加速器电磁元件设计微机软件库的初步研制

描述了加速器电磁元件设计的微机软件库初步研制工作,包括ＰＯＩＳＳＯＮ／ＳＵＰＥＲＦＩＳＨ、ＲＥ－ＬＡＸ３Ｄ、ＴＲＡＮＳＰＯＲＴ、ＲＡＹＴＲＡＣＥ、ＭＡＤ、ＰＡＲＭＥＬＡ系统软件和软件包括移植、开发、以及自行开发的软件ＤＥ２Ｄ、ＤＥ３Ｄ、ＣＹＣＣＡＥ、ＣＹＣＣＥＮ的整理工作。介绍了这些软件在一定实际工程中的应用情况。     

NJOY—WIMS程序系统的研制

ＮＪＯＹ－ＷＩＭＳ程序系统是在开发ＮＪＯＹ（包括ＷＩＭＳＲ）、ＷＩＭＳ等程序，增加有关管理模块ＳＣＮ和ＣＣＭＩＴ的基础上建立起来的。这个系统可以用来进行ＷＩＭＳ库的制作及临界安全计算。作者由微观ＪＥＦ－１出发，应用该系统给出了ＴＲＸ－１、２，ＢＡＰＬ－ＵＯ２－１、２、３的计算结果。经比较表明，该系统是可靠的。     

NaI（Tl）整体测量γ谱的连续谱本底自动处理法

本文介绍了一种应用“数字内插和蚀峰”原理，从ＮａＩ（Ｔｌ）整体测量γ谱中自动生成连续本底的计算方法。用此模拟本底来代替测量“本底人”的本底。该方法已用ＢＡＳＩＣ和ＦＯＲＴＲＡＮ－７７语言编写出运算程序ＡＴＢＫ，分别在ＴＲＳ－８０和ＣＯＰＡＭ－ＰＣ－２８６Ｓ微机上做了运行和试验。通过自编的转换程序ＲＥＡＤ，使ＡＴＢＫ与ＮＤ－ＡＳＡＰ谱分析软件相联合，组成了一个人体测量数据自动分析系统。     

NaI（TI）整体测量γ谱的连续谱本底自动处理法

本文介绍了一种应用“数字内插和蚀峰”原理，从ＮａＩ（ＴＩ）整体测量γ谱中自动生成连续本底的计算方法，用此模拟本底来测量“本底人”的本底。该方法已用ＢＡＳＩＣ和ＦＯＲＴＲＡＮ－７７语言编写出运算程序ＡＴＢＫ，分别在ＴＲＳ－８０和ＣＯＰＡＭ－ＰＣ－２８６Ｓ微机上做了运行和试验，通过自编的转换程序ＲＥＡＤ，使ＡＴＢＫ与ＮＤ－ＡＳＡＰ谱分析软件相联合，组成了一个人体测量数据自动分析系统。     

RETRAN 02程序的改进及在NOS／VE系统上的移置

ＲＥＴＲＡＮ０２程序是迄今为止核反应堆事故瞬态分析的主要工具之一。但由于包含有汇编和非标准ＦＯＲＴＲＡＮ语言（约占整个程序的１／３），ＲＥＴＲＡＮ０２程序向其他操作系统或其他机器的移置十分困难。我们利用标准ＦＯＲＴＲＡＮ和Ｃ语言替换和修改了包含汇编和非标准ＦＯＲ－ＴＲＡＮ部分，实现了原程序的全部功能。选用ＲＥＴＲＡＮ标准例题和低温核供热堆例题校核了程序的初始运行和再启动运行模块，用Ｃ语言实现了绘图模块功能。     

NJOY－WIMS程序系统的研制

ＮＪＯＹ－ＷＩＭＳ程序系统是在开发ＮＪＯＹ（包括ＷＩＭＳＲ）、ＷＩＭＳ等程序，增加有关管理模块ＳＣＮ和ＣＣＭＩＴ的基础上建立起来的。这个系统习以用来进行ＷＩＭＳ库的制作及临界安全计算。作者由微观ＪＥＦ－１出发，应用该系统给出了ＴＲＸ－１、２，ＢＡＰＬ－ＵＯ＿２－１、２、３的计算结果。经比较表明，该系统是可靠的。     

用于压水堆核电站瞬态分析的微机程序──MACONP

提出了一个能在微机上运行的ＰＷＲ核电站瞬态分析程序ＭＡＣＯＮＰ。该程序可对有关运行瞬态和大部分设计基准事故进行分析计算。计算精度高、速度快、程序操作简单、使用方便，并给出了几种ＡＴＷＳ瞬态工况的分析结果，与大程序ＲＥＬＡＰ５／ＭＯＤ２和ＲＥＴＲＡＮ０２／ＭＯＤ２计算结果相比较，两者符合良好。     

RADTRAN4.0程序的移植和应用

本文介绍了将ＲＡＤＴＲＡＮ４．０程序向ＭＩＣＲＯ ＶＡＸ－Ⅱ／ＶＭＳ操作系统和ＬＸ－３８６ ＳＸ微机上移植过程中所做的工作。经过例题运算分析，证明该程序的移植是成功的，移植后的程序已初步应用于放射性物质运输的实例分析中。     

RESIDUAL GAS IONIZATION BEAM PROFILE MONITOR ON 40MeV H~－ BEAM TRANSPORT LINE

ＲＥＳＩＤＵＡＬＧＡＳＩＯＮＩＺＡＴＩＯＮＢＥＡＭＰＲＯＦＩＬＥＭＯＮＩＴＯＲＯＮ４０ＭｅＶＨ~－ＢＥＡＭＴＲＡＮＳＰＯＲＴＬＩＮＥＸｕＷｅｉｐｅｎｇ（徐伟鹏）（ＩｎｓｔｉｔｕｔｅｏｆＮｕｃｌｅａｒＲｅｓｅａｒｃｈ，ｔｈｅＣｈｉｎｅｓｅＡｃａｄｅｍｙ...     

PANAMA程序及其在10MW高温气冷实验堆安全分析中的应用

ＰＡＮＡＭＡ程序是德国在高温气冷堆安全研究中开发的一个实用程序，可以用来计算ＴＲＩＳＯ－包覆燃料颗粒在事故条件下的破损率，本文简介ＰＡＮＡＭＡ模型，着重开发了ＰＡＮＡＭＡ程序中ＳｉＣ压力容器失效模式，并利用１０ＭＷ高温气冷实验堆（ＨＴＲ－１）包覆燃料颗粒的设计参数，计算了燃耗，温度，核芯直径以及各包覆层厚度对颗粒破损率的影响，结果分析表明破损率阻燃耗，温度和核志直径的增大面而增长较快，对缓冲层和Ｓ     

Features and performance of a coupled three-dimensional thermal-hydraulic/kinetics TRAC-PF1/NEM pressurized water reactor (PWR) analysis code

This paper summarizes the current status of the Pennsylvania State University (PSU) version of the coupled three-dimensional (3-D) thermal-hydraulic/kinetics TRAC-PF1/NEM code for pressurized water reactor (PWR) transient and accident analysis and describes applications to reactivity insertion accident (RIA) simulations as well as recent developments. The TRAC-PF1/NEM methodology utilizes closely coupled 3-D thermal-hydraulics and 3-D core neutronics transient models to simulate the vessel and a 1-D simulation of the primary system. An efficient and flexible cross-section generation procedure was developed and implemented into TRAC-PF1/NEM. These features make the coupled code capable of modeling PWR reactivity transients, including boron dilution transients, in a reasonable amount of computer time. Three-dimensional studies on hot zero power (HZP) rod ejection and main steam line break (MSLB) transients in a PWR, as well as a large break loss-of-coolant-accident (LBLOCA) and boron dilution transients, were accomplished using TRAC-PF1/NEM. The results obtained demonstrate that this code is appropriate for analysis of the space-dependent neutronics and thermal-hydraulic coupled phenomena related to most current safety issues.     

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.     

压水堆最佳估算程序TRAC-PFI——流体动力学模型及特点

本文在简介了TRAC-PF1程序的基础上,着重论证和讨论了TRAC-PF1的流体动力学模型和数值解方法,分析了该模型的特点和应用范围。最后简述了TRAC-PF1在流体动力学模型上对TRAC-PD2的改进及其结果。     

Monte Carlo calculations on high speed machines

1. 1. Monte Carlo Calculations by a Modified Vector Processor.At Japan Atomic Energy Research Institute, four Monte Carlo codes KENO-IV, MORSE-DD, and VIM and MCNP were vectorized to examine the adaptability of vector processors for these codes. The performances and vectorization rates of the vectorized versions were not good except KENO-IV on SX-2 vector processor, on which the vectorized version attained three times faster speed compared to its scalar version. According to the experience with the vectorization, some additional features specialized for Monte Carlo calculations will improve the vectorization rates of the four codes. Functions of the features and anticipated effects are presented.

2. 2. Speedup of Monte Carlo Criticality Calculation by a Parallel Processor.The authors implemented the Monte Carlo code KENO-IV on a parallel processor system Topology 1000 with three processing units under control of SUN workstation. The Hansen-Roach 16-group cross section set was used for the calculation. A computation for a bare sphere of highly enriched uranium metal showed that the system attained 2.7 times speedup compared to that of one processing unit.

     

Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway.     

Analysis of TRAC-PF1 calculated core heat transfer for a CCTF test

TRAC-PF1 posttest calculation for CCTF test C1-5 (Run 14) was performed to assess the core thermal-hydraulic models in the TRAC-PF1 code during the reflood phase of a PWR LOCA. TRAC showed good agreement with data for heater rods turnaround temperatures and turnaround times in the lower half of the core. However, TRAC overpredicted turnaround times and underpredicted quench times in the upper part of the core. Even though heat transfer correlations have a strong dependency on the local void fraction in TRAC, TRAC-predicted void fraction profiles showed poor agreement with CCTF data that have been inferred from differential pressure measurements. From these comparisons with CCTF data, the following areas for future improvements of TRAC-PF1 should be considered: (1) the core hydraulic modeling used to calculate the void fraction profile in the core. (2) the method for evaluating heat transfer within the core.     

安全分析程序SAC-PREARS的开发

介绍了SAC-PREARS程序的基本数学模型。利用MISAP程序和通用程序TRAC-PFI对SAC-PREARS计算结果进行了验证。结果表明：SAC-PREARS程序能够有效地计算和分析核供热堆PRHRS的稳态和瞬态热工水力特性。     

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.     

TRAC—PF1在大破口LOCA分析中的应用

文章简述了TRAC-PF1与大破口LOCA分析有关的功能和特点。针对大破口LOCA分析做出了秦山核电厂核蒸汽系统的适用于TRAC-PF1的模型。给出了对系统的稳态模拟结果和大破口LOCA分析的基本假设、事故过程及瞬态曲线。最后对结果进行了分析,指出为实际得到秦山核电厂大破口LOCA分析结果,在此基础上尚需获得并核实的关键数据。本文的意义在于介绍了一种应用TRAC-PF1进行大破口LOCA分析的方法。