Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Fynan, Douglas A.,Ahn, Kwang-Il Elsevier 2016 Nuclear engineering and design Vol.310 No.-

<P><B>Abstract</B></P> <P>Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.</P> <P><B>Highlights</B></P> <P> <UL> <LI> Pressure drop-flow rate curves for superheated steam in U-tubes were generated. </LI> <LI> Forward flow of hot steam is favored in the longer and taller U-tubes. </LI> <LI> Reverse flow of cold steam is favored in short U-tubes. </LI> <LI> Steam generator U-tube bundle geometry and tube diameter are important. </LI> <LI> Need for correlation development for natural convention heat transfer coefficient. </LI> </UL> </P>

Preliminary study for a PWR steam generator with CUPID/MARS multi-scale thermal-hydraulics simulation

Lee, Jae Ryong,Yoon, Han Young Pergamon 2017 Applied thermal engineering Vol. No.

<P><B>Abstract</B></P> <P>For the thermal hydraulics analysis of nuclear reactor components of LWRs (Light Water Reactors) such as reactor vessel, steam generator, containment, a multi-dimensional two-phase flow code, named CUPID, has been being developed. The CUPID code pursues a capability of multi-physics and multi-scale thermal hydraulics analysis. In the present study, multi-scale simulation was performed by coupling with system-scale code, MARS. The coupled code was assessed to visualize the flow behavior of the steam generator of the Advanced Power Reactor (APR1400). The primary side of the steam generator and remaining Reactor Coolant System (RCS) is modeled by MARS and secondary side is by CUPID. For the secondary side simulation by the CUPID part, a porous media approach was adopted to two-fluid model and conductor model to simplify the complicated geometry of the steam generator. In order to obtain a porosity of a given computing cell, a special algorithm was employed to directly calculate volume ratio by mapping the 3D CAD file onto the grid system. Besides, the proper constitutive relationships for U-tubes are considered further. To treat the complex thermo-hydraulic phenomena on the shell side of a steam generator, a set of constitutive models available in the literature for a two-phase flow map, interfacial heat and mass transfer, interfacial drag, wall friction, wall heating, and heat partitioning in flows over tube bundles were applied to close the numerical model. This paper presents the description of the coupling method, porous media approach to simplify the steam generator, and the simulation results using the coupled code.</P>

소듐냉각 고속로용 증기발생기 기술분석 및 개념개발

남호윤(Ho-Yun Nam),김종범(Jong-Bum Kim),이재한(Jae-Han Lee),박창규(Chang-Gyu Park) 대한기계학회 2011 대한기계학회 춘추학술대회 Vol.2011 No.4

소듐냉각 고속로를 개발함에 있어 최대 현안 중 하나가 증기발생기에서의 소듐-물 반응사고 가능성이다. 이를 개선하기 위해 지금까지 수십 종 이상 연구개발 되었지만 국가마다 그 사양이 다르고, 동일한 기종이 후속기에 다시 활용되지 못할 정도로 기술이 안정화 상태에 도달하지 못하였다. 최근 개발되고 있는 증기발생기의 공통적 목표는 소듐-물 반응사고의 조기감지 및 제어, 증기발생기의 검사 및 보수가 쉽게 용접개수를 줄이고 경제성을 높인 Benson 증기사이클을 적용하는 것이다. 이 논문에서는 지금까지 설계 또는 활용한 증기발생기들의 사양과 문제점을 비교, 분석하였고, 이를 토대로 현안 극복방안을 제시하였다. Steam generator is one of the great issues in developing a sodium-cooled fast reactor due to sodium-water reaction problems. Up to now, several kinds of steam generators have been developed, but the specifications of the steam generators differ from each country, and even if a country had developed a steam generator, it had not been used in the subsequent reactor because the current techniques were not stabilized to select the proper steam generator. As a common development direction, the Benson steam cyc1e may be adopted considering with few welding locations, and with high economical efficiency. Also the design is dwelled on the convenience of inspection, detection, control, maintenance for the sodium-water reaction. A review of the specifications of the designed steam generators and an analysis of the current technical issues of steam generators had been performed. The concepts were proposed to overcome the current technical issues of steam generators.

헬륨가열계통을 활용한 고온수전해 증기/공기 공급 장치

홍성덕(Sung Deok Hong),김신엽(Sin Yeob Kim),강경준(Kyung Jun Kang),박병하,김찬수(Chan Soo Kim),Hong Sik Lim 대한기계학회 2023 대한기계학회 춘추학술대회 Vol.2023 No.11

Hydrogen production efficiency using High Temperature Steam Electrolysis (HTSE) improves as the temperature of the supplied steam increases. A Lab-scale HTSE system is now under constructing at the Korea Atomic Energy Research Institute. An experiment facility to generate high-temperature steam of 820℃ is designed and constructed for 30㎾ HTSE using solid-oxide electrolyte cell stacks. The experiment facility equipped a 77㎾ heating system to heat the helium up to 950℃, and a Multi-stream heat exchanger (MHX) is designed and manufactured to heat the steam and air up to 800℃ with the heated helium. To control the mass flow rate and pressure of superheated steam into the MHX, shell and helical-tube type steam generator with core tank is separately designed. Pressurized pure water at a constant flow rate is injected into the shell side of the steam generator, and helium from the MHX passes through the tube side of steam generator. In this study, the stable generation of both superheated steam of 15.4 ㎏/hr at 805℃ and air of 11.5 ㎏/hr at 761℃ are tested through the helium heated steam/air supply system.

응축기 순환수 압력에 따른 스팀 생산 히트펌프의 성능에 관한 연구

강대훈(Dae Hoon Kang),나선익(Sun-Ik Na),김민수(Min Soo Kim) 대한기계학회 2018 대한기계학회 춘추학술대회 Vol.2018 No.12

A steam generation heat pump cycle is an effective method for utilizing waste heat and supplying steam to industrial processes. The boiling water is fed into a plate heat exchanger and circulated to a flash tank to generate steam. Unlike conventional hot water generation heat pump, steam generation involves a large volumetric change and therefore requires a flash tank. Optimal control of flash tank affects the performance of the system and steam generation rate. In this study, the condenser outlet pressure and mass flow rate of water was controlled by a valve to compare the steam generation rate. When the same condensing pump power is used, the higher outlet pressure makes the higher condensation heat. This also increased the system COP and the steam generation rate.

적응제어기법에 의한 PWR STEAM GENERATOR의 자동화 운전에 관한 연구

卓漢浩,李寅龍 진주산업대학교 농업기술연구소 1996 農業技術硏究所報 Vol.9 No.-

본 논문에서는 PWR steam generator의 레벨제어 시스템을 모델링하였다. 실제어문제에서 증기발생 시스템의 수축, 팽창 현상으로 15%이하에서 기존의 PID 시스템으로 제어하기가 곤란하다는 것을 증명하였다. 그리고 R 다변수(28개) 시스템이므로 간이화 개념을 도입하여 시스템을 구성하였고, 제어기와 시스팀은 Irving이 제시한 모델이 적합지 않아 제시한 모델을 부분 수정하고 제어기는 직접 구성하였다. Bihoreaux이 제시한 기본적인 적응 알고리즘과 관측기 및 K-step 예측기등의 필요성과 등을 추정을 통하여 확인함으로써 전 출력운전 범위에서 자동조정이 가능함을 소개하였으며 제어기 및 적응시스템의 변수값 설정에는 많은 시뮬레이션을 통하여 적합한 값을 산출하였고 시뮬레이션은 연속시 알고리즘으로 수행하였다. 향후 계획은 다음과 같다. 1) 발전소 plant를 대상으로 한 보강된 제어기 설계와 2) plant 설비에 적합한 보다 나은 적응시스템의 모델링과 3) 제어기와 적응 시스템의 상태변수를 쉽게 구하는 방법 4) 다양한 입력을 가하는 알고리즘의 개방등으로 여러 원자력 발전모델에 공통적으로 적용할 수 있도록 하면 전체 시스템의 신뢰도와 안정성에 기여할 수 있으리라 생각된다. Due to the problems of the steam generator is presented with swell and shrink thus a sim plified mathematical modeling. The practical control problem shows that the control variable is not fully utilized by the PID system designed in the classical way. The simplified adaptive control mathode is set out which avoids the different problems encountered. This paper description necessity of state observer and a K-step-ahead prodictor. To check through the estmate will be realized in the full automatic control of the steam generator of the PWR reactor from low to high loads. The numerical results description that output responses and water level rate due to a step in the water flow with dependent on the load, the step responses of the water level are strongly dependent on the load.

다변량 로지스틱 회귀분석을 이용한 증기발생기 전열관 ODSCC의 POD곡면 분석

이재봉(Jae Bong Lee),박재학(Jai Hak Park),김홍덕(Hong-Deok Kim),정한섭(Han-Sub Chung) 대한기계학회 2007 대한기계학회 춘추학술대회 Vol.2007 No.5

Steam generator tubes play an important role in safety because they constitute one of the primal)' barriers between the radioactive and non-radioactive sides of the nuclear power plant. For tins reason, the integrity of file tubes is essential in minimizing the leakage possibility of radioactive water, The integrity of the tubes is evaluated based on NDE (non-destructive evaluation) inspection results. Especially ECT (eddy current test) method is usually used for detecting the flaws in steam generator tubes. However, detection capacity of the NDE is not perfect and all of the real flaws which actually existing in steam generator tubes is not known by NDE results. Therefore reliability of NDE system is one of tile essential parts in assessing tile integrity of steam generators. In this study POD (probability of detection) of ECT system for ODSCC in steam generator tubes is evaluated using multivariate logistic regression. The cracked tube specimens are made using the withdrawn steam generator tubes. Therefore the cracks are not artificial but real. Using the multivariate logistic regression method, continuous POD surfaces are evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive evaluation of tile cracked tubes. Length and depth of cracks are considered in multivariate logistic regression and their effects on detection capacity are evaluated.

고효율 인쇄기판형 열교환장치의 증기발생기분야 적용을 위한 예비크기 산정

권길성(Kilsung Kwon),김상지(Sang Ji Kim) 대한설비공학회 2018 대한설비공학회 학술발표대회논문집 Vol.2018 No.6

A steam generator is one of vital parts for the steam power cycle. Its performance and reliability should be improved to increase the economics of power plants. The printed circuit steam generator (PCSG), which is produced by a photochemical etching and diffusion bonding, has been considered as an alternative of the existing shell-and-tube steam generator owing to the great durability under high pressure and temperature. In this study, we dealt with the integration of PCSG into sodium-cooled fast reactor (SFR). The outlet temperature of sodium and inlet temperature of water was computed by the thermal approach. Also, the size of PCSG was determined by the basic scaling method applicable to the once-through type steam generator.

신형경수로 증기발생기 마모손상 억제를 위한 설계최적화

임혁순(Hyuk-Soon LIM),박영섭(Young-Sheop PARK),이광한(Kwang-Han LEE),이석호(Seok-Ho LEE),정대율(Dae-Yul CHUNG) 대한기계학회 2004 대한기계학회 춘추학술대회 Vol.2004 No.4

Inconel-600 alloy has been used as steam generator tube material for current pressurized water reactors (PWRs). The long-term operation of steam generators showed that the use of this material induced localized corrosion damages and increased tube wear of steam generator. To protect these problems, steam generator tube material is being changed to Inconel-690 alloy. Based on the current trend, we have chosen Inconel 690 as the Advanced Power Reactor 1400 (APR1400) steam generator(SG) tube material and performed the design optimization of preventive measure against tube fretting wear for the APR1400 steam generator. In this paper, we examined the technical consideration in this modification : the selection of material, wear characteristics, effect of the Egg-crate Flow Distribution Plate installation, and effect analysis of vertical strip installation.

Automated Analysis Technique Developed for Detection of ODSCC on the Tubes of OPR1000 Steam Generator

김인철,남민우 한국비파괴검사학회 2013 한국비파괴검사학회지 Vol.33 No.6

A steam generator(SG) tube is an important component of a nuclear power plant(NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC(outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS(computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.