Extension of fuel performance code package FEMAXI; Development and validation of reactivity-initiated accident analysis module RANNS for light water reactor fuels

Tasaki, Yudai; Udagawa, Yutaka

JAEA-Data/Code 2024-012, 76 Pages, 2024/12

JAEA-Data-Code-2024-012.pdf:9.25MB

Japan Atomic Energy Agency (JAEA) has been developing a fuel performance code, FEMAXI, to evaluate the behavior of LWR fuels under normal operation and transient conditions. In March 2019, FEMAXI-8, the first systematically validated and performance evaluated code, was released. Since then, the code has undergone various improvements. In parallel, since the 2000s, JAEA has been developing the RANNS module as a branch for design basis accident (DBA) analysis, with a particular emphasis on computational stability, so that fuel behavior can be tracked even for very steep transients, in this case mainly reactivity-initiated accidents (RIAs). The specific models include boiling heat transfer, fission gas release by grain boundary failure, and cladding failure determination based on fracture mechanics parameters, which are essential for predicting such transient behavior. In this report, prior to the release of RANNS, we present a description of the models for accident behavior analysis, the relationship with FEMAXI-8 in terms of the design and structure of the program, and the results of a large-scale validation using the extensive database of RIA experiments conducted and accumulated by JAEA, to evaluate the overall RIA analysis performance. The code will be made available to users as a packaged FEMAXI/RANNS, enabling them to analyze fuel behavior under various conditions. The model parameter sets determined through the above validation analyses are also presented in this report, and by referring to them, the analysis can be easily performed with almost no change in usability from the previously released FEMAXI-8.

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

https://doi.org/10.1080/00223131.2019.1637795

no abstracts in English

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Analysis on pellet-clad mechanical interaction process of high-burnup PWR fuel rods by RANNS code in Reactivity-initiated accident conditions

Suzuki, Motoe; Fuketa, Toyoshi; Saito, Hiroaki*

Nuclear Technology, 155(3), p.282 - 292, 2006/09

https://doi.org/10.13182/NT06-A3762

Experimental analyses were performed for the RIA-simulated tests, OI-10 and OI-11 of high burnup PWR rods, in the NSRR by the RANNS code. The rod conditions were calculated by the fuel performance code FEMAXI-6 following the actual power history from the beginning to the end of irradiation in PWR and the results were given to the RANNS code as pre-test conditions. The RANNS analysis was conducted on the basis of such test conditions in the NSRR as the pre-test conditions, pulse power enthalpy and coolant temperature. The predicted quantities such as temperature of pellet stack and cladding, stress-strain distribution in cladding, and interactions among them during pulse irradiation were discussed in terms of PCMI process and compared with the experimental observations. In the OI-10 rod, calculated cladding permanent strain has a reasonable agreement with strain profile obtained in PIE, while locally enhanced strain of cladding was pointed out. In the OI-11 rod, the process from crack initiation to split failure was accounted for by the plastic strain occurrence in cladding.

RANNS code analysis on the local mechanical conditions of cladding of high burnup fuel rods under PCMI in RIA-simulated experiments in NSRR

Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10

The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.

Modification of ring tensile test for LWR fuel cladding

Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi

Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.912 - 932, 2005/10

High burnup fuel cladding can fail due to mechanical interaction with expanding fuel pellet under reactivity initiated accident (RIA) conditions. In order to evaluate the cladding failure limit, investigations to modify ring tensile test have been performed to measure mechanical properties of Zircaloy cladding properly. JAERI developed the test method and geometry minimizing undesirable effects of friction and bending moment in the specimen. Using the modified test method, mechanical properties of unirradiated Zircaloy-4 cladding were evaluated as functions of hydrogen concentration and temperature. For hydrogen concentrations above 700 ppm, obvious increase of ductility is observed with the temperature increase from 300 to 473 K. For hydrogen concentrations below 500 ppm, on the other hand, temperature dependence of ductility is relatively small in the present temperature range from 300 to 573 K.

RIA-simulating experiments on high burnup PWR fuel rods with advanced cladding alloys

Sugiyama, Tomoyuki; Fuketa, Toyoshi; Ozawa, Masaaki*; Nagase, Fumihisa

Proceedings of 2004 International Meeting on LWR Fuel Performance, p.544 - 550, 2004/09

Two pulse irradiation experiments simulating reactivity initiated accidents were performed on high burnup (60 GWd/t) PWR UO rods with advanced cladding alloys. Test OI-10 was performed on an MDA cladded rod with large-grain (25 m) fuel pellets with a peak fuel enthalpy condition of 435 J/g, and resulted in a peak residual hoop strain of 0.7%. On the other hand, Test OI-11 on a ZIRLO cladded rod with conventional pellets resulted in a fuel failure at a fuel enthalpy of 500 J/g due to the pellet-cladding mechanical interaction (PCMI). A long axial split was generated on the cladding over the active length. The fuel pellets were fragmented and dispersed into the coolant water. The fuel enthalpy at failure is higher than the PCMI failure criterion of 209 J/g at the corresponding burnup. The experimental results suggest that the rods with improved corrosion resistance have much safety margin against the PCMI failure compared to the conventional Zircaloy-4 rod.

Light water reactor fuel analysis code FEMAXI-6, 1; Detailed structure and user's manual

Suzuki, Motoe; Saito, Hiroaki*

JAERI-Data/Code 2003-019, 423 Pages, 2003/12

JAERI-Data-Code-2003-019.pdf:17.7MB

A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version with a number of improvements. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, permitting an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and the coupling with burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code is a versatile tool not only in the normal operation but also in transient conditions. This report describes the design, basic theory, models and numerical method, improvements, and model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output, and a sample output in an actual form are included.

NSRR tests under simulated power oscillation conditions of BWRs

Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi

HPR-359, Vol.2, p.34_1 - 34_16, 2002/09

In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.

Phenomenon identification and ranking tables (PIRTs) for rod ejection accidents in pressurized water reactors containing high burnup fuel

Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Deveney, R. C.*; Dunn, B. M.*; Fuketa, Toyoshi; Higar, K. E.*; Hochreiter, L. E.*; Langenbuch, S.*; et al.

NUREG/CR-6742, 263 Pages, 2001/09

no abstracts in English