Hydraulics codes

Passive safety features for the MHR include ceramic, coated-particle fuel and an annular graphite core with high heat capacity and low power density. Recently, INL has used the ATHENA thermal hydraulic code to model the response of the MHR during loss-of-flow and loss-of-coolant accidents and has confirmed these passivity safety features work to maintain fuel temperatures well below failure thresholds [8]. 

Kelly JM, Stewart CW, Cuta JM (1992) VIPRE-02 - A two-fluid thermal-hydraulics code for reactor core and vessel analysis Mathematical modeling and solution methods. Nucear Technology 100 246-259... 

Although the pebble bed neutronics/thermal hydraulics code VSOP of KFA Jfllich is available through the NEA Databank now, ECN decided to develop it s own code system. The dynamic neutronics code PANTHER has been acquired from AEA, UK. This code is being coupled to the thermal hydraulics code THERMIX-DIREKT, kindly delivered to ECN by KFA Jfllich in a cooperation framework. 

After a familiarization phase, the Dutch HTR program is well under way. First decisions have been made on design features for the HTR conceptual design in The Netherlands. The activities will be concentrated at the development of a small size pebble bed HTR for combined heat and power production with a closed cycle gas turbine. An independent neutronics and thermal hydraulics code system is being developed. An optimized design based on the GHR-20 and the PAP-GT is plaimed for 1996. 

Calculation tools of this kind are very useful to the designer or to the overall system analyst (even if they leave the true specialists of the branch rather puzzled), as they allow the study of many cases and for transient times as long as are desired. It has been observed, with reference to the Three MUe Island accident, that if the time length of the calculated transients had been prolonged beyond the intervention time of the safety systems, the adopted thermal-hydraulic codes (RELAP and so on) could have shown the danger of getting to a situation where the pressurizer is substantially full of liquid while the reactor vessel is nearly empty. As it is known, this situation may cause the operators to erroneously think that all of the primary system is full and therefore make them shut off the safety injection systems. In fact, the calculations performed were stopped precisely at the moment of their intervention. 

To perform analyses described above the advanced thermal hydraulic code ATHLET was applied, this code is used at the NRI since 1993. The code has been verified for the WWER accident analyses, in the first place by computations performed by the code-developer organization — GRS. 

Event-tree and fault-tree models are built on the basis of results from physical models as implemented in a thermal-hydraulics or integral code. For instance, calculations with a deterministic thermal-hydraulics code provide information on the minimum requirements for safety and emergency functions, i.e. on the minimum number n of safety systems needed to cope... 

Moreover, the calculations with a thermal-hydraulics code provide information on the time windows available to perform required human actions. This information is used to determine human error probabilities, i.e. the probabilities that the required tasks can be performed within the available time windows. 

For some aspects of model uncertainties, well-known quantification methods are available. A Bayesian approach might be practicable, for instance, to quantify the uncertainty on the probability model to apply for the stochastic failure behaviour of system components. Monte Carlo analysis might be appropriate to quantify the uncertainty resulting from the application of thermal-hydraulics codes. 

OECD NUCLEAR ENERGY AGENCY, Report of Uncertainty Methods Study for Advanced Best Estimate Thermal Hydraulic Code Apphcations, Rep. NEA/ CSNI/R(97)35,2 Vols, OECD/NEA, Paris (1998). 

The computer codes which are used to carry out the anticipated operational occurrences and DBA analysis should be properly verified and validated. This includes the codes used to predict the behaviour of the reactor core, thermal-hydraulic codes and the radiological release and consequence codes. In addition, the analysts and users of the codes should be suitably qualified, experienced and trained. 

Thermal-hydraulic codes which model the behaviour of the reactor core and the associated coolant systems during normal operation and following accidents,... 

As shown in Table 4.2, large break LOCA events involve the most physical phenomena and, therefore, require the most extensive analysis methods and tools. Typically, 3D reactor space-time kinetics physics calculation of the power transient is coupled with a system thermal hydraulics code to predict the response of the heat transport circuit, individual channel thermal-hydraulic behavior, and the transient power distribution in the fuel. Detailed analysis of fuel channel behavior is required to characterize fuel heat-up, thermochemical heat generation and hydrogen production, and possible pressure tube deformation by thermal creep strain mechanisms. Pressure tubes can deform into contact with the calandria tubes, in which case the heat transfer from the outside of the calandria tube is of interest. This analysis requires a calculation of moderator circulation and local temperatures, which are obtained from computational fluid dynamics (CFD) codes. A further level of analysis detail provides estimates of fuel sheath temperatures, fuel failures, and fission product releases. These are inputs to containment, thermal-hydraulic, and related fission product transport calculations to determine how much activity leaks outside containment. Finally, the dispersion and dilution of this material before it reaches the public is evaluated by an atmospheric dispersion/public dose calculation. The public dose is the end point of the calculation. 

TRACE (TRAC/RELAP Advanced Computational Engine) is recent thermal-hydraulics code designed to consolidate and extend the capabilities of safety codes— TRAC-P, TRAC-B, and RELAP. It is intended for analysis of large- and small-break LOCAs and system transients in both PWRs and BWRs. The code has the capability to model thermal-hydraulic phenomena in both ID and 3D spaces. 

Validation of power conversion unit operational performance (physical model for validation of primary thermal-hydraulic code)... 

Neutronic and thermo-hydraulic code development and reactor conceptual design. 

This work evaluated the CHF for LABGENE through experimental design and by simulating the test section under reactor operational conditions through COBRAIIIc/MIT-1 code (Jackson Todreas 1981). COBRA is a thermal-hydraulic code widely used to evaluate the core of PWRs by... 

The development of methods of calculation for thermal-hydraulic design has proceeded in close association with the work on nuclear design methods. The complete system of digital codes now available for steady state performance analysis is known as PATRIARCH (see Pig. 1). Descriptions of the functions of the reactor physics codes appear in a companion paper. The approaches used in the thermal-hydraulic codes will be reviewed briefly below. In the space available, it is only possible to outline the basis of the more important codes. 

It is evident from the preceding discussion that there must be Iterations between the nuclear and thermal-hydraulic codes of the PATRIARCH scheme to achieve consistent solutions. The thermal codes need power distributions which are in turn dependent on the coolant density distribution. 

The methods used include validated thermal-hydraulic codes (see Section 16.10), system and component reliability analysis, and PRAs. 

Krepper, E., Prasser, H.-M., 1999. Natural circulation experiments at the ISB-VVER integral test facility and calculations using the thermal-hydraulic code ATHLET. Nuclear Technology 128, 75-86. 

The analyzed reactor core consists of 336 high-efficiency reentrant fuel channels. The inlet temperature of the coolant is 350°C at a pressure of 25 MPa, and the outlet temperature is 625°C. As a conservative approach, the thermal power corresponding to a fuel channel with the maximum thermal power was used in order to calculate the fuel centerline and sheath temperatures with the use of a one-dimensional thermal-hydraulic code. The temperature variation of the fuel hottest element in the radial direction is shown in Fig. 18.23. The maximum fuel centerline temperature of the UO2 fuel reaches 2196°C in the hottest fuel element of a fuel channel with a maximum thermal power of 10.23 MWn,. The temperature profiles of the coolant and the cladding (ie, CLaDding Temperature (CLDT)), as well as the Heat Transfer Coefficient (HTC) are shown in Fig. 18.24. 

In Phebus, unlike in a reactor, most of the pipe walls are heated to a fixed temperature so the thermal hydraulic codes do not have to calculate them. The first exception is the part of the vertical line just above the bundle erit which is expected to be at such a high temperatiu e that heater elements placed there woidd burn out. 

Clearly haring the wall temperatures defined makes the job easier for the thermal hydraulics codes. Furthermore, in FPTO, the coolant flow will always be single phase vapour. Originally, as we shall see, it hadbeenproposedto allow condensationinthesteam generator but t scenario has now been postponed to a later test. 

During this exerdse it was realized that none of the major thermal hydraulic codes had adequate models for wall condensation in the presence of hydrogen. This was one factor in the postponement of a cold steam generator till a later experiment. It is thou t that advantage will be taken of such a delay to develop adequate models and to check them on a PHEBUS scale 1 companion fadlity. 

There was a large scatter in the thermal hydraulic code results. This was due partly to the codes themselves. Another reason was that the boundary conditions used by the code runners were not always exactly the same. 

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