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Large-break loss-of-collant accident analysis of a direct-cycle supercritical-pressure light water reactor
Abstract
Large-break loss-of-coolant accident (LOCA) was analyzed in the course of the design study concerning direct-cycle supercritical-pressure light water reactor (SCLWR). The advantages of SCLWR are a higher thermal efficiency and simpler reactor system than the current light water reactors (LWRs). A computer code was prepared for the analysis of the blowdown phase from the supercritical pressure. The calculation was connected to the REFLA-TRAC code when the system pressure decreased to around atmospheric pressure. The analyzed accidents are 100, 75, 50 and 25% cold-leg and 100% hot-leg breaks. First, blowdown and heatup phases without an emergency core cooling system (ECCS) were evaluated. A low-pressure coolant injection system (LPCI) was designed to fill the core with water before the cladding (stainless-steel) temperature reached a limit of 1260°C. The LPCI consists of four units, each of which has the capacity 805 kg/s. An automatic depressurization system (ADS) was designed to release the steam generated in the core in the case of cold-leg breaks and to permit operation of LPCI in the case of LOCAs of less than 100% break. For all cases analyzed, the peak cladding temperatures were lower than the limit when the designed ECCS is implemented.
... The safety systems of the reference plant are mainly based on active systems. In Figure 2-9 the layout of the reference plant with the safety systems is represented [Koshi94a], [Lee98a]. It consists of the automatic de-pressurization system (ADS), the low pressure coolant injection system (LPCI), the turbine-driven auxiliary feed-water system (AFS), steam dump valves, etc. ...
... The blow-down phase of a large break located in the steam line (0.22 m²) close to the RPV is analyzed with RELAP5. The following assumptions were considered according to [Koshi94a], [Lee94a], [Lee98a]: Scram is caused once the feedwater mass flow rate reduces below 90 % of its nominal value. ...
... Tab. 2-1 Comparison of main operational parameters of different reactor types 10 Tab. 2-2 ECCS of the reference reactor design[Oka00b],[Koshi94a],[Lee98a] 12 Tab. 3-1 Available RELAP5-versions with different steam tables Differences between the inventory in scenario A and B at steady state conditions 46 ...
... 3) The core which had been analyzed by Koshizuka et al. was used for the validation calculation. 6) The modified SPRAT-DOWN is also compared with these 2 codes. The calculations of the SPRAT-DOWN and the SCRELA start at 25 MPa while the REFLA-TRAC calculation starts at 17 MPa because the REFLA-TRAC can not calculate supercritical pressure condition. ...
LOCA analysis of the supercritical-pressure light water cooled thermal reactor with downward-flow water rods (called Super LWR) are carried out to clarify its characteristics. 1–100% hot/cold leg breaks are analyzed. At the cold-leg large break, excessive core heat-up is mitigated by the automatic depressurization system (ADS) during blowdown because reactor depressurization induces core coolant flow. The coolant inventory in the top dome and the water rods is effectively used for core cooling. After blowdown, the core is slowly re-flooded by the low-pressure ECCS like PWR. The highest cladding temperature of the large LOCA is lower than the criterion (1,260°C) by about 430°C which appears during the reflooding phase. Small break of the cold-leg gives the higher cladding temperatures than that of the large break because the ADS are not actuated in the analysis. The highest cladding temperature is lower than the criterion by about 260°C. If the ADS actuation is assumed by the “drywell pressure high” signal, the cladding temperature is lower. The hot-leg break is less severe than the cold-leg break because it increases the core coolant flow rate and forced flooding is expected after blowdown.
Supercritical Water-Cooled Reactors (SCWRs) conception has been developing as promising advanced nuclear systems in recent decades around the globe because of its many desirable features. It has many potential advantages include high thermal efficiency (~45% in comparison to ~33% efficiency of present Light Water Reactors), relative simplicity in plant construction with high power density, which eventually would decrease the investment expenses so that the reactor economy will be improved. SCWR is the only direct upgrade version of the Generation III water-cooled reactor using water as a coolant material amongst the six reactor types being studied by the GEN-IV International Forum (GIF). Generally, SCWR design is broadly categorized as (i) pressure-vessel type and (ii) pressure-tube type. Japan first suggested the pressure-vessel type concept and then by the Euratom partnership. The pressure tube idea was proposed by Canada first, and therefore it is named Canadian SCWR. These two concepts have several similar characteristics; outer pressure and temperatures, selection of steam cycle, features of materials, and heat transfer. In the present work, a critical analysis of existing SCWR conceptual designs was performed. An assessment of these land-based designs and a review of current models of nuclear submarine and ship were conducted. Also, considering the operating conditions and performance requirements of performance for ships, an optimal design of SCWR for ships was proposed. So far, pressurized water-saturated steam nuclear power plants have been demonstrated to be reliable marine power plants. Further technology and design improvements have been made to make LWR plants competitive for a large number of marine applications. The conceptual design of SCWR for ship application, which is preliminarily discussed in this paper, could represent potential progress in marine life technology. This advancement is supposed to result from a significant reduction in plant size, weight, and hopefully, capital and operating costs. The confirmation of this design necessitates extensive analysis and experiments.
Within the 5th Framework Programme of the European Commission a project for studying the potential of water reactors using supercritical water/steam conditions was performed. Among the different activities performed within this project emphasis was also set on the major characteristics of a plant configuration. Since the HPLWR has to be considered as a long term development project, the requirements applied for the design are expected to be a combination of existing ones (like EUR) and such discussed for future designs (like Generation IV reactors). This means among others that in such a design, passive means and means for mitigation of severe accidents have to be incorporated. Concerning the primary system the principles of the core configuration and arrangement were outlined as a first conceptual approach. Concerning the fuel assemblies both hexagonal and quadratic types are analyzed. Regarding the containment design, the proposal is to consider a modern BWR containment and in order to introduce passive features into the design, the containment should be provided with a core flooding pool and emergency condensers. The safety systems configuration is intended to provide a passive core flooding pool to replace the amount of water existing in the primary circuits of LWR's. This provision since in case of the HPLWR the mass of water within the RPV is about 1/10 of the primary system water of a BWR or PWR only. Therefore, in case of transients like "loss of offsite power", the safety philosophy of such reactors to maintain the primary pressure and consequently use the heat capacity of the existing water is not appropriate in case of an HPLWR. Instead of using a high pressure feedwater injection it is proposed to initiate the automatic depressurization system (ADS) and use the existing low pressure injection system and the passive flooding system. With this proposed containment and safety concept it is considered that an extensive use of passive systems is reached. A preliminary economic evaluation of the HPLWR plant concept which was performed, indicates that such nuclear plants may have the potential to reach electricity generation costs which can significantly increase the economic advantages of nuclear plants compared with fossil fueled plants. This statement especially holds true if one considers in the future a further increase in fossil fuel prices and the possibility of getting deeper insight into the design of HPLWR's in order to explore the potential of cost reduction more precisely and in more detail.
CATHARE is the French best estimate code used for safety analysis of Light Water Reactors. Two-phase flow is modelled with a two-fluid six-equation model in 0-D, 1-D and 3-D modules. A few months ago, the range of applications was limited to the sub-critical two-phase region, but thanks to the extension of the water and steam properties up to 26 MPa and recent modifications of the standard code version, it is now possible to simulate transients where both supercritical and sub-critical regimes are encountered. This paper describes these recent developments in the modelling of supercritical water flows using the two-fluid model implemented into the CATHARE code and shows the results of the preliminary assessment of this model successfully achieved by running several calculations for 1-D applications: - Steady state in supercritical conditions followed by blow-down - Stationary isenthalpic expansion - Stationary heat-up of supercritical fluid Then, using the modified CATHARE version, first steady-state and transient calculations of a Supercritical Light Water Reactor operating above the critical point were carried out, among them, Loss of Coolant Accidents (LOCAs) considering simple sequences of events. The predicted transient behaviours seem to be correct from supercritical to sub-critical regimes, which confirms the CATHARE capabilities.
Core designs and safety analyses of the Super LWR and Super FR are described in Sects. 2.1 and 2.2. The single-pass core of the Super LWR adopts the fuel assembly with large water rods having a thermal insulator. The upper core structure allows for simplified refueling procedures like light water reactors (LWRs). The single-pass core of a Super FR adopts the blanket fuel assembly with mixed oxide fuel rods in the lower part. Safety characteristics at abnormal transients, accidents and anticipated transients without scram (ATWS) of the Super LWR and Super FR are described. The total loss of flow accident and loss of coolant accidents (LOCAs) are important. Transient sub-channel analysis predicts a lower fuel cladding temperature for accidents and abnormal transients than the single-channel model for the Super FR where peaking is small. Transient subchannel analysis is described in Sect. 2.3.
New rod-type spacer was developed and is described in Sect. 2.4. Transmutation of long life radioactive fission products (LLFPs) is studied from the viewpoints of environmental risk and human risk. The high breeding core of the Super FR is developed with the tightly packed fuel rod fuel assembly. They are described in Sects. 2.5 and 2.6 respectively. Nuclear calculation of the fast and thermal neutron coupled core is described in Sect. 2.7.
The author of this chapter except Sect. 2.4 is Yoshiaki Oka with assistance of Qingjie Liu and Sutanto. Shinichi Morooka is the author of Sect. 2.4.
Safety characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with upward flow core cooling in two pass is investigated for the abnormal transients and accidents at supercritical pressure. Upward flow cooling has advantage of simplifying the upper core structure in comparison with the downward flow scheme that part of the coolant flows downward in the blanket fuel assemblies from the top dome of reactor pressure vessel. It also has advantages that flow stagnation does not occur at loss of coolant flow events due to the buoyancy of the coolant. The coolant flow scheme of this design is the all blanket fuel assemblies and part of the seed fuel assemblies are cooled with upward flow first, the coolant flows radially above the core and flow downward in the gap between the core and the shroud to the lower mixing plenum and cools the rest of seed fuel assemblies with upward flow till the upper mixing plenum before core outlet. To evaluate the safety performance, eleven transients and four accidents at supercritical-pressure are analyzed. Safety analysis results show that the safety criteria are satisfied with large margins for all the selected transients and accidents. But in the total loss of coolant flow accident the MCST (maximum cladding surface temperature) is still high. Because of this flow scheme, it is found that the MCST is sensitive to the volume of the gap between two pass. Actuating depressurization valves with low flow single at total loss of flow events is effective to induce flow for once-through SCWR and therefore improves safety performance.
Anticipated transient without scram (ATWS) of the supercritical-pressure light water-cooled fast reactor (Super FR) with single flow pass core is analyzed to clarify its safety characteristics. Abnormal reactivity due to loss of feedwater heating and uncontrolled withdrawal of a control rod is mitigated by density and Doppler reactivity feedback. Increase of the coolant density at the upstream of the fuel channel leads the core flow rate to decrease which in turn decreases the power especially due to the density reactivity feedback. Meanwhile, ATWS accidents which are followed by trip of both RCPs are important. Decreased feedwater flow rate leads directly to a decrease of core flow rate in all fuel assemblies. If the ADS is not used, the MCST increases rapidly due to high power density feature of the Super FR. Coolant expansion effect does not appear in the single flow pass core of Super FR while it appears in the second pass of the two-flow pass core leading to mild increase of MCST of the second pass. However, signal of flow low level 3 (6%) is provided for the ADS actuation and the MCST of each fuel channel is effective to be mitigated due to large core coolant flow rate induced by the ADS operation. The power is also decreased by density and Doppler reactivity feedbacks. MCST and core pressure of all the analyzed events satisfy the criteria.
The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST.
The supercritical water cooled fast reactor with all-upward two-pass flow scheme has been designed for simplifying the upper core structure. It is also possible to avoid the stagnation effect of the downward flow channel in loss of coolant flow events. To evaluate the safety performance and correspondingly improve the core and safety design for current flow scheme with different characteristics, safety analysis with regard to accidents and abnormal transients has been conducted. Safety analysis results show that the safety criteria are satisfied with large margins for all the selected accidents and transients. Furthermore, current all-upward two-pass flow core design with a downward gap space exhibits unique flow behavior of feedwater-dominant first-pass flow and valve-dominant second-pass flow because of the large volume of gap space between the first and second pass. Actuating depressurization valves at total loss of flow events is effective to induce flow for once-through supercritical water cooled reactor (SCWR) and therefore improves safety performance.
Total loss of flow accident characteristics of the supercritical-pressure light water-cooled fast reactor (Super FR) with all upward flow core cooling in two paths is investigated at supercritical pressure. Upward flow cooling has advantage that flow stagnation does not occur at total loss of coolant flow accident due to the buoyancy of the coolant. It is found that maximum cladding surface temperature (MCST) is sensitive to the volume of the gap between two paths the. By open the automatic depressurization system (ADS), depressurization will induce flow in the RPV, and it can improve the MCST in this accident.
A conceptual design of a direct-cycle supercritical-pressure light water reactor with water rods is presented. Three types of water rods are analyzed: single, semi-double and full-double tubes. A water rod replaces seven fuel rods in a triangular lattice. The coolant density change in the water rods and the fuel channel is calculated using a code developed in the present study. The full double tube is most superior in terms of the distribution of the moderator. The number of fuel rods to water rods is 198:19, which makes optimum moderation. The average enrichment becomes 4.13%. The axial power flattening is finally achieved by partial length fuel rods and enrichment split of 0.25%. The discharge burnup is 45 GWd/t.
Dynamic performance studies in a channel model of a supercritical-pressure, direct-cycle light water reactor is carried out in this paper. In this supercritical-pressure system where some parameters are temperature dependent within the channel, neutronic and thermal-hydraulic calculations are carried out to define the distributions of heat transfer coefficient and power generated along the supercritical channel model. A linearized mathematical nodal model representing the heat transfer system is given for the supercritical channel. A digital-analogue simulation technique is used to study the dynamic performance of the channel model from the transient response of the fuel and coolant temperatures at different nodes in the supercritical channel model. Transient responses are displayed for different modes of perturbation of the system.
Core design of a direct-cycle, supercritical-pressure, light water reactor (SCLWR) is carried out. Double tube water rods are used for enhancing the moderation. Whole coolant flows upward in the inner tube and downward in the outer tube of the water rod, and then the fuel is cooled. The double tube water rods flatten the axial power distribution easier than the single tube water rods. Since the hot coolant from the fuel channel does not mix with the cold coolant from the water rod and keeps high temperature, the thermal efficiency is not deteriorated. Whole coolant flows through the fuel channel and the mass flow rate is high. The critical heat flux, hence, increases and the enthalpy difference between the inlet and the outlet becomes large. The core height decreases to 3.7m from the core using the single tube water rod of 5.7m. The axial power distribution was further flattened by changing the gadolinia concentration and the fuel enrichment. Clustered control rods like PWR are adopted as the primary reactivity control system. Borated water injection system is adopted as the backup system. Each system attains cold shutdown independently. The designed plant has thermal efficiency of 40.8%, gross electric power of 1,006 MW.
The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP.
The conceptual design of a direct-cycle fast breeder reactor (FBR) core cooled by supercritical water is carried out as a step toward a low-cost FBR plant. The supercritical water does not exhibit change of phase. The turbines are directly driven by the core outlet coolant. In comparison with a boiling water reactor (BWR), the recirculation systems, steam separators, and dryers are eliminated. The reactor system is much simpler than the conventional steam-cooled FBRs, which adopted Loeffler boilers and complicated coolant loops for generating steam and separating it from water. Negative complete and partial coolant void reactivity are provided without much deterioration in the breeding performances by inserting thin zirconium-hydride layers between the seeds and blankets in a radially heterogeneous core. The net electric power is 1245 MW (electric). The estimated compound system doubling time is 25 yr. The discharge burnup is 77.7 GWd/t, and the refueling period is 15 months with a 73% load factor. The thermal efficiency is high (41.5%), an improvement of 24% relative to a BWR's. The pressure vessel is not thick at 30.3 cm.
The supercritical-pressure light water cooled reactors aim at cost reduction by system simplification and higher thermal efficiency, and have flexibility for the fuel cycle due to technical feasibility for various neutron spectrum reactors. Since loss-of-coolant accident (LOCA) behavior at supercritical pressure conditions cannot be analyzed with the existing codes for the current light water reactors, a LOCA analysis code for the supercritical-pressure light water cooled reactor is developed in this study. This code, which is named SCRELA, is composed of two parts: the blowdown and reflood analysis modules. The blowdown analysis module is designed based on the homogeneous equilibrium model. The reflood analysis module is modeled by the thermal equilibrium relative velocity model. SCRELA is validated by the REFLA-TRAC code, which is developed in the Japan Atomic Energy Research Institute based on TRAC-PF1. A large break LOCA of the thermal neutron spectrum reactor (SCLWR) is analyzed by SCRELA. The result shows that the peak clad temperature (PCT) is nearly 980°C about 60s after the break and the PCT position is quenched at 170s This means that PCT is sufficiently lower than the safety limit of 1260°C. In conclusion, the developed code shows the safety of SCLWR under the large break LOCA, and is expected to be applied to LOCA analysis of other types of the supercritical-pressure light water cooled reactors.
Design studies of supercritical-pressure light-water-cooled reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is cooled by supercritical water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedwater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.
A high temperature reactor cooled and moderated by supercritical light water (SCLWR-H) is designed for assessing its technical feasibility and potential economic improvement. The plant system is the once-through direct cycle. Whole core coolant flows to the turbine. When adopting the conventional ascending flow type water rods, the cold coolant flowing up in the water rod is mixed with that of the fuel channel coolant above the core. Although the maximum surface temperature of the Ni-base alloy cladding is 620°C, the average core outlet temperature is low, 455°C. It, however, reaches 508°C in the core with the descending flow type water rods since no mixing of the coolant occurs above the core. The fuel enrichment is also decreased. The flow rate of each fuel assembly is controlled by an orifice to maximize the outlet coolant temperature. The reactor achieves high thermal efficiency of 44.0% and electric power of 1610 MW. The coolant flow rate per generated electricity is 24% lower than that of an Advanced Boiling Water Reactor (ABWR). This will reduce the size of the balance of plant (BOP) system. The burnup reactivity can be compensated by changing coolant flow rate ratio in the descending flow type water rods to the whole core through the fuel cycle. This reduces the number of control rods. The inlet and outlet coolant temperature and total flow rate need not be changed for keeping the power constant.
The concepts of a direct-cycle supercritical-pressure, light-water-cooled reactors were developed. The conceptual design of the thermal and fast reactors was carried out. The plant system is identical among them. Breeding is possible in the tight lattice core. Waste problems are improved due to the high thermal efficiency. Power generation cost will be greatly reduced from the current LAIR.
Conceptual design studies of advanced power reactors are summarized. The concept of power reactors cooled by supercritical water was developed. The coolant system is once through type like a supercritical fossil-fired power plant. It attains higher thermal efficiency and simpler reactor system than the light water reactors. Breeding is possible in the tight lattice core. The coolant void reactivity of the fast reactors becomes negative by placing a thin zirconium hydride layer between seed and blanket. The basic safety principles follows those of LWRs. Heterogeneous cores of the liquid metal cooled fast breeder reactor was designed for assessing the effectiveness of coolant orificing. Overpower transients after withdrawing a peripheral control rod were analysed by a three-dimensional space-time kinetics code which was developed for the study. The space-time power variations were more influential on boiling behavior in the heterogeneous cores. Conceptual design of the fusion-fission hybrid reactors and the inertial confinement fusion reactor were described for comparing with the fission power systems. The differences in heat generation profiles and technical features are discussed.
The concept of a direct cycle light water reactor (LWR) operating at supercritical pressure is presented. It is attractive for improving the thermal efficiency of LWRs and for simplifying the reactor system(1). The critical pressure of water is 22.1 MPa (221 bars). The density of water changes continuously above it and the concept of boiling does not exist. The system is conceptually depicted in Fig. 1. The water coolant which flows into the core is heated up and can be directly fed to turbines. Compared with the current BWRs, the recirculation system and steam separators and dryers will be eliminated. The vessel size will be greatly reduced. This will reduce the capital cost. The thermal efficiency will be also revolutionally improved. The feasibility of the system was studied through the conceptual design of the core.The axial temperature and density distribution of the coolant in the core were calculated at 25 MPa (250 bars) using a single channel model of a fuel and coolant. The density decreases continuously from the inlet (0.725 g/cm3) to the outlet (0.137 g/cm3) and the temperature increases from 583 K (310°C) to 689 K (416°C). The heat is efficiently removed due to the high specific heat of water around 658 K (385°C). The density change is mild at 250 bars than 225 bars which is close to the critical pressure. It is desirable to choose 250 bars for the stability of the flow in the core.
A reflood model for two fluid model code was developed referring to physical model of the REFLA code to improve the predictability of the TRAC-PF1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes. (1) the void fraction correlation in saturated two-phase and transition flow regions, (2) the drag coefficient correlation acting to drops, (3) the correlation for the heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation were selected to be implemented into the TRAC-PF1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. It was confirmed by the assessment study for various reflood tests that the core thermal hydraulic behaviors during the reflood in a PWR LOCA can be predicted excellently with the modified TRAC-PF1 code.
A reflood model for two fluid model code was developed referring to physical model of the REFLA code to improve the predictability of the TRAC-PF1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes, (1) the void fraction correlation in saturated two-phase and transition flow regions, (2) the drag coefficient correlation acting to drops, (3) the correlation for the heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation were selected to be implemented into the TRAC-PF1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. It was confirmed by the assessment study for various reflood tests that the core thermal hydraulic behaviors during the reflood in a PWR LOCA can be predicted excellently with the modified TRAC-PF1 code.