Transient reactor analysis code

TRAC (Transient Reactor Analysis Code)-PWR is an advanced best-estimate systems code designed primarily for analysis of large-break LOCAs (LB-LOCAs) in PWRs, although its versatility allows for the analysis of a wide range of scenarios. This code was developed at Los Alamos National Laboratory under sponsorship of the US Nuclear Regulatory Commission (USNRC). 

The analysis has been conducted using a transient reactor analysis code (TRAC) coupled with the heat transfer models for steam condensation with the presence of a non-condensable gas in the IC condenser tube. The non-condensable gas, which initially fills the PCV space, will be absorbed into the IC condenser tubes through the broken feedwater pipe. 

TRAC-B Transient Reactor Analysis Code-Boiling water reactor... 

Because it is based on well-established LWR technology and implements benchmarked thermal hydraulic safety analysis codes, and because each module is a relatively small reactor, it is possible that all of the information required for design certification may be obtained using a full-scale demonstration of a single module. Of significant interest for design certification would be the impact of neutronic feedback on flow stability, particularly during plant transients. 

VIPRE-01 is a finite-volume sub-channel analysis code enable of three-dimensional modelling of reactor cores and other similar geometries in steady-state and transient conditions. VIPRE-01 calculates the detailed steady-state and operational transient core flow distributions, coolant conditions, fuel rod temperature and departure from nucleate boiling ratio (DNBR). 

The natural recourse for this situation is to use numerical solution methods for both steady-state and transient analyses. The systems-analysis codes that are used for design and safety studies can all handle the coupled NCL case. At the same time, local special- and general-purpose natural-circulation models, methods, and codes are being developed for applications to Gen IV reactor concepts. 

The configuration related to the CEFR reactor protection system is also introduced in this simulation. The various rupture positions and all the transient parameters of the primary loop, as well as the accident sequence are obtained with the help of the system safety analysis code OASIS. This code is based on the original French code OASIS, modified to simulate the CEFR. 

One-dimensional bum-up and two-dimensional R-Z core calculation procedures for the fast reactor are found in refs. [108,109]. The LOCA analysis code, SCRELA was described in ref. [110]. The procedure for a simplified PSA is also described there. The single channel thermal-hydrauUc calculation code SPROD, the two-dimensional coupled core calculation scheme of the thermal spectram reactor with water rods and transient and accident analysis code at supercritical-pressure, SPRAT are described in ref. [111]. 

Analysis of the heat transfer deterioration mechanism by numerical simulation using the k-s turbulence model is in ref. [112]. Transient and accident analysis code for fast reactors, SPRAT-F, and calculation of the Oka-Koshizuka heat transfer correlation for the safety analysis at supercritical pressure are described in ref. [113]. 

The Light Water Reactor Fuel Analysis Code FEMAXI-6 developed by researchers at JAEA is used for the fuel rod analysis [35]. It is capable of obtaining a complete coupled solution of the thermal analysis and mechanical analysis, enabling an accurate prediction of pellet-clad gap size and pellet-clad mechanical interaction (PCMI) in high bumup fuel rods not only in normal operation but also in transient conditions. It is based on a deterministic method and the main features of its calculation models are as follows ... 

From the beginning of the conceptual study on supercritical water cooled reactors, several plant transient analysis codes have been developed, modified, and applied to them [1-9]. The general name of these codes is Supercritical Pressure Reactor Accident and Transient analysis code (SPRAT). SPRAT mainly calculates mass and energy conservations, fuel rod heat conduction, and point kinetics. The relation among these calculations is shown in Fig. 4.1. SPRAT can deal with flow, pressure, and reactivity induced transients and accidents at supercritical pressure. The flow chart is shown in Fig. 4.2. 

The plant transient analysis code SPRAT-F, introduced in Sect. 7.9, is used to calculate the flow distribution and the MCST in the three hot channels with respect to the core power and feedwater flow rate. The nodahzation is shown again in Fig. 7.81 [32]. Since the core power and feedwater flow rate are raised very slowly at the power raising phase, the reactor can be practically treated as in a steady state... 

The analysis of transient flows is necessary for safety analysis of nuclear reactors. Such efforts usually result in the development of large computer codes (e.g., RELAP-5, RETRAN, COBRA, TRAC). Rather than going into the details of such codes, this section gives the principles and basic models involved in the analysis. 

This Safety Report is intended for use in the performance of safety analyses of nuclear power plants both under construction and in operation. While focusing on the performance of the reactor and its systems, including the accident localization system (ALS), during transients and accidents, this Safety Report takes account of best estimate analysis and conservative analysis. The application of best estimate codes that use well grounded acceptance criteria and conservative input data provides a more reasonable assessment of the safety margins in various situations. Adequate conservatism in input data is normally achieved by setting the parameter values at the worst boundary of the range of deviations allowed by the technical specifications of the nuclear power plant [7]. 

Sodium release to RGB under CD A has been estimated at about 1.5 t, based on the approach followed for FFTF reactor. The important input to this analysis are the transient and quasi-static pressure of the sodium after slug impact beneath the top shield and the fraction of the sodium mass in the reactor assembly which has potential to get ejected. These parameters are obtained from the detailed fast transient fluid-structure interaction analysis using an in-house computer code called FUSTIN. A preliminary estimate is also made on the transient pressure and temperature rise in the RGB for the 1.5 t of sodium release and the values are 30 kPa and 80 K respectively. 

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. 

Analysis of the design basis and beyond design basis accidents for NPPs based on a floating power unit with the KLT-20 reactor is being performed using a set of calculation codes developed by OKBM and proven in calculations of stationary and transient modes of ice-breaker reactor operation. 

The TWINKLE code is used to predict the kinetic behaviour of a reactor for transients that cause a major perturbation in the spatial neutron flux distribution. TWINKLE was used in the analysis performed in support of the Sizewell B PCSR (Reference 5.7). There is therefore a high degree of confidence that an acceptable verification statement can be made in the context of the UK regulatory regime. 

The results of the analysis demonstrate that the peak reactor coolant system pressure reached during the transient is less than that which causes stresses to exceed the faulted condition stress limits of the ASME Code, Section III. Also, the peak cladding surface temperature is considerably less than 1482°C. These results represent the most limiting conditions with respect to the locked rotor event or the pump shaft break. With the reactor tripped, a stable plant condition is eventually attained. Normal plant shutdown may then proceed. 

The relief capacities of the pressuriser safety valve is determined from the postulated overpressure transient conditions in conjunction with the action of the reactor protection system. An overpressure protection report is prepared according to Article NB-7300 of Section III of the ASME code. Reference 6.2 describes the analytical model used in the analysis ofthe overpressure protection system and the basis for its validity. 

The modeling is based on adaptation of the equations in the previous section to the coupled loops case. The model equations developed herein will be written for the case of single-phase flow in the primary and secondary loops. Both steady-state and off-normal transient conditions in the Gen IV nuclear reactor case involve two-phase fluid states. Safety-grade analyses of design and off-normal states will generally be handled by systems-analysis models and codes that easily accommodate generalized geometry, fluid states, and flow directions. 

The main pressure pipe break accident analysis was conducted using the French OASIS Code, which is a dynamic system simulation program especially for the pool type sodium cooled FBR. It can simulate the thermal-hydraulics of the whole FBR plant circuits and reactor control and protection system, including the regulation system. So OASIS is a good simulation tool to study the operation and accident transients in a FBR Plant. An introduction and the physical models of the OASIS code are presented in the Ref. [1]. 

The SSC-K is based on the methods and models of SSC-L [5], which was originally developed to analyze loop-type liquid metal reactor transients. Because of the inherent difference between the pool and loop designs, major modification to the SSC-L has been made for the analysis of the thermal hydraulic behaviour within a pool-type reactor. Now, the SSC-K code has the capability to analyze both, loop and pool type liquid metal cooled reactors. 

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