Help:WebKORIGEN
From NucleonicaWiki
webKORIGEN: A New web-based KORIGEN Package for Nuclide Depletion Calculations in Nucleonica
(webKORIGEN module last updated 24 June 2008)
(webKORIGEN module updated 28 Feb. 2008)
H.W. Wiese, A. Schwenk-Ferrero
Forschungszentrum Karlsruhe Technik und Umwelt, Postfach 3640, 76021 Karlsruhe, Germany
mailto: Aleksandra.Schwenk-Ferrero@iket.fzk.de
Contents
|
The KORIGEN code and its nuclear data libraries
Origin and utilization
The KORIGEN code, developed in the Karlsruhe Research Centre, is a stand alone-package which serves to calculate the fuel depletion during irradiation(burn-up) and decay [1]. KORIGEN, originating from the ORNL ORIGEN code [2] can be used to determine some characteristics of spent nuclear fuel such as time-dependent nuclide masses, radioactivities, decay heat, radiation sources and radiotoxicities.
Fields of application
Industrial power plants as pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are charged with Uranium OXide and/or Mixed Plutonium OXide and Uranium Oxide fuel subassemblies. Under nominal operation conditions, the fuel undergoes depletion during in-core residence which is caused by fission processes generating fission products and by neutron capture processes transmuting U to Pu and Pu to Am, Cm and heavier transplutonium isotopes.The discharged spent fuel changes its composition with time due to radioactive decay and spontaneous fission processes, the decay being accompanied by strong release of heat. Reliable assessment of nuclide inventories and decay heat from actinides, fission products and activated fuel impurities is a key issue in the design of core cooling systems and evaluation of their performance as well as for the safe handling, reprocessing and storage of spent fuel. Moreover, integrated accident consequence analyses of nuclear facilities require additional information on the fuel elemental inventories and their isotopic break down as a function of time to allow the evaluation of eventual volatile releases and radiological dose to members of the public.
Calculation procedure
Isotopic summation codes like KORIGEN explicitly calculate time dependent nuclide concentrations by solving the underlying system of coupled linear differential burn-up and decay equations for a large set of isotopes. Principally, the equation coefficients depend on time since, during irradiation they are dynamic functions dependent on the already reached depletion level and the actual neutron flux density.This problem can be overcome by skillfully subdividing the total irradiation time into appropriately short time intervals, in which the system coefficients can be approximated well by constant average values. The number of equations is equal to the number of considered isotopes, i.e. in KORIGEN to about 1400 . Integral quantities such as the total, decay heat, n- and γ-emissions are calculated in the code by summing up the contributions from each individual radionuclide. Thus the application of summation codes like KORIGEN requires in particular complete and adequate nuclear data libraries available for fissionable nuclides, fission products and structural elements.
Associated data libraries
The accuracy of the prediction of time dependent fuel properties is determined for each nuclide by the ability to manage during burn-up and decay calculations accounting of its complete decay chain, comprising metastable levels, modes of decay (branchings) and half-lives for daughter nuclides, and the recoverable energy per decay (Q value) for each generated radionuclide. These data are stored in associated libraries and their accuracy warrants the performance of the code. KORIGEN is equipped with three basic ORFI data libraries containing the complete and adequate nuclear data available for actinides and their fission product yields, fission products and structural/activation materials. Three further basic ORFI libraries provide the code with photon emission spectra. Reactor-specific neutron reaction cross section data for important nuclides, especially in thermal reactors burn-up dependent for actinides , are gathered in dedicated KORI libraries. Data needed for calculating neutron emission from (alpha, n)-reactions are available in a file called ANDA.
Upgrades and updates
Over the past three decades, at Karlsruhe a considerable effort was made to upgrade and extend the nuclear data and analysis capabilities of the KORIGEN code mainly in the field of calculations for German power plants: PWRs and BWRs [1].
A special effort concerned the updating of neutron reaction cross-sections by generating dedicated KORIGEN libraries, specific to particular thermal system. The required one-group effective cross-sections were generated by coupling multigroup cell-calculations and burn-up calculations in a time dependent manner [3]. The strategy followed within one KORIGEN calculation run is to replace with these cross sections those present in the master libraries, by using burn-up dependent data for important actinides (235U to 244Cm) and burn-up averaged for about 70 fission product nuclides. To assure the quality assessment the updated cross section sets were extensively benchmarked against experiments [1].
In addition, the fission product yields from neutron-induced thermal and fast fissions for 19 actinides including 232Th, 233-236,238 U, 237,238 Np, 238-242 Pu, 241,242m,243 Am, 243,244,245 Cm were implemented in the code. As a new option, the determination of fission products from spontaneous fissions of 242,244Cm and 252Cf was enabled. The evaluations were done based on JEF-2.2 yields.The adoption of the new fission yields required decay data for new fission products not yet included in the basic library. The missing data were taken from JEF-2.2 and EAF-3.1 evaluated data.The photon library was revised as well using both JEF-2.2 discrete (line-energy) and continuous-energy emission spectra.
In concern with nuclear waste, the code was upgraded to calculate the neutron emission of vitrified high-active waste(HAW), taking account of neutrons from (alpha, n)-reactions of light nuclides and of spontaneous-fission neutrons.
Recent developments i.e. a further KORIGEN code extension and a new cross section library enabled the code to predict accurately the decay heat of advanced fuels. The KORIGEN results for transmutation (inert matrix ) fuels irradiated in fast spectrum were validated in the framework of a benchmark for European First Industrial Transmuter (EFIT)[5].
In brief summary, the major modifications and extensions of KORIGEN versus ORIGEN features including input and output aspects are:
- Input and output blocks in the original ORIGEN were restricted to 10 time steps which, in case of long irradiation histories, especially in multi-zone calculations, required a large number of blocks ( so called KORIGEN phases). The allowed number of time steps per such a phase was encreased to 36 for faster input preparation and simplified handling of output in webKORIGEN
- for comfortable evaluation, results are stored in a compact manner in a dedicated output file
-for a user friendly performance, both the interactive input pre-processor to KORIGEN POKI and the output manager were written.
WebKORIGEN a web-based version of KORIGEN driven by Nucleonica
The KORIGEN supervising staff at the Research Centre Karlsruhe together with the Nucleonica developers form the European Commission's Joint Centre have created an extended version of KORIGEN called webKORIGEN, driven by the EU-webportal Nucleonica. The objective was to facilitate the input preparation, pre-processing, running, post-processing and to offer a fast graphical output generation for webKORIGEN users. WebKORIGEN supports calculations for a set of standardized problems, trimmed to three major classes of nuclear plants whereas more complex and general tasks can be solved only with the KORIGEN original code.
The nuclear systems supported by webKORIGEN are: the thermal power plants deployed worldwide as Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR)and a future extension to the current industrial technology : the European Fast Reactor (EFR).
The user of Nucleonica can launch the application "webKORIGEN" , by activating the webKORIGEN option in Nucleonica. An easy-to-use WebKORIGEN MS WINDOWS GUI (Graphical User Interface)will be activated and menus, toolbars and online help that support the user while preparing the input file will be offered. The pre-processor execution run supplements the full-extend KORIGEN input with further problem dependent data necessary to perform the calculations.
In webKORIGEN four operational modes can be run. They constitute only a subset of full variety of KORIGEN features, dedicated data libraries and miscellaneous options.
To address efficienly the needs of Nucleonica community, webKORIGEN offers an analyses of four case studies. They require the both the diverse methodology and a different rapid input set up. Therefore in STEP1 the computation mode has to be specified. The modes are depicted as diagrams versus time: fuel power history versus irradiation time, decay heat or decreasing mass concentration versus decay time.
webKORIGEN input structure
STEP 1: selecting the computation mode
Four operational modes can be run:
Mode 1: Irradiation of fresh nuclear fuel or a target (one KORIGEN phase)
Mode 2: Decay of a single-nuclide radioactive material (one KORIGEN phase)
Mode 3: Mode 1 with subsequent period of the decay storage(two KORIGEN phases)
Mode 4: Mode 3 with subsequent reprocessing and decay storage of waste (three KORIGEN phases)
Each mode requires the both a diverse method of calculation and a different input parameter set up
STEP 2 : set up of mode-dependent input parameters
The first phase in mode 1, 3 and 4 is an irradiation phase, for which a fresh fuel, the facility in which it is going to be irradiated, and the irradiation history have to be set up.
Description of input parameters in STEP 2/Mode 1
Irradiation of reactor fuel
1. Facility: the multi-choice options comprise at present: PWR, BWR, THERMAL and EFR
The option THERMAL was added to permit an irradiation calculation with 2200m/s cross-sections for comparison with results from irradiation in light-water reactor neutron spectra(PWR, BWR)
2. Fuel type and composition: Present options UOX (UO2) and MOX (UO2+PuO2). For UOX fuel the initial enrichnment in weight-percent of 235U, UE, has to be specified, whereas for MOX the break-down of plutonium isotopes from 238Pu to 242Pu (plutonium composition) and in order to account for plutonium aging, the weight fraction of 241Am/Pu has to be input activating the "Define" button. For MOX, additionally the content of fissile plutonium in the initial heavy metal IHM=U+Pu: Pu fiss=(239Pu+241Pu)/IHM, and the enrichment of 235U in the carrier matrix,UCE, (depleted or natural uranium) is weight-percent are required.
Recommended aged plutonium vector compositions (238Pu, 239Pu, 240Pu, 241Pu, 242Pu, 241Am):
LWR (2.6, 50.5, 27.8,11.5, 7.6, 1.0)
EFR (2.0, 53.8, 26.0,10.1, 8.1, 2.0)
Ranges:
PWR: 3.8 <= UE <= 4.7 ; LWR: UCE=.72
BWR: 4.1 <= UE <= 4.3 ; EFR: UCE=0.25
PWR: 3.3 <= Pu fiss <= 4.2
BWR: 3.6 <= Pu fiss <= 3.8
These ranges correspont in the validity limits of restricted number of enrichment dependent cross section libraries available in webKORIGEN.
Out-of-range parameters will lead to non-reliable results!
3. Facility parameters related to irradiation history
Mass of the total initial heavy metal (IHM) [t]: IHM may refer to the total reactor core, a single batch, a fuel subassemly, or may be set up to 1t for upscaling.
Discharge burn-up (B) [MWth d/tIHM]; LWR: B<=50/60 dependent on enrichment, EFR: B<= 165
Specific power (P) [MWth/tIHM, P= constant during full power operation
LWR: 35 <= P <= 40, EFR: 80 <= P <= 90
Number of irradiation cycles (NCI); NCI <=6
Length of cycle (CL)[s,m,h,d or y]; 350d <= CL <=380d
Load factor(LF) [%], LF=cycle full power days/cycle length [d] x 100; 80 <= LF <= 95
These operational parameters are related by B = P x NCIxCLxLF/100, whereas CL is specified in days. The irradiation history is specified either by NCI, LF, CL, B or by NCI, LF, CL, P.
Irradiation of a target
Calculations of target irradiations are restricted to small samples of a single-nuclide material in larger facilities, in this way the neutron spectrum in the target is dominated by the facility spectrum, i.e. for the neutron cross sections sigma(targer) ~ sigma(facility) is valid. Since the depletion of target accompanied by the formation of daughter nuclides is mainly due to decay and neutron-induced reactions, the latter given by sigma x neutron flux density, the neutron föux density is here a parameter of choice. The input parameters are:
1.Type of initial nuclides (NUIR), example 239Pu, 209Bi (the latter for formation of 210PO)
2.Initial mass of the target (MIR) [g]
3.Facilities PWR, BWR, THERMAL or EFR
4.Neutron flux density (PHI) [n/s/(cm x cm)]; only PHI = constant in [0,TIR] is allowed
LWR: 2E14 <= PHI <=5E15, EFR: 1E15 <= PHI <= 5E15
5.Irradiation time (TIR) is determined by TIR=NCI x LF/100 x CL with LF=100% since anyway NCI and CL are available in this menu.
Decay storage after fuel irradiation
In modes 3 and 4, a decay phase follows the irradiation phase. The corresponding input parameters are:
Decay time(TDC) [s, m, h, d or y]
Number of equal-length time intervals(NDC) in [0,TDC]
Decay of a single-nuclide radioactive material
In mode2, the decay of a radioactive, single-nuclide material is performed and the inventories of daughter nuclides are assessed. The input parameters are
1. Name of nuclide(NUDC), e.q. Pu236; NUDC must not be stable
2. Amount of nuclide(RMDC) [g]
3. Decay time (TDC) [s,m,h,d or y]
4. Number of equal-length time intervals (NDC) [] in [0,TDC]
Reprocessing and decay of waste after fuel irradiation and pre-decay
The input parameters are:
1.Time of decay after reprocessing (TWD) [s,m,h,d or y]
2.Number of time intervals (NDW) in [0,TDW] time bins equidistant
time scale linear, if TDW <1000, else logarithmic
3.Elements to be reprocessed (ERP1, ERP2, ...) Example: U,Pu
4.Reprocessing losses (ERPL1, ERPL2,...) [weight-percent]
ERPLi % of isotopes of elements ERPi, will be contained in the waste, all other isotopes are considered 100% as waste
STEP 3: Display of input summary und start of the execution
The input data log (input reproduction) shown in this menu provides a printed record of the input as it has been accepted and processed by the pre-procesor. If non-default values have been used quick check-up of settings in respect with their compatibility to permitted parameter ranges should be done
Clicking the "Run" button starts the execution run of the case considered
WebKORIGEN output structure:
After case execution, the results can be viewed by pressing the “display results” button. If the KORIGEN log file is of interest (for reviewing input) “ display log button “ should be selected to browse the content of the log file.
The results are displayed in tables and/or graphs, the latter being activated by clicking the "Plot" buttons. In the right-up combo box, the case sensitive quantities to be displayed for can be selected.
Calculated quantities
Depending on the mode, for isotopes and elements the following quantities sre available:
• Masses [g]
• Radioactivities [Bq]
• Decay heat [W]
• Gamma heat [W]
• Inhalation toxicity [Sv]
• Ingestion toxicity [Sv]
Quantity dependent top-20 nuclides and elements
The output for isotopes and elements is restricted to top-20 main contributors to the total, e.q. in the isotopes activity table the first isotope contributes the most to the total radioactivity, the second isotope the second most etc. The nuclear properties are listed or plotted versus the irradiation or the ; the top-20, in generall dependent on time with regard to their major contribution, reffer to the end of irradiation or decay time. The top-20 depend on the considered quantity:masses, activities,etc.: the heaviest in general is not the most radioactive!
Displaying the top-20 goes stepwise: first the top1-to-10, then the top-11-to-20 (switch down-right)
Totals of each quantity
Furthermore, for each quantity a table is provided and a plot of the total sum over all contributing isotopes.
Last-phase output
To be easy to survey, in modes 1, 2 and 3, output will only be given from the last KORIGEN phase, which implicitely is fixed by the mode selected in Step 1. Thus, choose mode for obtaining in-core results, mode 3 for fuel or target properties after discharge etc. In mode 4, the output is provided for phases 3 and 4, i.e. for decay prior and after the reprocessing; TDW will be added to TDC. In mode 1, only masses and the infinite neutron multiplication factor k-inf are provided. In modes 2,3 and 4, additionally neutron and gamma emission spectra can be assessed. The spectra are presented as tables and can be plotted as bar diagrams.
Time at which the output is sampled
Output of results is sampled at the end of all time intervals of the output phase only if there are no more than 36 intervals in total. Otherwise the results will be given in a collapsed manner.
In mode 1, in addition the neutron multiplication factor k-inf versus time can be printed out and analyzed. Kinf is assessed in KORIGEN during the phase of fuel in-core residence. Corresponding Kinf values are plotted at the beginning of each consecutive time interval and represents a ratio of the number of neutron produced due to fission to the number of losses caused by neutron capture. As the medium is assumed infinite no leakage must be taken into account.
Case dependent, the neutron emission rate from (α,n)-reactions and spontaneous fission, broken down into contributions from actinides and fission product, may be output together with gamma emission rates. KORIGEN calculates the gamma power from the photon spectra by summing up the absolute gamma intensities (photon/decay) of the respective nuclide group i multiplied with the mean energy of group i (the spectra of the actinides are divided into 18 groups, that of the fission products and of light elements into 12 groups).
References
[1]U. Fischer, H.W. Wiese: Verbesserte konsistente Berechnung des nuklearen Inventars abgebrannter DWR-Brennstoffe auf der Basis von Zell-Abbrand-Verfahren mit KORIGEN, KfK 3014 (1983), ORNL-tr-5043
[2]M.J. Bell:ORIGEN - The Oak Ridge Isotope Generation and Depletion Code, ORNL-4628(1973)
A.G. Croff:ORIGEN2 - -A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code, ORNL-5621(1980)
[3]C.H.M. Broeders:Entwicklungsarbeiten für die neutronenphysikalische Auslegung von Fortschrittlichen Druckwasserreaktoren(FDWR) mit kompakten Dreiecksgittern in hexagonalen Brennelementen, KfK 5072 (1992)
[4]A. Schwenk-Ferrero, A. Rineiski, H.W. Wiese, W. Maschek: Validation of KORIGEN Decay Heat Assessment for EFIT Cores, JKT 2007 Karlsruhe (2007)
