• Nuclear Engineering and Technology
• /
• 제55권12호
• /
• pp.4518-4526
• /
• 2023

NECP-SARAX is a neutronics analysis code system for advanced reactor developed by Nuclear Engineering Computational Physics Laboratory of Xi'an Jiaotong University. In past few years, improvements have been implemented in TULIP code which is the cross-section generation module of NECP-SARAX, including the treatment of resonance interface, considering the self-shielding effect in non-resonance energy range, hyperfine group method and nuclear library with thermal scattering law. Previous studies show that NECP-SARAX has high performance in both fast and thermal spectrum system analysis. The accuracy of TULIP code in fast and thermal spectrum system analysis is demonstrated preliminarily. However, a systematic verification and validation is still necessary. In order to validate the applicability of TULIP code for full energy range, 147 fast spectrum critical experiment benchmarks and 170 thermal spectrum critical experiment benchmarks were selected from ICSBEP and used for analysis. The keff bias between TULIP code and reference value is less than 300 pcm for all fast spectrum benchmarks. And that bias keeps within 200 pcm for thermal spectrum benchmarks with neutron-moderating materials such as polyethylene, beryllium oxide, etc. The numerical results indicate that TULIP code has good performance for the analysis of fast and thermal spectrum system.

• Nuclear Engineering and Technology
• /
• 제40권1호
• /
• pp.87-92
• /
• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Nuclear Engineering and Technology
• /
• 제33권2호
• /
• pp.209-224
• /
• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Current Optics and Photonics
• /
• 제3권6호
• /
• pp.531-540
• /
• 2019

Spectral Amplitude Coding Optical Code Division Multiple Access (SAC OCDMA) is an advanced technique in asynchronous environments. This paper proposes design and implementation of a novel quantum logic gate (QLG) code, with code construction algorithm generated without following any code mapping procedures for SAC system. The proposed code has a unitary matrices property with maximum overlap of one chip for various clients and no overlaps in spectra for the rest of the subscribers. Results indicate that a single algorithm produces the same length increment for codes with weight greater than two and follows the same signal to noise ratio (SNR) and bit error rate (BER) calculations for a higher number of users. This paper further examines the performance of a QLG code based SAC-OCDMA system with NAND and direct detection techniques. BER analysis was carried out for the proposed code and results were compared with existing MDW, RD and GMP codes. We demonstrate that the QLG code based system performs better in terms of cardinality, which is followed by improved BER. Numerical analysis reveals that for error free transmission (10^-9), the suggested code supports approximately 170 users with code weight 4. Our results also conclude that the proposed code provides improvement in the code construction, cross-correlation and minimization of noises.

• 한국산업융합학회 논문집
• /
• 제25권6_2호
• /
• pp.1091-1095
• /
• 2022

Dynamic stability of the main spindle system shall be ensured when operating the planer miller for remanufacturing the planer miller. This paper explains the analysis process that determines the stability of the planer miller spindle system in the design stage using ANSYS, an analysis program. First, the dynamic stability of the main spindle system is verified through risk speed analysis in the rated RPM range of the planer miller through ANSYS Modal Analysis, and second, the stability and durability of the main spindle system are verified through ANSYS nCode Analysis.

• Nuclear Engineering and Technology
• /
• 제43권1호
• /
• pp.45-62
• /
• 2011

The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(1)
• /
• pp.91-96
• /
• 1995

In this paper, research efforts to develop computer code system for analysis of the transmutation reactors at KAIST are described Especially the computer code HANCELL for assembly calculation of fast reactors is mainly described. Features and function of the code are identified md current status of the code development is provided

• 한국정보처리학회:학술대회논문집
• /
• 한국정보처리학회 2013년도 춘계학술발표대회
• /
• pp.972-975
• /
• 2013

This work developed a code analysis & extraction tool named Function Network Analyzer (FNA) to reduce the cost of software safety analysis. FNA analyzes functions and variables which a given function depends on, and extracts subset of code that can be compiled of automotive operating system, final resulting a well-ordered code sequence that can be compiled for model checking technique. And the experimental result illustrates that FNA can get 100% accurate rate and over 96% reduction rate by testing API functions from trampoline system.

• 한국안전학회지
• /
• 제35권4호
• /
• pp.84-91
• /
• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• 한국통신학회논문지
• /
• 제31권12C호
• /
• pp.1135-1142
• /
• 2006

본 논문에서는 전송률이 서로 다른 다양한 서비스들을 지원하기 위한 새로운 비동기 MC-CDMA 시스템을 제안한다. 제안한 시스템에서는 W-CDMA 하향링크에 사용되는 채널화 코드인 OVSF 코드의 생성특성을 이용하여 발생된 OVSF 코드를 다중전송률 MC-CDMA의 확산부호로 사용한다. 사용자의 전송률에 따라 길이가 다른 OVSF 코드를 확산부호로 사용하며 OVSF 코드트리에서 같은 가지에 속한 코드들을 사용하는 사용자들을 모아 그룹화를 한다. 코드그룹화 간섭제거방식을 사용하여 그룹간 간섭을 일차적으로 제거하는데, 이때 간섭을 일으키는 다른 사용자들에 대한 정보가 요구되지 않는다. 제안된 다중 전송률 비동기 MC-CDMA 시스템을 위한 모델을 제시하고 이에 따른 시스템 성능분석을 행한다. 제안된 시스템과 직교부호를 확산부호로 사용하는 단일 전송률 MC-CDMA시스템의 성능과 비교 분석한다.