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At least 1,405 records · Page 78
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Simulation evaluation of NIST air-kerma rate calibration standard for electronic brachytherapy: Simulation of NIST air-kerma rate eBT calibration standard

Purpose: Dosimetry for the model S700 50 kV electronic brachytherapy (eBT) source (Xoft, Inc., a subsidiary of iCAD, San Jose, CA) was simulated using Monte Carlo (MC) methods by Rivard et al. [“Calculated and measured brachytherapy dosimetry parameters in water for the Xoft Axxent x-ray source: An electronic brachytherapy source,” Med. Phys. 33, 4020–4032 (2006)] and recently by Hiatt et al. [“A revised dosimetric characterization of the model S700 electronic brachytherapy source containing an anode-centering plastic insert and other components not included in the 2006 model,” Med. Phys. 42, 2764–2776 (2015)] with improved geometric characterization. Additionally, while these studies examined the dose distribution in water, there have not previously been reports of the eBT source calibration methods beyond that recently reported by Seltzer et al. [“New national air-kerma standard for low-energy electronic brachytherapy sources,” J. Res. Natl. Inst. Stand. Technol. 119, 554–574 (2014)]. Therefore, the motivation for the current study was to provide an independent determination of air-kerma rate at 50 cm in air K air (d = 50 cm) using MC methods for the model S700 eBT source. Methods: Using CAD information provided by the vendor and disassembled sources, an MC model was created for the S700 eBT source. Simulations were run using the mcnp6 radiation transport code for the NIST Lamperti air ionization chamber according to specifications by Boutillon et al. [“Comparison of exposure standards in the 10-50 kV x-ray region,” Metrologia 5, 1–11 (1969)], in air without the Lamperti chamber, and in vacuum without the Lamperti chamber. K air (d = 50 cm) was determined using the *F4 tally with NIST values for the mass energy-absorption coefficients for air. Photon spectra were evaluated over 2π azimuthal sampling for polar angles of 0° ≤ θ ≤ 180° every 1°. Volume averaging was averted through tight radial binning. Photon energy spectra were determined over all polar angles in both air and vacuum using the F4 tally with 0.1 keV resolution. A total of 10 11 simulated histories were run for the Lamperti chamber geometry (statistical uncertainty of 0.14%), with 10 10 histories for the in-air and in-vacuum simulations (statistical uncertainty of 0.04%). The total standard uncertainty in the calculated air-kerma rate determination amounted to 6.8%. Results: MC simulations determined the air-kerma rate at 50 cm from the source with the modeled Lamperti chamber to be (1.850 ± 0.126) × 10 -4 Gy/s, which was within the range of K air (d = 50 cm) values (1.67–2.11) × 10 -4 Gy/s measured by NIST. The ratio of the photon spectra in air and in vacuum were in good agreement above 13 keV, and for θ < 150° where the influence of the Kovar sleeve and the Ag epoxy components caused increased scatter in air. Below 13 keV, the ratio of the photon spectra in air to vacuum exhibited a decrease that was attributed to increased attenuation of the photons in air. Across most of the energy range on the source transverse plane, there was good agreement between the authors' simulated spectra and that measured by NIST. Discrepancies were observed above 40 keV where the NIST spectrum had a steeper fall-off towards 50 keV. Conclusions: Through MC simulations of radiation transport, this study provided an independent validation of the measured air-kerma rate at 50 cm in air at NIST for the model S700 eBT source, with mean results in agreement within 3.3%. Lastly, this difference was smaller than the range (i.e., 23%) of the measured values.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2026

Measuring the Stress Factors for Photovoltaic (PV) Backsheet Degradation

Back sheet failure has resulted in power loss and large-scale recall of photovoltaic modules, resulting in billions of dollars in lost revenue. The light exposure on the backside of a photovoltaic module comes primarily from reflected light which alters the distribution of natural sunlight. Because of this, modelling the backside exposure and duplicating the exposure is much more difficult than modeling the frontside exposure. This project aims to study how various back sheets and junction box materials degrade under different conditions and to develop Python code to help model and predict degradation. The stress factors for back-sheet degradation must be quantified to extrapolate accelerated stress tests to the field. Test samples were placed in the A3, A4, and A5 conditions, as defined in IEC 62788-7-2, to assess the temperature and humidity dependence of ultraviolet (UV) induced degradation. We are utilizing a custom chamber with exposure from 0.5 UV-suns to 5 UV-suns to understand the dependence of degradation on light intensity. A group of samples put in the A3 condition had glass filters with 50% UV cut-offs of 320 nm, 335 nm, and 360 nm to assess the wavelength dependence of UV degradation. All this data is necessary to assess the impact of non-standard UV light exposure. The material evaluation tests include gloss measurements, attenuated total internal reflectance Fourier transform infrared spectroscopy (ATR-FTIR), UV-visible reflectance/transmittance utilizing a Cary Ci7000 spectrophotometer, and a nano-indenter for surface hardness and modulus measurements. Alongside the experimental work, there is a computational effort using raytracing and Python open-source tools in PVDeg , PVLib, and Bifacial_Radiance. This code will create specific exposure scenarios and enable the evaluation of chamber degradation relative to field degradation. Equation 1 is a strawman equation used to model degradation on the backside of a PV module. We will create simplified code, based on the results of ray-tracing calculations, which uses a view factor approach to provide fast calculations for the most common exposure scenarios.

14 SOLAR ENERGY↗
DOE OSTI2022

First-collision source treatment for ray effect mitigation in discrete-ordinate radiation transport solutions

Deterministic transport codes play a fundamental role in the modelling and simulation of neutron transport. One of the most common deterministic methods is the method of discrete ordinates, also known as the S method. While offering significant advantages over other deterministic methods or stochastic methods like Monte Carlo, the method of discrete ordinates suffers from non-physical artifacts in its local solution due to its discretization of angle. These artifacts, referred to as ray effects because of their ray-like appearance, tend to be worse in problems with small sources in areas with little scattering. Significant effort has gone into developing methods to mitigate ray effects, such as the first-collision source treatment, which separates the angular flux into the uncollided and collided fluxes and solving them using non-traditional techniques such as ray tracing. One such code capable of doing this is Lawrence Livermore National Laboratory's deterministic transport code ARDRA. Current ray tracing methods typically trace to a set of points inside a zone to compute an overall flux. However, this approach has significant drawbacks, such as a low order of convergence and not being conservative. Therefore, a new method has been developed that traces instead to a set of points on each of a zone's surfaces and computing the currents, before using these to obtain the flux. A comparison between these two ray tracing methods showed significant advantages to the new surface method, including inherent conservation, a higher convergence rate, and an increase in calculable information like leakage. This work performed under the auspices of the U.S. Department of Energy by Lawrence Liver- more National Laboratory under Contract DE-AC52-07NA27344. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Characterization of sub-picosecond laser-produced fast electrons by modeling angularly resolved bremsstrahlung measurements with 3D hybrid particle-in-cell code

Hard x-rays produced by intense laser-produced fast electrons interacting with solids are a vital source for producing radiographs of high-density objects and implosion cores for inertial confinement fusion. Accurate calculation of hard x-ray sources requires a three-dimensional (3D) simulation geometry that fully models the electron transport dynamics, including electron recirculation and the generation of absolute photon yields. To date, 3D simulations of laser-produced bremsstrahlung photons over tens of picoseconds and code benchmarking have not been performed definitively. In this study, we characterize sub-picosecond laser-produced fast electrons by modeling angularly resolved bremsstrahlung measurements for refluxing and non-refluxing targets using the 3D hybrid particle-in-cell (PIC), Large Scale Plasma code. Bremsstrahlung radiation and escaped electron data were obtained by focusing a 50-TW Leopard laser (15 J, 0.35 ps, 2 × 10 19 W/cm 2 ) on a 100-μm-thick Cu foil and a Cu with a large plastic backing (Cu–CH target). Data for both the Cu and Cu–CH targets were reproduced for simulations with a given set of electron parameters. Comparison of the simulations revealed that the hard x-ray emission from the Cu target was significantly longer in duration than that from the Cu–CH target. Here, the benchmarked hybrid PIC code could prove to be a powerful tool in the design and optimization of time- and angular-dependent bremsstrahlung sources for flash x-ray and gamma-ray radiography.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Add-a-Source Analysis Method for Neutron Coincidence Counting of Waste: Laboratory Exercises

In this session you will setup the INCC code to perform calibration and verification measurements on simulated waste drums using the Waste Drum Assay System (WDAS) detector, which uses the add-a-source (AAS) method. The basis of the AAS method is to measure the change in the counting rate from a small 252 Cf source (~3x10 4 n/s) located outside of the drum and to use this information to correct the count rate from material inside the drum for any changes caused by the matrix inside the drum. The count rate of the material inside the drum is corrected to the empty drum case so that only a single mass calibration curve has to be measured (for the empty drum configuration). In the WDAS system, the AAS neutron source has one location, centered beneath the waste drum. The sample matrix has two primary effects on the neutrons: (1) energy reduction by scattering reactions and (2) neutron absorption of the low-energy neutrons. In general, the counter is designed with an optimum moderator (CH 2 ) thickness to be relatively insensitive to the energy reduction; however, as the hydrogen density in the drum increases, the adsorption process significantly reduces the measured neutron signal.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2015

Optimization of the beam extraction systems for the Linac4 H{sup −} ion source

The development of the Linac 4 and its integration into CERN’s acceleration complex is part of the foreseen luminosity upgrade of the Large Hadron Collider (LHC). The goal is to inject a 160 MeV H{sup −} beam into the CERN PS Booster (PSB) in order to increase the beam brightness by a factor of 2 compared to the 50 MeV proton linac, Linac 2, that is currently in operation. The requirements for the ion source are a 45 keV H{sup −} beam of 80 mA intensity, 2 Hz repetition rate and 0.5 ms pulse length within a normalized rms-emittance of 0.25 mm· mrad. The previously installed beam extraction system has been designed for an H{sup −} ion beam intensity of 20 mA produced by an RF-volume source with an electron to H{sup −} ratio of up to 50. For the required intensity upgrades of the Linac4 ion source, a new beam extraction system is being produced and tested; it is optimized for a cesiated surface RF-source with a nominal beam current of 40 mA and an electron to H{sup −} ratio of 4. The simulations, based on the IBSIMU code, are presented. At the Brookhaven National Laboratory (BNL), a peak beam current of more than 100 mA was demonstrated with a magnetron H{sup −} source at an energy of 35 keV and a repetition rate of 2 Hz. A new extraction system is required to operate at an energy of 45 keV; simulation of a two stage extraction system dedicated to the magnetron is presented.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

A variable-order time-dependent neutron transport method for nuclear reactor kinetics using analytically-integrated space-time characteristics

A new time-dependent neutron transport method based on the method of characteristics (MOC) has been developed. Whereas most spatial kinetics methods treat time dependence through temporal discretization, this new method treats time dependence by defining the characteristics to span space and time. In this implementation regions are defined in space-time where the thickness of the region in time fulfills an analogous role to the time step in discretized methods. The time dependence of the local source is approximated using a truncated Taylor series expansion with high order derivatives approximated using backward differences, permitting the solution of the resulting space-time characteristic equation. To avoid a drastic increase in computational expense and memory requirements due to solving many discrete characteristics in the space-time planes, the temporal variation of the boundary source is similarly approximated. This allows the characteristics in the space-time plane to be represented analytically rather than discretely, resulting in an algorithm comparable in implementation and expense to one that arises from conventional time integration techniques. Furthermore, by defining the boundary flux time derivative in terms of the preceding local source time derivative and boundary flux time derivative, the need to store angularly-dependent data is avoided without approximating the angular dependence of the angular flux time derivative. The accuracy of this method is assessed through implementation in the neutron transport code DeCART. The method is employed with variable-order local source representation to model a TWIGL transient. The results demonstrate that this method is accurate and more efficient than the discretized method. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

V2XConnect: Harmonizing the Landscape of Bidirectional Charging Codes, Standards, and Communication Protocols

The utility grid is constantly evolving with new generation sources and diverse loads from backup generators, backup batteries, rooftop photovoltaics (PV), and flexible charging loads. Each of these technologies presents consumers with an opportunity to work with their local utility to promote grid stability, reliability, and affordability. Through a wide range of programs, utilities are incentivizing the use of these resources to benefit the grid. However, especially for devices with inverters that supply power back to the grid, it is important that these assets comply with local grid codes to ensure human safety, power quality, and grid reliability. One of the most promising of these new technologies is the bidirectional electric vehicle (EV) coupled with bidirectional electric vehicle supply equipment (EVSE). This report introduces the technologies that enable bidirectional charging and how these systems operate. It then outlines the necessary interconnection codes and industry standards that govern the operations of grid-tied inverters, including those in bidirectional charging. A key challenge is then outlined from the complex landscape of communication protocols and proprietary systems designed to coordinate these grid assets. Finally, a single gateway solution is proposed to harmonize communications between all assets across an entire site for a cohesive response to grid codes and utility programs that is flexible enough to support a range of market options.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2022

Using Calibrated Water Data for Preliminary Validation of the SRT Code for Advanced Reactors

Recent interest and corresponding progress worldwide regarding advanced nuclear reactors has renewed focus on their performance and related safety assessments. Specifically, the U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic approaches to source term analysis for advanced reactor licensing applications, which attempt to realistically account radionuclide transport and retention phenomena. Further model development is required due to the numerous and complex physical and chemical phenomena associated with mechanistic source term analyses. Reflecting the need for modeling advancement, Argonne National Laboratory developed a mechanistic source term analysis tool for sodium fast reactors. The Simplified Radionuclide Transport (SRT) code describes fuel pin failure (for simulating the initial condition at the point of fuel pin breach), bubble scrubbing, deposition, leakage and following environmental impact. In the current work, a validation study of the SRT bubble scrubbing model is performed using a water-loop experiment performed at the University of Wisconsin-Madison. Through the analysis, the approach and fundamental bubble scrubbing models in SRT, which examine the removal of aerosols within the bubble as it is transported through a pool, have been widely evaluated. The results of the assessment demonstrate a high level of agreement in the regions of greater aerosol size. In the parameter range of minimal aerosol removal, the simulation slightly underpredicts the experiment results; however, considering the scale of plots and huge uncertainties inherently included, the deviation can be judged to be minor and would produce a conservative result. In addition, uncertainty analysis has been further refined to reflect the experimental distribution of parameters including aerosol sizes, which induces a span of performance for each representative aerosol size. Based upon the initial validation results along with uncertainty effects, SRT is expected to provide meaningful insights for the analysis of bubble scrubbing. Future sodium-loop tests will provide further validation basis.

Kam, Dong Hoon↗
DOE OSTI2016

Technical Note: Effect of explicit M and N-shell atomic transitions on a low-energy x-ray source

Purpose: In EGSnrc, atomic transitions to and from the M and N-shells are treated in an average way by default. This approach is justified in which the energy difference between explicit and average M and N-shell binding energies is less than 1 keV, and for most applications can be considered negligible. However, for simulations of low energy x-ray sources on thin, high-Z targets, characteristic x-rays can make up a significant portion of the source spectra. As of release V4-2.4.0, EGSnrc has included an option to enable a more complete algorithm of all atomic transitions available in the EADL compilation. In this paper, the effect of M and N-shell averaging on the calculation of half-value layer (HVL) and relative depth dose (RDD) curve of a 50 kVp intraoperative x-ray tube with a thin gold target was investigated. Methods: A 50 kVp miniature x-ray source with a gold target (The INTRABEAM System, Carl Zeiss, Germany) was modeled with the EGSnrc user code cavity, both with and without M and N-shell averaging. From photon fluence spectra simulations, the source HVLs were determined analytically. The same source model was then used with egs-chamber to calculate RDD curves in water. Results: A 4% increase of HVL was reported when accounting for explicit M and N-shell transitions, and up to a 9% decrease in local relative dose for normalization at 3 mm depth in water. Conclusions: The EGSnrc default of using averaged M and N-shell binding energies has an observable effect on the HVL and RDD of a low energy x-ray source with high-Z target. For accurate modeling of this class of devices, explicit atomic transitions should be included.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Energy Savings Analysis of the Proposed Revision of the Washington D.C. Non-Residential Energy Code

This report presents the results of an assessment of savings for the proposed Washington D.C. energy code relative to ASHRAE Standard 90.1-2010. It includes annual and life cycle savings for site energy, source energy, energy cost, and carbon dioxide emissions that would result from adoption and enforcement of the proposed code for newly constructed buildings in Washington D.C. over a five year period.

Commercial Energy Code↗
DOE OSTI2015

Study of two different radioactive sources for prostate brachytherapy treatment

In this study we evaluated two radioactive sources for brachytherapy treatments. Our main goal was to quantify the absorbed doses on organs and tissues of an adult male patient, submitted to a brachytherapy treatment with two radioactive sources. We evaluated a {sup 192}Ir and a {sup 125}I radioactive sources. The {sup 192}Ir radioactive source is a cylinder with 0.09 cm in diameter and 0.415 cm long. The {sup 125}I radioactive source is also a cylinder, with 0.08 cm in diameter and 0.45 cm long. To evaluate the absorbed dose distribution on the prostate, and other organs and tissues of an adult man, a male virtual anthropomorphic phantom MASH, coupled in the radiation transport code MCNPX 2.7.0, was employed.We simulated 75, 90 and 102 radioactive sources of {sup 125}I and one of {sup 192}Ir, inside the prostate, as normally used in these treatments, and each treatment was simulated separately. As this phantom was developed in a supine position, the displacement of the internal organs of the chest, compression of the lungs and reduction of the sagittal diameter were all taken into account. For the {sup 192}Ir, the higher doses values were obtained for the prostate and surrounding organs, as the colon, gonads and bladder. Considering the {sup 125}I sources, with photons with lower energies, the doses to organs that are far from the prostate were lower. All values for the dose rates are in agreement with those recommended for brachytherapy treatments. Besides that, the new seeds evaluated in this work present usefulness as a new tool in prostate brachytherapy treatments, and the methodology employed in this work may be applied for other radiation sources, or treatments. (authors)

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Griffin Software Development Plan

Griffin is a MOOSE-based reactor physics application for advanced reactor multiphysics modeling and simulation. The application is developed in a consistent multiphysics environment with strong software quality assurance. Griffin inherited most of the capabilities of MAMMOTH/Rattlesnake and is adopting the capabilities from PROTEUS that are needed in the code. The toolset includes a variety of deterministic radiation transport solvers for fixed source, k-eigenvalue, ad-joint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code contains the cross-section preparation capabilities applicable to fast and thermal reactors, including TRISO-fueled reactors. Core management capabilities include core performance, fuel depletion and shuffling, equilibrium core calculation, pebble-bed reactor run-in and equilibrium core, molten-salt reactor delayed neutron precursor drift, and control rod and drum movement with cusping correction. This software development plan presents the current and future capabilities and features in Griffin for the design and analysis of non-light-water reactor systems in steady-state and transient conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Simulation of ion beam transport through the 400 Kv ion implanter at Michigan Ion Beam Laboratory

The Michigan Ion Beam Laboratory houses a 400 kV ion implanter. An application that simulates the ion beam trajectories through the implanter from the ion source to the target was developed using the SIMION Registered-Sign code. The goals were to have a tool to develop an intuitive understanding of abstract physics phenomena and diagnose ion trajectories. Using this application, new implanter users of different fields in science quickly understand how the machine works and quickly learn to operate it. In this article we describe the implanter simulation application and compare the parameters of the implanter components obtained from the simulations with the measured ones. The overall agreement between the simulated and measured values of magnetic fields and electric potentials is {approx}10%.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Testing the Activation Analysis for Fusion in OpenMC

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It can perform fission simulations such as fixed-source, k-eigenvalue, and subcritical multiplication calculations on models built using either a constructive solid geometry or CAD representation. To explore the use of OpenMC for fusion activation analysis, a detailed model of the Fusion Neutronics Science Facility (FNSF) was first developed for comparisons against an existing SERPENT model. A 90-degree model of FNSF in Standard-Triangle-Language (STL) CAD format was converted to Constructive Solid Geometry (CSG) using each code's built-in functions, and the geometries were validated by ensuring no cells overlapped and no particles were lost during simulations. The neutron fluxes were calculated and compared for multiple components close to the plasma. The results show differences mostly below 1% in fluxes and averaged 8% for activity and decay heat. Here, the work described in this study tests the CAD-based geometry using the DagMC toolkit in OpenMC and compares the activation analysis of OpenMC to SERPENT code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

X-ray microlaminography with polycapillary optics

We demonstrate layer-by-layer x-ray microimaging using polycapillary optics. The depth resolution is achieved without sample or source rotation and in a way similar to classical tomography or laminography. The method takes advantage from large angular apertures of polycapillary optics and from their specific microstructure, which is treated as a coded aperture. The imaging geometry is compatible with polychromatic x-ray sources and with scanning and confocal x-ray fluorescence setups.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

HDRMC, an accelerated Monte Carlo dose calculator for high dose rate brachytherapy with CT-compatible applicators

Purpose: To present a new accelerated Monte Carlo code for CT-based dose calculations in high dose rate (HDR) brachytherapy. The new code (HDRMC) accounts for both tissue and nontissue heterogeneities (applicator and contrast medium). Methods: HDRMC uses a fast ray-tracing technique and detailed physics algorithms to transport photons through a 3D mesh of voxels representing the patient anatomy with applicator and contrast medium included. A precalculated phase space file for the{sup 192}Ir source is used as source term. HDRM is calibrated to calculated absolute dose for real plans. A postprocessing technique is used to include the exact density and composition of nontissue heterogeneities in the 3D phantom. Dwell positions and angular orientations of the source are reconstructed using data from the treatment planning system (TPS). Structure contours are also imported from the TPS to recalculate dose-volume histograms. Results: HDRMC was first benchmarked against the MCNP5 code for a single source in homogenous water and for a loaded gynecologic applicator in water. The accuracy of the voxel-based applicator model used in HDRMC was also verified by comparing 3D dose distributions and dose-volume parameters obtained using 1-mm{sup 3} versus 2-mm{sup 3} phantom resolutions. HDRMC can calculate the 3D dose distribution for a typical HDR cervix case with 2-mm resolution in 5 min on a single CPU. Examples of heterogeneity effects for two clinical cases (cervix and esophagus) were demonstrated using HDRMC. The neglect of tissue heterogeneity for the esophageal case leads to the overestimate of CTV D90, CTV D100, and spinal cord maximum dose by 3.2%, 3.9%, and 3.6%, respectively. Conclusions: A fast Monte Carlo code for CT-based dose calculations which does not require a prebuilt applicator model is developed for those HDR brachytherapy treatments that use CT-compatible applicators. Tissue and nontissue heterogeneities should be taken into account in modern HDR brachytherapy planning.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

SCALE Shielding Calculations for Advanced Reactor Accident Scenarios

The study results presented in this paper demonstrate the capabilities of the SCALE computer code for non–light water reactor (non-LWR) radiation source term and shielding calculations in support of the U.S. Nuclear Regulatory Commission non-LWR fuel cycle demonstration project. Representative non-LWR types, including the sodium-cooled fast reactor (SFR), the molten salt reactor (MSR), and the heat pipe microreactor (HPMR), were analyzed to evaluate dose rates associated with postulated accident scenarios (SFR and MSR), reactor operation (MSR and HPMR), and fuel self-protecting characteristics (HPMR). New features were implemented in SCALE depletion codes to better simulate MSR operation.

MSR↗