TRAC-PF1/MOD1 calculations and data comparisons for mist small-break LOCA, feed-and-bleed and steam-generator tube-rupture experiments

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (B&W) nuclear plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 × 4 [two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps] representation of B&W lowered-loop reactor systems. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from the MIST facility.Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for three transients run in the MIST facility. The energy removal and depressurization mechanisms in these tests are identified and the phenomena occurring in these tests compared. The tests analyzed are MIST Test 3109AA, the nominal small-break LOCA, Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. TRAC was able to predict these phenomena although the timing and magnitude of events were not always in good agreement.The MIST test have demonstrated the thermal-hydraulic phenomena expected to occur during transients in B&W nuclear plants. Because of scaling atypicalities, test results cannot be extrapolated directly to plant conditions. Although the phenomena were demonstrated in the MIST tests, there may be differences in the timing, magnitude and sequences of events in plant transients. Assessment calculations, three of which are presented here, have shown that the TRAC computer code can predict the major trends and phenomena occurring during the MIST tests with reasonable qualitative agreement. This includes complex sequences of events. Reasonable qualitative agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. These assessment results, taken with assessment results from other facilities at a wide range of scales, provide us with confidence that the TRAC code can adequately simulate the transient phenomena possible in B&W nuclear plants.     

A phenomenological model of the thermal hydraulics of convective boiling during the quenching of hot rod bundles. Part III: Model assessment using Winfrith steady-state, post-CHF, void-fraction and heat-transfer measurements and Berkeley transient-reflood test data

Alter completion and assessment of the thermal-hydraulic model developed in companion papers, we performed further independent assessment calculations using recently available Winfrith steady-state, post-CHF, void-fraction and heat-transfer data, and Berkeley transient-reflood test data. This paper discusses the results of those calculations.The thermal-hydraulic model in the Transient Reactor Analysis Code, Modification 2 (TRAC/PF1-MOD2, or simply TRAC), was used to predict the axial variation of void fraction as measured in Winfrith post-CHF tests. The predictions for reflood calculations were reasonable. The model correctly predicted the trends in void fraction due to the effect of pressure and power, with the effect of pressure being more apparent than that of heat flux. The effects of pressure, test-section power, and flow rate on the axial variation of tube wall temperature are predicted reasonably well for a large range of operating parameters.The predicted precursory cooling rates in Berkeley transient-reflood tests were in reasonably good agreement with the measured data. For high-test section input powers, oscillations associated with the void-fraction predictions observed. The predicted quench-front velocities (rewetting velocities) and their variations along the test tube were also found to be in reasonably good agreement with measured data.     

Experimental Study on Difference in Reflood Core Heat Transfer among CCTF,FLECHT-SET and Predicted with FLECHT Correlation

It has been reported that the core heat transfer coefficients measured in the CCTF tests, which were conducted under the conditions expected to appear during the refiooding period in a PWR, can not be predicted well with the FLECHT correlation, which has been used in the safety evaluation. In order to investigate the reason for this, a CCTF test was conducted under the typical FLECHT-SET experimental conditions. Investigating results from both tests, the following has been clarified:

The FLECHT correlation can not describe the heat transfer for the refiooding situations with the initial Accumulator injection period, which is expected to appear in a PWR, and gives much lower values than the measured. The core heat transfer in the FLECHT-SET is similar to that in the CCTF, and they are well predicted with the Murao-Sugimoto correlation. When there is some core radial power distribution, which strongly affects the heat transfer in a large vScale core, the heat transfer coefficients in the CCTF can be well predicted with taking account of this effect in addition to the Murao-Sugimoto correlation.     

Tritium Separation by CO2 Laser Multiphoton Dissociation of Trifluoromethane

Integral system tests with the Cylindrical Core Test Facility (CCTF) were performed to investigate the effect of the initial clad temperature on the reflood phenomena in a PWRLOCA. The initial peak clad temperatures in these three tests were 871, 968 and 1,047 K, respectively. The feedback of the system on the core inlet mass flow rate was estimated to be little influenced by the variation of the initial clad temperature except for the first 20 s in the transient. The observed temperature rise from the reflood initiation was lower with the higher initial clad temperature. This qualitatively agreed with the results of the small scale forced feed reflood experiments. However, the magnitude of the temperature rise in CCTF was significantly low due to the high initial core inlet mass flow rate. Also observed were the multi-dimensional thermal behaviors for the three cases in the CCTF wide core. The analysis codes REFLA and TRAC reasonably predicted the effect of the initial clad temperature on the core thermo-hydraulics under the simulated core inlet flow conditions. However, the calculated temperature rise of the maximum powered rod based on the one-dimensional core analysis was higher than that of the average powered rod, which contradicts the tendency observed in CCTF tests.     

Scaling effects predicted by COBRA/TRAC for UPI plant best estimate LOCA

The COBRA/TRAC-MOD7A, Rev. 1 code is currently licensed for best estimate LOCA analyses of 3 and 4 loop PWRs. As part of a licensing effort to extend the code application to plants equipped with upper plenum injection (UPI), scaling effects predicted by the code are investigated. The scaling effects of UPI tests were obtained through data analyses and summarized in Damerell and Simons (1993). The scaling subjects are: breakthrough flow area, downflow rate into the core, hot leg water carryover, and liquid level in upper plenum. The test facilities that supplied the data include UPTF and CCTF. In this report, the scaling trend is obtained from COBRA/TRAC analyses of CCTF Run 72 and Run 76 (scaling factor 0.091), UPTF Tests 20A, 20B, and 20C (scaling factor 2.1), and a typical UPI plant (scaling factor 1.0). The predicted scaling trend is found to agree well with the test data.     

Application of Simplified Condensation Model to PWR LBLOCA Transient Analysis with TRAC-PF1 Code

In the system analyses of a large-break loss-of-coolant accident (LBLOCA) of pressurized water reactors (PWRs), the TRAC-PF1 code predicted an unrealistic depressurization and required much computational time due to the problem of the condensation model. To eliminate the unrealistic depressurization, the TRAC-PF1 code was improved using a simplified condensation model that determined the total condensation rate at cold leg. Through the assessment calculations for CCTF, UPTF and LOFT tests, it was confirmed that the simplified model could eliminate the unrealistic depressurization and reduce the computational time. It was also confirmed that the model could improve the accuracy of the system calculation for the core inlet flow rate and clad temperature as the result of the elimination of the unrealistic depressurization. It has been concluded that the simplified condensation model is useful for the system calculation of the PWR LBLOCA.     

Application of the COBRA/TRAC best-estimate methodology to the AP600 large-break LOCA analysis

The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently United States Nuclear Regulatory Commission (USNRC)-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the COBRA/TRAC code to model the AP600 unique features was validated against cylindrical core test facility (CCTF) and upper plenum test facility (UPTF) downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models, as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95th percentile peak cladding temperature (PCT^95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.     

含弯曲棒的全长燃料棒束沸腾传热CFD计算分析

采用计算流体动力学（CFD）分析方法模拟了含一根弯曲燃料棒（简称“弯曲棒”）的5×5全长燃料棒束内的沸腾传热现象。基于欧拉两流体模型和改进的壁面沸腾模型进行计算,并基于压水堆子通道和棒束实验( PSBT )基准题中的试验数据对计算方法进行了验证,计算所得截面平均空泡份额与试验数据吻合良好,说明了现有计算方法的可靠性。基于计算结果考察了弯曲棒对棒束通道内流场、温度场、空泡份额等关键参数的影响。研究结果表明,弯曲棒的存在对截面横向流动、流体温度、空泡份额等均未产生显著影响,但弯曲棒表面温度增加,气泡也易发生聚集,增加了发生临界热流密度（CHF）的风险。      

CFD analysis on subcooled boiling phenomena in PWR coolant channel

Eulerian two-fluid model coupled with wall boiling model was employed to calculate the three dimensional flow field and heat transfer characteristics in a hot channel with vaned spacer grid in PWR. The heat transfer from pellet-gap-cladding to coolant was also taken into account by a system coupled code MpCCI. The wall boiling model utilized in this study was validated by Bartolomei experiment data, and a good agreement can be observed. By solving the governing equation in a two-way coupled method, the distribution of temperature in the pellet-gap-cladding region and the distribution of temperature, void fraction and velocity of two-phase flow in coolant channel can be obtained. The influences of spacer grid and mixing vane on the thermal-hydraulic characteristics were analyzed. The heat transfer capacity was strongly improved by the spacer grid and mixing vane, while the flow resistance was also enlarged. Localized volume fraction of vapor phase decreased due to mixing vane, which will decrease the possibility of the departure from nucleate boiling (DNB) and increase the critical heat flux (CHF). By analyzing the temperature and void fraction at cladding outer surface, the critical regions where hot spot may occur were determined.     

基于两相CFD的非均匀加热圆管CHF预测方法研究

为建立非均匀加热工况临界热流密度（CHF）预测方法，以对换热系统的安全分析提供新的辅助手段，本研究采用欧拉两流体模型和壁面沸腾模型，对非均匀加热圆管的CHF进行预测。通过数值计算得到不同热流密度下近壁面空泡份额和壁面温度的分布，将壁面温度出现二次峰值和此时近壁面空泡份额的峰值位置分别作为CHF发生的依据和CHF发生的点，并用此方法对2种不同功率分布圆管的CHF进行研究。研究结果表明，预测得到临界时的平均热流密度及临界发生的位置都与实验结果符合较好。因此，本研究建立的数值预测方法能够用于非均匀加热圆管CHF的预测。      

Quantitative Evaluation of Heat Transfer Enhancement due to Radial Power Distribution during Reflood Phase of PWR-LOCA

Experimental studies using Slab Core Test Facility (SCTF) and Cylindrical Core Test Facility (CCTF) indicated that the degree of heat transfer enhancement due to the radial power distribution during the reflood phase of a PWR-LOCA was governed mainly by the radial power ratio itself and less dependent on the shape of radial power distribution within the maximum power ratio of 1.36. The experimental condition covering the wide ranges of the reflood phase and the scale of core radius from 1/4.6 to 1/1 had little effect on the two- dimensional heat transfer behavior. The heat transfer coefficient under nonuniform radial power distribution was expressed as a sum of the heat transfer coefficient obtained under a complete mixing condition and an additional value given by an empirical correlation based on the SCTF results. The temperature rise at the peak power rod calculated with this expression tended to be lower than that calculated with the complete mixing model used in a reflood analysis code REFLA. That is, the complete mixing model was proved to give a conservative result under a nonuniform radial power distribution condition in a full size core.     

200MW供热堆余热排出过程的分析

２００ＭＷ供热堆采用自然循环的余热排出系统，具有非能安全的特点，程序ＴＲＡＣ－ＰＦ１采用了带有不凝气体场的两相二流体可非平衡态的流体力学模型，被用于余热排出系统的分析，但是，供热堆系统和压水堆核电站不完全相同，在将这部程序用于供热堆分析时，做了一些修改和补充，例如补充了流体横掠冲帽管束的传热计算式和阻力损失计算式等等，分析结果表明：自然循环的余热排出系统能够保证供热堆的停堆安全。     

Vapor generation model for flashing in the initial blowdown phase

A vapor generation model for flashing in the initial blowdown phase is proposed based on a wall nucleation theory and a bubble transport model. Comparisons are made between the proposed model and the TRAC-PF1 model by using the MINCS code through analyses of three blowdown experiments with different scales. The present model well predicts the pressure undershoot in the vessel, while the TRAC model can not predict this typical thermodynamic nonequilibrium phenomenon.     

Application of REFLA Core Model to Safety Evaluation Code

In the analysis of the core thermal-hydraulic behavior during the reflood phase of a PWR-LOCA, current safety evaluation codes like WREM code system are usually limited to use in narrow region where the employed empirical correlations are validated. In order to make a safety evaluation code more flexibly applicable, the empirical correlations in TOODEE2 code in the WREM code system was replaced with the core model built in the REFLA code. By changing the multiplication factor for the calculated heat transfer coefficient for the region above the quench front, the predicted clad surface temperatures were compared with those measured in Cylindrical Core Test Facility (CCTF) tests.

It was found that the multiplication factor 0.9 gives always a conservative prediction against CCTF data.     

Numerical Investigation of Heat Transfer Enhancement Phenomenon during the Reflood Phase of PWR-LOCA

The heat transfer in higher power bundles was enhanced in large-scale reflood tests at Japan Atomic Energy Research Institute. The heat transfer enhancement in the core under a radial power distribution is very important to quantify the safety margin in PWR-LOCA. In this study, we analyzed the physical mechanism by numerical simulations with a multi-dimensional two-fluid model code, REFLA/TRAC, using data from the large scale reflood test. The heat transfer enhancement is caused by the increase of local upward liquid velocity resulting from the formation of flow circulation in the core. The flow circulation is generated by a radial difference of waterhead below quench front under a radial power distribution. The upward liquid velocity depends on the bundle power and the cross flow resistance. The higher power and the smaller cross flow resistance give the higher upward liquid velocity, which increases the magnitude of the heat transfer enhancement. Through the present study, some guidelines were obtained for the multi-dimensional analyses to predict the heat transfer enhancement phenomenon with high accuracy.     

Investigation of correlation diagram between heat transfer coefficient and void fraction under sodium–water reaction

Sodium-water reaction (SWR) in a steam generator of sodium-cooled fast reactor (SFR) is a significant phenomenon for safety assessment of the system. One of the top concerns in the SWR is an overheating rupture phenomenon in which a neighbor heat transfer tube fails instantaneously because of a deterioration of structural integrity under a high temperature condition. Hence, the heat transfer coefficient on the tube surface is of importance. Since hydrogen gas is generated in the SWR and liquid water will evaporate quickly due to depressurization, the reaction region is covered with a multi-phase flow structure, and thus the value of the heat transfer coefficient will vary widely. In the present paper, a correlation diagram has been developed between the heat transfer coefficient and the void fraction based on one dimensional homogeneous flow simulation. Furthermore, the transient of void fraction in SWAT-1R experiment is investigated using the diagram.     

Comparisons of the RALIZA-2/02 two-phase flow model with experimental data

The RALIZA-2 computer program was designed for thermal-hydraulic analysis of flow channel and fuel element of PWR/BWR at steady-state and transient conditions. A nonhomogeneous, nonequilibrium model of a two-phase flow and a two-dimensional heat conduction model of fuel pin are used in the program. A fully implicit integration scheme for both models is used. The steady-state constitutive correlations set is used. The void fraction, pre- and post-DNB heat transfer mechanism are compared with data. Also a loss of flow experiment was calculated and compared with nuclear heated rod bundle experimental data for typical PWRs. A very good agreement was obtained.     

SCDAP/RELAP5 Analysis of Station Blackout with Pump Seal LOCA in Surry Plant

During a station blackout of PWR, the pump seal will fail due to loss of the seal cooling. This particular transient-LOCA sequence designated as S3-TMLB' analyzed by SNL with MELPROG/TRAC for Surry plant showed that the depressurization due to the pump seal LOCA would result in early accumulator injection and subsequent core cooling which lead to the delay of reactor pressure vessel (RPV) meltthrough. The present analysis was performed with SCDAP/RELAP5 to evaluate this scenario shown in the MELPROG/TRAC analyses. Addition-ally, the calculated results were compared with the similar experimental studies of JAERI's ROSA-IV program.

The present analyses showed that: (1) During S3-TMLB', the loop seal clearing would occur and cause a slight delay of accident progression. (2) It is unlikely that the accumulator injection, which leads to the delay of RPV meltthrough by approximately 60 min, is initiated automatically during S3-TMLB'. Accordingly, an intentional depressurization using PORVs is recommended for the mitigation of the accident consequences. (3) The present SCDAP/RELAP5 analyses did not show significant delay of accident progression. It was found that non-realistic lower heat generation and higher core cooling models used in the MELPROG/TRAC analysis are attributed to this discrepancy.     

压水堆子通道欠热沸腾数值验证及交混翼研究

为改善压水堆交混翼格架在欠热沸腾工况下的热工水力特性,以子通道为研究对象验证了所使用的欠热沸腾数值模型在不同工况下的有效性。基于已验证的数值模型,对含不同偏折角交混翼格架的子通道模型在不同工况下进行了两相流数值模拟,研究交混翼及其偏折角对子通道中两相流动、传热及气泡分布的影响。结果表明：交混翼在增大压降的同时明显强化了冷却剂的交混、降低了近壁面气泡份额、提高了换热效率,且在一定范围内偏折角越大影响越明显。相对较高的气泡份额将导致更大的压力损失、减弱冷却剂的交混、降低传热效率。当交混翼偏折角达25°时,继续增大其偏折角对降低近壁面气泡份额和提高传热效率的作用不再明显,反而造成压降的快速增大,因此建议其偏折角在25°左右。