Results for “Software Verification” — Page 3

DOE OSTI2022

2021 Small Turbine Certification Awardee: Intergrid LLC

Currently, there are no U.S., fully certified inverters for wind turbines in the 10-15-kilowatt (kW) range. The federal government's Competitiveness Improvement Project (CIP) and Small Business Innovation Research programs present the most viable paths for manufacturers in the distributed wind energy industry to conduct research, develop their products, and, ultimately, commercialize them. Through CIP, Intergrid LLC is reducing the cost, size, and weight of its legacy inverter for distributed wind turbines while increasing efficiency, reliability, and serviceability. To accomplish this, Intergrid will address three areas related to power electronics for distributed wind turbines: (1) Software verification and functional safety certification; (2) Simulation for grid-interconnecting testing; (3) Component alternatives that reduce cost and improve reliability and serviceability. The first two areas relate to long-term management of inverter software, which, for grid-connected inverters, must comply with three standards - one issued by Underwriters Laboratories (UL) and two issued by the Institute of Electrical and Electronics Engineers.

CIP↗

DOE OSTI2016

Quantitative Analysis Guidance in Safety Software Validation

Quantitative analysis in software development has been a challenging topic over the years in safety software. While qualitative processes for vigorous safety software verification and validation (V and V) have been provided through standard such as IEEE 1012, quantitative analysis for V and V tasks such as correctness, completeness or accuracy evaluation have not been formally defined. In practice, there are quantitative analyses in terms of using software metrics. Many of these software metrics are useful in analyzing the test results. Some of them are used for providing guidance in code evaluation activities e.g. numbers of line of codes not exceeding certain limits, recursive calls are not allowed, etc. However, few can be used for providing quantitative analysis guidance for the V and V tasks such as completeness evaluation. AREVA Federal Services (AFS) has provided services for many nuclear projects in US. Through compliance with IEEE 1012, AFS V and V processes have provided vigorous qualitative processes ensuring quality level of the software products used in the project meeting the required nuclear software level. In addition, quantitative analysis is developed to provide structured validation framework ensuring 'zero defect' in the software products. This paper describes some aspects of the quantitative approaches using a sample safety application. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 1

This is Volume 1 of a two-volume document that collects the verification and validation (V&V) works done for the PLTEMP/ANL code during the years of its development and improvement. Volume 1 describes the V&V of sixteen capabilities of the PLTEMP/ANL code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis. Volume 2 describes the V&V of developments and improvements since the release of PLTEMP/ANL Version 4.3. Each chapter of the document focuses on verifying or validating a specific part of the software that calculates a specific phenomenon, e.g., channel flow calculation, coolant property calculation, heat transfer calculation, and flow instability calculation. Software verification is performed by comparing the code with hand calculation, Microsoft spreadsheet calculation, Mathematica calculation, or MATLAB calculation. The software validation is done by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL V&V works that are available in the open literature are simply cited in Volume 1 of the document. PLTEMP/ANL has been used in the safety analysis reports of several US and foreign research reactors licensed and converted from highly enriched uranium fuel to low-enriched uranium fuel. A list of such reactors is given in Volume 1 of the document.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors, Volume 2

This is Volume 2 of a two-volume document that collects the verification and validation (V&V) works done for the PLTEMP/ANL code during the years of its development and improvement. Volume 1 describes the V&V of sixteen capabilities of the PLTEMP/ANL code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis. Volume 2 describes the V&V of developments and improvements since the release of PLTEMP/ANL Version 4.3. Each chapter of the document focuses on verifying or validating a specific part of the software that calculates a particular phenomenon, e.g., channel flow calculation, coolant property calculation, heat transfer calculation, and flow instability calculation. Software verification is performed by comparing the code with a hand calculation, Microsoft spreadsheet calculation, Mathematica calculation, or MATLAB calculation. The software validation is achieved by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL V&V works that are available in the open literature are simply referenced in Volume 1 of the document. PLTEMP/ANL has been used in safety analysis reports of several US and foreign research reactors licensed and converted from highly enriched uranium fuel to low-enriched uranium fuel. A list of such reactors is given in Volume 1 of the document.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Code Coverage Status of ARC Code-DIF3D

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. DIF3D plays a pivotal role in the ARC system as the primary homogenized assembly neutronic calculation methodology for fast reactor problems. Over its 40 years history, ARC software usage with DIF3D has been applied to numerous fast and thermal spectrum reactor analysis projects with good to excellent comparison against experiments. With continued improvement of computation resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. In recent work, the DIF3D software verification was completed for DIF3D-FD and DIF3D-VARIANT on the geometry options used in the Versatile Test Reactor project. While we can be confident that these capabilities of DIF3D are well used and thus trusted, it does not demonstrate that all possible input options of DIF3D are actually working, but just those that were tested as part of VTR are and that they are correct. Thus, the purpose of the present work is to identify a set of test problems for DIF3D and assess the code coverage of DIF3D for those test problems. The goal is to document what parts of the existing DIF3D code are touched by the set of test problems and which are not. Because the verification work done on DIF3D-VARIANT and DIF3D-FD was focused on the most common uses of DIF3D for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that DIF3D relies upon. The DIF3D-Nodal capability will also be inspected for code coverage as part of this work to further ensure that regular regression testing of DIF3D will trap any likely errors the end user might experience with the DIF3D software. The code coverage analysis of DIF3D was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of DIF3D. The detailed coverage tables are given for each submodule of DIF3D separately, and for the submodules which are primarily developed for DIF3D, most of the source files could be at least partially touched. Most of the uncovered parts/files could be easily ignored, because they are either for error message and debugging output or obviously not needed by DIF3D. Out of the entire source codes of DIF3D, only a few uncovered modules deserve further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Code Coverage Status of the ARC Code DIF3D

The Argonne Reactor Code (ARC) software system supports users in their fast reactor design goals by providing neutronic, thermal-hydraulic, and structural analysis capabilities. DIF3D plays a pivotal role in the ARC system as the primary homogenized assembly neutronic calculation methodology for fast reactor problems. Over its 40 years history, ARC software usage with DIF3D has been applied to numerous fast and thermal spectrum reactor analysis projects with good to excellent comparison against experiments. With continued improvement of computation resources, many of the geometry modeling capabilities in DIF3D that were primarily used in low order schemes are not really needed anymore. Today, the diffusion and transport capabilities of DIF3D-VARIANT are primarily used in the reactor design process with some scattered usage of DIF3D-FD and DIF3D-Nodal. In recent work, the DIF3D software verification was completed for DIF3D-FD and DIF3D-VARIANT on the geometry options used in the Versatile Test Reactor project. While we can be confident that these capabilities of DIF3D are well used and thus trusted, it does not demonstrate that all possible input options of DIF3D are actually working, but just those that were tested as part of VTR are and that they are correct. Thus, the purpose of the present work is to identify a set of test problems for DIF3D and assess the code coverage of DIF3D for those test problems. The goal is to document what parts of the existing DIF3D code are touched by the set of test problems and which are not. Because the verification work done on DIF3D-VARIANT and DIF3D-FD was focused on the most common uses of DIF3D for fast reactor analysis, the code coverage assessment of those capabilities is the highest priority. This will ensure that nothing is being missed by the existing verification test problems that DIF3D relies upon. The DIF3D-Nodal capability will also be inspected for code coverage as part of this work to further ensure that regular regression testing of DIF3D will trap any likely errors the end user might experience with the DIF3D software. The code coverage analysis of DIF3D was performed with the Code Coverage Tool of the Intel Fortran compiler which requires modifications to the compilation of DIF3D. The detailed coverage tables are given for each submodule of DIF3D separately, and for the submodules which are primarily developed for DIF3D, most of the source files could be at least partially touched. Most of the uncovered parts/files could be easily ignored, because they are either for error message and debugging output or obviously not needed by DIF3D. Out of the entire source codes of DIF3D, only a few uncovered modules deserve further investigation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

An operation verification strategy against network attack of DCS in nuclear power plant - 323

Digital Instrumentation and control (I and C) system serves as the 'senses and nervous system' of a nuclear power plant (NPP). It plays an important role in the safe and economic operation in the NPPs. Digital I and C system has a series of advantages compared with analogue ones, but the introduction of the operation system platform, the application software, and the network communication structure are causing security risk to the power plant. Different kinds of procedures have already been adopted against these risks such as the communication isolation, but the security attacks still occur from time to time in the industrial fields around the world. This paper presents two computer security methods based on the thought of software verification and validation with step-by-step operation identification. First, three scenarios of security attack are proposed according to the structure of a distributed control system (DCS). Then a method with two network bypasses and a security computer are adopted to obtain the operation commands, the actuation signal, and the operation feedback in case of the first two scenarios by identifying the consistency of the obtained data. Furthermore, the method is optimized by combining the operation procedure and the camera shot to identify the anomaly of the information flow, thus to improve the effectiveness against the computer security issues. In the end, the influence and feasibility of this method is discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

SU-E-T-49: A Multi-Institutional Study of Independent Dose Verification for IMRT

Purpose: AAPM TG114 does not cover the independent verification for IMRT. We conducted a study of independent dose verification for IMRT in seven institutes to show the feasibility. Methods: 384 IMRT plans in the sites of prostate and head and neck (HN) were collected from the institutes, where the planning was performed using Eclipse and Pinnacle3 with the two techniques of step and shoot (S&S) and sliding window (SW). All of the institutes used a same independent dose verification software program (Simple MU Analysis: SMU, Triangle Product, Ishikawa, JP), which is Clarkson-based and CT images were used to compute radiological path length. An ion-chamber measurement in a water-equivalent slab phantom was performed to compare the doses computed using the TPS and an independent dose verification program. Additionally, the agreement in dose computed in patient CT images between using the TPS and using the SMU was assessed. The dose of the composite beams in the plan was evaluated. Results: The agreement between the measurement and the SMU were −2.3±1.9 % and −5.6±3.6 % for prostate and HN sites, respectively. The agreement between the TPSs and the SMU were −2.1±1.9 % and −3.0±3.7 for prostate and HN sites, respectively. There was a negative systematic difference with similar standard deviation and the difference was larger in the HN site. The S&S technique showed a statistically significant difference between the SW. Because the Clarkson-based method in the independent program underestimated (cannot consider) the dose under the MLC. Conclusion: The accuracy would be improved when the Clarkson-based algorithm should be modified for IMRT and the tolerance level would be within 5%.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

SU-E-T-229: Craniospinal Radiotherapy Planning with VMAT, Two First Years Experience

Purpose: To describe how we moved to VMAT in the craniospinal radiotherapy planning process, the actual procedure details, and the results for the patients treated. Methods: Twelve patients underwent craniospinal irradiation with the new procedure, based on the paper by Lee et al. (IJROBP 82, 2012), with some additional modifications. Patients were treated in supine position in Varian Clinac iX linacs with 6 MV RapidArc; prescription doses ranged from 23.4 to 40 Gy (13 to 20 fractions); depending on the PTV length, 2 or 3 isocenters were used, all coordinates being equal except the longitudinal one, setting a few centimeter-long overlapping region; 2 arcs (RA) sharing isocentre for the cranial region, RA1 encompassing cranium and superior spinal region, and RA2 intended to improve conformity, only for cranium; for spine, 1 or 2 isocenters were employed; optimization was performed with Eclipse (V 13.0) using AAA algorithm, establishing sets of optimization parameters to give high conformity while sparing OAR. In pediatric patients, homogeneous irradiation of the vertebrae was also required.Conformity (CI) and heterogeneity (HI) indices (same as Lee et al.), and mean and maximum doses for OAR were calculated. Several pre-treatment verification methods were used: Octavius4D (PTW) for each isocentre, point dose at the junction region, Portal Dosimetry (when possible), and independent MU verification software (Diamond, PTW). Results: CI median value was 1.02 (0.99–1.07) and HI, 1.07 (1.06–1.09); a great reduction was observed for CI and OAR mean doses with respect to Lee et al. data; median maximum eye lens dose was 7.3 Gy (4.0–12.0); mean LungV20Gy was 1.9%; in children, vertebrae were homogeneously irradiated (D95%=20.8 Gy, Dmean= 23.2 Gy).All pre-treatment verifications were found within our action levels except for Portal Dosimetry. Conclusion: A RapidArc planning process for craniospinal axis irradiation has been implemented with significant advantages on conformity, homogeneity, feasibility and efficiency. and increase brain tissue sparing. Variations in volume decrease may be related to shape or location of the tumor.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2024

Gamma Source Verification for GAMSRC and GAMSOR

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interactions. The GAMSOR program was first built in the mid 1980s to properly account for the gamma heating in an operating reactor core on core internals. The GAMSOR code is sequence of DIF3D calculations to compute the neutron and gamma flux and combine them to define both the neutron and gamma heating throughout the modeled domain. The goal of this manuscript is to present the software verification of GAMSOR. The first step of the GAMSOR sequence of calculations involves of a modified version of DIF3D (called DIF3D-GAMSOR) which generates a gamma source distribution for the follow-on DIF3D gamma transport calculation (step 2). This modified version of DIF3D increases the burden of maintenance and verification work on GAMSOR as one must reverify the DIF3D capabilities which is undesirable. Because the calculation of the gamma source is the only unique aspect of GAMSOR beyond the regular DIF3D capabilities, that part was put in a standalone code called GAMSRC such that one can use the verified DIF3D code in step 1 followed by GAMSRC to carry out the same GAMSOR calculation step. As a consequence, this manuscript is focused on verification of the gamma source files generated by GAMSRC. The verification of the modified version of DIF3D (DIF3D-GAMSOR) will be done less rigorously in that it will be verified that it produces the same output that GAMSRC does and thus GAMSRC is equivalent to GAMSOR on the problems studied here. Hand calculations and independent numerical calculations of the gamma source generation are used for the verification work. This work follows the same methodology of GAMSRC. As expected, the results agree well with those calculated by GAMSRC as will be shown.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Gamma Source Verification for GAMSRC and GAMSOR

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interactions. The GAMSOR program was first built in the mid 1980s to properly account for the gamma heating in an operating reactor core on core internals. The GAMSOR code is sequence of DIF3D calculations to compute the neutron and gamma flux and combine them to define both the neutron and gamma heating throughout the modeled domain. The goal of this manuscript is to present the software verification of GAMSOR. The first step of the GAMSOR sequence of calculations involves of a modified version of DIF3D (called DIF3D-GAMSOR) which generates a gamma source distribution for the follow-on DIF3D gamma transport calculation (step 2). This modified version of DIF3D increases the burden of maintenance and verification work on GAMSOR as one must reverify the DIF3D capabilities which is undesirable. Because the calculation of the gamma source is the only unique aspect of GAMSOR beyond the regular DIF3D capabilities, that part was put in a standalone code called GAMSRC such that one can use the verified DIF3D code in step 1 followed by GAMSRC to carry out the same GAMSOR calculation step. As a consequence, this manuscript is focused on verification of the gamma source files generated by GAMSRC. The verification of the modified version of DIF3D (DIF3D-GAMSOR) will be done less rigorously in that it will be verified that it produces the same output that GAMSRC does and thus GAMSRC is equivalent to GAMSOR on the problems studied here. Hand calculations and independent numerical calculations of the gamma source generation are used for the verification work. This work follows the same methodology of GAMSRC. As expected, the results agree well with those calculated by GAMSRC as will be shown.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2024

Verification of EvaluateFLux Utility Program

The EvaluateFlux program is a post-processing utility program for DIF3D, specifically DIF3D-VARIANT which handles Cartesian and hexagonal geometries. The EvaluateFlux program was developed to allow users to obtain flux and power traverses through the geometry domain, and its initial purpose was to facilitate foil analysis by evaluating the flux solution from DIF3D-VARIANT and combining it with foil cross section data. The EvaluateFlux program can calculate the neutron flux, as well as the reaction rates, at any user provided evaluation point. It does this by identifying the spatial mesh associated with the evaluation point and then evaluates the polynomial based neutron flux moments stored in the NHFLUX file at that point. The output of EvaluateFlux varies depending on the input setup. The maximum output includes the neutron flux and microscopic and macroscopic reaction rates at each evaluation point. The purpose of this work is to verify the outputs of EvaluateFlux. Simple models that have hand calculatable results are first defined and used to verify the EvaluateFlux outputs. More complex cases are then added where a duplicate program of EvaluateFlux that uses PrintTables outputs of the binary files is used to verify the EvaluateFlux outputs. In those complex cases, hand calculations of selected evaluation points were also displayed to confirm the software verification. For all the tests done, the hand calculations agreed well with those calculated by EvaluateFlux. For the larger complex problems, the duplicate program that can process hundreds of evaluation points was able to identify that zero points within some meshes have large errors. This aspect was attributed to the truncation error on the input provided to the duplicate program and is not a concern for the accuracy of the EvaluateFlux software.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2024

Verification of the EvaluateFlux Utility Program

The EvaluateFlux program is a post-processing utility program for DIF3D, specifically DIF3D-VARIANT which handles Cartesian and hexagonal geometries. The EvaluateFlux program was developed to allow users to obtain flux and power traverses through the geometry domain, and its initial purpose was to facilitate foil analysis by evaluating the flux solution from DIF3D-VARIANT and combining it with foil cross section data. The EvaluateFlux program can calculate the neutron flux, as well as the reaction rates, at any user provided evaluation point. It does this by identifying the spatial mesh associated with the evaluation point and then evaluates the polynomial based neutron flux moments stored in the NHFLUX file at that point. The output of EvaluateFlux varies depending on the input setup. The maximum output includes the neutron flux and microscopic and macroscopic reaction rates at each evaluation point. The purpose of this work is to verify the outputs of EvaluateFlux. Simple models that have hand calculatable results are first defined and used to verify the EvaluateFlux outputs. More complex cases are then added where a duplicate program of EvaluateFlux that uses PrintTables outputs of the binary files is used to verify the EvaluateFlux outputs. In those complex cases, hand calculations of selected evaluation points were also displayed to confirm the software verification. For all the tests done, the hand calculations agreed well with those calculated by EvaluateFlux. For the larger complex problems, the duplicate program that can process hundreds of evaluation points was able to identify that zero points within some meshes have large errors. This aspect was attributed to the truncation error on the input provided to the duplicate program and is not a concern for the accuracy of the EvaluateFlux software.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

A new ensemble-based consistency test for the Community Earth System Model (pyCECT v1.0)

Abstract. Climate simulation codes, such as the Community Earth System Model (CESM), are especially complex and continually evolving. Their ongoing state of development requires frequent software verification in the form of quality assurance to both preserve the quality of the code and instill model confidence. To formalize and simplify this previously subjective and computationally expensive aspect of the verification process, we have developed a new tool for evaluating climate consistency. Because an ensemble of simulations allows us to gauge the natural variability of the model's climate, our new tool uses an ensemble approach for consistency testing. In particular, an ensemble of CESM climate runs is created, from which we obtain a statistical distribution that can be used to determine whether a new climate run is statistically distinguishable from the original ensemble. The CESM ensemble consistency test, referred to as CESM-ECT, is objective in nature and accessible to CESM developers and users. The tool has proven its utility in detecting errors in software and hardware environments and providing rapid feedback to model developers.

58 GEOSCIENCES↗

DOE OSTI2014

SU-E-T-65: Characterization of a 2D Array for QA and Pretreatment Plan Verification

Purpose: The OCTAVIUS detector729 is a 2D array of 729 air vented cubic plane parallel ion chambers used for pretreatment verification and QA. In this study we investigated dosimetric characteristics of this system for clinical photon beam dosimetry. Methods: Detector performance evaluation included determination of the location of the effective point of measurement (EPM), sensitivity, linearity, and reproducibility of detector response, as well as output factor, dose rate, and source to surface distance (SSD) dependence. Finally, assessment of wedge modulated fields was carried out. All the evaluations were performed five times for low and high photon energies. For reference measurements, a 0.6 cc ionization chamber was used. Data analysis and comparison of the OCTAVIUS detector with reference ion chamber data was performed using the VeriSoft patient plan verification software. Results: The reproducibility and stability of the measurements are excellent, the detector showed same signal with a maximum deviation of less than 0.5% in short and long term. Results of sensitivity test showed same signal with a maximum deviation of approximately 0.1%. As the detector 729 response is linear with dose and dose rate, it can be used for the measurement at regions of high dose gradient effectively. The detector agrees with the ionization chamber measurement to within 1% for SSD range of 75 to 125 cm. Also, its measured wedge modulated profiles matched very well with ion chamber dose profiles acquired in a water tank. Conclusions: As the response of the detector 729 is linear with dose and dose rate, it can be used for the measurements in the areas of dose gradients effectively. Based on the measurements and comparisons performed, this system is a reliable and accurate dosimeter for QA and pretreatment plan verification in radiotherapy.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

SU-E-T-762: Toward Volume-Based Independent Dose Verification as Secondary Check

Purpose: Lung SBRT plan has been shifted to volume prescription technique. However, point dose agreement is still verified using independent dose verification at the secondary check. The volume dose verification is more affected by inhomogeneous correction rather than point dose verification currently used as the check. A feasibility study for volume dose verification was conducted in lung SBRT plan. Methods: Six SBRT plans were collected in our institute. Two dose distributions with / without inhomogeneous correction were generated using Adaptive Convolve (AC) in Pinnacle3. Simple MU Analysis (SMU, Triangle Product, Ishikawa, JP) was used as the independent dose verification software program, in which a modified Clarkson-based algorithm was implemented and radiological path length was computed using CT images independently to the treatment planning system. The agreement in point dose and mean dose between the AC with / without the correction and the SMU were assessed. Results: In the point dose evaluation for the center of the GTV, the difference shows the systematic shift (4.5% ± 1.9 %) in comparison of the AC with the inhomogeneous correction, on the other hands, there was good agreement of 0.2 ± 0.9% between the SMU and the AC without the correction. In the volume evaluation, there were significant differences in mean dose for not only PTV (14.2 ± 5.1 %) but also GTV (8.0 ± 5.1 %) compared to the AC with the correction. Without the correction, the SMU showed good agreement for GTV (1.5 ± 0.9%) as well as PTV (0.9% ± 1.0%). Conclusion: The volume evaluation for secondary check may be possible in homogenous region. However, the volume including the inhomogeneous media would make larger discrepancy. Dose calculation algorithm for independent verification needs to be modified to take into account the inhomogeneous correction.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2016

SU-F-T-288: Impact of Trajectory Log Files for Clarkson-Based Independent Dose Verification of IMRT and VMAT

Purpose: To investigate the effect of the trajectory files from linear accelerator for Clarkson-based independent dose verification in IMRT and VMAT plans. Methods: A CT-based independent dose verification software (Simple MU Analysis: SMU, Triangle Products, Japan) with a Clarksonbased algorithm was modified to calculate dose using the trajectory log files. Eclipse with the three techniques of step and shoot (SS), sliding window (SW) and Rapid Arc (RA) was used as treatment planning system (TPS). In this study, clinically approved IMRT and VMAT plans for prostate and head and neck (HN) at two institutions were retrospectively analyzed to assess the dose deviation between DICOM-RT plan (PL) and trajectory log file (TJ). An additional analysis was performed to evaluate MLC error detection capability of SMU when the trajectory log files was modified by adding systematic errors (0.2, 0.5, 1.0 mm) and random errors (5, 10, 30 mm) to actual MLC position. Results: The dose deviations for prostate and HN in the two sites were 0.0% and 0.0% in SS, 0.1±0.0%, 0.1±0.1% in SW and 0.6±0.5%, 0.7±0.9% in RA, respectively. The MLC error detection capability shows the plans for HN IMRT were the most sensitive and 0.2 mm of systematic error affected 0.7% dose deviation on average. Effect of the MLC random error did not affect dose error. Conclusion: The use of trajectory log files including actual information of MLC location, gantry angle, etc should be more effective for an independent verification. The tolerance level for the secondary check using the trajectory file may be similar to that of the verification using DICOM-RT plan file. From the view of the resolution of MLC positional error detection, the secondary check could detect the MLC position error corresponding to the treatment sites and techniques. This research is partially supported by Japan Agency for Medical Research and Development (AMED)

60 APPLIED LIFE SCIENCES↗

DOE OSTI2015

SU-E-T-48: A Multi-Institutional Study of Independent Dose Verification for Conventional, SRS and SBRT

Purpose: To show the results of a multi-institutional study of the independent dose verification for conventional, Stereotactic radiosurgery and body radiotherapy (SRS and SBRT) plans based on the action level of AAPM TG-114. Methods: This study was performed at 12 institutions in Japan. To eliminate the bias of independent dose verification program (Indp), all of the institutions used the same CT-based independent dose verification software (Simple MU Analysis, Triangle Products, JP) with the Clarkson-based algorithm. Eclipse (AAA, PBC), Pinnacle{sup 3} (Adaptive Convolve) and Xio (Superposition) were used as treatment planning system (TPS). The confidence limits (CL, Mean±2SD) for 18 sites (head, breast, lung, pelvis, etc.) were evaluated in comparison in dose between the TPS and the Indp. Results: A retrospective analysis of 6352 treatment fields was conducted. The CLs for conventional, SRS and SBRT were 1.0±3.7 %, 2.0±2.5 % and 6.2±4.4 %, respectively. In conventional plans, most of the sites showed within 5 % of TG-114 action level. However, there were the systematic difference (4.0±4.0 % and 2.5±5.8 % for breast and lung, respectively). In SRS plans, our results showed good agreement compared to the action level. In SBRT plans, the discrepancy between the Indp was variable depending on dose calculation algorithms of TPS. Conclusion: The impact of dose calculation algorithms for the TPS and the Indp affects the action level. It is effective to set the site-specific tolerances, especially for the site where inhomogeneous correction can affect dose distribution strongly.

60 APPLIED LIFE SCIENCES↗