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Ito, Hiroto; Kato, Daisuke*; Onizawa, Kunio; Shibata, Katsuyuki
JAEA-Data/Code 2006-001, 33 Pages, 2006/02
JAEA-Data-Code-2006-001.pdf:1.33MB
As a part of the aging and structural integrity research for LWR components, new probabilistic fracture mechanics (PFM) analysis code PASCAL-EC (PFM Analysis of Structural Components in Aging LWR - Erosion/-Corrosion) has been developed. This code evaluates the failure probability of an aged piping with a wall thinning by Monte Carlo method. PASCAL-EC treats the wall thinning due to erosion/corrosion or flow accelerated corrosion (FAC) in piping. The development of this code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent developments in the structural mechanics and aging researches. This report provides the user's manual and theoretical background of PASCAL-EC.
Shirai, Nobutoshi; Inano, Masatoshi
JAEA-Data/Code 2006-002, 5 Pages, 2006/03
JAEA-Data-Code-2006-002.pdf:0.49MB
This report describes basic data related to civil engineering, construction, operation, electric power, steam, chemical reagents and radioactive wastes of Tokai Reprocessing Plant for environmental burden evaluation.
Takeda, Seiji; Kanno, Mitsuhiro*; Sasaki, Toshihisa*; Minase, Naofumi*; Kimura, Hideo
JAEA-Data/Code 2006-003, 137 Pages, 2006/02
JAEA-Data-Code-2006-003.pdf:7.4MB
no abstracts in English
Tsutsumi, Hideaki*; Sugino, Hideharu*; Onizawa, Kunio; Mori, Kazunari*; Yamada, Hiroyuki*; Shibata, Katsuyuki; Ebisawa, Katsumi*
JAEA-Data/Code 2006-004, 167 Pages, 2006/03
JAEA-Data-Code-2006-004.pdf:6.41MB
EBISA (Equipment Base Isolation System Analysis) code evaluates the effectiveness of seismic isolation for the important components in the seismic safety, and consists of the three codes, probabilistic seismic hazard code (SHEAT), seismic dynamic response analysis code (RESP) and seismic failure probability and frequency evaluation code. In these codes, RESP code is used for the calculation of the dynamic response behavior of a nuclear component with seismic isolation devices. This report describes the overall explanation of EBISA, and user's guide of RESP code including the analysis function, input manual and sample problem.
Sugino, Hideharu*; Onizawa, Kunio
JAEA-Data/Code 2006-005, 43 Pages, 2006/03
JAEA-Data-Code-2006-005.pdf:1.52MB
Regression models of group delay time have been established by analyzing Earthquake Observation database obtained from Vertical Instrument Arrays at the Oarai Research and Development Center of Japan Atomic Energy Agency. It is known that the mean value and standard deviation of group delay time obtained by differentiating the Fourier phase spectrum show the center of gravity position and continuous time in earthquake time history. Therefore the regression model of the mean value and standard deviation of group delay time was made as function of the magnitude and the epicentral distance. This model includes the source and propagation of earthquake around the site. It is expected that the development of seismic motion prediction method considering the site's phase property by using this regression model.
Ando, Masaki; Iijima, Susumu*; Oigawa, Hiroyuki; Sakurai, Takeshi; Nemoto, Tatsuo*; Okajima, Shigeaki; Osugi, Toshitaka*; Ono, Akio; Hayasaka, Katsuhisa; Sodeyama, Hiroshi
JAEA-Data/Code 2006-006, 67 Pages, 2006/03
JAEA-Data-Code-2006-006.pdf:6.08MB
As a part of research and development of an advanced fueled fast reactor, we carried out benchmark experiments in the FCA-XVII-1 core with MOX simulating fuel to obtain reference data to be compared with those measured in the FCA-XVI-1 and XVI-2 cores simulating metallic fueled FBR. Following nuclear characteristics were measured in the experiments: Criticality, reaction rate ratio, sample reactivity worth, sodium void reactivity effect and 
U Doppler effect. Extra measurements were performed in modified FCA-XVII-1 cores to obtain experimental data for various reactor types: (1) Measurement of sodium void reactivity effect in various plutonium isotope compositions, (2) Measurement of sodium void reactivity effect in a core where axial blanket was replaced with a sodium layer and (3) Measurement of various nuclear characteristics in a nitride fuel region. This report describes methods and results of the above experiments and method of analysis.
Kureta, Masatoshi; Tamai, Hidesada; Liu, W.; Sato, Takashi; Watanabe, Hironori; Onuki, Akira; Akimoto, Hajime
JAEA-Data/Code 2006-007, 90 Pages, 2006/03
JAEA-Data-Code-2006-007.pdf:14.83MB
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests as one of essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor which aims to achieve a high breeding ratio and super high burn-up by innovative performance-up of water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition(BT)(Subjects:BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the base case test section. The thermal-hydraulic characteristics using the large scale test section were obtained for the critical power, the pressure drop and the wall heat transfer under a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Effects of local peaking factor on the critical power were also obtained.
Tochigi, Yoshikatsu; Sasamoto, Hiroshi; Shibata, Masahiro; Sato, Haruo; Yui, Mikazu
JAEA-Data/Code 2006-008, 16 Pages, 2006/03
JAEA-Data-Code-2006-008.pdf:1.15MB
The development of the database system has been started to manage with the generally used. The database system has been constructed based on datasheets of the effective diffusion coefficient of the nuclides in the rock matrix in order to be applied on the "H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan". In this document, the examination and expansion of the datasheet structure and the process of construction of the database system and conversion of all data existing on datasheets are described. As the first step of the development of the database, this database system and its data will continue to be updated and the interface will be revised to improve the availability. The developed database system is attached on the CD-ROM as the file format of Microsoft Access.
Ueta, Shohei; Izumiya, Toru; Umeda, Masayuki; Ishigaki, Yoshinobu; Ohashi, Jumpei*; Iyoku, Tatsuo
JAEA-Data/Code 2006-009, 129 Pages, 2006/03
JAEA-Data-Code-2006-009.pdf:12.35MB
The fabrication of the 2nd loading fuel for the HTTR has been started from October 2002 and completed successfully by March 2005. While the fabrication of the 2nd loading fuel for the HTTR, the fuel kernels, coated fuel particles and fuel compacts has inspected to compared with fuel fabrication specification. As the result of these inspections, we confirmed a good quality of the 2nd loading fuel for the HTTR as similarly the first loading fuel for the HTTR. This report describes fabrication data of fuel kernels, coated fuel particle and fuel compacts of HTTR 2nd loading fuel. The fabrication data of this report could be the basis of the study of fuel and fission product release behavior etc. on the HTTR operation, and contribute to post irradiation examinations of HTTR fuel in the future.
Lim, D. H.
JAEA-Data/Code 2006-010, 19 Pages, 2006/03
JAEA-Data-Code-2006-010.pdf:3.37MB
A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a heterogeneous flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canisters configuration for nuclide migration in a water-saturated HLW repository [1,2], while the dependency was assumed to be either independent [3-5] or perfectly dependent [6,7] in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results.
Suzuki, Yuji*; Kato, Tomoko; Makino, Hitoshi; Oi, Takao
JAEA-Data/Code 2006-011, 277 Pages, 2006/03
JAEA-Data-Code-2006-011.pdf:23.39MB
The project to establish a technical basis for HLW disposal in Japan (H12) has used "reference biosphere" methodology. The radiation dose which was an index as a basis of the safety assessment was calculated. (In fact, the flux-to-dose conversion factors were calculated as coefficient reduced flux of radionuclide caused by geological disposal to radiation dose in biosphere.) And then, ICRP has introduced the new system for radiological protection in the publication 60 published in 1990. New regulations in Japan have been issued in 2001 and it defines new dose conversion factors of ingestion and inhalation as effective dose coefficients based on the publication 60. And so, nuclides for biosphere assessment of HLW disposal, the flux-to-dose conversion factors for biosphere assessment were calculated based on the new regulations. On the other hands, recent safety assessment on a waste of geological disposal equivalent in radioactive waste containing transuranic (TRU) nuclide was carried out the safety assessment based on the biosphere assessment model constructed by the methodology other than the reference biosphere. "Co-disposal concept" disposed of by also juxtaposing TRU waste with geological disposal site of high-level radioactive waste, was examined the geological disposal of TRU waste, and the consistency of models and assessment system in the geological disposal of HLW and TRU waste would have been required. Moreover, the environmental conditions in future various biosphere based on climatic change are considered, the importance of discussion considering the variation of the biosphere system has been indicated. The data set for appropriate models that should be considered in geological disposal safety assessment of TRU waste was also set, and flux-to-dose conversion factors were calculated for the new regulation based on the same concept and models as HLW disposal.
CHayashi, Maki*; Satake, Kenji*; Sasamoto, Hiroshi
JAEA-Data/Code 2006-012, 25 Pages, 2006/03
JAEA-Data-Code-2006-012.pdf:2.2MB
An international interlaboratory study (ILS) was coordinated by Argonne National Laboratory (ANL) to evaluate the precision and bias of a single-pass flow-through (SPFT) test method that can be used to measure the forward dissolution rate of borosilicate glass. In this report we present the results of tests conducted at 70C using the LRM glass prepared by ANL. Based on measurements of the concentrations of glass components (i.e., Si, B and Na) in effuluent solutions and the solution flow rate, glass dissolution rates were calculated under the steady-state concentrations of glass components. The rates in the absence of glass components were estimated by extrapolation of the experimental results to zero concentration of glass components. Results show that the average pH of effuluent solutions was 11.68
0.23 (0.23 refers to standard deviation) at room temperature (21.784.03C, 4.03 means standard deviation). The estimated dissolution rate of LRM glass, based on variations in Si, B and Na concentrations in the absence of solution-feedback effects, is 2.11, 1.99 and 1.93 (g/m
d), respectively. The reliability of these estimates is questionable, however, due to considerable scatter in the aqueous concentration data. Based on these results, it is suggested that carefully controlled and constant flow rates are required to obtain reliable data using the SPFT test method.
Nakajima, Kunihiko*; Makino, Hitoshi
JAEA-Data/Code 2006-013, 31 Pages, 2006/06
JAEA-Data-Code-2006-013.pdf:3.63MB
In the safety assessment of geological disposal system, data uncertainty is inherent by the heterogeneity of the field in the natural geological environment and the change of the geological environment condition with the progress in the time and insufficient of understanding and information, and it is difficult to perfectly remove them. It is important to safety assessment that understanding influence of the uncertainty of parameter on the result of the Monte Carlo simulation considering data uncertainty. In our study, examining the range of application and the accuracy of the decision tree analysis technique by various approaches in the sensitivity analysis, and examining the object various results including the maximum value of all doses (others, the time of the maximum value of all doses, the identification of dominant nuclides). As a result, it was shown to be applicable for various purposes using the decision tree analysis technique to the sensitivity analysis.
Shiraishi, Akemi; Sekiguchi, Masato; Tachibana, Haruo; Yoshizawa, Michio; Komuro, Yuji*; Nemoto, Kiyoko*; Okawa, Ikuko*
JAEA-Data/Code 2006-014, 36 Pages, 2006/06
JAEA-Data-Code-2006-014.pdf:2.46MB
In Japan Atomic Energy Research Institute (JAERI), individual monitoring and dose data recording for radiation workers have been conducted since 1957, the next year of which JAERI was established. This report compiles the statistics of individual doses, such as average doses, collective doses, the number of radiation workers and dose distributions, over the past 48 years from 1957 to 2005, when JAERI merged with Japan Nuclear Cycle Development Institute into Japan Atomic Energy Agency. Transition of the statistics showed the history of radiation works in JAERI and many efforts for dose reduction based on the ALARA principle recommended by ICRP. In addition, it was found from the analysis of cumulative distributions that, in recent years, there was a specific work-group exposed to significantly high dose compared with other workers.
Nishizaki, Chihiro*; Endo, Akira; Takahashi, Fumiaki
JAEA-Data/Code 2006-015, 56 Pages, 2006/06
JAEA-Data-Code-2006-015.pdf:2.12MB
no abstracts in English
Tamai, Hidesada; Kureta, Masatoshi; Liu, W.; Sato, Takashi; Watanabe, Hironori; Onuki, Akira; Akimoto, Hajime
JAEA-Data/Code 2006-016, 134 Pages, 2006/11
JAEA-Data-Code-2006-016.pdf:49.7MB
Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT)(Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one.
Isogai, Takeshi*; Sasamoto, Hiroshi; Shibata, Masahiro
JAEA-Data/Code 2006-017, 37 Pages, 2006/07
JAEA-Data-Code-2006-017.pdf:4.39MB
We have been developping the technique for measuring the chemical composition of porewater using low-decolorant pH test papers and high-absorbancy pads embedded in compacted bentonite. Previously, we reported that the pH of porewater near the infiltration surface slightly decreased with time in compacted bentonite contacted with distilled water. Preliminary thermodynamic calculations have suggested that the porewater pH decrease near the infiltration surface observed in experiments using distilled water may be due to partial oxidation of trace amounts of pyrite in Kunigel V1. Such interpretation has not been verified experimentally by using Kunipia F which does not include pyrite as the accessory minerals in bentonite, however. Additionally, it is also suspected that the material used in the experiments (i.e., ceramic filter) and the experimental condition (i.e., bentonite contacts with the static solution) can affect the variations of porewater pH near the infiltration surface. Thus the following experiments with distilled water were conducted to identify the reason of temporal pH variations near the infiltration surface. (1) Experiments using Kunipia F, (2) Experiments using the alternative filter (i.e., combination of plastic filter with membrane filter), (3) Experiments using the alternative experimental condition (i.e., bentonite contacts with the stirred solution) Results indicate that the decrease of porewater pH near the infiltration surface is observed, although the Kunipia F is used. Difference of variations is not observed by using the alternative filter and experimental condition. Therefore, it would be suggested that the pH decrease near the infiltration surface may not be due to these factors but be due to another factor which affects the porewater pH variations.
Hyodo, Hideaki*; Tatsumi, Masahiro*
JAEA-Data/Code 2006-018, 120 Pages, 2006/08
JAEA-Data-Code-2006-018.pdf:8.42MB
In order to utilize the measured burnup data for improvement on accuracy in reactor core design, it is important to minimize the methodological errors to retrieve physical meanings from experimental data. The system for neutronics analyses that has been developed as the JUPITER standard analysis method assumes geometry of critical assembly, thus the system has not been maintained in functionalities for analysis of composition change of fuel materials. Therefore, there is a potential restriction for the purpose of detail analysis due to extreme inefficiency that comes from variety of limitation on its functionalities. It is not sufficient to follow a predefined analysis sequence in burnup analysis for reactor core; it is also needed to change analysis sequence to examine modeling error in analysis or to retrieve calculated values in an intermediate computation step for interpretation of physical meanings. Therefore it is not complete with a simple join of each function; it is needed to develop a new system for burnup analysis of reactor cores with flexibility on composition and decomposition of analysis components such as cell and core calculations. In this work, necessary conditions are examined for a new burnup analysis system targeted to actual reactor cores from the results of a research on the current working set in burnup analysis. With results in the research, a set of conceptual and fundamental designs were done.
Takahashi, Fumiaki
JAEA-Data/Code 2006-019, 83 Pages, 2006/09
JAEA-Data-Code-2006-019.pdf:5.77MB
In a criticality accident, a person near fissile material can receive extremely high dose which can cause acute health effect. For such a case, medical treatment should be carried out for the exposed person, according to severity of the exposure. Then, radiation dose should be rapidly assessed soon after an outbreak of an accident. Dose assessment based upon the quantity of induced 
Na in human body through neutron exposure is expected as one of useful dosimetry techniques in a criticality accident. A dose assessment program, called RADAPAS (RApid Dose Assessment Program from Activated Sodium in Criticality Accidents), was therefore developed to assess rapidly radiation dose to exposed persons from activity of induced Na. RADAPAS consists of two parts. One of them is a database, which contains compendiums of energy spectra and dose conversion coefficients from specific activity of Na induced in human body. The other part has execution files for dose calculation from specific activity of induced Na. Information for criticality configuration or characteristics of radiation in the accident field is to be interactively given with interface displays in the dose calculation. RADAPAS can rapidly derive radiation dose to the exposed person from the given information and measured Na specific activity by using the conversion coefficient in database. This report describes data for dose conversion and dose calculation in RADAPAS and explains how to use the program.
Osakabe, Kazuya; Kato, Daisuke*; Onizawa, Kunio; Shibata, Katsuyuki
JAEA-Data/Code 2006-020, 371 Pages, 2006/09
JAEA-Data-Code-2006-020.pdf:11.17MB
As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics analysis code PASCAL has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel under transient conditions such as pressurized thermal shock. PASCAL Ver.1 has functions of optimized sampling in stratified Monte Carlo simulation and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack and others. A generalized analysis method is proposed based on the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2.
Yamane, Yuichi; Sakai, Mikio*; Abe, Hitoshi; Yamamoto, Toshihiro*; Okuno, Hiroshi; Miyoshi, Yoshinori
JAEA-Data/Code 2006-021, 75 Pages, 2006/10
JAEA-Data-Code-2006-021.pdf:24.75MB
Propety data of MOX, Zinc Stearate, etc. were investigated and examined as part of the development for criticality accident evaluation method for MOX fuel fabrication facility. Property data gathered for the powders of MOX, UO
, Zinc Stearate, Tungsten and their mixture were density, specific heat, thermal conductivity and etc. as well as the data concerning fluidization or degree of mixing.
Nojiri, Naoki; Tochio, Daisuke; Hamamoto, Shimpei; Umeda, Masayuki; Fujimoto, Nozomu; Iyoku, Tatsuo; Takeda, Tetsuaki
JAEA-Data/Code 2006-022, 61 Pages, 2006/10
JAEA-Data-Code-2006-022.pdf:5.68MB
For the future HTGR development and the management of the High Temperature engineering Test Reactor (HTTR), the HTTR operation data base is constructed. The data base consists of the sorted or evaluated data based on the measured values from the HTTR's operation such as excess reactivity of the core, temperature at facilities of the core and the plant, impurity in coolant and so on. The data base also consists of some sub-databases which have objects related to the future HTGR development or the HTTR's operational management in order to manage the important operation data systematically on a long term. This paper describes the outline and structure of the HTTR operation data base. Also, as an example, some part of the HTTR common data-base, the HTTR nuclear characteristics data-base and the Helium purity control data-base are described.
Satoh, Daiki; Sato, Tatsuhiko; Shigyo, Nobuhiro*; Ishibashi, Kenji*
JAEA-Data/Code 2006-023, 43 Pages, 2006/11
JAEA-Data-Code-2006-023.pdf:1.76MB
The Monte Carlo based computer code SCINFUL-QMD has been developed to evaluate response function and detection efficiency of a liquid organic scintillator for neutrons from 0.1 MeV to 3 GeV. This code is a modified version of SCINFUL that was developed at Oak Ridge National Laboratory in 1988, to provide a calculated full response anticipated for neutron interactions in a scintillator. The upper limit of the applicable energy was extended from 80 MeV to 3 GeV by introducing the quantum molecular dynamics incorporated with the statistical decay model (QMD+SDM) in the high-energy nuclear reaction part. This report serves as not only introduction of the physical model and computational flow but also user manual of the code.
Masukawa, Fumihiro; Abe, Teruo*; Hayashi, Katsumi*; Handa, Hiroyuki*; Nakashima, Hiroshi
JAEA-Data/Code 2006-024, 98 Pages, 2006/11
JAEA-Data-Code-2006-024.pdf:7.9MB
We have developed SHINE3, simple code for high energy neutron skyshine dose evaluation, for the shielding designs of high energy accelerator facilities. This code uses the 4-parametric fitting function to the skyshine dose response by neutron and secondary 
-ray, which were predicted by PHITS, the general-purpose Particle and Heavy Ion Transport Monte Carlo code system. This code can be applied for skyshine dose evaluation of neutron up to 3 GeV in energy, and distance ranging between 10 and 2000 meters from source point, with equivalent accuracy as those by Monte Carlo Method.
Ichihara, Akira; Kunieda, Satoshi; Iwamoto, Osamu
JAEA-Data/Code 2006-025, 51 Pages, 2006/12
JAEA-Data-Code-2006-025.pdf:2.53MB
A computer code, POD-S, was developed for the nuclear data evaluations. Energy spectra of the reactions (n,), (n,n'), (n,p),(n,
), (n,d), (n,t), (n,
He), angular distributions of the neutron elastic- and inelastic-scattering, and their integrated cross sections are calculated within the statistical model. The computational methods and input parameters are explained in this report, with sample inputs and outputs.
Tsukui, Rota*; Nishiki, Tsukasa*; Higashinaka, Motonori*; Tsu, Nobuhiro*
JAEA-Data/Code 2006-026, 103 Pages, 2007/02
JAEA-Data-Code-2006-026.pdf:16.81MB
For the 1st phase of Horonobe Underground Research Laboratory Project, Japan Atomic Energy Agency, Geological Isolation Research and Development Directorate, Horonobe Underground Research Unit (before October 1, 2005: Japan Nuclear Cycle Development Institute Horonobe Underground Research Center) was carried out following three geophysical surveys in FY2004. *Seismic Reflection Survey *Multi-Offset VSP Survey *Gravity Survey / This document presents the outline and results of these geophysical surveys.
Saegusa, Jun
JAEA-Data/Code 2006-027, 79 Pages, 2007/02
JAEA-Data-Code-2006-027.pdf:11.37MB
Radioactivity measurements for volume samples are usually based on the -ray spectroscopy method, and the measurements involve the efficiency calibration for a radioactivity measuring instrument. A novel method which enables precise efficiency calibrations with a standard point source has been proposed, and a calculation code for implementing the method has been newly developed. The code is named CREPT-MCNP (Calibration Code for the Representative Point Calibration Method with MCNP). The code finds out the position of the representative point which is intrinsic to each shape of volume sample and gives the self-absorption factors to make correction on the efficiencies measured at the representative point with a standard point source. It can deal with -rays between 20 keV and 2 MeV with p- or n-type germanium semiconductor detectors. This report describes features of CREPT-MCNP code, accompanied by its user's manual and validation examples.