Astrophysics Data System (ADS)

Monte Carlo calculations were performed for a selected cycle of the Paks NPP Unit II to test a computational model. In the model the source term was calculated by the core design code KARATE and the neutron transport calculations were performed by the MCNP. Different forms of the source specification were examined. The calculated results were compared with measurements and in most cases fairly good agreement was found.

Hordósy, G.; Hegyi, Gy.; Keresztúri, A.; Maráczy, Cs.; Temesvári, E.; Vértes, P.; Zsolnay, É.

2003-06-01

408

Gross Thermodynamics of 2-component Core Convection

NASA Astrophysics Data System (ADS)

We model the inner core by an alloy of iron and 8 percent sulphur or silicon and the outer core by the same mix with an additional 8 percent oxygen. When the liquid core freezes S and Si remain with the Fe to form the solid and excess O is ejected into the liquid. Properties of Fe, diffusion constants for S, Si, O, and chemical potentials are calculated by first principles methods under the assumption that S, O, Si react with the Fe but not with each other. This gives the parameters required to calculate the power supply to the geodynamo as the Earth's core cools. Compositional convection, driven by light O released at the inner core boundary on freezing, accounts for almost half the entropy balance and 13 percent of the heat balance. This means the same magnetic field can be generated with about half the heat throughput needed if the geodynamo were driven by heat alone. Chemical effects are small. Cooling rates below 69 K/Gyr are too low to maintain thermal convection everywhere; when the cooling rate lies between 30 and 69 K/Gyr convection at the top of the core is maintained compositionally against a stabilising temperature gradient; below 30 K/Gyr the dynamo fails completely. All cooling rates freeze the inner core in less than 1.5 Gyr. A very large amount of radioactive heating is needed to extend the life of the inner core to 3.5 Ga.

Gubbins, D.; Alfe, D.; Alfe, D.; Masters, G.; Price, D.; Gillan, M.

2001-12-01

409

Enhancements have been made to the REBUS-3/DIF3D code suite to facilitate its use for the design and analysis of prismatic Very High Temperature Reactors (VHTRs). A new cross section structure, using table-lookup, has been incorporated to account for cross section changes with burnup and fuel and moderator temperatures. For representing these cross section dependencies, three new modules have been developed using FORTRAN 90/95 object-oriented data structures and implemented within the REBUS-3 code system. These modules provide a cross section storage procedure, construct microscopic cross section data for all isotopes, and contain a single block of banded scattering data for efficient data management. Fission products other than I, Xe, Pm, and Sm, can be merged into a single lumped fission product to save storage space, memory, and computing time without sacrificing the REBUS-3 solution accuracy. A simple thermal-hydraulic (thermal-fluid) feedback model has been developed for prismatic VHTR cores and implemented in REBUS-3 for temperature feedback calculations. Axial conduction was neglected in the formulation because of its small magnitude compared to radial (planar) conduction. With the simple model, the average fuel and graphite temperatures are accurately estimated compared to reference STAR-CD results. The feedback module is currently operational for the non-equilibrium fuel cycle analysis option of REBUS-3. Future work should include the extension of this capability to the equilibrium cycle option of the code and additional verification of the feedback module. For the simulation of control rods in VHTR cores, macroscopic cross section deviations (deltas) have been defined to account for the effect of control rod insertion. The REBUS-3 code has been modified to use the appropriately revised cross sections when control rods are inserted in a calculation node. In order to represent asymmetric core blocks (e.g., fuel blocks or reflector blocks containing asymmetric absorber rods), surface-dependent discontinuity factors based on nodal equivalence theory have been introduced into the nodal diffusion theory option of the DIF3D code (DIF3D-nodal) to improve modeling accuracy. Additionally, the discontinuity factors based on the Simplified Equivalence Theory (SET) have been incorporated as an alternative and may be employed for both the DIF3D-nodal and DIF3D-VARIANT (nodal transport) solution options. Two- and three-dimensional core calculations have been performed using the routines developed and modified in this work, along with cross sections generated from single fuel block and one-dimensional or two-dimensional fuel-reflector model. Generally, REBUS-3/DIF3D results for the core multiplication factor and power distribution are found to be in good agreement with reference results (generated with MCNP continuous energy calculations) particularly when discontinuity factors are applied. The DIF3D-VARIANT option was found to provide a more accurate solution in its diffusion approximation than the DIF3D-nodal option. Control rod worths can be estimated with acceptably small errors compared to MCNP results. However, estimation of the core power tilt needs to be improved by introducing the surface-dependent discontinuity factor capability in DIF3D-VARIANT.

Lee, C. H.; Zhong, Z.; Taiwo, T.A.; Yang, W.S.; Khalil, H.S.; Smith, M.A.; Nuclear Engineering Division

2006-10-13

410

Academic Rigor: The Core of the Core

ERIC Educational Resources Information Center

Some educators see the Common Core State Standards as reason for stress, most recognize the positive possibilities associated with them and are willing to make the professional commitment to implementing them so that academic rigor for all students will increase. But business leaders, parents, and the authors of the Common Core are not the only…

Brunner, Judy

2013-01-01

411

An efficient computational technique for light water reactor core dynamics

By combining a modified version of the so-called ''adiabatic'' method for reactor dynamic calculations with a simplified flow redistribution scheme, an efficient method for predicting three-dimensional core behavior has been developed for pressurized water reactor transients. Both the simplified core reactivity and the flow redistribution calculations are shown to yield close approximations of the results obtained by more rigorous approaches.

C. D. Wu; J. Weisman

1988-01-01

412

NASA Astrophysics Data System (ADS)

The available oxygen potential data of LWR-fuels by the EFM-method have been reviewed and compared with thermodynamic data of equivalent simulated fuels and mixed oxide systems, combined with the analysis of lattice parameter data. Up to burn-ups of 70-80 GWd/tM the comparison confirmed traditional predictions anticipating the fuels to remain quasi stoichiometric along irradiation. However, recent predictions of a fuel with average burn-up around 100 GWd/tM becoming definitely hypostoichiometric were not confirmed. At average burn-ups around 80 GWd/tM and above, it is shown that the fuels tend to acquire progressively slightly hyperstoichiometric O/M ratios. The maximum derived O/M ratio for an average burn-up of 100 GWd/tM lies around 2.001 and 2.002. Though slight, the stoichiometry shift may have a measurable accelerating impact on fission gas diffusion and release.

Spino, J.; Peerani, P.

2008-03-01

413

The paper describes basic neutron-physics models developed in the Division of Advanced Nuclear Power Systems of the Institute of Nuclear Reactors, Russian Research Center Kurchatov Institute, as design models intended for calculating the characteristics of block fuel assemblies of a high-temperature gas-cooled reactor GT-MHR, namely, models for calculating burnup of fuel and isotopes of burnable neutron absorbers and calculating fuel assemblies at fixed points with respect to burnup with preparation of the neutron constants in a preassigned number of energy groups for full-scale design of a reactor. A model problem for investigation of calculated approximations is proposed. The outcome of this investigation is a developed stage-by-stage procedure of preparing group homogeneous cross sections of a fuel assembly and its parts that has been introduced into the practice of design calculations of a GT-MHR reactor.

Boyarinov, V. F., E-mail: Boyarinov@dhtp.kiae.ru; Fomichenko, P. A. [Russian Research Centre Kurchatov Institute (Russian Federation)

2011-12-15

414

Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod was analyzed by a fuel performance code FEMAXI-6. The code was developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using finite element method (FEM).During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed

Motoe Suzuki; Hiroshi Uetsuka; Hiroaki Saitou

2004-01-01

415

Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent fuel casks used for transportation and storage. Recently, analyses have demonstrated the technical feasibility and estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This report summarizes the extension of the previous PWR technical feasibility assessment to boiling water reactor (BWR) fuel. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. Two different aspects of fuel characterization were considered:l first, the generation of burn- up dependent material interaction probabilities; second, the prediction of material inventories over time (depletion). After characterizing the spent fuel at various stages of exposure and decay, three dimensional (3-D) models for an infinite array of assemblies and, in several cases, infinite arrays of assemblies in a typical shipping cask basket were analyzed. Results for assemblies without a basket provide reactivity control requirements as a function of burnup and decay, while results including the basket allow assessment of typical basket configurations to provide sufficient reactivity control for spent BWR fuel. Resulting basket worths and reactivity trends over time are then evaluated to determine whether burnup credit is needed and feasible in BWR applications.

Broadhead, B.L.

1991-08-01

416

A series of ex-reactor heating tests on low burnup U-26wt.%Pu-10wt.%Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel\\/cladding metallurgical interaction and its effect on cladding integrity at elevated temperatures. The reaction between the fuel and cladding caused liquid-phase formation and dissolution of the inner surface of the cladding. The rate of cladding penetration was below the existing

Hanchung Tsai; Yung Y. Liu; Da-Yung Wang; J. M. Kramer

1991-01-01

417

Current risk assessments of spent fuel in storage and transportation casks use the properties of LWR fuel below 45 GWd/MTU. Fuel is being driven to higher burnups that may influence the source term in cask accidents. To achieve these burnups the manufacturers are introducing new assembly designs and cladding alloys. As a result, at the higher burnups (? 50 GWd/MTU) some of the characteristics of the fuel pellets, cladding, and assembly design used in the safety analysis have changed. The fuel pellet has developed a fine grained, Pu rich rim zone on its exterior surface. The source term may increase by 1 – 3 orders of magnitude depending on the fracture characteristics of the rim. The cladding may acquire hydrogen contents up to 700 wppm during the increased exposure. Embrittlement with subsequent lose of ductility may occur, especially if there is hydride reorientation. As a result, there may be a greater propensity for fracture of the rods upon impact with subsequent release of fuel particulate and gas. Significantly improved source terms can be developed if additional data on fuel rim fracture as a function of impact energy, the dependence of cladding ductility for Zircaloy and the newer cladding alloys as a function of hydride reorientation, and release characteristics for fractured rods were obtained. CRUD spallation characteristics only make a significant contribution to the source term if the rods do not fracture in the accident or if a fire only accident occurs.

Einziger, Robert E.; Beyer, Carl E.

2007-08-01

418

Microstructure and fission gas bubbles in irradiated mixed carbide fuels at 2 to 11 a/o burnup

NASA Astrophysics Data System (ADS)

An analysis of the defect structure and of small fission gas bubbles has been performed on mixed carbide fuels with burn-ups between 1.8 and 11 a/o by transmission electron microscopy (TEM). A complex defect structure consisting of dislocations, loops and at least 3 types of solid fission product precipitates was observed. Na-bonded carbides develop predominantly a dislocation network increasing in density with burn-up whereas He-bonded carbides showed mainly a corresponding network of crystallographic needle precipitates. Locally the nucleation and growth of small fission gas bubbles with 1 to 20 nm diameters (bubble population P 1) is closely related to their dislocation or needle environment, larger bubbles with diameters 30 to 50 nm appear to be mostly associated with platelike precipitates or dislocation boundaries. The local swelling contribution ? 1 of bubble population P 1 is ? 0.5% and its fission gas content G 1 is 4 to 5% of the total amount of gas created over the whole burn-up range investigated.

Ray, I. L. F.; Blank, H.

1984-05-01

419

Neutron transport and diffusion theory space- and energy-dependent reactor kinetics calculations were performed in the frequency domain to determine the sensitivity of an ex-core neutron detector to in-core vibrations and coolant boiling in a PWR. Results of these calculations indicate that the ex-core detectors are sensitive to neutron sources, to vibrations, and to boiling occurring over large regions of the

F. J. Sw