Analysis of Reactivity Accidents in Modular HTGRs

2020 ◽  
Author(s):  
Jacob P. Gorton ◽  
Nicholas R. Brown
Keyword(s):  
Heat Transfer ◽  
Kinetic Parameters ◽  
Nuclear Regulatory Commission ◽  
Sensitivity Study ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Kinetic ◽  
Regulatory Commission ◽  
The Impact

Abstract A control rod withdrawal (CRW) is a possible reactivity initiated accident (RIA) in modular high-temperature gas-cooled reactors (mHTGRs). The purpose of this study is to perform a sensitivity analysis of a CRW event in an mHTGR model using the systems code RELAP5-3D with point kinetics feedback to demonstrate the impact of uncertainty in heat transfer and reactor kinetic parameters. The adaptive Sobol decomposition method in the uncertainty quantification code RAVEN is used to perform the sensitivity study and to determine which input parameters have the greatest impact on the figures of merit, which in this case are peak reactor power and maximum fuel temperature. This study addresses a need highlighted by the Nuclear Regulatory Commission (NRC) for transient fuel testing by quantifying the impact of uncertainty in heat transfer and reactor kinetic parameters and by generating potential boundary conditions for transient testing of conventional mHTGR fuel.

Human Factor Data in Nuclear Power Plant Applications

1980 ◽  
Vol 24 (1) ◽  
pp. 123-123
Author(s):  
Linda O. Hecht
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Reliability Modeling ◽  
Plant Safety ◽  
Modeling Techniques ◽  
Nuclear Systems ◽  
Factor Data ◽  
Regulatory Commission ◽  
The Impact

Due to the concern for safety the nuclear power industry in the United States has fostered the use of reliability analysis to assess system performance and the impact of system failure on overall plant safety. The need for system and component failure rate data has been recognized and has spurred such efforts as NPRDS (Nuclear Power Research Data System) and IEEE's Std 500 (The Reliability Data Manual). Reliability modeling techniques have been developed for application to nuclear systems and are presently being considered by the Nuclear Regulatory Commission for licensing purposes.

Effects of Weld Geometry on Residual Stress and Crack Driving Force for Centerhole Control Rod Drive Mechanism Nozzles: Part I — Weld Residual Stress

2005 ◽  
Author(s):  
Wentao Cheng ◽  
David L. Rudland ◽  
Gery Wilkowski ◽  
Wallace Norris
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Driving Force ◽  
Nuclear Regulatory Commission ◽  
Crack Driving Force ◽  
Control Rod ◽  
Drive Mechanism ◽  
Weld Geometry ◽  
Weld Residual Stress ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) has undertaken a program to assess the integrity of control rod drive mechanism (CRDM) nozzles in existing plants that are not immediately replacing their RPV heads. This two-part paper summarizes some of the efforts undertaken on the behalf of the U.S.NRC for the development of detailed residual stress and circumferential crack-driving force solutions to be used in probabilistic determinations of the time from detectable leakage to failure. In this first paper, the finite element (FE) simulations were conducted to investigate the effects of weld geometry on the residual stresses in the J-weld for a centerhole CRDM nozzle. The variables of weld geometry included three weld heights (weld sizes) and three groove angles for each weld height while keeping the same weld size. The analysis results indicate that the overall weld residual stress decreases as the groove angle increases and higher residual stress magnitude is associated with certain weld height. The results also reveal that the axial residual stresses in the Alloy 600 tube are very sensitive to the weld height, and that the tube hoop stresses above the J-weld root increase with the increasing weld height.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core

2010 ◽  
Author(s):  
Minggang Lang ◽  
Yujie Dong
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
System Pressure ◽  
Temperature Limitation

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1242.4°C. It was a little higher than 1230°C–the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

Lumped Capacitance and 3D CFD Conjugate Heat Transfer Modelling of an Automotive Turbocharger

2015 ◽  
Author(s):  
R. Burke ◽  
C. Copeland ◽  
T. Duda ◽  
M. A. Reyes-Belmonte
Keyword(s):  
Heat Transfer ◽  
High Speed ◽  
Weak Link ◽  
Three Dimensional ◽  
Sensitivity Study ◽  
Heat Fluxes ◽  
Strategy Development ◽  
Inlet Temperature ◽  
Compression Process ◽  
The Impact

One dimensional wave-action engine models have become an essential tool within engine development including stages of component selection, understanding system interactions and control strategy development. Simple turbocharger models are seen as a weak link in the accuracy of these simulation tools and advanced models have been proposed to account for phenomena including heat transfer. In order to run within a full engine code, these models are necessarily simple in structure yet are required to describe a highly complex 3D problem. This paper aims to assess the validity of one of the key assumptions in simple heat transfer models, namely, that the heat transfer between the compressor casing and intake air occurs only after the compression process. Initially a sensitivity study was conducted on a simple lumped capacity thermal model of a turbocharger. A new partition parameter was introduced αA, which divides the internal wetted area of the compressor housing into pre and post compression. The sensitivity of heat fluxes to αA was quantified with respect to the sensitivity to turbine inlet temperature (TIT). At low speeds, the TIT was the dominant effect on compressor efficiency whereas at high speed αA had a similar influence to TIT. However, modelling of the conduction within the compressor housing using an additional thermal resistance caused changes in heat flows of less than 10%. Three dimensional CFD analysis was undertaken using a number of cases approximating different values of αA. It was seen that when considering a case similar to αA=0, meaning that heat transfer on the compressor side is considered to occur only after the compression process, significant temperature could build up in the impeller area of the compressor housing, indicating the importance of the pre-compression heat path. The 3D simulation was used to estimate a realistic value for αA which was suggested to be between 0.15 and 0.3. Using a value of this magnitude in the lumped capacitance model showed that at low speed there would be less than 1% point effect on apparent efficiency which would be negligible compared to the 8% point seen as a result of TIT. In contrast, at high speeds, the impact of αA was similar to that of TIT, both leading to approximately 1% point apparent efficiency error.

Impact of Undetected Fabrication Flaws on LBB Risk

2017 ◽  
Author(s):  
Robert Kurth ◽  
Cédric Sallaberry ◽  
Elizabeth Kurth ◽  
Frederick Brust
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Leak Detection ◽  
Nuclear Regulatory Commission ◽  
Acceptance Testing ◽  
Analysis Tool ◽  
Risk Of Rupture ◽  
Asme Code ◽  
Regulatory Commission ◽  
The Impact

On-going assessments of the impact of active degradation mechanisms in US nuclear power plants previously approved for leak before break (LBB) relief are being performed with, among other methods, the extremely low probability of rupture (xLPR) code being developed under a memorandum of understanding between the US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) [1]. This code finished with internal acceptance testing in July of 2016 and is undergoing sensitivity and understanding analyses and studies in preparation for its general release. One of the key components of the analysis tool is the integration of inspection methods for damage and the impact of leak detection on the risk of a pipe rupture before the pipe is detected to be leaking. While it is not impossible to detect a crack or defect that is less than 10% of the pipe wall thickness current ASME code does not give credit for inspections identifying a crack of this size. This study investigates the impact of combining the probabilistic analysis results from xLPR with a pre-existing flaw to determine the change, if any, to the risk. Fabrication flaws were found to have a statistically significant impact on the risk of rupture before leak detection.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core Under the Loss of the System Pressure

2014 ◽  
Author(s):  
Minggang Lang
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
Design Basis ◽  
System Pressure

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and to use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basis accidents and beyond design basis accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake under the loss of the system pressure. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted. At the same time, the system pressure was supposed to lose by some reason. Thus the reactor power increased and the temperature of the core increased. And the protection system warns with two scram signal: too high of the negative varying rate of the system pressure and too high of the reactor power, which should induce the reactor to scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure continued to increase but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1203.4°C. It was a little bit lower than 1230°C — the fuel temperature limitation of HTR-10 and there is no release of any radioactivity. So the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10.

Comparison of Component Fatigue Test Data With Proposed EAF Methodologies for Austenitic Stainless Steel

2009 ◽  
Author(s):  
Rim Nayal ◽  
Hasan Charkas
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Fatigue Test ◽  
Low Cycle Fatigue ◽  
Nuclear Regulatory Commission ◽  
Fatigue Curve ◽  
Asme Code ◽  
Regulatory Commission ◽  
License Renewal ◽  
The Impact

The U.S. Nuclear Regulatory Commission (NRC) currently requires evaluation of the effect of environmental fatigue for both license renewal and new plants. NRC required the use of methodology in EPRI MRP-47, Rev. 1 addressing NUREG/CR-5704, be used for license renewal of stainless steel (SS) components, and NUREG/CR-6909 for use in new plants. These two methodologies are based on applying an environmental correction factor (Fen) on the number of in-air design cycles. These factors are applied to the fatigue usage from each individual range of stress (or range of strain). The focus of this paper is to compare the two aforementioned methodologies; this includes comparison of the fatigue curve as well as the comparison of the environmental correction factors (Fen). Fatigue test results data reported by others are also compared with these two methodologies. It is important to evaluate the impact of using any of those methodologies on the design fatigue life of the components. It is concluded that NUREG/CR-5704 is more severe than NUREG/CR-6909 in the LCF (low-cycle fatigue) regime, while NUREG/CR-6909 is more severe elsewhere, and both NUREG’s extremely underestimate fatigue life in PWR environment. It is also concluded that the current ASME-code fatigue curve for stainless steel reasonably estimates fatigue life in an LWR environment with reasonable margins.

The Impact of Geometric Discontinuities on Alloy 617 Creep-Rupture Behavior

2020 ◽  
Author(s):  
Ryann E. Rupp ◽  
Michael D. McMurtrey
Keyword(s):  
Rupture Life ◽  
Nuclear Regulatory Commission ◽  
Creep Rupture ◽  
Multiaxial Stress ◽  
Alloy 617 ◽  
Multiaxial Creep ◽  
Term Creep ◽  
Regulatory Commission ◽  
The Right ◽  
The Impact

Abstract The Nuclear Regulatory Commission (NRC) is currently reviewing Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code for potential endorsement. A prior NRC-sponsored review on the application of Section III, Division 5 design rules for elevated-temperature reactors have identified issues currently not in the code that need to be addressed. These include an insufficient knowledge of the effect of notches, structural discontinuities, and multiaxial stress. This work qualitatively evaluates Alloy 617 as notch strengthening or notch weakening and investigates the impact of multiaxial stress on Alloy 617 creep behavior. Multiaxial stress does not degrade Alloy 617’s short-term creep-rupture properties. This was concluded from the following observations: 1) multiaxial creep resided on or to the right of the Larson-Miller curve for uniaxial creep-rupture data, 2) Alloy 617 exhibited notch strengthening, and 3) creep-rupture life increased with a stronger multiaxial stress. Long-term and intermediate creep-rupture tests are in progress in order to determine whether notch behavior will cross from strengthening to weakening at lower stresses and longer rupture lifetimes. A nondestructive characterization methodology using X-ray computed tomography is explored as a means to identify the failure location on the specimen prior to rupture.

scholarly journals ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM

2016 ◽  
Vol 19 (2) ◽  
pp. 75
Author(s):  
Syarip, Khoirul Anam, Dwi Priyantoro
Keyword(s):  
High Temperature ◽  
Power Reactor ◽  
Operation Time ◽  
Reactor Power ◽  
Fuel Temperature ◽  
Reactor Operation ◽  
Control Rod ◽  
Negative Reactivity ◽  
Control Rods ◽  
High Temperature Gas

ANALISISPENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. ABSTRAK ANALISIS PENGATURAN POSISI CONTROL RODS PADA KONSEP REAKTOR DAYA EKSPERIMENTAL INDONESIA PASCA REACTOR SCRAM. Telah dilakukan analisis simulasi pengaturan posisi batang-batang kendali untuk melanjutkan operasi reaktor daya eksperimental (RDE) paska scram setelah beroperasi pada periode waktu tertentu. Pengendalian reaktivitas pada reaktor RDE yang akan dibangun di Indonesia dengan rujukan high temperature gas reactor (HTR) 10 MWt, dilakukan dengan 10  pasang batang-batang kendali atau control rod (CR). Apabila terrjadi kondisi abnormal maka CR secara otomatis akan jatuh tersisip ke dalam reflektor  reaktor sehingga reaktor scram dan berada pada kondisi subkritis. Untuk melanjutkan operasi reaktor pasca scram diperlukan analisis terkait pengaruh reaktivitas negatif dari Xenon dan suhu. Pada makalah ini disajikan hasil simulasi yang dilakukan untuk penentuan posisi CR paling optimum untuk melanjutkan operasi reaktor, menggunakan simulator PCTRAN-HTR. Simulasi dilakukan pada variasi 70%, 85% dan 100% dari tingkat daya penuh dan dengan variasi waktu operasi 50 s, 10.000 s, dan 20.000 s di mana setelah reaktor beroperasi pada tingkat-tingkat daya dan waktu operasi tersebut reaktor mengalami scram. Untuk melanjutkan operasi lagi maka CR harus dinaikkan lagi dan diatur ke posisi tertentu sampai   reaktor mencapai kondisi kritis lagi pada tingkat daya nominal tersebut. Hasil yang telah diperoleh menunjukkan bahwa dengan posisi CR naik 52 % sudah bisa menghasilkan kondisi kritis dan mampu mengatasi reaktivitas negatif peracunan xenon maupun suhu. Kata kunci: RDE, HTR, operasi reaktor, batang kendali, reaktivitas, scram ABSTRACT POST REACTOR SCRAM CONTROL RODS POSITION ADJUSTMENT ANALYSIS FOR THE INDONESIAN EXPERIMENTAL POWER REACTOR CONCEPT. Analytical study using PC-based simulator has been carried out on control rods position adjustment of the Indonesian experimental power reactor concept or reaktor daya ekperimental (RDE) in a post reactor scram to continue operation after a certain operation period. Reactivity control of the RDE uses 10 pairs of control rods (CRs), which is based on that applied in the high temperature gas reactor (HTR) 10 MW(t). If an abnormal operating condition occurs, these control rods automatically dropped to the reflector that bring the reactor into a scram and subcritical condition. To continue reactor operation after a period of time, the CRs should be withdrawn to achieve recriticality. Prior to any CRs withdrawal, an analysis of negative reactivity effects of Xenon (poissoning) and fuel temperature coefficient should be done. Simulations using PCTRAN-HTR simulator to determine the optimum CRs positions in achieving reactor criticality for continuation of reactor operation is presented in this paper. The simulations were conducted by varying the reactor power levels at 70%, 85% and 100% of full power, respectively. The reactor operation time was varied at 50s, 10000s, and 20000 s prior to the reactor scram. Adjustment of CRs position should be done to continue reactor operation at those nominal power levels by withdrawing the CRs to the proper positions. The simulation results show that recriticality can be achieverd by whitdrawing the CRs 52% of farther and the negative reactivity from xenon poisoning and temperature could be overcome. Keywords : RDE, HTR, reactor operation, control rod, reactivity, scram.

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