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Sample records for reactor dosimetric evaluation

  1. Dosimetric evaluation of semiconductor detectors for application in neutron dosimetry and microdosimetry in nuclear reactor and radiosurgical facilities

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio Nahuel

    2010-01-01

    The main objective of this research is the dosimetric evaluation of semiconductor components (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB - Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation methodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established. (author)

  2. Epithermal beam development at the BMRR [Brookhaven Medical Research Reactor]: Dosimetric evaluation

    International Nuclear Information System (INIS)

    Saraf, S.K.; Fairchild, R.G.; Kalef-Ezra, J.; Laster, B.H.; Fiarman, S.; Ramsey, E.; Ioannina Univ.; Brookhaven National Lab., Upton, NY; State Univ. of New York, Stony Brook, NY

    1989-01-01

    The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al 2 O 3 ) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs

  3. Internal dosimetric evaluation due to uranium aerosols

    International Nuclear Information System (INIS)

    Garcia Aguilar Juan; Delgado Avila Gustavo

    1991-01-01

    The present work has like object to carry out the internal dosimetric evaluation to the occupationally exposed personnel, due to the inhalation of aerosols of natural uranium and enriched in the pilot plant of nuclear fuel production of the National Institute of Nuclear Research

  4. Dosimetric evaluation of proton stereotactic radiosurgery

    International Nuclear Information System (INIS)

    Min, Byung Jun; Shin, Dong Ho; Yoo, Seung Hoon; Jeong, Hojin; Lee, Se Byeong

    2011-01-01

    Surgical excision, conventional external radiotherapy, and chemotherapy could prolong survival in patients with small intracranial tumors. However, surgical excision for meningiomas located in the region of the base of skull or re-resection is often difficult. Moreover, treatment is needed for patients with recurrent tumors or postoperative residual tumors. Conventional external radiotherapy is popular and has significantly increased for treating brain tumors. Stereotactic radiosurgery is an effective alternative treatment technique to microsurgical resection such as benign brain tumor or vestibular Schwannomas. In general, the dose to OAR of 3D conformal plan is lower than that of conformal arc and dynamic conformal arc plans. However, any of OARs was not reached to tolerance dose. Although mean dose of the healthy brain tissue for 3D conformal plan was slightly higher than that of arc plans, the doses of the healthy brain tissue at V10 and V20 were significantly low for dynamic conformal arc plan. The dosimetric differences were the greatest at lower doses. In contrast, 3D conformal plan was better spare at higher doses. In this study, a dosimetric evaluation of proton stereotactic radiosurgery for brain lesion tumors was using fixed and arc beams. A brass block fitted to the PTV structure was modeled for dynamic conformal collimator. Although all treatment plans offer a very good coverage of the PTV, we found that proton arc plans had significantly better conformity to the PTV than static 3D conformal plan. The V20 dose of normal brain for dynamic conformal arc therapy is dramatically reduced compare to those for other therapy techniques.

  5. Dosimetric evaluation program for dental radiology practices

    International Nuclear Information System (INIS)

    Gregori, B.; Milat, J.; Fernandez, J.; Micinquevich, S.; Andrieu, J.

    1992-01-01

    The preliminary results of a program undertaken to estimate the doses to patients associated with dental radiology practices in Argentine, are presented. Information collected from the search demonstrated that the Dieck and coronal techniques are the most commonly used practices, while all the examinations are performed by using a circular collimator. For both practices, the dosimetric studies were carried out on a Rando Alderson phantom. All dose measurements were made using thermoluminescent detectors LiF and Ca 2 F. In addition, a mathematical model was developed by applying the Monte Carlo method to a MIRD-V phantom. Circular and rectangular collimators were used. Absorbed dose distribution on head and neck, as well as surface dose distribution, were estimated. The comparison of the performance of both collimators shows that the use of the rectangular one allows for a dose reduction of 80%. Besides, a good correlation between the physical and mathematical models applied was found. (author)

  6. Dosimetric evaluation of mammary tomosynthesis procedures

    International Nuclear Information System (INIS)

    Silva, Rayre Janaína Vieira; Perini, Ana Paula; Santos, William de Souza; Vedovato, Uly P.; Neves, Lucio Pereira

    2017-01-01

    This work presents the results of the research on the evaluation of radiation doses usually applied in mammary procedures, using the Monte Carlo method. A virtual environment was created, to mimic the procedures room, including the room, its components, patient and source. The spectrum was obtained from the literature. The percentage of energy deposited compared to energy deposited in the breast was determined, and the scattered radiation was absorbed in specific areas. The regions of the head and neck were the most affected by scattered radiation. (author)

  7. Evaluation of reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-15

    Although the operation of nuclear reactors has a remarkably good record of safety, the prevention of possible reactor accidents is one of the major factors that atomic planners have to contend with. At the same time, excessive caution may breed an attitude that hampers progress, either by resisting new development or by demanding unnecessarily elaborate and expensive precautions out of proportion to the actual hazards involved. The best course obviously is to determine the possible dangers and adopt adequate measures for their prevention, providing of course, for a reasonable margin of error in judging the hazards and the effectiveness of the measures. The greater the expert understanding and thoroughness with which this is done, the narrower need the margin be. This is the basic idea behind the evaluation of reactor safety

  8. Numerical reactor evaluation

    International Nuclear Information System (INIS)

    Venter, A.M.

    1973-08-01

    A short discussion is given of the physics of a nuclear reactor and the parameters which are used in the study of neutron transport. The mathematical formulation and detailed derivation is given of the neutron diffusion and transport equations. A description is given of the computer programmes, FIRE-5 and PELSN, developed at Pelindaba for the evaluation of both thermal and fast reactor systems. It is indicated how these computer programmes have been applied in the study of the PELINDUNA-O and other known critical facilities. The application of Lie-series to the solution of the neutron diffusion equation is discussed in detail. The time dependence of the variables is removed by means of a Laplacetransformation and the semi-analytical solution is written in terms of a transfer matrix. A complete set of recursion formulae, applicable to both homogeneous and heterogeneous reactor systems, is derived. The method used in the evaluation of the effective multiplication factor, k-eff, and the alpha-eigen-value is described. A computer programme was written to solve the neutron diffusion equation in terms of the Lie-series. The results are compared with the TIMOC and PELSN computer programmes. A method is suggested in which the Lie-series are used to solve the neutron transport equation. The transfer matrix for this case, is derived. A complete discussion is given of the solution to the space and time dependent diffusion equation in the presence of a delta source [af

  9. Dosimetric evaluation of whole-breast radiation therapy: Clinical experience

    International Nuclear Information System (INIS)

    Osei, Ernest; Darko, Johnson; Fleck, Andre; White, Jana; Kiciak, Alexander; Redekop, Rachel; Gopaul, Darin

    2015-01-01

    Radiation therapy of the intact breast is the standard therapy for preventing local recurrence of early-stage breast cancer following breast conservation surgery. To improve patient standard of care, there is a need to define a consistent and transparent treatment path for all patients that reduces significance variations in the acceptability of treatment plans. There is lack of consistency among institutions or individuals about what is considered an acceptable treatment plan: target coverage vis-à-vis dose to organs at risk (OAR). Clinical trials usually resolve these issues, as the criteria for an acceptable plan within the trial (target coverage and doses to OAR) are well defined. We developed an institutional criterion for accepting breast treatment plans in 2006 after analyzing treatment data of approximately 200 patients. The purpose of this article is to report on the dosimetric review of 623 patients treated in the last 18 months to evaluate the effectiveness of the previously developed plan acceptability criteria and any possible changes necessary to further improve patient care. The mean patient age is 61.6 years (range: 25.2 to 93.0 years). The mean breast separation for all the patients is 21.0 cm (range: 12.4 to 34.9 cm), and the mean planning target volume (PTV-eval) (breast volume for evaluation) is 884.0 cm"3 (range: 73.6 to 3684.6 cm"3). Overall, 314 (50.4%) patients had the disease in the left breast and 309 (49.6%) had it in the right breast. A total of 147 (23.6%) patients were treated using the deep inspiration breath-hold (DIBH) technique. The mean normalized PTV-eval receiving at least 92% (V_9_2_% _P_D) and 95% (V_9_5_% _P_D) of the prescribed dose (PD) are more than 99% and 97%, respectively, for all patients. The mean normalized PTV-eval receiving at least 105% (V_1_0_5_% _P_D) of the PD is less than 1% for all groups. The mean homogeneity index (HI), uniformity index (UI), and conformity index (CI) for the PTV-eval are 0.09 (range: 0

  10. Dosimetric evaluation of Radiotherapy units wit 60Co

    International Nuclear Information System (INIS)

    Leon, B. Salinas de; Tovar M, V.; Becerril V, A.

    2000-01-01

    The SSDL network of the IAEA performs, every year, quality audit tests for radiotherapy services ( 60 Co units and linear accelerators), and for national SSDL as well. Because of the SSDL-Mexico results in these tests and due to our enthusiasm and confidence in our work, a parallel test has been done , which is described in this talk as well as the results. Nowadays, a second parallel test goes up, which could confirm our optimism and open the possibility to our country to start a national dosimetric audit of 60 Co radiotherapy units. (Author)

  11. Internal dosimetric evaluation due to uranium aerosols; Evaluacion dosimetrica interna debido a aerosoles de uranio

    Energy Technology Data Exchange (ETDEWEB)

    Juan, Garcia Aguilar; Gustavo, Delgado Avila [Instituto Nacional de Investigaciones Nucleares, Salazar (Mexico)

    1991-07-01

    The present work has like object to carry out the internal dosimetric evaluation to the occupationally exposed personnel, due to the inhalation of aerosols of natural uranium and enriched in the pilot plant of nuclear fuel production of the National Institute of Nuclear Research.

  12. Evaluation of Specific Absorption Rate as a Dosimetric Quantity for Electromagnetic Fields Bioeffects

    OpenAIRE

    Panagopoulos, Dimitris J.; Johansson, Olle; Carlo, George L.

    2013-01-01

    PURPOSE: To evaluate SAR as a dosimetric quantity for EMF bioeffects, and identify ways for increasing the precision in EMF dosimetry and bioactivity assessment. METHODS: We discuss the interaction of man-made electromagnetic waves with biological matter and calculate the energy transferred to a single free ion within a cell. We analyze the physics and biology of SAR and evaluate the methods of its estimation. We discuss the experimentally observed non-linearity between electromagnetic exposu...

  13. Dosimetric quantities and basic data for the evaluation of generalised derived limits

    International Nuclear Information System (INIS)

    Harrison, N.T.; Simmonds, J.R.

    1980-12-01

    The procedures, dosimetric quantities and basic data to be used for the evaluation of Generalised Derived Limits (GDLs) in environmental materials and of Generalised Derived Limits for discharges to atmosphere are described. The dosimetric considerations and the appropriate intake rates for both children and adults are discussed. In most situations in the nuclear industry and in those institutions, hospitals and laboratories which use relatively small quantities of radioactive material, the Generalised Derived Limits provide convenient reference levels against which the results of environmental monitoring can be compared, and atmospheric discharges can be assessed. They are intended for application when the environmental contamination or discharge to atmosphere is less than about 5% of the Generalised Derived Limit; above this level, it will usually be necessary to undertake a more detailed site-specific assessment. (author)

  14. Use of VAP3D software in the construction of pathological anthropomorphic phantoms for dosimetric evaluations

    International Nuclear Information System (INIS)

    Lima, Lindeval Fernandes de; Lima, Fernando R.A.

    2011-01-01

    This paper performs a new type of dosimetric evaluation, where it was used a phantom of pathological voxels (representative phantom of sick person). The software VAP3D (Visualization and Analysis of Phantoms 3D) were used for, from a healthy phantom (phantom representative of healthy person), to introduce three dimensional regions to simulate tumors. It was used the Monte Carlo ESGnrc code to simulate the X ray photon transport, his interaction with matter and evaluation of absorbed dose in organs and tissues from thorax region of the healthy phantom and his pathological version. This is a computer model of typical exposure for programming the treatments in radiodiagnostic

  15. Dosimetric evaluation in panoramic and tele-radiography procedures

    International Nuclear Information System (INIS)

    Oliveira, Georgge Gomes

    2004-01-01

    The present work had as an objective to evaluate the skin surface entrance dose in panoramic and tele radiography procedures in three clinics in Recife - Pernambuco - Brazil, and to contribute with data for the determination of reference levels for super cited extra oral procedures, for this purpose, operational conditions in 3 clinics were evaluated in Recife, aiming to evaluate the existence and integrity of the radioprotection equipment, manner and conditions of image processing; and radiographic equipment parameters such as the dimension of the irradiation filed, the total filtration, the exposure time and the potential applied to the X ray tube. For an estimation of the skin entrance dose of the patient, the phantom Randon Alderson and thermoluminescence dosemeters were used. From these values and the conversion factors determined by the Monte Carlo technique, with the phantom MAX it was possible to estimate the dose absorbed in the organ due to the tele radiography procedures. Regarding panoramic radiography the study showed that the more elevated doses occurred in the parotid gland region which is near rotational venters. In the case of tele radiography the highest dose value occurred in the regions corresponding to the temporal lobe of the brain, followed by linfonodes, ears and parotid glands. The doses absorbed in the eyes and the thyroid gland were, 0.037 mGy and 0.002 mGy in Clinic A and 0.062 mGy and 0.003 mGy in Clinic C, respectively. Regarding equipment test, inadequacy was found in the beam collimation in Clinic A and in the reproducibility of the X ray exposure in Clinic C. The total filtration in both clinics was inadequate.(author)

  16. Dosimetric evaluation of cone beam computed tomography scanning protocols

    International Nuclear Information System (INIS)

    Soares, Maria Rosangela

    2015-01-01

    It was evaluated the cone beam computed tomography, CBCT scanning protocols, that was introduced in dental radiology at the end of the 1990's, and quickly became a fundamental examination for various procedures. Its main characteristic, the difference of medical CT is the beam shape. This study aimed to calculate the absorbed dose in eight tissues / organs of the head and neck, and to estimate the effective dose in 13 protocols and two techniques (stitched FOV e single FOV) of 5 equipment of different manufacturers of cone beam CT. For that purpose, a female anthropomorphic phantom was used, representing a default woman, in which were inserted thermoluminescent dosimeters at several points, representing organs / tissues with weighting values presented in the standard ICRP 103. The results were evaluated by comparing the dose according to the purpose of the tomographic image. Among the results, there is a difference up to 325% in the effective dose in relation to protocols with the same image goal. In relation to the image acquisition technique, the stitched FOV technique resulted in an effective dose of 5.3 times greater than the single FOV technique for protocols with the same image goal. In the individual contribution, the salivary glands are responsible for 31% of the effective dose in CT exams. The remaining tissues have also a significant contribution, 36%. The results drew attention to the need of estimating the effective dose in different equipment and protocols of the market, besides the knowledge of the radiation parameters and equipment manufacturing engineering to obtain the image. (author)

  17. Evaluation of Torsatrons as reactors

    International Nuclear Information System (INIS)

    Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

    1994-03-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio

  18. Dosimetric evaluation of fattening filter free photon beams

    International Nuclear Information System (INIS)

    Lechner, W.

    2014-01-01

    Very recently, conventional treatment machines able to deliver flattening filter free photon have been introduced into clinical practice. An Elekta Pricise linear accelerator LINAC which was able to deliver 6MV and 10 MV flattened (FF) and flattening filter free (FFF) photon beams was used throughout this work. The investigated modalities were 9-Field intensity modulated radiotherapy (Intensity modulated radiotherapy (IMRT)) and 360° - single arc volumetric modulated arc therapy (Volumetric modulated arc therapy (VMAT)) with flattened and unflattened photon beams. For treatment plan evaluation the concept of Pareto optimal fronts was employed. Additionally, the efficiency of these modalities was assessed. The evaluation of the treatment plan quality showed no significant difference between FF- and FFF-beams. With respect to treatment plan efficiency, a significant decrease of delivery time of IMRT treatment plans without flattening filter compared to those with filter was found. On average, the delivery time decreased by 18% and 4% for prostate and head-and-neck cases, respectively. In contrast to that, the delivery time of VMAT treatment plans without flatting filter was significantly increased by 22% and 16% for prostate and head- and-neck cases, respectively. The applicability of 14 different detectors to small field dosimetry in FF- as well as in FFF-beams was investigated by measuring output factors of 6MV and 10 MV FF- and FFF- beams. An additional MLC was attached to the treatment machine in order to generate field sizes between 0.6x0.6 cm 2 and 10x10 cm 2 . Alanine pellets were used as reference detectors for the calculation of correction factors for small field dosimetry. Compared to alanine, the solid sate detectors and the liquid filled ionization chamber generally overestimated the output factors of small fields, whereas the air filled ionzation chambers underestimated the output factors. For the shielded diodes the correction factors ranged between 8

  19. Development and dosimetric evaluation of radiochromic PCDA vesicle gel dosimeters

    International Nuclear Information System (INIS)

    Sun, P.; Fu, Y.C.; Hu, J.; Hao, N.; Huang, W.; Jiang, B.

    2016-01-01

    The gel dosimeter has the unique capacity in recording radiation dose distribution in three dimensions (3D), which has the specific advantages in dosimetry measurements where steep dose gradients exist, such as in intensity-modulated radiation therapy (IMRT), brachytherapy and so on. Some 3D dosimeters, such as Fricke gel dosimeters, polymer gel dosimeters, the PRESAGE plastic dosimeters and micelle gel dosimeters have appeared recently. However, there are several disadvantages of these 3D dosimeters limit their application in radiotherapy dose verification. In this study, a novel radiochromic gel dosimeter for 3D dose verification of radiotherapy was developed by dispersing nanovesicles self-assembled by 10,12-pentacosadiynoic acid (PCDA) into the tissue equivalence gel matrix. The characteristics of radiochromic PCDA vesicle gel dosimeters were evaluated. The results indicate that these radiochromic gel dosimeters have good linear dose response to X-ray irradiation in the dose range of 2–100 Gy. In addition, the radiochromic gel dosimeters breakthrough the limitations of the existing gel dosimeters such as diffusion effect, post-radiation effect, and poor forming ability. The response of the gel dosimeter does not show any dose rate dependence, energy dependence and temperature effect, and there was no obvious difference in the gel response between single and cumulative dose of fractional irradiation. Hence, the radiochromic PCDA vesicle gel dosimeters developed in this study could be generally applied to 3D dose verification in radiotherapy. - Highlights: • A novel radiochromic gel dosimeter was developed by dispersing PCDA nanovesicles into the tissue equivalence gel matrix. • This nanovesicle overcomes the dose image blurring caused by the diffusion of monomer molecules. • This nanovesicle limits the polymer chain growth, so as to reduce the post-radiation effect. • The gel matrixes possess excellent tissue equivalence and elastic strength, which

  20. Dosimetric evaluation of a novel polymer gel dosimeter for proton therapy

    Energy Technology Data Exchange (ETDEWEB)

    Zeidan, O. A.; Sriprisan, S. I.; Lopatiuk-Tirpak, O.; Kupelian, P. A.; Meeks, S. L.; Hsi, W. C.; Li, Z.; Palta, J. R.; Maryanski, M. J. [M. D. Anderson Cancer Center Orlando, Orlando, Florida 32806 (United States); University of Florida Proton Therapy Institute, Jacksonville, Florida 32206 (United States); MGS Research, Inc., Madison, Connecticut 06443 (United States)

    2010-05-15

    Purpose: The aim of this study is to evaluate the dosimetric performance of a newly developed proton-sensitive polymer gel formulation for proton therapy dosimetry. Methods: Using passive scattered modulated and nonmodulated proton beams, the dose response of the gel was assessed. A next-generation optical CT scanner is used as the readout mechanism of the radiation-induced absorbance in the gel medium. Comparison of relative dose profiles in the gel to ion chamber profiles in water is performed. A simple and easily reproducible calibration protocol is established for routine gel batch calibrations. Relative stopping power ratio measurement of the gel medium was performed to ensure accurate water-equivalent depth dose scaling. Measured dose distributions in the gel were compared to treatment planning system for benchmark irradiations and quality of agreement is assessed using clinically relevant gamma index criteria. Results: The dosimetric response of the gel was mapped up to 600 cGy using an electron-based calibration technique. Excellent dosimetric agreement is observed between ion chamber data and gel. The most notable result of this work is the fact that this gel has no observed dose quenching in the Bragg peak region. Quantitative dose distribution comparisons to treatment planning system calculations show that most (>97%) of the gel dose maps pass the 3%/3 mm gamma criterion. Conclusions: This study shows that the new proton-sensitive gel dosimeter is capable of reproducing ion chamber dose data for modulated and nonmodulated Bragg peak beams with different clinical beam energies. The findings suggest that the gel dosimeter can be used as QA tool for millimeter range verification of proton beam deliveries in the dosimeter medium.

  1. Dosimetric evaluation of a novel polymer gel dosimeter for proton therapy

    International Nuclear Information System (INIS)

    Zeidan, O. A.; Sriprisan, S. I.; Lopatiuk-Tirpak, O.; Kupelian, P. A.; Meeks, S. L.; Hsi, W. C.; Li, Z.; Palta, J. R.; Maryanski, M. J.

    2010-01-01

    Purpose: The aim of this study is to evaluate the dosimetric performance of a newly developed proton-sensitive polymer gel formulation for proton therapy dosimetry. Methods: Using passive scattered modulated and nonmodulated proton beams, the dose response of the gel was assessed. A next-generation optical CT scanner is used as the readout mechanism of the radiation-induced absorbance in the gel medium. Comparison of relative dose profiles in the gel to ion chamber profiles in water is performed. A simple and easily reproducible calibration protocol is established for routine gel batch calibrations. Relative stopping power ratio measurement of the gel medium was performed to ensure accurate water-equivalent depth dose scaling. Measured dose distributions in the gel were compared to treatment planning system for benchmark irradiations and quality of agreement is assessed using clinically relevant gamma index criteria. Results: The dosimetric response of the gel was mapped up to 600 cGy using an electron-based calibration technique. Excellent dosimetric agreement is observed between ion chamber data and gel. The most notable result of this work is the fact that this gel has no observed dose quenching in the Bragg peak region. Quantitative dose distribution comparisons to treatment planning system calculations show that most (>97%) of the gel dose maps pass the 3%/3 mm gamma criterion. Conclusions: This study shows that the new proton-sensitive gel dosimeter is capable of reproducing ion chamber dose data for modulated and nonmodulated Bragg peak beams with different clinical beam energies. The findings suggest that the gel dosimeter can be used as QA tool for millimeter range verification of proton beam deliveries in the dosimeter medium.

  2. Dosimetric evaluation of a MOSFET detector for clinical application in photon therapy.

    Science.gov (United States)

    Kohno, Ryosuke; Hirano, Eriko; Nishio, Teiji; Miyagishi, Tomoko; Goka, Tomonori; Kawashima, Mitsuhiko; Ogino, Takashi

    2008-01-01

    Dosimetric characteristics of a metal oxide-silicon semiconductor field effect transistor (MOSFET) detector are studied with megavoltage photon beams for patient dose verification. The major advantages of this detector are its size, which makes it a point dosimeter, and its ease of use. In order to use the MOSFET detector for dose verification of intensity-modulated radiation therapy (IMRT) and in-vivo dosimetry for radiation therapy, we need to evaluate the dosimetric properties of the MOSFET detector. Therefore, we investigated the reproducibility, dose-rate effect, accumulated-dose effect, angular dependence, and accuracy in tissue-maximum ratio measurements. Then, as it takes about 20 min in actual IMRT for the patient, we evaluated fading effect of MOSFET response. When the MOSFETs were read-out 20 min after irradiation, we observed a fading effect of 0.9% with 0.9% standard error of the mean. Further, we applied the MOSFET to the measurement of small field total scatter factor. The MOSFET for dose measurements of small field sizes was better than the reference pinpoint chamber with vertical direction. In conclusion, we assessed the accuracy, reliability, and usefulness of the MOSFET detector in clinical applications such as pinpoint absolute dosimetry for small fields.

  3. Dosimetric evaluation of the interplay effect in respiratory-gated RapidArc radiation therapy

    International Nuclear Information System (INIS)

    Riley, Craig; Yang, Yong; Li, Tianfang; Zhang, Yongqian; Heron, Dwight E.; Huq, M. Saiful

    2014-01-01

    Purpose: Volumetric modulated arc therapy (VMAT) with gating capability has had increasing adoption in many clinics in the United States. In this new technique, dose rate, gantry rotation speed, and the leaf motion speed of multileaf collimators (MLCs) are modulated dynamically during gated beam delivery to achieve highly conformal dose coverage of the target and normal tissue sparing. Compared with the traditional gated intensity-modulated radiation therapy technique, this complicated beam delivery technique may result in larger dose errors due to the intrafraction tumor motion. The purpose of this work is to evaluate the dosimetric influence of the interplay effect for the respiration-gated VMAT technique (RapidArc, Varian Medical Systems, Palo Alto, CA). Our work consisted of two parts: (1) Investigate the interplay effect for different target residual errors during gated RapidArc delivery using a one-dimensional moving phantom capable of producing stable sinusoidal movement; (2) Evaluate the dosimetric influence in ten clinical patients’ treatment plans using a moving phantom driven with a patient-specific respiratory curve. Methods: For the first part of this study, four plans were created with a spherical target for varying residual motion of 0.25, 0.5, 0.75, and 1.0 cm. Appropriate gating windows were applied for each. The dosimetric effect was evaluated using EDR2 film by comparing the gated delivery with static delivery. For the second part of the project, ten gated lung stereotactic body radiotherapy cases were selected and reoptimized to be delivered by the gated RapidArc technique. These plans were delivered to a phantom, and again the gated treatments were compared to static deliveries by the same methods. Results: For regular sinusoidal motion, the dose delivered to the target was not substantially affected by the gating windows when evaluated with the gamma statistics, suggesting the interplay effect has a small role in respiratory-gated Rapid

  4. Dosimetric evaluation of the interplay effect in respiratory-gated RapidArc radiation therapy.

    Science.gov (United States)

    Riley, Craig; Yang, Yong; Li, Tianfang; Zhang, Yongqian; Heron, Dwight E; Huq, M Saiful

    2014-01-01

    Volumetric modulated arc therapy (VMAT) with gating capability has had increasing adoption in many clinics in the United States. In this new technique, dose rate, gantry rotation speed, and the leaf motion speed of multileaf collimators (MLCs) are modulated dynamically during gated beam delivery to achieve highly conformal dose coverage of the target and normal tissue sparing. Compared with the traditional gated intensity-modulated radiation therapy technique, this complicated beam delivery technique may result in larger dose errors due to the intrafraction tumor motion. The purpose of this work is to evaluate the dosimetric influence of the interplay effect for the respiration-gated VMAT technique (RapidArc, Varian Medical Systems, Palo Alto, CA). Our work consisted of two parts: (1) Investigate the interplay effect for different target residual errors during gated RapidArc delivery using a one-dimensional moving phantom capable of producing stable sinusoidal movement; (2) Evaluate the dosimetric influence in ten clinical patients' treatment plans using a moving phantom driven with a patient-specific respiratory curve. For the first part of this study, four plans were created with a spherical target for varying residual motion of 0.25, 0.5, 0.75, and 1.0 cm. Appropriate gating windows were applied for each. The dosimetric effect was evaluated using EDR2 film by comparing the gated delivery with static delivery. For the second part of the project, ten gated lung stereotactic body radiotherapy cases were selected and reoptimized to be delivered by the gated RapidArc technique. These plans were delivered to a phantom, and again the gated treatments were compared to static deliveries by the same methods. For regular sinusoidal motion, the dose delivered to the target was not substantially affected by the gating windows when evaluated with the gamma statistics, suggesting the interplay effect has a small role in respiratory-gated RapidArc therapy. Varied results were

  5. TL and LOE dosimetric evaluation of diamond films exposed to beta and ultraviolet radiation

    International Nuclear Information System (INIS)

    Preciado F, S.; Melendrez, R.; Chernov, V.; Barboza F, M.; Schreck, M.; Cruz Z, E.

    2005-01-01

    The diamond possesses a privileged position regarding other materials of great technological importance. Their applications go from the optics, microelectronics, metals industry, medicine and of course as dosemeter, in the registration and detection of ionizing and non ionizing radiation. In this work the results of TL/LOE obtained in two samples of diamond of 10 μm thickness grown by the chemical vapor deposition method (CVD) assisted by microwave plasma. The films were deposited in a silicon substrate (001) starting from a mixture of gases composed of CH 4 /H 2 and 750 ppm of molecular nitrogen as dopant. The samples were exposed to beta radiation (Sr 90 / Y 90 ) and ultraviolet, being stimulated later on thermal (TL) and optically (LOE) to evaluate their dosimetric properties. The sample without doping presented high response TL/LOE to the ultraviolet and beta radiation. The TL glow curve of the sample without doping showed two TL peaks with second order kinetics in the range of 520 to 550 K, besides a peak with first order kinetics of more intensity around 607 K. The TL efficiency of the non doped sample is bigger than the doped with nitrogen; however the LOE efficiency is similar in both samples. The results indicate that the CVD diamond possesses excellent perspectives for dosimetric applications, with special importance in radiotherapy due to it is biologically compatible with the human tissue. (Author)

  6. Dosimetric evaluation of tomography and four-box field conformal radiotherapy in locally advanced rectal cancer

    International Nuclear Information System (INIS)

    Yu, Mina; Lee, Hyo Chun; Chung, Mi Joo; Kim, Sung Hwan; Lee, Jong Hoon; Jang, Hong Seok; Jeon, Dong Min; Cheon, Geum Seong

    2013-01-01

    To report the results of dosimetric comparison between intensity-modulated radiotherapy (IMRT) using Tomotherapy and four-box field conformal radiotherapy (CRT) for pelvic irradiation of locally advanced rectal cancer. Twelve patients with locally advanced rectal cancer who received a short course preoperative chemoradiotherapy (25 Gy in 5 fractions) on the pelvis using Tomotherapy, between July 2010 and December 2010, were selected. Using their simulation computed tomography scans, Tomotherapy and four-box field CRT plans with the same dose schedule were evaluated, and dosimetric parameters of the two plans were compared. For the comparison of target coverage, we analyzed the mean dose, Vn Gy, Dmin, Dmax, radical dose homogeneity index (rDHI), and radiation conformity index (RCI). For the comparison of organs at risk (OAR), we analyzed the mean dose. Tomotherapy showed a significantly higher mean target dose than four-box field CRT (p 0.001). But, V26.25 Gy and V27.5 Gywere not significantly different between the two modalities. Tomotherapy showed higher Dmax and lower Dmin. The Tomotherapy plan had a lower rDHI than four-box field CRT (p = 0.000). Tomotherapy showed better RCI than four-box field CRT (p = 0.007). For OAR, the mean irradiated dose was significantly lower in Tomotherapy than four-box field CRT. In locally advanced rectal cancer, Tomotherapy delivers a higher conformal radiation dose to the target and reduces the irradiated dose to OAR than four-box field CRT.

  7. Dosimetric evaluation of tomography and four-box field conformal radiotherapy in locally advanced rectal cancer

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Mina; Lee, Hyo Chun; Chung, Mi Joo; Kim, Sung Hwan; Lee, Jong Hoon [Dept. of Radiation Oncology, St. Vincent' s Hospital, The Catholic University of Korea College of Medicine, Suwon (Korea, Republic of); Jang, Hong Seok; Jeon, Dong Min; Cheon, Geum Seong [Dept. of Radiation Oncology, Seoul St. Mary' s Hospital, The Catholic University of Korea College of Medicine, Seoul (Korea, Republic of)

    2013-12-15

    To report the results of dosimetric comparison between intensity-modulated radiotherapy (IMRT) using Tomotherapy and four-box field conformal radiotherapy (CRT) for pelvic irradiation of locally advanced rectal cancer. Twelve patients with locally advanced rectal cancer who received a short course preoperative chemoradiotherapy (25 Gy in 5 fractions) on the pelvis using Tomotherapy, between July 2010 and December 2010, were selected. Using their simulation computed tomography scans, Tomotherapy and four-box field CRT plans with the same dose schedule were evaluated, and dosimetric parameters of the two plans were compared. For the comparison of target coverage, we analyzed the mean dose, Vn Gy, Dmin, Dmax, radical dose homogeneity index (rDHI), and radiation conformity index (RCI). For the comparison of organs at risk (OAR), we analyzed the mean dose. Tomotherapy showed a significantly higher mean target dose than four-box field CRT (p 0.001). But, V26.25 Gy and V27.5 Gywere not significantly different between the two modalities. Tomotherapy showed higher Dmax and lower Dmin. The Tomotherapy plan had a lower rDHI than four-box field CRT (p = 0.000). Tomotherapy showed better RCI than four-box field CRT (p = 0.007). For OAR, the mean irradiated dose was significantly lower in Tomotherapy than four-box field CRT. In locally advanced rectal cancer, Tomotherapy delivers a higher conformal radiation dose to the target and reduces the irradiated dose to OAR than four-box field CRT.

  8. Daily tritium intakes by people living near a heavy-water research reactor facility: dosimetric significance

    International Nuclear Information System (INIS)

    Trivedi, A.; Cornett, R.J.; Galeriu, D.; Workman, W.; Brown, R.M.

    1997-02-01

    We have estimated the relative daily intakes of tritiated water (HTO) and organically bound tritium (OBT), and have measured HTO-in-urine, in an adult population residing in the town of Deep River, Ontario, near a heavy-water research reactor facility at Chalk River. The daily intake of elevated levels of atmospheric tritium has been estimated from its concentration in environmental and biological samples, and various food items from a local tritium-monitoring program. Where the available data were inadequate, we used estimates generated by an environmental tritium-transfer model. From these data and estimates, we calculated a total daily tritium intake of about 55 Bq. Of this amount, 2.5 Bq is obtained from OBT-in-diet. Inhalation of HTO-in-air (15 Bq d -1 ) and HTO-in-drinking water (15 Bq d -1 ) accounts for more than half of the HTO intake. Skin absorption of HTO from air and bathing or swimming (for 30 min d -1 ) accounts for another 9 Bq d -1 and 0.1 Bq d -1 , respectively. The remaining intake of HTO is from food as tissue-free water tritium. The International Commission on Radiological Protection's recommended two-compartment metabolic model for tritium predicts an equilibrium body burden of about 900 Bq from HTO (818 Bq) and OBT (83 Bq) in the body, which corresponds to an annual tritium dose of 0.41 μSv. The model-predicted urinary excretion of HTO (∼18 Bq L -1 ) agrees well with measured HTO-in-urine (range, 10-32 Bq L -1 ). The OBT dose contribution to the total tritium dose is about 16%. We conclude that for the people living near the Chalk River research reactor facility, the bulk of the tritium dose is due to HTO intake. (author)

  9. Dosimetric evaluation of indigenously developed non-lead bilayered radiation protective aprons

    International Nuclear Information System (INIS)

    Senthilkumar, S.

    2018-01-01

    Radiation shielding garments are commonly used to protect medical patients and radiation workers from X-radiation exposure during diagnostic imaging in hospitals. Originally, protective aprons consisted of lead-impregnated vinyl with a shielding equivalent given in millimeters of lead. All contained up to 2 mm of lead. While lead has long been used to shield patients from X-rays, its toxicity poses a health threat if the protective apron containing the metal wear out or the lead gets damaged. However, lead garments must be treated as hazardous waste for disposal and are heavy, causing back strain and other orthopedic problems for those who must wear them for long periods of time. The main purpose of this work was to indigenously develop light weight non lead based bilayered radiation protective aprons and evaluate dosimetrically with different combination of fabricated non lead materials and commercially available lead based aprons

  10. Development of an algorithm simulator of the planar radioactive source for dosimetric evaluations in accidents with radiopharmaceuticals used in nuclear medicine

    International Nuclear Information System (INIS)

    Claudino, Gutemberg L. Sales; Vieira, Jose Wilson; Leal Neto, Viriato; Lima, Fernando R. Andrade

    2013-01-01

    Objective of this work is to develop an algorithm simulator for dosimetric evaluation of accidents that may happen in Nuclear Medicine using PDF NT (Probability Density Functions). A software was developed using C# and WPF technology, in the integrated environment of Microsoft Visual Studio to organize and present the dosimetric results

  11. Evaluation of specific absorption rate as a dosimetric quantity for electromagnetic fields bioeffects.

    Directory of Open Access Journals (Sweden)

    Dimitris J Panagopoulos

    Full Text Available PURPOSE: To evaluate SAR as a dosimetric quantity for EMF bioeffects, and identify ways for increasing the precision in EMF dosimetry and bioactivity assessment. METHODS: We discuss the interaction of man-made electromagnetic waves with biological matter and calculate the energy transferred to a single free ion within a cell. We analyze the physics and biology of SAR and evaluate the methods of its estimation. We discuss the experimentally observed non-linearity between electromagnetic exposure and biological effect. RESULTS: WE FIND THAT: a The energy absorbed by living matter during exposure to environmentally accounted EMFs is normally well below the thermal level. b All existing methods for SAR estimation, especially those based upon tissue conductivity and internal electric field, have serious deficiencies. c The only method to estimate SAR without large error is by measuring temperature increases within biological tissue, which normally are negligible for environmental EMF intensities, and thus cannot be measured. CONCLUSIONS: SAR actually refers to thermal effects, while the vast majority of the recorded biological effects from man-made non-ionizing environmental radiation are non-thermal. Even if SAR could be accurately estimated for a whole tissue, organ, or body, the biological/health effect is determined by tiny amounts of energy/power absorbed by specific biomolecules, which cannot be calculated. Moreover, it depends upon field parameters not taken into account in SAR calculation. Thus, SAR should not be used as the primary dosimetric quantity, but used only as a complementary measure, always reporting the estimating method and the corresponding error. Radiation/field intensity along with additional physical parameters (such as frequency, modulation etc which can be directly and in any case more accurately measured on the surface of biological tissues, should constitute the primary measure for EMF exposures, in spite of similar

  12. Evaluation of specific absorption rate as a dosimetric quantity for electromagnetic fields bioeffects.

    Science.gov (United States)

    Panagopoulos, Dimitris J; Johansson, Olle; Carlo, George L

    2013-01-01

    To evaluate SAR as a dosimetric quantity for EMF bioeffects, and identify ways for increasing the precision in EMF dosimetry and bioactivity assessment. We discuss the interaction of man-made electromagnetic waves with biological matter and calculate the energy transferred to a single free ion within a cell. We analyze the physics and biology of SAR and evaluate the methods of its estimation. We discuss the experimentally observed non-linearity between electromagnetic exposure and biological effect. WE FIND THAT: a) The energy absorbed by living matter during exposure to environmentally accounted EMFs is normally well below the thermal level. b) All existing methods for SAR estimation, especially those based upon tissue conductivity and internal electric field, have serious deficiencies. c) The only method to estimate SAR without large error is by measuring temperature increases within biological tissue, which normally are negligible for environmental EMF intensities, and thus cannot be measured. SAR actually refers to thermal effects, while the vast majority of the recorded biological effects from man-made non-ionizing environmental radiation are non-thermal. Even if SAR could be accurately estimated for a whole tissue, organ, or body, the biological/health effect is determined by tiny amounts of energy/power absorbed by specific biomolecules, which cannot be calculated. Moreover, it depends upon field parameters not taken into account in SAR calculation. Thus, SAR should not be used as the primary dosimetric quantity, but used only as a complementary measure, always reporting the estimating method and the corresponding error. Radiation/field intensity along with additional physical parameters (such as frequency, modulation etc) which can be directly and in any case more accurately measured on the surface of biological tissues, should constitute the primary measure for EMF exposures, in spite of similar uncertainty to predict the biological effect due to non-linearity.

  13. Dosimetric evaluation of the conformation of the multileaf collimator to irregularly shaped fields

    International Nuclear Information System (INIS)

    Frazier, Arthur; Du, Maria; Wong, John; Vicini, Frank; Taylor, Roy; Yu, Cedric; Matter, Richard; Martinez, Alvaro; Yan Di

    1995-01-01

    Purpose: The goal of this study was to evaluate the dosimetric characteristics of geometric MLC prescription strategies and compare them to those of conventional shielding block. Methods and Materials: Circular fields, square fields, and 12 irregular fields for patients with cancer of the head and neck, lung, and pelvis were included in this study. All fields were shaped using the MLC and conventional blocks. A geometric criterion was defined as the amount of area discrepancy between the MLC and the prescription outline. The 'least area discrepancy' (LAD) of the MLC conformation was searched by selecting the collimator angle, meanwhile keeping a preselected position along the width of the leaf into the prescribed field. Five LAD conventions were studied. These included the LAD-0, LAD-(1(3)), LAD-(1(2)), and LAD-(2(3)) that inserted the leaves at the 0, (1(3)), (1(2)), and (2(3)) of the leaf end into the prescription field, respectively. In addition, the LAD optimization was applied to the transecting (TRN) approach for leaf conformation that prescribed an equal area of overblocking and underblocking under each leaf. Film dosimetry was performed in a 20 cm polystyrene phantom at 10 cm depth 100 cm from source to axis distance (SAD) for both 6 and 18 MV photons with each of the above MLC conformations and conventional blocks. The field penumbra width, defined as the mean of the separation between the 20% and 80% isodose lines along the normal of the prescription field edge, was calculated using both the MLC and conventional block film dosimetry and compared. In a similar way, the d20 is defined as the mean separation between the 20% isodose line and the prescription field edge, and the d80 is defined as the mean separation between the 80% isodose line and the prescription field edge. Results: The field penumbra width for all MLC conventions was approximately 2 mm larger than that of the conventional block. However, there was a larger variation of the separation

  14. Online dosimetric evaluation of larynx SBRT: A pilot study to assess the necessity of adaptive replanning.

    Science.gov (United States)

    Mao, Weihua; Rozario, Timothy; Lu, Weiguo; Gu, Xuejun; Yan, Yulong; Jia, Xun; Sumer, Baran; Schwartz, David L

    2017-01-01

    We have initiated a multi-institutional phase I trial of 5-fraction stereotactic body radiotherapy (SBRT) for Stage III-IVa laryngeal cancer. We conducted this pilot dosimetric study to confirm potential utility of online adaptive replanning to preserve treatment quality. We evaluated ten cases: five patients enrolled onto the current trial and five patients enrolled onto a separate phase I SBRT trial for early-stage glottic larynx cancer. Baseline SBRT treatment plans were generated per protocol. Daily cone-beam CT (CBCT) or diagnostic CT images were acquired prior to each treatment fraction. Simulation CT images and target volumes were deformably registered to daily volumetric images, the original SBRT plan was copied to the deformed images and contours, delivered dose distributions were re-calculated on the deformed CT images. All of these were performed on a commercial treatment planning system. In-house software was developed to propagate the delivered dose distribution back to reference CT images using the deformation information exported from the treatment planning system. Dosimetric differences were evaluated via dose-volume histograms. We could evaluate dose within 10 minutes in all cases. Prescribed coverage to gross tumor volume (GTV) and clinical target volume (CTV) was uniformly preserved; however, intended prescription dose coverage of planning treatment volume (PTV) was lost in 53% of daily treatments (mean: 93.9%, range: 83.9-97.9%). Maximum bystander point dose limits to arytenoids, parotids, and spinal cord remained respected in all cases, although variances in carotid artery doses were observed in a minority of cases. Although GTV and CTV SBRT dose coverage is preserved with in-room three-dimensional image guidance, PTV coverage can vary significantly from intended plans and dose to critical structures may exceed tolerances. Online adaptive treatment re-planning is potentially necessary and clinically applicable to fully preserve treatment

  15. Natural radioactivity measurements and dosimetric evaluations in soil samples with a high content of NORM

    Science.gov (United States)

    Caridi, F.; Marguccio, S.; Durante, G.; Trozzo, R.; Fullone, F.; Belvedere, A.; D'Agostino, M.; Belmusto, G.

    2017-01-01

    In this article natural radioactivity measurements and dosimetric evaluations in soil samples contaminated by Naturally Occurring Radioactive Materials (NORM) are made, in order to assess any possible radiological hazard for the population and for workers professionally exposed to ionizing radiations. Investigated samples came from the district of Crotone, Calabria region, South of Italy. The natural radioactivity investigation was performed by high-resolution gamma-ray spectrometry. From the measured gamma spectra, activity concentrations were determined for 226Ra , 234-mPa , 224Ra , 228Ac and 40K and compared with their clearance levels for NORM. The total effective dose was calculated for each sample as due to the committed effective dose for inhalation and to the effective dose from external irradiation. The sum of the total effective doses estimated for all investigated samples was compared to the action levels provided by the Italian legislation (D.Lgs.230/95 and subsequent modifications) for the population members (0.3mSv/y) and for professionally exposed workers (1mSv/y). It was found to be less than the limit of no radiological significance (10μSv/y).

  16. Dosimetric evaluation in heterogeneous tissue of anterior electron beam irradiation for treatment of retinoblastoma

    International Nuclear Information System (INIS)

    Kirsner, S.M.; Hogstrom, K.R.; Kurup, R.G.; Moyers, M.F.

    1987-01-01

    A dosimetric study of anterior electron beam irradiation for treatment of retinoblastoma was performed to evaluate the influence of tissue heterogeneities on the dose distribution within the eye and the accuracy of the dose calculated by a pencil beam algorithm. Film measurements were made in a variety of polystyrene phantoms and in a removable polystyrene eye incorporated into a tissue substitute phantom constructed from a human skull. Measurements in polystyrene phantoms were used to demonstrate the algorithm's ability to predict the effect of a lens block placed in the beam, as well as the eye's irregular surface shape. The eye phantom was used to measure dose distributions within the eye in both the sagittal and transverse planes in order to test the algorithm's ability to predict the dose distribution when bony heterogeneities are present. Results show (1) that previous treatment planning conclusions based on flat, uniform phantoms for central-axis depth dose are adequate; (2) that a three-dimensional heterogeneity correction is required for accurate dose calculations; and (3) that if only a two-dimensional heterogeneity correction is used in calculating the dose, it is more accurate for the sagittal than the transverse plane

  17. Dosimetric evaluation of anti-CD20 labelled with 188Re

    International Nuclear Information System (INIS)

    Barrio, Graciela; Osso Junior, Joao A.

    2011-01-01

    Radioimmunotherapy has the potential to deliver lethal radiation energy directly to malignant cells via targeting of radioisotope-conjugated monoclonal antibodies (MAbs) to specific antigens. B-cell lymphoma is a particularly good candidate for radioimmunotherapy because the disease is inherently radiosensitive, malignant cells in the blood, bone marrow, spleen and lymphonodes are accessible, and MAbs have been developed to B-cell surface antigens that do not shed or modulate. Rituximab (RTX), the human IgG1-type chimeric form of the parent murine antibody ibritumomab, is specifically targeted against CD20, a surface antigen expressed by pre-B and mature human B lymphocytes. The use of rhenium-188 from a 188 W/ 188 Re generator system represents an attractive alternative radionuclide for therapy. 188 Re is produced from beta decay of the 188 W parent. In addition to the emission of high-energy electrons (Eβ= 2118 keV), 188 Re also decays with emission of a gamma photon with an energy of 155 keV in 15% abundance. Besides the therapeutic usefulness of 188 Re, the emission of gamma photon is an added advantage since the biodistribution of 188 Re-labeled antibodies can be evaluated in vivo with a gamma camera. Also, rhenium has chemical properties similar to technetium. Thus, both can be conjugated to antibodies using similar chemistry methods. The objective of this work is to prove the usefulness of this radiopharmaceutical based on dosimetric studies, that are also required by the Brazilian Regulatory Agency (ANVISA). (author)

  18. Retrospective evaluation of dosimetric quality for prostate carcinomas treated with 3D conformal, intensity modulated and volumetric modulated arc radiotherapy

    Energy Technology Data Exchange (ETDEWEB)

    Crowe, Scott B [Science and Engineering Faculty, Queensland University of Technology, Brisbane, Queensland (Australia); Kairn, Tanya [Science and Engineering Faculty, Queensland University of Technology, Brisbane, Queensland (Australia); Premion, Wesley Medical Centre, Brisbane, Queensland (Australia); Middlebrook, Nigel; Hill, Brendan; Christie, David R H; Knight, Richard T [Premion, Wesley Medical Centre, Brisbane, Queensland (Australia); Kenny, John [Australian Clinical Dosimetry Services, Australian Radiation Protection and Nuclear Safety Agency, Melbourne, Victoria (Australia); Langton, Christian M; Trapp, Jamie V [Science and Engineering Faculty, Queensland University of Technology, Brisbane, Queensland (Australia)

    2013-12-15

    This study examines and compares the dosimetric quality of radiotherapy treatment plans for prostate carcinoma across a cohort of 163 patients treated across five centres: 83 treated with three-dimensional conformal radiotherapy (3DCRT), 33 treated with intensity modulated radiotherapy (IMRT) and 47 treated with volumetric modulated arc therapy (VMAT). Treatment plan quality was evaluated in terms of target dose homogeneity and organs at risk (OAR), through the use of a set of dose metrics. These included the mean, maximum and minimum doses; the homogeneity and conformity indices for the target volumes; and a selection of dose coverage values that were relevant to each OAR. Statistical significance was evaluated using two-tailed Welch's T-tests. The Monte Carlo DICOM ToolKit software was adapted to permit the evaluation of dose metrics from DICOM data exported from a commercial radiotherapy treatment planning system. The 3DCRT treatment plans offered greater planning target volume dose homogeneity than the other two treatment modalities. The IMRT and VMAT plans offered greater dose reduction in the OAR: with increased compliance with recommended OAR dose constraints, compared to conventional 3DCRT treatments. When compared to each other, IMRT and VMAT did not provide significantly different treatment plan quality for like-sized tumour volumes. This study indicates that IMRT and VMAT have provided similar dosimetric quality, which is superior to the dosimetric quality achieved with 3DCRT.

  19. Dosimetric Evaluation of Intensity Modulated Radiotherapy and 4-Field 3-D Conformal Radiotherapy in Prostate Cancer Treatment

    Directory of Open Access Journals (Sweden)

    Bora Uysal

    2013-03-01

    Full Text Available Objective: The purpose of this dosimetric study is the targeted dose homogeneity and critical organ dose comparison of 7-field Intensity Modulated Radiotherapy (IMRT and 3-D 4-field conformal radiotherapy. Study Design: Cross sectional study. Material and Methods: Twenty patients with low and moderate risk prostate cancer treated at Gülhane Military Medical School Radiation Oncology Department between January 2009 and December 2009 are included in this study. Two seperate dosimetric plans both for 7-field IMRT and 3D-CRT have been generated for each patient to comparatively evaluate the dosimetric status of both techniques and all the patients received 7-field IMRT. Results: Dose-comparative evaluation of two techniques revealed the superiority of IMRT technique with statistically significantly lower femoral head doses along with reduced critical organ dose-volume parameters of bladder V60 (the volume receiving 60 Gy and rectal V40 (the volume receiving 40 Gy and V60. Conclusion: It can be concluded that IMRT is an effective definitive management tool for prostate cancer with improved critical organ sparing and excellent dose homogenization in target organs of prostate and seminal vesicles.

  20. Fractionated stereotactic radiotherapy: A method to evaluate geometric and dosimetric uncertainties using radiochromic films

    Energy Technology Data Exchange (ETDEWEB)

    Coscia, Gianluca; Vaccara, Elena; Corvisiero, Roberta; Cavazzani, Paolo; Ruggieri, Filippo Grillo; Taccini, Gianni [S. C. Fisica Sanitaria, E.O. Ospedali Galliera di Genova, Via Alessandro Volta, 8 16128 Genova (Italy); S. C. Radioterapia, E.O. Ospedali Galliera di Genova, Via Alessandro Volta, 8 16128 Genova (Italy); S. C. Fisica Sanitaria, E.O. Ospedali Galliera di Genova, Via Alessandro Volta, 8 16128 Genova (Italy)

    2009-07-15

    In the authors' hospital, stereotactic radiotherapy treatments are performed with a Varian Clinac 600C equipped with a BrainLAB m3 micro-multileaf-collimator generally using the dynamic conformal arc technique. Patient immobilization during the treatment is achieved with a fixation mask supplied by BrainLAB, made with two reinforced thermoplastic sheets fitting the patient's head. With this work the authors propose a method to evaluate treatment geometric accuracy and, consequently, to determine the amount of the margin to keep in the CTV-PTV expansion during the treatment planning. The reproducibility of the isocenter position was tested by simulating a complete treatment on the anthropomorphic phantom Alderson Rando, inserting in between two phantom slices a high sensitivity Gafchromic EBT film, properly prepared and calibrated, and repeating several treatment sessions, each time removing the fixing mask and replacing the film inside the phantom. The comparison between the dose distributions measured on films and computed by TPS, after a precise image registration procedure performed by a commercial piece of software (FILMQA, 3cognition LLC (Division of ISP), Wayne, NJ), allowed the authors to measure the repositioning errors, obtaining about 0.5 mm in case of central spherical PTV and about 1.5 mm in case of peripheral irregular PTV. Moreover, an evaluation of the errors in the registration procedure was performed, giving negligible values with respect to the quantities to be measured. The above intrinsic two-dimensional estimate of treatment accuracy has to be increased for the error in the third dimension, but the 2 mm margin the authors generally use for the CTV-PTV expansion seems adequate anyway. Using the same EBT films, a dosimetric verification of the treatment planning system was done. Measured dose values are larger or smaller than the nominal ones depending on geometric irradiation conditions, but, in the authors' experimental

  1. Fractionated stereotactic radiotherapy: A method to evaluate geometric and dosimetric uncertainties using radiochromic films

    International Nuclear Information System (INIS)

    Coscia, Gianluca; Vaccara, Elena; Corvisiero, Roberta; Cavazzani, Paolo; Ruggieri, Filippo Grillo; Taccini, Gianni

    2009-01-01

    In the authors' hospital, stereotactic radiotherapy treatments are performed with a Varian Clinac 600C equipped with a BrainLAB m3 micro-multileaf-collimator generally using the dynamic conformal arc technique. Patient immobilization during the treatment is achieved with a fixation mask supplied by BrainLAB, made with two reinforced thermoplastic sheets fitting the patient's head. With this work the authors propose a method to evaluate treatment geometric accuracy and, consequently, to determine the amount of the margin to keep in the CTV-PTV expansion during the treatment planning. The reproducibility of the isocenter position was tested by simulating a complete treatment on the anthropomorphic phantom Alderson Rando, inserting in between two phantom slices a high sensitivity Gafchromic EBT film, properly prepared and calibrated, and repeating several treatment sessions, each time removing the fixing mask and replacing the film inside the phantom. The comparison between the dose distributions measured on films and computed by TPS, after a precise image registration procedure performed by a commercial piece of software (FILMQA, 3cognition LLC (Division of ISP), Wayne, NJ), allowed the authors to measure the repositioning errors, obtaining about 0.5 mm in case of central spherical PTV and about 1.5 mm in case of peripheral irregular PTV. Moreover, an evaluation of the errors in the registration procedure was performed, giving negligible values with respect to the quantities to be measured. The above intrinsic two-dimensional estimate of treatment accuracy has to be increased for the error in the third dimension, but the 2 mm margin the authors generally use for the CTV-PTV expansion seems adequate anyway. Using the same EBT films, a dosimetric verification of the treatment planning system was done. Measured dose values are larger or smaller than the nominal ones depending on geometric irradiation conditions, but, in the authors' experimental conditions, always

  2. Dosimetric evaluation of anti-CD20 labelled with {sup 188}Re

    Energy Technology Data Exchange (ETDEWEB)

    Barrio, Graciela; Osso Junior, Joao A., E-mail: gracielabarrio@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    Radioimmunotherapy has the potential to deliver lethal radiation energy directly to malignant cells via targeting of radioisotope-conjugated monoclonal antibodies (MAbs) to specific antigens. B-cell lymphoma is a particularly good candidate for radioimmunotherapy because the disease is inherently radiosensitive, malignant cells in the blood, bone marrow, spleen and lymphonodes are accessible, and MAbs have been developed to B-cell surface antigens that do not shed or modulate. Rituximab (RTX), the human IgG1-type chimeric form of the parent murine antibody ibritumomab, is specifically targeted against CD20, a surface antigen expressed by pre-B and mature human B lymphocytes. The use of rhenium-188 from a {sup 188}W/{sup 188}Re generator system represents an attractive alternative radionuclide for therapy. {sup 188}Re is produced from beta decay of the {sup 188}W parent. In addition to the emission of high-energy electrons (E{beta}= 2118 keV), {sup 188}Re also decays with emission of a gamma photon with an energy of 155 keV in 15% abundance. Besides the therapeutic usefulness of {sup 188}Re, the emission of gamma photon is an added advantage since the biodistribution of {sup 188}Re-labeled antibodies can be evaluated in vivo with a gamma camera. Also, rhenium has chemical properties similar to technetium. Thus, both can be conjugated to antibodies using similar chemistry methods. The objective of this work is to prove the usefulness of this radiopharmaceutical based on dosimetric studies, that are also required by the Brazilian Regulatory Agency (ANVISA). (author)

  3. Dosimetric response evaluation of tooth enamel for accelerator-based neutron radiation

    International Nuclear Information System (INIS)

    Khan, R.F.H.; Rink, W.J.; Boreham, D.R.

    2003-01-01

    To study the neutron response of human tooth enamel, a number of experiments with an accelerator-based neutron source have been designed. The neutron beam was produced with the low gamma yield, 7 Li(p,n) 7 Be type thick target, using the 3 MV McMaster K.N. Van de Graaff accelerator. The dosimetry was done using a pre-calibrated snoopy type neutron dosimeter. Neutron irradiation induces a dosimetric signal in the tooth enamel at the same defect site as gamma produced damage with the same g-values (g parallel =1.9973, width 0.4 mT g perpendicular =2.002, width 0.3 mT). The dosimetric signal grows linearly with neutron dose from 6-35 Gy tissue dose. Dosimetric response in two different grain sizes (300-500 μm, and grains <4 mm) has shown increased dosimetric amplitude in the larger grains. Dose build up effect on tooth inside the mouth due to cheek was simulated by placing a 4 mm thick paraffin wax layer between the beam and tooth, but had little effect. These results show that for mean neutron energy of 280 keV, the relative neutron response of the human tooth enamel ranges from 8% to 12% of the equivalent gamma ray response

  4. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  5. SU-E-T-323: Dosimetric Evaluation of Small Fields for SBRT Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, R; Eldib, A; Wang, B; Ma, C; Li, J [Fox Chase Cancer Center, Philadelphia, PA (United States)

    2015-06-15

    Purpose: Stereotactic body radiation therapy (SBRT) is commonly employed to treat small targets for effective tumor control with radiation beams of small field sizes. The goal of this work was to evaluate dosimetrically a treatment planning system (TPS) by comparing the calculated dose for SBRT treatment with ion-chamber measurements. Methods: 3D images of a solid-water phantom with a pinpoint ion-chamber (0.015cm3) inside were acquired with a CT scanner. Active volume of the ion-chamber was delineated on CT images. Targets with a diameter of 1.5cm, 2cm, 3cm, 4cm and 5cm were drawn around the chamber. 3DCRT plans were generated for each target size with centrally opened 6MV beams and off-axis beams by changing the isocenter location, respectively, using a TPS with the Analytical Anisotropic Algorithm. A 21iX linear accelerator was employed for plan delivery. The measured and calculated doses were compared. To evaluate the dose calculations in heterogeneity for small fields SBRT treatment, similar plans were also generated and delivered on a heterogeneous thoracic phantom for 5 different size targets in the lung. Results: Dose comparisons between measurements and calculations showed 5.2%, 1.88%, 1.34%, 1.01% and 0.85% difference for SBRT plans with small central axis beams and 0.96%, 0.15%, 0.58%, 0.22% and 0.77% difference for plans with off-axis beams for five different size targets. For the thoracic phantom, the differences on dose between measurements and calculations are bigger, which are 8%, 5.9%, 4.5%, 3.9% and 4.5%, respectively. Conclusion: Dose verification for small fields used in the SBRT treatment has been performed based on ion-chamber measurements in both homogenous and heterogeneous phantoms. More than a 5% difference has been observed in the heterogeneous phantom, especially for very small fields. To meet the ICRU recommendation on a dose difference of no more than 5%, some corrections on the commissioning parameters of the TPS are needed.

  6. Dosimetric evaluation of three adaptive strategies for prostate cancer treatment including pelvic lymph nodes irradiation.

    Science.gov (United States)

    Cantin, Audrey; Gingras, Luc; Lachance, Bernard; Foster, William; Goudreault, Julie; Archambault, Louis

    2015-12-01

    The movements of the prostate relative to the pelvic lymph nodes during intensity-modulated radiation therapy treatment can limit margin reduction and affect the protection of the organs at risk (OAR). In this study, the authors performed an analysis of three adaptive treatment strategies that combine information from both bony and gold marker registrations. The robustness of those treatments against the interfraction prostate movements was evaluated. A retrospective study was conducted on five prostate cancer patients with 7-13 daily cone-beam CTs (CBCTs). The clinical target volumes (CTVs) consisting of pelvic lymph nodes, prostate, and seminal vesicles as well as the OARs were delineated on each CBCT and the initial CT. Three adaptive strategies were analyzed. Two of these methods relied on a two-step patient positioning at each fraction. First step: a bony registration was used to deliver the nodal CTV prescription. Second step: a gold marker registration was then used either to (1) complete the dose delivered to the prostate (complement); (2) or give almost the entire prescription to the prostate with a weak dose gradient between the targets to compensate for possible motions (gradient). The third method (COR) used a pool of precalculated plans based on images acquired at previous treatment fractions. At each new fraction, a plan is selected from that pool based on the daily position of prostate center-of-mass. The dosimetric comparison was conducted and results are presented with and without the systematic shift in the prostate position on the CT planning. The adaptive strategies were compared to the current clinical standard where all fractions are treated with the initial nonadaptive plan. The minimum daily prostate D95% is improved by 2%, 9%, and 6% for the complement, the gradient, and the COR approaches, respectively, compared to the nonadaptive method. The average nodal CTV D95% remains constant across the strategies, except for the gradient approach

  7. Dosimetric evaluation of three adaptive strategies for prostate cancer treatment including pelvic lymph nodes irradiation

    International Nuclear Information System (INIS)

    Cantin, Audrey; Gingras, Luc; Archambault, Louis; Lachance, Bernard; Foster, William; Goudreault, Julie

    2015-01-01

    Purpose: The movements of the prostate relative to the pelvic lymph nodes during intensity-modulated radiation therapy treatment can limit margin reduction and affect the protection of the organs at risk (OAR). In this study, the authors performed an analysis of three adaptive treatment strategies that combine information from both bony and gold marker registrations. The robustness of those treatments against the interfraction prostate movements was evaluated. Methods: A retrospective study was conducted on five prostate cancer patients with 7–13 daily cone-beam CTs (CBCTs). The clinical target volumes (CTVs) consisting of pelvic lymph nodes, prostate, and seminal vesicles as well as the OARs were delineated on each CBCT and the initial CT. Three adaptive strategies were analyzed. Two of these methods relied on a two-step patient positioning at each fraction. First step: a bony registration was used to deliver the nodal CTV prescription. Second step: a gold marker registration was then used either to (1) complete the dose delivered to the prostate (complement); (2) or give almost the entire prescription to the prostate with a weak dose gradient between the targets to compensate for possible motions (gradient). The third method (COR) used a pool of precalculated plans based on images acquired at previous treatment fractions. At each new fraction, a plan is selected from that pool based on the daily position of prostate center-of-mass. The dosimetric comparison was conducted and results are presented with and without the systematic shift in the prostate position on the CT planning. The adaptive strategies were compared to the current clinical standard where all fractions are treated with the initial nonadaptive plan. Results: The minimum daily prostate D 95% is improved by 2%, 9%, and 6% for the complement, the gradient, and the COR approaches, respectively, compared to the nonadaptive method. The average nodal CTV D 95% remains constant across the strategies

  8. Dosimetric Evaluation of Automatic Segmentation for Adaptive IMRT for Head-and-Neck Cancer

    International Nuclear Information System (INIS)

    Tsuji, Stuart Y.; Hwang, Andrew; Weinberg, Vivian; Yom, Sue S.; Quivey, Jeanne M.; Xia Ping

    2010-01-01

    Purpose: Adaptive planning to accommodate anatomic changes during treatment requires repeat segmentation. This study uses dosimetric endpoints to assess automatically deformed contours. Methods and Materials: Sixteen patients with head-and-neck cancer had adaptive plans because of anatomic change during radiotherapy. Contours from the initial planning computed tomography (CT) were deformed to the mid-treatment CT using an intensity-based free-form registration algorithm then compared with the manually drawn contours for the same CT using the Dice similarity coefficient and an overlap index. The automatic contours were used to create new adaptive plans. The original and automatic adaptive plans were compared based on dosimetric outcomes of the manual contours and on plan conformality. Results: Volumes from the manual and automatic segmentation were similar; only the gross tumor volume (GTV) was significantly different. Automatic plans achieved lower mean coverage for the GTV: V95: 98.6 ± 1.9% vs. 89.9 ± 10.1% (p = 0.004) and clinical target volume: V95: 98.4 ± 0.8% vs. 89.8 ± 6.2% (p 3 of the spinal cord 39.9 ± 3.7 Gy vs. 42.8 ± 5.4 Gy (p = 0.034), but no difference for the remaining structures. Conclusions: Automatic segmentation is not robust enough to substitute for physician-drawn volumes, particularly for the GTV. However, it generates normal structure contours of sufficient accuracy when assessed by dosimetric end points.

  9. Cooperation in reactor design evaluation and licensing

    International Nuclear Information System (INIS)

    Kaufer, B.; Wasylyk, A.

    2014-01-01

    In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

  10. Cooperation in reactor design evaluation and licensing

    Energy Technology Data Exchange (ETDEWEB)

    Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

    2014-07-01

    In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

  11. An evaluation of the dosimetric performance characteristics of N-vinylpyrrolidone-based polymer gels

    Energy Technology Data Exchange (ETDEWEB)

    Papadakis, A E [Department of Medical Physics, Faculty of Medicine, University of Crete, PO Box 2208, Iraklion 71003, Crete (Greece); Maris, T G [Department of Medical Physics, Faculty of Medicine, University of Crete, PO Box 2208, Iraklion 71003, Crete (Greece); Zacharopoulou, F [Department of Medical Physics, Faculty of Medicine, University of Crete, PO Box 2208, Iraklion 71003, Crete (Greece); Pappas, E [Department of Medical Physics, Faculty of Medicine, University of Crete, PO Box 2208, Iraklion 71003, Crete (Greece); Zacharakis, G [Institute of Electronic Structure and Laser (IESL), Foundation for Research and Technology-Hellas (FORTH), PO Box 1527, Iraklion, Crete (Greece); Damilakis, J [Department of Medical Physics, Faculty of Medicine, University of Crete, PO Box 2208, Iraklion 71003, Crete (Greece)

    2007-08-21

    The aim of this work was to investigate the dosimetric performance properties of the N-vinylpyrrolidone argon (VIPAR) based polymer gel as a dosimetric tool in clinical radiotherapy. VIPAR gels with a larger concentration of gelatin than the standard recipe were manufactured and irradiated up to 68 Gy using a 6 and 18 MV linear accelerator. Using MRI, the R2-dose response was recorded at different imaging sessions within a 34 day time period post-irradiation. The R2-dose response was found to be linear between 5 and 68 Gy. Although dose sensitivity did not show significant variation with time, the measured R2-dose values showed an increasing trend, which was less evident beyond 17 days. At one day post-irradiation, calculated dose standard uncertainties at 20 Gy and 56 Gy were 2.2% and 1.7%, providing a dose resolution of 0.45 Gy and 0.97 Gy, respectively. Although these values fulfilled the 2% limit of ICRU, when gels were imaged at one day post-irradiation, it was shown that the temporal evolution of the R2 values deteriorated the per cent standard uncertainty and the dose resolution by {approx}57%, when imaged 17 days post-irradiation. Variation in the coagulation temperature of the gels did not impact the R2-dose sensitivity. This study has shown that the VIPAR gel has the properties of a dosimetric tool required in clinical radiotherapy, especially in applications where a wide dose dynamic range is employed. For results with the lowest per cent uncertainty and the optimum dose resolution, the dosimetry gels used in this work should be MR scanned at one day post-irradiation. Furthermore, a preliminary study on the R2-dose response of a new normoxic N-vinylpyrrolidone-based polymer gel showed that it could potentially replace the traditional VIPAR gel formulation, while preserving the wide dynamic dose response inherent to that monomer.

  12. Some dosimetric properties of the LiF:Mg,Ti evaluated by the automatic 6600 thermoluminescent reader

    Energy Technology Data Exchange (ETDEWEB)

    Ben-Shachar, B; Weinstein, M; German, U [Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev

    1996-12-01

    Some dosimetric properties of the new LiF:Mg,Ti TLD cards were checked, when evaluated by the new automatic 6600 TLD reader. The cards were calibrated to a dose of 1.0 mGy by five identical irradiations, and the TL-dose response was measured for a range of 75 - 1100 mGy. A very high accuracy was found for the three kind of chips measured (TLD-100, TLD-700 and TLD-600) and a low minimum measurable dose (MMD) was found, too. There is a good fit between the analytical evaluation and the theoretical calculation of the MMD. The results obtained are much better than those of the LiF:Mg,Ti cards evaluated by the older automatic 2271 reader used in the last two decades (authors).

  13. Reactor accident analysis and evaluation

    International Nuclear Information System (INIS)

    Chang, J.W.

    1983-01-01

    Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)

  14. SU-F-T-05: Dosimetric Evaluation and Validation of Newlydeveloped Well Chamber for Use in the Calibration of Brachytherapy Sources

    Energy Technology Data Exchange (ETDEWEB)

    Saminathan, S; Godson, H; Ponmalar, R; Manickam, R [Kidwai Memorial Institute of Oncology, Bangalore, Karnataka (India); Mazarello, J [Rosalina India private limited, Mumbai, Maharastra (India)

    2016-06-15

    Purpose: To evaluate the dosimetric characteristics of newly developed well type ionization chamber and to validate the results with the commercially available calibrated well chambers that are being used for the calibration of brachytherapy sources. Methods: The newly developed well type ionization chamber (BDS 1000) has been designed for the convenient use in brachytherapy which is open to atmospheric condition. The chamber has a volume of 240 cm3 and weight of 2.5 Kg. The calibration of the radioactive source with activities from 0.01 mCi to 20 Ci can be carried out using this chamber. The dosimetric parameters such as leakage current, stability, scattering effect, ion collection efficiency, reference air kerma rate and nominal response with energy were carried out with the BDS 1000 well type ion chamber. The evaluated dosimetric characteristics of BDS1000 well chamber were validated with two other commercially available well chambers (HDR 1000 plus and BTC/3007). Results: The measured leakage current observed was negligible for the newly developed BDS 1000 well type ion chamber. The ion collection efficiency was close to 1 and the response of the chamber was found to be very stable. The determined sweet spot was at 42 mm from bottom of the chamber insert. The reference air kerma rate was found to be 4.634 × 105 Gym2hr-1A-1 for the BDS 1000 well chamber. The overall dosimetric characteristics of BDS 1000 well chamber was in good agreement with the dosimetric properties of other two well chambers. Conclusion: The dosimetric study shows that the newly developed BDS 1000 well type ionization chamber is high sensitive and reliable chamber for reference air kerma strength calibration. The results obtained confirm that this chamber can be used for the calibration of HDR and LDR brachytherapy sources.

  15. Operational safety evaluation for minor reactor accidents

    International Nuclear Information System (INIS)

    Wang, O.S.

    1981-01-01

    The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

  16. Evaluation of dosimetric characteristics of graphene oxide/PVC nanocomposite for gamma radiation applications

    Energy Technology Data Exchange (ETDEWEB)

    Feizi, Shahzad; Malekie, Shahryar; Ziaie, Farhood [Nuclear Science and Technology Research Institute (NSTRI), Karaj (Iran, Islamic Republic of). Radiation Application Research School; Rahighi, Reza; Tayyebi, Ahmad [Univ. of Technology, Tehran (Iran, Islamic Republic of). Dept. of Physics

    2017-04-01

    Graphene oxide-polyvinyl chloride composite was prepared using tetrahydrofuran solvent-assisted dispersion of characterized nano flakes of graphene oxide in polymer matrix. Electrical percolation threshold of GO/PVC nanocomposite was determined via a finite element simulation method with a 2D model and compared with experimental results. A conductive cell with two silver coated walls was designed and fabricated for exploring dosimetric properties of the composite. Some characteristics of the new nanocomposite such as linearity of dose response, repeatability, sensitivity and angular dependence are investigated. According to 2D proposed method, obtained data associated to electrical conductivity of the GO/polymer composite for PVC matrix plotted in different GO weight percentages and had good compatibility (validity) with experimental data. The dose response is linear in the 17-51 mGy dose range and it can be introduced for gamma radiation dosimetry in diagnostic activities.

  17. Design, manufacture, and evaluation of an anthropomorphic pelvic phantom purpose-built for radiotherapy dosimetric intercomparison

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, K. M.; Ebert, M. A.; Kron, T.; Howlett, S. J.; Cornes, D.; Hamilton, C. S.; Denham, J. W. [Department of Radiation Oncology, Calvary Mater Newcastle, Waratah, New South Wales 2298, Australia and School of Physics, University of Newcastle, New South Wales 2308 (Australia); Department of Radiation Oncology, Sir Charles Gairdner Hospital, Western Australia, Australia and School of Physics, University of Western Australia, Western Australia 6009 (Australia); Department of Physical Sciences, Peter MacCallum Cancer Centre, Victoria 8006 (Australia); Australiasian College of Physical Scientists and Engineers in Medicine, Sydney, New South Wales 2020 (Australia); Trans-Tasman Radiation Oncology Group, Calvary Mater Newcastle, New South Wales 2298 (Australia); Heidelberg Repatriation Hospital, Victoria 3081 (Australia); Department of Radiation Oncology, Calvary Mater Newcastle, Waratah, New South Wales 2298, Australia and School of Medicine and Population Health, University of Newcastle, New South Wales 2308 (Australia)

    2011-10-15

    Purpose: An anthropomorphic pelvic phantom was designed and constructed to meet specific criteria for multicenter radiotherapy dosimetric intercomparison. Methods: Three dimensional external and organ outlines were generated from a computed tomography image set of a male pelvis, forming the basis of design for an anatomically realistic phantom. Clinically relevant points of interest were selected throughout the dataset where point-dose values could be measured with thermoluminescence dosimeters and a small-volume ionization chamber. Following testing, three materials were selected and the phantom was manufactured using modern prototyping techniques into five separate coronal slices. Time lines and resource requirements for the phantom design and manufacture were recorded. The ability of the phantom to mimic the entire treatment chain was tested. Results: The phantom CT images indicated that organ densities and geometries were comparable to those of the original patient. The phantom proved simple to load for dosimetry and rapid to assemble. Due to heat release during manufacture, small air gaps and density heterogeneities were present throughout the phantom. The overall cost for production of the prototype phantom was comparable to other commercial anthropomorphic phantoms. The phantom was shown to be suitable for use as a ''patient'' to mimic the entire treatment chain for typical external beam radiotherapy for prostate and rectal cancer. Conclusions: The phantom constructed for the present study incorporates all characteristics necessary for accurate Level III intercomparison studies. Following use in an extensive Level III dosimetric comparison over a large time scale and geographic area, the phantom retained mechanical stability and did not show signs of radiation-induced degradation.

  18. Dosimetric evaluation of a novel high dose rate (HDR) intraluminal / interstitial brachytherapy applicator for gastrointestinal and bladder cancers

    Science.gov (United States)

    Aghamiri, Seyyed Mahmoud Reza; Najarian, Siamak; Jaberi, Ramin

    2010-01-01

    High dose rate (HDR) brachytherapy is one of the accepted treatment modalities in gastro‐intestinal tract and bladder carcinomas. Considering the shortcoming of contact brachytherapy routinely used in gastrointestinal tract in treatment of big tumors or invasive method of bladder treatment, an intraluminal applicator with the capability of insertion into the tumor depth seems to be useful. This study presents some dosimetric evaluations to introduce this applicator to the clinical use. The radiation attenuation characteristics of the applicator were evaluated by means of two dosimetric methods including well‐type chamber and radiochromic film. The proposed 110 cm long applicator has a flexible structure made of stainless steel for easy passage through lumens and a needle tip to drill into big tumors. The 2 mm diameter of the applicator is thick enough for source transition, while easy passage through any narrow lumen such as endoscope or cystoscope working channel is ensured. Well‐chamber results showed an acceptably low attenuation of this steel springy applicator. Performing absolute dosimetry resulted in a correlation coefficient of R=0.9916(p‐value≈10−7) between standard interstitial applicator and the one proposed in this article. This study not only introduces a novel applicator with acceptable attenuation but also proves the response independency of the GAFCHROMIC EBT films to energy. By applying the dose response of the applicator in the treatment planning software, it can be used as a new intraluminal / interstitial applicator. PACS number: 87.53.Bn, 87.53.Jw, 29.40.Cs

  19. Evaluating the dosimetric effect of treatment-induced changes in virally mediated head and neck cancer patients

    International Nuclear Information System (INIS)

    Brown, Elizabeth; Owen, Rebecca; Mengersen, Kerrie; Harden, Fiona; Porceddu, Sandro

    2013-01-01

    Patients with virally mediated head and neck cancer (VMHNC) often present with advanced nodal disease that is highly radioresponsive as demonstrated by tumour and nodal regression during treatment. The resultant changes may impact on the planned dose distribution and so adversely affect the therapeutic ratio. The aim of this study was to evaluate the dosimetric effect of treatment-induced anatomical changes in VMHNC patients who had undergone a replan. Thirteen patients with virally mediated oropharyngeal or nasopharyngeal cancer who presented for definitive radiotherapy between 2005 and 2010 and who had a replan generated were investigated. The dosimetric effect of anatomical changes was quantified by comparing dose–volume histograms (DVH) of primary and nodal gross target volumes and organs at risk (OAR), including spinal cord and parotid glands, from the original plan and a comparison plan. Eleven three-dimensional conformal radiation therapy (3DCRT) and two intensity modulated radiation therapy (IMRT) plans were evaluated. Dose to the spinal cord and brainstem increased by 4.1% and 2.6%, respectively. Mean dose to the parotid glands also increased by 3.5%. In contrast, the dose received by 98% of the primary and nodal gross tumour volumes decreased by 0.15% and 0.3%, respectively, when comparing the initial treatment plan to the comparison plan. In this study, treatment-induced anatomical changes had the greatest impact on OAR dose with negligible effect on the dose to nodal gross tumour volumes. In the era of IMRT, accounting for treatment-induced anatomical changes is important as focus is placed on minimizing the acute and long-term side effects of treatment

  20. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

    Long, V.H.; Lam, P.V.; An, T.K.

    1989-01-01

    After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

  1. Dosimetric evaluation of lung tumor immobilization using breath hold at deep inspiration

    International Nuclear Information System (INIS)

    Barnes, Elizabeth A.; Murray, Brad R.; Robinson, Donald M.; Underwood, Lori J.; Hanson, John; Roa, Wilson H.Y.

    2001-01-01

    Purpose:To examine the dosimetric benefit of self-gated radiotherapy at deep-inspiration breath hold (DIBH) in the treatment of patients with non-small-cell lung cancer (NSCLC). The relative contributions of tumor immobilization at breath hold (BH) and increased lung volume at deep inspiration (DI) in sparing high-dose lung irradiation (≥20 Gy) were examined. Methods and Materials:Ten consecutive patients undergoing radiotherapy for Stage I-IIIB NSCLC who met the screening criteria were entered on this study. Patients were instructed to BH at DI without the use of external monitors or breath-holding devices (self-gating). Computed tomography (CT) scans of the thorax were performed during free breathing (FB) and DIBH. Fluoroscopy screened for reproducible tumor position throughout DIBH, and determined the maximum superior-inferior (SI) tumor motion during both FB and DIBH. Margins used to define the planning target volume (PTV) from the clinical target volume included 1 cm for setup error and organ motion, plus an additional SI margin for tumor motion, as determined from fluoroscopy. Three conformal treatment plans were then generated for each patient, one from the FB scan with FB PTV margins, a second from the DIBH scan with FB PTV margins, and a third from the DIBH scan with DIBH PTV margins. The percent of total lung volume receiving ≥20 Gy (using a prescription dose of 70.9 Gy to isocenter) was determined for each plan. Results:Self-gating at DIBH was possible for 8 of the 10 patients; 2 patients were excluded, because they were not able to perform a reproducible DIBH. For these 8 patients, the median BH time was 23 (range, 19-52) s. The mean percent of total lung volume receiving ≥20 Gy under FB conditions (FB scan with FB PTV margins) was 12.8%. With increased lung volume alone (DIBH scan with FB PTV margins), this was reduced to 11.0%, tending toward a significant decrease in lung irradiation over FB (p=0.086). With both increased lung volume and tumor

  2. Evaluation of the dosimetric performance characteristic of fluoroscopy system used in medicine

    International Nuclear Information System (INIS)

    Qi Xuesong; Wei Kedao; Cheng Yuxi; Zhou Qifu; Ge Lijuan; Hou Changsong

    2001-01-01

    Objective: To discuss establishment of diagnostic reference dose value in fluoroscopic examinations for survey of 16 different types of fluoroscopy systems. Methods: Choosing dosimetric characteristic parameters including: IIESDR, ESDR (typical value) and ESDR max (ESDR maximum), and DAP, which was calibrated in situ on the X-ray unit. Results: Results of dose survey are summarized in three tables, from these we could get wide changes in accordance with those in many other countries resulting from maximum and minimum of IIESDR, ESDR and ESDRmax when measurements were performed at same entrance field size on I.I. Image Intensifier of the 15 fluoroscopy systems and under conditions of ABC. And also we could get less changes of DAP mean values, though differences for patient weight, technological parameters of fluoroscopic exam setting, fluoroscopic time and number of film were more remarkable. Conclusions: Measurements on IIESDR, ESDR (typical value) and ESDRmax (ESDR maximum) are not satisfied as diagnostic reference level. But it is suggested that DAP values, in fluoroscopic exam, are used as a tool to achieve this. (author)

  3. Study and evaluation of the Siemens virtual wedge factor: dosimetric monitor system and variable field effects

    Energy Technology Data Exchange (ETDEWEB)

    Sendon Rio, J R Sendon; Martinez, C Otero; GarcIa, M Sanchez; Busto, R Lobato; Vega, V Luna; Sueiro, J Mosquera; Camean, M Pombar [Servizo de Radiofisica e Proteccion Radioloxica, Complexo Hospitalario Universitario de Santiago de Compostela (CHUS), Santiago de Compostela (Spain)], E-mail: jose.ramon.sendon.del.rio@sergas.es

    2008-03-07

    In the year 1997 Siemens introduced the virtual wedge in its accelerators. The idea was that a dose profile similar to that of a physical wedge can be obtained by moving one of the accelerator jaws at a constant speed while the dose rate is changing. This work explores the observed behaviour of virtual wedge factors. A model is suggested which takes into account that at any point in time, when the jaw moves, the dose at a point of interest in the phantom is not only due to the direct beam. It also depends on the scattered radiation in the phantom, the head scatter and the behaviour of the monitoring system of the accelerator. Measurements are performed in a Siemens Primus accelerator and compared to the model predictions. It is shown that the model agrees reasonably well with measurements spanning a wide range of conditions. A strong dependence of virtual wedge factors on the dosimetric board has been confirmed and an explanation has been given on how the balance between different contributions is responsible for virtual wedge factors values.

  4. Dosimetric and geometric evaluation of a hybrid strategy of offline adaptive planning and online image guidance for prostate cancer radiotherapy

    International Nuclear Information System (INIS)

    Liu Han; Wu Qiuwen

    2011-01-01

    For prostate cancer patients, online image-guided (IG) radiotherapy has been widely used in clinic to correct the translational inter-fractional motion at each treatment fraction. For uncertainties that cannot be corrected online, such as rotation and deformation of the target volume, margins are still required to be added to the clinical target volume (CTV) for the treatment planning. Offline adaptive radiotherapy has been implemented to optimize the treatment for each individual patient based on the measurements at early stages of treatment process. It has been shown that offline adaptive radiotherapy can effectively reduce the required margin. Recently a hybrid strategy of offline adaptive replanning and online IG was proposed and the geometric evaluation was performed. It was found that the planning margins can further be reduced by 1-2 mm compared to online IG only strategy. The purpose of this study was to investigate the dosimetric benefits of such a hybrid strategy on the target and organs at risk. A total of 420 repeated helical computed tomography scans from 28 patients were included in the study. Both low-risk patients (LRP, CTV = prostate) and intermediate-risk patients (IRP, CTV = prostate + seminal vesicles, SV) were included in the simulation. Two registration methods, based on center-of-mass shift of prostate only and prostate plus SV, were performed for IRP. The intensity-modulated radiotherapy was used in the simulation. Criteria on both cumulative and fractional doses were evaluated. Furthermore, the geometric evaluation was extended to investigate the optimal number of fractions necessary to construct the internal target volume (ITV) for the hybrid strategy. The dosimetric margin improvement was smaller than its geometric counterpart and was in the range of 0-1 mm. The optimal number of fractions necessary for the ITV construction is 2 for LRPs and 3-4 for IRPs in a hypofractionation protocol. A new cumulative index of target volume was proposed

  5. Dosimetric evaluation of simultaneous integrated boost during stereotactic body radiation therapy for pancreatic cancer

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Wensha, E-mail: wensha.yang@cshs.org [Department of Radiation Oncology, Cedars Sinai Medical Center, Los Angeles, CA (United States); Reznik, Robert; Fraass, Benedick A. [Department of Radiation Oncology, Cedars Sinai Medical Center, Los Angeles, CA (United States); Nissen, Nicholas [Department of Surgery, Cedars Sinai Medical Center, Los Angeles, CA (United States); Hendifar, Andrew [Department of Gastrointestinal Oncology, Cedars Sinai Medical Center, Los Angeles, CA (United States); Wachsman, Ashley [Department of Cross-Sectional Imaging Interventional Oncology, Cedars Sinai Medical Center, Los Angeles, CA (United States); Sandler, Howard; Tuli, Richard [Department of Radiation Oncology, Cedars Sinai Medical Center, Los Angeles, CA (United States)

    2015-04-01

    Stereotactic body radiation therapy (SBRT) provides a promising way to treat locally advanced pancreatic cancer and borderline resectable pancreatic cancer. A simultaneous integrated boost (SIB) to the region of vessel abutment or encasement during SBRT has the potential to downstage otherwise likely positive surgical margins. Despite the potential benefit of using SIB-SBRT, the ability to boost is limited by the local geometry of the organs at risk (OARs), such as stomach, duodenum, and bowel (SDB), relative to tumor. In this study, we have retrospectively replanned 20 patients with 25 Gy prescribed to the planning target volume (PTV) and 33~80 Gy to the boost target volume (BTV) using an SIB technique for all patients. The number of plans and patients able to satisfy a set of clinically established constraints is analyzed. The ability to boost vessels (within the gross target volume [GTV]) is shown to correlate with the overlap volume (OLV), defined to be the overlap between the GTV + a 1(OLV1)- or 2(OLV2)-cm margin with the union of SDB. Integral dose, boost dose contrast (BDC), biologically effective BDC, tumor control probability for BTV, and normal tissue complication probabilities are used to analyze the dosimetric results. More than 65% of the cases can deliver a boost to 40 Gy while satisfying all OAR constraints. An OLV2 of 100 cm{sup 3} is identified as the cutoff volume: for cases with OLV2 larger than 100 cm{sup 3}, it is very unlikely the case could achieve 25 Gy to the PTV while successfully meeting all the OAR constraints.

  6. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    Popp, P.

    1987-01-01

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  7. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

  8. Dosimetric performance evaluation regarding proton beam incident angles of a lithium-based AB-BNCT design

    International Nuclear Information System (INIS)

    Lee, Pei-Yi; Jiang, Shiang-Huei; Liu, Yuan-Hao

    2014-01-01

    The 7 Li(p,xn) 7 Be nuclear reaction, based on the low-energy protons, could produce soft neutrons for accelerator-based boron neutron capture therapy (AB-BNCT). Based on the fact that the induced neutron field is relatively divergent, the relationship between the incident angle of proton beam and the neutron beam quality was evaluated in this study. To provide an intense epithermal neutron beam, a beam-shaping assembly (BSA) was designed. And a modified Snyder head phantom was used in the calculations for evaluating the dosimetric performance. From the calculated results, the intensity of epithermal neutrons increased with the increase in proton incident angle. Hence, either the irradiation time or the required proton current can be reduced. When the incident angle of 2.5-MeV proton beam is 120 deg., the required proton current is ∼13.3 mA for an irradiation time of half an hour. The results of this study show that the BSA designs can generate neutron beams with good intensity and penetrability. Using a 20-mA, 2.5-MeV proton beam as the source, the required irradiation time, to induce 60 RBE-Gy of maximum tumour dose, is less than half an hour in any proton beam alignments. On the premise that the dosimetric performances are similar, the intensity of epithermal neutrons can be increased by using non-collinear (e.g. 90 deg., 120 deg.) incident protons. Thus, either the irradiation time or the required proton current can be reduced. The use of 120 deg. BSA model shows the possibility to reduce the required proton current to ∼13.3 mA when the goal of irradiation time is 30 min. The decrease of required proton beam current certainly will make the use of lithium target much easier. In June 2013, a 5-MeV, 30-mA radio frequency quadruple (RFQ) accelerator for BNCT was built at INFN-LNL (Legnaro National Laboratories, Italy), which shows a possibility to build a suitable RFQ accelerator for the authors' design. In addition, a 2.5-MeV, 30-mA Tandem accelerator was

  9. The Advanced Test Reactor Strategic Evaluation Program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1990-01-01

    A systematic evaluation of safety, environmental, and operational issues has been initiated at the Advanced Test Reactor (ATR). This program, the Strategic Evaluation Program (STEP), provides an integrated review of safety and operational issues against the standards applied to licensed commercial facilities. In the review of safety issues, 18 deviations were identified which required prompt attention. Resolution of these items has been accelerated in the program. An integrated living schedule is being developed to address the remaining findings. A risk evaluation is being performed on the proposed corrective actions and these actions will then be formally ranked in order of priority based on considerations of safety and operational significance. Once the final ranking is completed, an integrated schedule will be developed, which will include considerations of availability of funding and operating schedule. 3 refs., 2 figs

  10. Dosimetric evaluation of the staff working in a PET/CT department

    International Nuclear Information System (INIS)

    Dalianis, K.; Malamitsi, J.; Gogou, L.; Pagou, M.; Efthimiadou, R.; Andreou, J.; Louizi, A.; Georgiou, E.

    2006-01-01

    The dosimetric literature data concerning the medical personnel working in positron emission tomography/computed tomography (PET/CT) departments are limited. Therefore, we measured the radiation dose of the staff working in the first PET/CT department in Greece at the Diagnostic and Therapeutic Center of Athens HYGEIA-Harvard Medical International. As, for the time being, only 2-deoxy-2-[ 18 F]fluoro-d-glucose (FDG) PET studies are performed, radiation dose measurements concern those derived from dispensing of the radiopharmaceutical as well as from the patients undergoing FDG-PET imaging. Our aim is to develop more effective protective measures against radionuclide exposure. To estimate the effective dose from external exposure, all seven members of the staff (two nurses, two medical physicists, two technologists, one secretary) had TLD badges worn at the upper pocket of their overall, TLD rings on the right hand and digital dosimeters at their upper side pocket. In addition, isodose curves were measured with thermoluminescence detectors for distances of 20, 50, 70 and 100 cm away from patients who had been injected with 18 F-FDG. Dose values of the PET/CT staff were measured with digital detectors, TLD badges and TLD rings over the first 8 months for a total of 160 working days of the department's operation, consisting of a workload of about 10-15 patients/week who received 250-420 MBq of 18 F-FDG each. Whole - body collective doses and hand doses for the staff were the following: Nurse no. 1 received 1.6 mSv as a whole body dose and 2,1 as a hand dose, Nurse no. 2 received 1.9 and 2.4 mSv respectively. For medical physicist no. 1 the dose values were 1.45 mSv whole body and 1.7 mSv hand dose, for medical physicist no. 2 1.67 mSv wholebody dose and 1.55 mSv hand dose and for technologists no. 1 and no. 2 the whole body doses were 0.7 and 0.64 mSv respectively. Lastly, the secretary received 0.1 mSv whole body dose. These preliminary data have shown that the dose

  11. Dosimetric evaluation of scattered and attenuated radiation due to dental restorations in head and neck radiotherapy

    Directory of Open Access Journals (Sweden)

    Mona Azizi

    2018-01-01

    Full Text Available In radiotherapy of head and neck cancer, the presence of high density materials modifies photon dose distribution near these high density materials during treatment. The aim of this study is to calculate the backscatter and attenuation effects of a healthy tooth, Amalgam, Ni-Cr alloy and Ceramco on the normal tissues before and after these materials irradiated by 6 and 15 MV photon beams, respectively. All measurements were carried out in a water phantom with dimension of 50 × 50 × 50 cm3with an ionization chamber detector. Two points before and four points after the dental sample were considered to score the photon dose. The depth dose on the central beam axis was explored in a water phantom for source to surface distance (SSD of 100 cm in a 10 × 10 cm2 field size. The percentage dose change was obtained relative to the dose in water versus depth of water, tooth, Amalgam, Ni-Cr alloy and Ceramco for the photon beams. The absolute dose (cGy was measured by prescription of 100 cGy dose in the water phantom at depth of 2.0 and 3.1 cm for 6 and 15 MV photons, respectively. At depth of 0.6 cm, the maximum percentage dose increase was observed with values of 6.99% and 9.43%for Ni-Cr and lowest percentage dose increase of 1.49% and 2.63% are related to the healthy tooth in 6 and 15 MV photon beams, respectively. The maximum absolute dose of 95.58 cGy and 93.64 cGy were observed at depth of 0.6 cm in presence of Ni-Cr alloy for 6 and 15 MV photon beams, respectively. The presence of dental restorations can cause backscattering dose during head and neck radiation therapy. Introduction of compositions and electron density of high density materials can improve the accuracy of dosimetric calculations in treatment planning systems to deliver the relevant dose to target organ and reduce the backscattering dose in healthy tissues in the surrounding of tooth.

  12. Dosimetric evaluation of the staff working in a PET/CT department

    Energy Technology Data Exchange (ETDEWEB)

    Dalianis, K. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece)]. E-mail: k.dalianis@hygeia.gr; Malamitsi, J. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece); Gogou, L. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece); Pagou, M. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece); Efthimiadou, R. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece); Andreou, J. [Department of PET/CT, Diagnostic and Therapeutic Center of Athens Hygeia-Harvard Medical International, Erythrau Stavrou 4, 5123 Athens (Greece); Louizi, A. [Department of Medical Physics Medical School University of Athens, Athens (Greece); Georgiou, E. [Department of Medical Physics Medical School University of Athens, Athens (Greece)

    2006-12-20

    The dosimetric literature data concerning the medical personnel working in positron emission tomography/computed tomography (PET/CT) departments are limited. Therefore, we measured the radiation dose of the staff working in the first PET/CT department in Greece at the Diagnostic and Therapeutic Center of Athens HYGEIA-Harvard Medical International. As, for the time being, only 2-deoxy-2-[{sup 18}F]fluoro-d-glucose (FDG) PET studies are performed, radiation dose measurements concern those derived from dispensing of the radiopharmaceutical as well as from the patients undergoing FDG-PET imaging. Our aim is to develop more effective protective measures against radionuclide exposure. To estimate the effective dose from external exposure, all seven members of the staff (two nurses, two medical physicists, two technologists, one secretary) had TLD badges worn at the upper pocket of their overall, TLD rings on the right hand and digital dosimeters at their upper side pocket. In addition, isodose curves were measured with thermoluminescence detectors for distances of 20, 50, 70 and 100 cm away from patients who had been injected with {sup 18}F-FDG. Dose values of the PET/CT staff were measured with digital detectors, TLD badges and TLD rings over the first 8 months for a total of 160 working days of the department's operation, consisting of a workload of about 10-15 patients/week who received 250-420 MBq of {sup 18}F-FDG each. Whole - body collective doses and hand doses for the staff were the following: Nurse no. 1 received 1.6 mSv as a whole body dose and 2,1 as a hand dose, Nurse no. 2 received 1.9 and 2.4 mSv respectively. For medical physicist no. 1 the dose values were 1.45 mSv whole body and 1.7 mSv hand dose, for medical physicist no. 2 1.67 mSv wholebody dose and 1.55 mSv hand dose and for technologists no. 1 and no. 2 the whole body doses were 0.7 and 0.64 mSv respectively. Lastly, the secretary received 0.1 mSv whole body dose. These preliminary data have

  13. Evaluation of dosimetric techniques in positrons emission tomography and computerized tomography (PET/CT)

    International Nuclear Information System (INIS)

    Pinto, Gabriella Montezano

    2014-01-01

    Among diagnostic techniques PET/CT is one of those with the highest dose delivery to the patient as a cause of external exposure to X-rays, and the use of a radiopharmaceutical that results in a high energy gamma emission. The dosimetry of these two components becomes important in order to optimize and justify the technique. Various dosimetric techniques are found in literature without a consensus of the best to use. With the advances in technological and consequent equipment configuration changes, upgrades and variation in methodologies, particularly in computed tomography, a standardization of these techniques is required. Previous studies show that CT is responsible for 70 % of the dose delivered to the patient in PET/CT examinations. Thus, many researchers have been focused on CT dose optimization protocols studies. This work analyzes the doses involved in a PET/CT oncology protocol by using an Alderson female anthropomorphic phantom in a public hospital of Rio de Janeiro city. The dose estimate for PET examination resulting from the use of 18 F - FDG radiopharmaceutical was conducted through dose factors published in ICRP 106; the dose for CT was estimated and compared by calculation of the absorbed doses to patients according to four methods: thermoluminescent dosimetry (TL0100) distributed in critical organs of the Alderson phantom; measurements of CTOI according to AAPM number 96; correction factor for effective diameter SSOE (AAPM Number 204); and simulation by ImPACT program For CT, the results in terms of effective dose presented (TLO, CTOI and ImPACT) ± 5 % maximum variations between methodologies. Considering medium absorbed dose (TLO, SSOE and ImPACT) the results differed in ± 7 % from each other. These findings demonstrate that parameters provided by the manufacturer on the console can be used to have a primary approach of both, absorbed and effective doses to the patient since that a quality assurance program of these parameters are adopted in

  14. Dosimetric evaluation of spectrophotometric response of alanine gel solution for gamma, photons, electrons and thermal neutrons radiations

    International Nuclear Information System (INIS)

    Silva, Cleber Feijo

    2009-01-01

    Alanine Gel Dosimeter is a new gel material developed at IPEN that presents significant improvement on Alanine system developed by Costa. The DL-Alanine (C 3 H 7 NO 2 ) is an amino acid tissue equivalent that improves the production of ferric ions in the solution. This work aims to analyse the main dosimetric characteristics this new gel material for future application to measure dose distribution. The performance of Alanine gel solution was evaluated to gamma, photons, electrons and thermal neutrons radiations using the spectrophotometry technique. According to the obtained results for the different studied radiation types, the reproducibility intra-batches and inter-batches is better than 4% and 5%, respectively. The dose response presents a linear behavior in the studied dose range. The response dependence as a function of dose rate and incident energy is better 2% and 3%, respectively. The lower detectable dose is 0.1 Gy. The obtained results indicate that the Alanine gel dosimeter presents good performance and can be useful as an alternative dosimeter in the radiotherapy area, using MRI technique for tridimensional dose distribution evaluation. (author)

  15. Thermoluminescence emission spectrometry of glass display in mobile phones and resulting evaluation of the dosimetric properties of a specific type of display glass

    International Nuclear Information System (INIS)

    Discher, Michael; Woda, Clemens

    2014-01-01

    Glass displays of mobile phones are sensitive to ionizing radiation and can be used for retrospective dosimetry for the purpose of triage after a radiological accident or attack. In this study the two main types of glass display that are used in modern mobile phones were investigated using thermoluminescence (TL) emission spectrometry. A different TL spectrum was observed for the glass display of category A (lime-aluminosilicate glass) and category B (boron-silicate glass). Based on the spectral measurements an optimized detection window was chosen to re-evaluate the dosimetric properties (dose response, optical and long-term stability) of glass display category B. - Highlights: • Two display glass types show similar TL emission peaks but with strongly different relative intensities. • The intrinsic background TL signal peaks at similar wavelengths as the radiation induced signal. • Dosimetric properties of one display glass type were re-evaluated using an optimized detection window

  16. Dosimetric evaluation of 4 different treatment modalities for curative-intent stereotactic body radiation therapy for isolated thoracic spinal metastases

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jun [Department of Radiation Oncology, Chinese PLA General Hospital, 28 Fuxing Road, Beijing, 100853 (China); Department of Oncology, First Affiliated Hospital of Xinxiang Medical University, 88 Jiankang Road, Weihui, Henan, 453100 (China); Ma, Lin [Department of Radiation Oncology, Chinese PLA General Hospital, 28 Fuxing Road, Beijing, 100853 (China); Department of Radiation Oncology, Hainan Branch of Chinese PLA General Hospital, Haitang Bay, Sanya, 572000 (China); Wang, Xiao-Shen; Xu, Wei Xu; Cong, Xiao-Hu; Xu, Shou-Ping; Ju, Zhong-Jian [Department of Radiation Oncology, Chinese PLA General Hospital, 28 Fuxing Road, Beijing, 100853 (China); Du, Lei [Department of Radiation Oncology, Hainan Branch of Chinese PLA General Hospital, Haitang Bay, Sanya, 572000 (China); Cai, Bo-Ning [Department of Radiation Oncology, Chinese PLA General Hospital, 28 Fuxing Road, Beijing, 100853 (China); Yang, Jack [Department of Radiation Oncology, Monmouth Medical Center, 300 2nd Avenue, Long Branch, NJ 07740 (United States)

    2016-07-01

    To investigate the dosimetric characteristics of 4 SBRT-capable dose delivery systems, CyberKnife (CK), Helical TomoTherapy (HT), Volumetric Modulated Arc Therapy (VMAT) by Varian RapidArc (RA), and segmental step-and-shoot intensity-modulated radiation therapy (IMRT) by Elekta, on isolated thoracic spinal lesions. CK, HT, RA, and IMRT planning were performed simultaneously for 10 randomly selected patients with 6 body types and 6 body + pedicle types with isolated thoracic lesions. The prescription was set with curative intent and dose of either 33 Gy in 3 fractions (3F) or 40 Gy in 5F to cover at least 90% of the planning target volume (PTV), correspondingly. Different dosimetric indices, beam-on time, and monitor units (MUs) were evaluated to compare the advantages/disadvantages of each delivery modality. In ensuring the dose-volume constraints for cord and esophagus of the premise, CK, HT, and RA all achieved a sharp conformity index (CI) and a small penumbra volume compared to IMRT. RA achieved a CI comparable to those from CK, HT, and IMRT. CK had a heterogeneous dose distribution in the target as its radiosurgical nature with less dose uniformity inside the target. CK had the longest beam-on time and the largest MUs, followed by HT and RA. IMRT presented the shortest beam-on time and the least MUs delivery. For the body-type lesions, CK, HT, and RA satisfied the target coverage criterion in 6 cases, but the criterion was satisfied in only 3 (50%) cases with the IMRT technique. For the body + pedicle-type lesions, HT satisfied the criterion of the target coverage of ≥90% in 4 of the 6 cases, and reached a target coverage of 89.0% in another case. However, the criterion of the target coverage of ≥90% was reached in 2 cases by CK and RA, and only in 1 case by IMRT. For curative-intent SBRT of isolated thoracic spinal lesions, RA is the first choice for the body-type lesions owing to its delivery efficiency (time); the second choice is CK or HT; HT is the

  17. Needle migration and dosimetric impact in high-dose-rate brachytherapy for prostate cancer evaluated by repeated MRI.

    Science.gov (United States)

    Buus, Simon; Lizondo, Maria; Hokland, Steffen; Rylander, Susanne; Pedersen, Erik M; Tanderup, Kari; Bentzen, Lise

    To quantify needle migration and dosimetric impact in high-dose-rate brachytherapy for prostate cancer and propose a threshold for needle migration. Twenty-four high-risk prostate cancer patients treated with an HDR boost of 2 × 8.5 Gy were included. Patients received an MRI for planning (MRI1), before (MRI2), and after treatment (MRI3). Time from needle insertion to MRI3 was ∼3 hours. Needle migration was evaluated from coregistered images: MRI1-MRI2 and MRI1-MRI3. Dose volume histogram parameters from the treatment plan based on MRI1 were related to parameters based on needle positions in MRI2 or MRI3. Regression was used to model the average needle migration per implant and change in D90 clinical target volume, CTV prostate+3mm . The model fit was used for estimating the dosimetric impact in equivalent dose in 2 Gy fractions for dose levels of 6, 8.5, 10, 15, and 19 Gy. Needle migration was on average 2.2 ± 1.8 mm SD from MRI1-MRI2 and 5.0 ± 3.0 mm SD from MRI1-MRI3. D90 CTV prostate+3mm was robust toward average needle migration ≤3 mm, whereas for migration >3 mm D90 decreased by 4.5% per mm. A 3 mm of needle migration resulted in a decrease of 0.9, 1.7, 2.3, 4.8, and 7.6 equivalent dose in 2 Gy fractions for dose levels of 6, 8.5, 10, 15, and 19 Gy, respectively. Substantial needle migration in high-dose-rate brachytherapy occurs frequently in 1-3 hours following needle insertion. A 3-mm threshold of needle migration is proposed, but 2 mm may be considered for dose levels ≥15 Gy. Copyright © 2017 American Brachytherapy Society. Published by Elsevier Inc. All rights reserved.

  18. Dosimetric evaluation of 4 different treatment modalities for curative-intent stereotactic body radiation therapy for isolated thoracic spinal metastases

    International Nuclear Information System (INIS)

    Yang, Jun; Ma, Lin; Wang, Xiao-Shen; Xu, Wei Xu; Cong, Xiao-Hu; Xu, Shou-Ping; Ju, Zhong-Jian; Du, Lei; Cai, Bo-Ning; Yang, Jack

    2016-01-01

    To investigate the dosimetric characteristics of 4 SBRT-capable dose delivery systems, CyberKnife (CK), Helical TomoTherapy (HT), Volumetric Modulated Arc Therapy (VMAT) by Varian RapidArc (RA), and segmental step-and-shoot intensity-modulated radiation therapy (IMRT) by Elekta, on isolated thoracic spinal lesions. CK, HT, RA, and IMRT planning were performed simultaneously for 10 randomly selected patients with 6 body types and 6 body + pedicle types with isolated thoracic lesions. The prescription was set with curative intent and dose of either 33 Gy in 3 fractions (3F) or 40 Gy in 5F to cover at least 90% of the planning target volume (PTV), correspondingly. Different dosimetric indices, beam-on time, and monitor units (MUs) were evaluated to compare the advantages/disadvantages of each delivery modality. In ensuring the dose-volume constraints for cord and esophagus of the premise, CK, HT, and RA all achieved a sharp conformity index (CI) and a small penumbra volume compared to IMRT. RA achieved a CI comparable to those from CK, HT, and IMRT. CK had a heterogeneous dose distribution in the target as its radiosurgical nature with less dose uniformity inside the target. CK had the longest beam-on time and the largest MUs, followed by HT and RA. IMRT presented the shortest beam-on time and the least MUs delivery. For the body-type lesions, CK, HT, and RA satisfied the target coverage criterion in 6 cases, but the criterion was satisfied in only 3 (50%) cases with the IMRT technique. For the body + pedicle-type lesions, HT satisfied the criterion of the target coverage of ≥90% in 4 of the 6 cases, and reached a target coverage of 89.0% in another case. However, the criterion of the target coverage of ≥90% was reached in 2 cases by CK and RA, and only in 1 case by IMRT. For curative-intent SBRT of isolated thoracic spinal lesions, RA is the first choice for the body-type lesions owing to its delivery efficiency (time); the second choice is CK or HT; HT is the

  19. EXPERIMENTAL EVALUATION OF DOSIMETRIC CHARACTERIZATION OF GAFCHROMIC EBT3 AND EBT-XD FILMS FOR CLINICAL CARBON ION BEAMS.

    Science.gov (United States)

    Yonai, Shunsuke; Arai, Chinatsu; Shimoyama, Kaoru; Fournier-Bidoz, Nathalie

    2018-02-03

    Radiochromic film is a very useful tool for 2D dosimetric measurements in radiotherapy because it is self-developing and has very high-spatial resolution. However, considerable care has to be taken in ion beam radiotherapy owing to the quenching effect of high-linear energy transfer (LET) radiation. In this study, the dose responses of GAFchromic EBT3 and EBT-XD films were experimentally investigated using the clinical carbon ion beam at the Heavy Ion Medical Accelerator in Chiba. Results showed that the relations between absorbed dose and net optical density could be expressed well using an equation proposed by Reinhardt (2015). The quenching effect was evaluated by determining their relative efficiencies for photon irradiation as a function of LET. A correction equation derived in this study allowed the absorbed dose to be determined in the small irradiation field used for carbon ion radiotherapy eye treatments. This study contributes to establishing an absolute dosimetry procedure for heavy ion beams using radiochromic film. © The Author(s) 2018. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  20. Dosimetric pens: evaluation of calibration results in the Laboratorio Nacional de Metrologia das Radiacoes Ionizantes do Instituto de Radioprotecao e Dosimetria (IRD/LNMRI), RJ, Brazil

    International Nuclear Information System (INIS)

    Quaresma, D.S.; Ramos, M.M.O.; Cabral, T.S.; Peixoto, J.G.P.

    2005-01-01

    Dosimetric pens are direct reading personal dosemeters that are used in the practices of radiation protection in industries, hospitals, universities, and research institutes in the country. Quality control of measurements made with these instruments must include their periodical calibration in one of the laboratories of the Calibration Laboratory Network for Ionizing Radiation with the aim to compare the behavior of the measurements made in dosimetric pens of different models and manufacturers, submitted for calibration in the LNMRI/IRD/CNEN (Brazilian Lab for Metrology of Ionizing Radiations of the Institute for Radioprotection and Dosimetry of the Brazilian Nuclear Energy Commission), RJ or national reference laboratory and a member of the Network, in the years of 2000 to 2002. The parameters considered for the purpose of this work were: accuracy and linearity of response and measurement uncertainty evaluated. The results show that among the analyzed models there are changes in behavior

  1. Dosimetric and geometric evaluation of a novel stereotactic radiotherapy device for breast cancer: The GammaPod Trade-Mark-Sign

    Energy Technology Data Exchange (ETDEWEB)

    Mutaf, Yildirim D.; Yi, Byong Yong; Prado, Karl; D' Souza, Warren D.; Regine, William F.; Feigenberg, Steven J. [Department of Radiation Oncology, University of Maryland School of Medicine, Baltimore, Maryland 21201 (United States); Zhang Jin [Xcision Medical Systems, Columbia, Maryland 21045 (United States); Yu, Cedric X. [Department of Radiation Oncology, University of Maryland School of Medicine, Baltimore, Maryland 21201 and Xcision Medical Systems, Columbia, Maryland 21045 (United States)

    2013-04-15

    Purpose: A dedicated stereotactic gamma irradiation device, the GammaPod Trade-Mark-Sign from Xcision Medical Systems, was developed specifically to treat small breast cancers. This study presents the first evaluation of dosimetric and geometric characteristics from the initial prototype installed at University of Maryland Radiation Oncology Department. Methods: The GammaPod Trade-Mark-Sign stereotactic radiotherapy device is an assembly of a hemi-spherical source carrier containing 36 {sup 60}Co sources, a tungsten collimator, a dynamically controlled patient support table, and the breast immobilization system which also functions as a stereotactic frame. The source carrier contains the sources in six columns spaced longitudinally at 60 Degree-Sign intervals and it rotates together with the variable-size collimator to form 36 noncoplanar, concentric arcs focused at the isocenter. The patient support table enables motion in three dimensions to position the patient tumor at the focal point of the irradiation. The table moves continuously in three cardinal dimensions during treatment to provide dynamic shaping of the dose distribution. The breast is immobilized using a breast cup applying a small negative pressure, where the immobilization cup is embedded with fiducials also functioning as the stereotactic frame for the breast. Geometric and dosimetric evaluations of the system as well as a protocol for absorbed dose calibration are provided. Dosimetric verifications of dynamically delivered patient plans are performed for seven patients using radiochromic films in hypothetical preop, postop, and target-in-target treatment scenarios. Results: Loaded with 36 {sup 60}Co sources with cumulative activity of 4320 Ci, the prototype GammaPod Trade-Mark-Sign unit delivers 5.31 Gy/min at the isocenter using the largest 2.5 cm diameter collimator. Due to the noncoplanar beam arrangement and dynamic dose shaping features, the GammaPod Trade-Mark-Sign device is found to deliver

  2. Dosimetric evaluation of the Fricke gel dosimeter using the spectrophotometric technique for application in electron and neutron dosimetry

    International Nuclear Information System (INIS)

    Mangueira, Thyago Fressatti

    2009-01-01

    In this work the main dosimetric characteristics of the Fricke Xylenol Gel (FXG) solution were established for further application in the measurement of dose distribution of clinical electron fields. The dose-response curves of the FXG in a neutron field were also evaluated for the research in Boron Neutron Capture Therapy (BNCT) and industrial electron fields. The standard reading technique was the spectrophotometric. For the clinical field, the intra and inter-batch reproducibility are better than 1.4% and 5.1 %, respectively, the response presents a linear behavior for doses ranging from 0.2 to 40 Gy independently of the energy and the dose rate in the studied ranges. Due to the effects of the FXG natural oxidation, the optimum elapsed time between FXG preparation and irradiation was established as 24h period and the behavior of the dose-response curve of the FXG using the variation in the absorbance relative to the non-irradiated dosimeter as a basis during the whole studied period were not altered. The dose-response to the industrial electron beam presented an exponential decreasing behavior and the neutron beam for research in BNCT presented a linear behavior for the complete studied dose range. According to the obtained results for the different types of radiation studied for the FXG, there was no change in the position of the characteristic bands of the absorption spectrum due to the interaction of these radiation types. Additional tests were performed to determine the digital photographic imaging of FXG analyses viability and the application of FXG dosimetry on intracavitary brachytherapy. The good performance of the FXG dosimeter in the tests that were carried out indicates that this dosimeter may be applied to the tri-dimensional dose evaluation in radiotherapic treatments using electrons and neutron beams. (author)

  3. Abreu System - a dosimetric system to evaluate basic functioning parameters of roentgenography equipment

    International Nuclear Information System (INIS)

    Feital, J.C.

    1988-01-01

    This work shows a system to evaluate the half-value thickness of X-ray bundle. This system consists in a card with an aluminium filter, thermoluminescent dosemeter of lithium fluoride and radiographics films. )C.G.C.) [pt

  4. Preliminary shielding design evaluation for reactor assembly of SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

  5. Evaluation of Medical and Dosimetric Monitoring of the Personnel Exposed to Ionizing Radiations in Industry

    International Nuclear Information System (INIS)

    Hammou, A.; Ben Hariz, N.; Ben Omrane, L.

    2008-01-01

    Increasing use of the ionizing radiations in industry, in particular in the field of the non destructive testing (NDT) exposes the operators to low radiation doses. Therefore Radiation protection measures in this field are needed. We report the results of a survey carried out on a sample of 50 workers in NDT in Tunisia; Our purpose is to evaluate the professional training levels in radiation protection of the operators, to determine their exposure dose rate. In case of over-exposure, to determine the causes, to evaluate the medical follow-up, and to propose adequate recommendations

  6. SU-E-T-446: Evaluation of the Dosimetric Properties of a Diode Detector to Proton Radiosurgery

    Energy Technology Data Exchange (ETDEWEB)

    Teran, A [Loma Linda University Medical Center, Loma Linda, CA (United States); San Diego State University, San Diego, CA (United States); McAuley, G; Slater, J M [Loma Linda University, Loma Linda, CA (United States); Slater, J D; Wroe, A [Loma Linda University Medical Center, Loma Linda, CA (United States)

    2014-06-01

    Purpose: To test the PTW PR60020 proton dosimetry diode in radiation fields relevant to proton radiosurgery applications and evaluate its suitability as a high resolution, real time dosimetry device. Methods: Data was collected using our standard nominal radiosurgery energies of 126 MeV and 155 MeV through a single stage scattering system, corresponding to a range of 9.7 and 15 cm in water respectively. Various beam modulations were tested as part of this study. Depth dose and beam profile measurements were completed with the PTW PR60020 dosimetry diode with comparative measurements using a PTW Markus ionization chamber and EBT2 Gafchromic film. Monte Carlo simulations were also completed for comparison. Results: The single 1 mm{sup 2} by 20 μm thick sensitive volume allowed for high spatial resolution measurements while maintaining sufficient sensitive volume to ensure that measurements could be completed without excessive beam delivery. Depth dose profiles exhibited negligible LET dependence which typically impacts film and other solid state dosimetry devices, while beam ranges measured with the PTW diode were within 1 mm of ion chamber data. In an edge on arrangement beam profiles were also measured within 0.5 mm full-width at half-maximum at various depths as compared to film and simulation data. Conclusion: The PTW PR60020 proved to be a very useful radiation metrology apparatus for proton radiosurgery applications. Its waterproof and rugged construction allowed for easy deployment in phantoms or water tanks that are commonly used in proton radiosurgery QA. Dosimetrically, the diode exhibited negligible LET dependence as a function of depth, while in edge on arrangement to the incident proton beam it facilitated the measurement of beam profiles with a spatial resolution comparable to both Monte Carlo and film measurements. This project was sponsored in part by funding from the Department of Defense (DOD# W81XWH-BAA-10-1)

  7. Dose-to-medium vs. dose-to-water: Dosimetric evaluation of dose reporting modes in Acuros XB for prostate, lung and breast cancer

    Directory of Open Access Journals (Sweden)

    Suresh Rana

    2014-12-01

    Full Text Available Purpose: Acuros XB (AXB dose calculation algorithm is available for external beam photon dose calculations in Eclipse treatment planning system (TPS. The AXB can report the absorbed dose in two modes: dose-to-water (Dw and dose-to-medium (Dm. The main purpose of this study was to compare the dosimetric results of the AXB_Dm with that of AXB_Dw on real patient treatment plans. Methods: Four groups of patients (prostate cancer, stereotactic body radiation therapy (SBRT lung cancer, left breast cancer, and right breast cancer were selected for this study, and each group consisted of 5 cases. The treatment plans of all cases were generated in the Eclipse TPS. For each case, treatment plans were computed using AXB_Dw and AXB_Dm for identical beam arrangements. Dosimetric evaluation was done by comparing various dosimetric parameters in the AXB_Dw plans with that of AXB_Dm plans for the corresponding patient case. Results: For the prostate cancer, the mean planning target volume (PTV dose in the AXB_Dw plans was higher by up to 1.0%, but the mean PTV dose was within ±0.3% for the SBRT lung cancer. The analysis of organs at risk (OAR results in the prostate cancer showed that AXB_Dw plans consistently produced higher values for the bladder and femoral heads but not for the rectum. In the case of SBRT lung cancer, a clear trend was seen for the heart mean dose and spinal cord maximum dose, with AXB_Dw plans producing higher values than the AXB_Dm plans. However, the difference in the lung doses between the AXB_Dm and AXB_Dw plans did not always produce a clear trend, with difference ranged from -1.4% to 2.9%. For both the left and right breast cancer, the AXB_Dm plans produced higher maximum dose to the PTV for all cases. The evaluation of the maximum dose to the skin showed higher values in the AXB_Dm plans for all 5 left breast cancer cases, whereas only 2 cases had higher maximum dose to the skin in the AXB_Dm plans for the right breast cancer

  8. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

  9. Abreu system - A dosimetric system to evaluate basic parameters of photofluorographic X-ray machine

    International Nuclear Information System (INIS)

    Feital, J.C.

    1987-01-01

    In Brazil, photofluorographic X-ray machines are used for cuberculosis mass screening throughout the country. The exact number of these X-ray equipment is unknown but it is estimated to be around 1000 operating units. Twelve million miniature chest radiographs are taken per year. In order to make local inspections speedier and also aiming at its postal use, a system has been developed wich evaluates the entrace exposure of the patient, the X-ray beam half-value layer ( leading to the evaluation of the tube's total filtration ) and the beam's field size. It consists of a piece of cardboard where filters, TLDs and X-ray films are inserted. So far the system has been tested in 53 X-ray machines in Rio de Janeiro. The results show that it can be used in a national survey program. (Author) [pt

  10. Evaluation of nuclear reactor based activation analysis techniques

    International Nuclear Information System (INIS)

    Obrusnik, I.; Kucera, J.

    1977-09-01

    A survey is presented of the basic types of activation analysis applied in environmental control. Reactor neutron activation analysis is described (including the reactor as a neutron source, sample activation in the reactor, methodology of neutron activation analysis, sample transport into the reactor and sample packaging after irradiation, instrumental activation analysis with radiochemical separation, data measurement and evaluation, sampling and sample preparation). Sources of environmental contamination with trace elements, sampling and sample analysis by neutron activation are described. The analysis is described of soils, waters and biological materials. Methods are shown of evaluating neutron activation analysis results and of their interpretation for purposes of environmental control. (J.B.)

  11. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  12. Dosimetric- and geometric evaluation of adaptive H&N IMRT using deformable image registration

    DEFF Research Database (Denmark)

    Eiland, R. B.; Behrens, C. F.; Sjöström, D.

    2012-01-01

    CT and the dose recalculated. DVH points (D50 for parotid glands and Dmax for spinal cord) were evaluated. Conformity index (CI), lesion coverage fraction (LCF) and normal tissue overdose fraction (NTOF) was evaluated with regard to target coverage. Results: The PTV volume was estimated larger for dCT than Re......CT was 11.2% (range  28.0; 16.7), 26.2% (range 42.1; 3.4) and 10.9% (range 33.3; 32.3) for parotid dxt, parotid sin and spinal cord, respectively. The median CMS was 0.51 cm (range 0.19; 2.22). DSC had a median of 0.47 (range 0.45; 0.85). The median relative deviation from ReCT in DVH points...... for parotid dxt, parotid sin and spinal cord was 8.3% (range 8.4; 25.3), 12.7% (range 28.6; 31.0), and 1.3% (range 5.4; 31.8), respectively. CI, LCF and NTOF are visualized in the figure. Ideal values of CI and LCF are unity and zero for NTOF. Conclusions: The DIR produced geometrical results similar...

  13. Abreu system - dosimetric system to evaluate the performance of the basic parameters of photofluorographic equipments

    International Nuclear Information System (INIS)

    Silva Feita, J.C. da.

    1986-01-01

    In Brazil, photofluorographic X-ray machines are used for tuberculosis mass screening throughout the country. The exact number of these X-ray equipment is unknown, but it is estimated to be around 1000 operating units. Twelve million miniature chest radiographs are taken per year. In order to make local inspections speedier and also aiming at its postal use, a system has been developed which evaluates the entrance exposure of the patients, the X-ray beam half-value layer (leading to the evaluation of the tube's total filtration) and the beam's field size. It consists of a piece of cardboard where filters, TLDs and X-ray films are inserted. So far the system has been tested in 53 X-ray machines. The results show that it can be used in a national survey program. The data collected were used for the calculation of doses and this showed the influence of field size and tube voltage on the dose to the thyroid, uterus, ovaries, bone marrow and lungs. Furthermore, the results can be used to estimate population doses and risks factors due to photofluorographic examinations. (author)

  14. Evaluation of TRIGA Mark II reactor in Turkey

    International Nuclear Information System (INIS)

    Bilge, Ali Nezihi

    1990-01-01

    There are two research reactors in Turkey and one of them is the university Triga Mark II reactor which was in service since 1979 both for education and industrial application purposes. The main aim of this paper is to evaluate the spectrum of the services carried by Turkish Triga Mark II reactor. In this work, statistical distribution of the graduate works and applications, by using Triga Mark II reactor is examined and evaluated. In addition to this, technical and scientific uses of this above mentioned reactor are also investigated. It was already showed that the uses and benefits of this reactor can not be limited. If the sufficient work and service is given, NDT and industrial applications can also be carried economically. (orig.)

  15. A study of the plan dosimetric evaluation on the rectal cancer treatment

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Hyun Hak; An, Beom Seok; Kim, Dae Il; Lee, Yang Hoon; Lee, Je Hee [Dept. of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of)

    2016-12-15

    In order to minimize the dose of femoral head as an appropriate treatment plan for rectal cancer radiation therapy, we compare and evaluate the usefulness of 3-field 3D conformal radiation therapy(below 3fCRT), which is a universal treatment method, and 5-field 3D conformal radiation therapy(below 5fCRT), and Volumetric Modulated Arc Therapy (VMAT). The 10 cases of rectal cancer that treated with 21EX were enrolled. Those cases were planned by Eclipse(Ver. 10.0.42, Varian, USA), PRO3(Progressive Resolution Optimizer 10.0.28) and AAA(Anisotropic Analytic Algorithm Ver. 10.0.28). 3fCRT and 5fCRT plan has 0 degrees, 270 degrees, 90 degrees and 0 degrees, 95 degrees, 45 degrees, 315 degrees, 265 degrees gantry angle, respectively. VMAT plan parameters consisted of 15MV coplanar 360 degrees 1 arac. Treatment prescription was employed delivering 54Gy to recum in 30 fractions. To minimize the dose difference that shows up randomly on optimizing, VMAT plans were optimized and calculated twice, and normalized to the target V100%=95%. The indexes of evaluation are D of Both femoral head and aceta fossa, total MU, H.I.(Homogeneity index) and C.I.(Conformity index) of the PTV. All VMAT plans were verified by gamma test with portal dosimetry using EPID. D of Rt. femoral head was 53.08 Gy, 50.27 Gy, and 30.92 Gy, respectively, in the order of 3fCRT, 5fCRT, and VMAT treatment plan. Likewise, Lt. Femoral head showed average 53.68 Gy, 51.01 Gy and 29.23 Gy in the same order. D of Rt. aceta fossa was 54.86 Gy, 52.40 Gy, 30.37 Gy, respectively, in the order of 3fCRT, 5fCRT, and VMAT treatment plan. Likewise, Lt. Femoral head showed average 53.68 Gy, 51.01 Gy and 29.23 Gy in the same order. The maximum dose of both femoral head and aceta fossa was higher in the order of 3fCRT, 5fCRT, and VMAT treatment plan. C.I. showed the lowest VMAT treatment plan with an average of 1.64, 1.48, and 0.99 in the order of 3fCRT, 5fCRT, and VMAT treatment plan. There was no significant difference on H

  16. Dosimetric Evaluation of High-Dose-Rate Interstitial Brachytherapy Boost Treatments for Localized Prostate Cancer

    International Nuclear Information System (INIS)

    Froehlich, Georgina; Agoston, Peter; Loevey, Jozsef; Somogyi, Andras; Fodor, Janos; Polgar, Csaba; Major, Tibor

    2010-01-01

    Purpose: to quantitatively evaluate the dose distributions of high-dose-rate (HDR) prostate implants regarding target coverage, dose homogeneity, and dose to organs at risk. Material and methods: treatment plans of 174 implants were evaluated using cumulative dose-volume histograms (DVHs). The planning was based on transrectal ultrasound (US) imaging, and the prescribed dose (100%) was 10 Gy. The tolerance doses to rectum and urethra were 80% and 120%, respectively. Dose-volume parameters for target (V90, V100, V150, V200, D90, D min ) and quality indices (DNR [dose nonuniformity ratio], DHI [dose homogeneity index], CI [coverage index], COIN [conformal index]) were calculated. Maximum dose in reference points of rectum (D r ) and urethra (D u ), dose to volume of 2 cm 3 of the rectum (D 2ccm ), and 0.1 cm 3 and 1% of the urethra (D 0.1ccm and D1) were determined. Nonparametric correlation analysis was performed between these parameters. Results: the median number of needles was 16, the mean prostate volume (V p ) was 27.1 cm 3 . The mean V90, V100, V150, and V200 were 90%, 97%, 39% and 13%, respectively. The mean D90 was 109%, and the D min was 87%. The mean doses in rectum and urethra reference points were 75% and 119%, respectively. The mean volumetric doses were D 2ccm = 49% for the rectum, D 0.1ccm = 126%, and D1 = 140% for the urethra. The mean DNR was 0.37, while the DHI was 0.60. The mean COIN was 0.66. The Spearman rank order correlation coefficients for volume doses to rectum and urethra were R(D r , D 2ccm ) = 0.69, R(D u , D 0.1ccm ) = 0.64, R(D u , D1) = 0.23. Conclusion: US-based treatment plans for HDR prostate implants based on the real positions of catheters provided acceptable dose distributions. In the majority of the cases, the doses to urethra and rectum were kept below the defined tolerance levels. For rectum, the dose in reference points correlated well with dose-volume parameters. For urethra dose characterization, the use of D1 volumetric

  17. SU-F-T-604: Dosimetric Evaluation of Intracranial Stereotactic Radiotherapy Plans On a LINAC

    Energy Technology Data Exchange (ETDEWEB)

    Sheth, N; Tabibian, A; Rose, J; Alvelo, M; Perel, C; Laiken, K; Kim, A [Bayonne Medical Center, Bayonne, NJ (United States)

    2016-06-15

    Purpose: To evaluate the dosimetry of cranial stereotactic radiotherapy (SRT) plans of varying techniques on linac that meets appropriate TG-142 tolerances using 1 cm leaf width multileaf collimator (MLC). Methods: Seventeen spherical targets were generated in the center of a head phantom with diameters ranging 8 mm to 40 mm. SRT plans used 100° non-coplanar arcs and 5 couch angles with 35° spacing. The field size was target plus 1 mm margin. Four plans were created for each target: symmetrical jaws blocking for 5 arcs with 0° collimator (J1C), symmetrical jaws blocking with 5 clockwise arcs with 0° collimator and 5 counter-clockwise arcs with 45° collimator (J2C), MLC blocking for 5 dynamic conformal arcs with 0° collimator (M1C), and MLC blocking for 5 clockwise dynamic conformal arcs with 0° collimators and 5 counter-clockwise dynamic conformal arcs with 45° collimator (M2C).Conformity was evaluated using a ratio of Rx to target volume (PITV). Heterogeneity was determined using a ratio of maximum dose to Rx dose. Falloff was scored using CGIg: difference of effective radii of spheres equal to half and full Rx volumes. Results: All plans met RTOG SRS criteria for conformity and heterogeneity. The mean PITV was 1.52±0.07, 1.49±0.08, 1.39±0.05, and 1.37±0.04 for J1C, J2C, M1C, and M2C plans respectively. The mean CGIg was 75.35±15.79, 74.19±16.66, 77.14±15.12, and 76.28±15.78 for J1C, J2C, M1C, and M2C plans respectively. The mean MDPD was 1.25±0.00 for all techniques. Conclusion: Clinically acceptable SRT plans for spherical targets were created on a linac with 1 cm MLC. Adding two collimator angles and MLC to arcs each improved conformity. The MLC improved the dose falloff while two collimator angles degraded it. This technique can expand the availability of SRT to patients especially to those who cannot travel to a facility with a dedicated stereotactic radiosurgery machine.

  18. Development of a computational system based in the code GEANT4 for dosimetric evaluation in radiotherapy

    International Nuclear Information System (INIS)

    Oliveira, Alex Cristovao Holanda de

    2016-01-01

    The incidence of cancer has grown in Brazil, as well as around the world, following the change in the age profile of the population. One of the most important techniques and commonly used in cancer treatment is radiotherapy. Around 60% of new cases of cancer use radiation in at least one phase of treatment. The most used equipment for radiotherapy is a linear accelerator (Linac) which produces electron or X-ray beams in energy range from 5 to 30 MeV. The most appropriate way to irradiate a patient is determined during treatment planning. Currently, treatment planning system (TPS) is the main and the most important tool in the process of planning for radiotherapy. The main objective of this work is to develop a computational system based on the MC code Geant4 for dose evaluations in photon beam radiotherapy. In addition to treatment planning, these dose evaluations can be performed for research and quality control of equipment and TPSs. The computer system, called Quimera, consists of a graphical user interface (qGUI) and three MC applications (qLinacs, qMATphantoms and qNCTphantoms). The qGUI has the function of interface for the MC applications, by creating or editing the input files, running simulations and analyzing the results. The qLinacs is used for modeling and generation of Linac beams (phase space). The qMATphantoms and qNCTphantoms are used for dose calculations in virtual models of physical phantoms and computed tomography (CT) images, respectively. From manufacturer's data, models of a Varian Linac photon beam and a Varian multileaf collimator (MLC) were simulated in the qLinacs. The Linac and MLC modelling were validated using experimental data. qMATphamtoms and qNCTphantoms were validated using IAEA phase spaces. In this first version, the Quimera can be used for research, radiotherapy planning of simple treatments and quality control in photon beam radiotherapy. The MC applications work independent of the qGUI and the qGUI can be used for

  19. SU-F-T-604: Dosimetric Evaluation of Intracranial Stereotactic Radiotherapy Plans On a LINAC

    International Nuclear Information System (INIS)

    Sheth, N; Tabibian, A; Rose, J; Alvelo, M; Perel, C; Laiken, K; Kim, A

    2016-01-01

    Purpose: To evaluate the dosimetry of cranial stereotactic radiotherapy (SRT) plans of varying techniques on linac that meets appropriate TG-142 tolerances using 1 cm leaf width multileaf collimator (MLC). Methods: Seventeen spherical targets were generated in the center of a head phantom with diameters ranging 8 mm to 40 mm. SRT plans used 100° non-coplanar arcs and 5 couch angles with 35° spacing. The field size was target plus 1 mm margin. Four plans were created for each target: symmetrical jaws blocking for 5 arcs with 0° collimator (J1C), symmetrical jaws blocking with 5 clockwise arcs with 0° collimator and 5 counter-clockwise arcs with 45° collimator (J2C), MLC blocking for 5 dynamic conformal arcs with 0° collimator (M1C), and MLC blocking for 5 clockwise dynamic conformal arcs with 0° collimators and 5 counter-clockwise dynamic conformal arcs with 45° collimator (M2C).Conformity was evaluated using a ratio of Rx to target volume (PITV). Heterogeneity was determined using a ratio of maximum dose to Rx dose. Falloff was scored using CGIg: difference of effective radii of spheres equal to half and full Rx volumes. Results: All plans met RTOG SRS criteria for conformity and heterogeneity. The mean PITV was 1.52±0.07, 1.49±0.08, 1.39±0.05, and 1.37±0.04 for J1C, J2C, M1C, and M2C plans respectively. The mean CGIg was 75.35±15.79, 74.19±16.66, 77.14±15.12, and 76.28±15.78 for J1C, J2C, M1C, and M2C plans respectively. The mean MDPD was 1.25±0.00 for all techniques. Conclusion: Clinically acceptable SRT plans for spherical targets were created on a linac with 1 cm MLC. Adding two collimator angles and MLC to arcs each improved conformity. The MLC improved the dose falloff while two collimator angles degraded it. This technique can expand the availability of SRT to patients especially to those who cannot travel to a facility with a dedicated stereotactic radiosurgery machine.

  20. Dosimetric Evaluation of High-Dose-Rate Interstitial Brachytherapy Boost Treatments for Localized Prostate Cancer

    Energy Technology Data Exchange (ETDEWEB)

    Froehlich, Georgina [Semmelweis Univ., Budapest (Hungary); Dept. of Radiotherapy, National Inst. of Oncology, Budapest (Hungary); Agoston, Peter; Loevey, Jozsef; Somogyi, Andras; Fodor, Janos; Polgar, Csaba; Major, Tibor [Dept. of Radiotherapy, National Inst. of Oncology, Budapest (Hungary)

    2010-07-15

    Purpose: to quantitatively evaluate the dose distributions of high-dose-rate (HDR) prostate implants regarding target coverage, dose homogeneity, and dose to organs at risk. Material and methods: treatment plans of 174 implants were evaluated using cumulative dose-volume histograms (DVHs). The planning was based on transrectal ultrasound (US) imaging, and the prescribed dose (100%) was 10 Gy. The tolerance doses to rectum and urethra were 80% and 120%, respectively. Dose-volume parameters for target (V90, V100, V150, V200, D90, D{sub min}) and quality indices (DNR [dose nonuniformity ratio], DHI [dose homogeneity index], CI [coverage index], COIN [conformal index]) were calculated. Maximum dose in reference points of rectum (D{sub r}) and urethra (D{sub u}), dose to volume of 2 cm{sup 3} of the rectum (D{sub 2ccm}), and 0.1 cm{sup 3} and 1% of the urethra (D{sub 0.1ccm} and D1) were determined. Nonparametric correlation analysis was performed between these parameters. Results: the median number of needles was 16, the mean prostate volume (V{sub p}) was 27.1 cm{sup 3}. The mean V90, V100, V150, and V200 were 90%, 97%, 39% and 13%, respectively. The mean D90 was 109%, and the D{sub min} was 87%. The mean doses in rectum and urethra reference points were 75% and 119%, respectively. The mean volumetric doses were D{sub 2ccm} = 49% for the rectum, D{sub 0.1ccm} = 126%, and D1 = 140% for the urethra. The mean DNR was 0.37, while the DHI was 0.60. The mean COIN was 0.66. The Spearman rank order correlation coefficients for volume doses to rectum and urethra were R(D{sub r}, D{sub 2ccm}) = 0.69, R(D{sub u}, D{sub 0.1ccm}) = 0.64, R(D{sub u}, D1) = 0.23. Conclusion: US-based treatment plans for HDR prostate implants based on the real positions of catheters provided acceptable dose distributions. In the majority of the cases, the doses to urethra and rectum were kept below the defined tolerance levels. For rectum, the dose in reference points correlated well with dose

  1. Dosimetric evaluation of high-dose-rate interstitial brachytherapy boost treatments for localized prostate cancer.

    Science.gov (United States)

    Fröhlich, Georgina; Agoston, Péter; Lövey, József; Somogyi, András; Fodor, János; Polgár, Csaba; Major, Tibor

    2010-07-01

    To quantitatively evaluate the dose distributions of high-dose-rate (HDR) prostate implants regarding target coverage, dose homogeneity, and dose to organs at risk. Treatment plans of 174 implants were evaluated using cumulative dose-volume histograms (DVHs). The planning was based on transrectal ultrasound (US) imaging, and the prescribed dose (100%) was 10 Gy. The tolerance doses to rectum and urethra were 80% and 120%, respectively. Dose-volume parameters for target (V90, V100, V150, V200, D90, D(min)) and quality indices (DNR [dose nonuniformity ratio], DHI [dose homogeneity index], CI [coverage index], COIN [conformal index]) were calculated. Maximum dose in reference points of rectum (D(r)) and urethra (D(u)), dose to volume of 2 cm(3) of the rectum (D(2ccm)), and 0.1 cm(3) and 1% of the urethra (D(0.1ccm) and D1) were determined. Nonparametric correlation analysis was performed between these parameters. The median number of needles was 16, the mean prostate volume (V(p)) was 27.1 cm(3). The mean V90, V100, V150, and V200 were 99%, 97%, 39%, and 13%, respectively. The mean D90 was 109%, and the D(min) was 87%. The mean doses in rectum and urethra reference points were 75% and 119%, respectively. The mean volumetric doses were D(2ccm) = 49% for the rectum, D(0.1ccm) = 126%, and D1 = 140% for the urethra. The mean DNR was 0.37, while the DHI was 0.60. The mean COIN was 0.66. The Spearman rank order correlation coefficients for volume doses to rectum and urethra were R(D(r),D(2ccm)) = 0.69, R(D(u),D0.(1ccm)) = 0.64, R(D(u),D1) = 0.23. US-based treatment plans for HDR prostate implants based on the real positions of catheters provided acceptable dose distributions. In the majority of the cases, the doses to urethra and rectum were kept below the defined tolerance levels. For rectum, the dose in reference points correlated well with dose-volume parameters. For urethra dose characterization, the use of D1 volumetric parameter is recommended.

  2. Dosimetric evaluation of lead and tungsten eye shields in electron beam treatment

    International Nuclear Information System (INIS)

    Shiu, Almon S.; Tung, Samuel S.; Gastorf, Robert J.; Hogstrom, Kenneth R.; Morrison, William H.; Peters, Lester J.

    1996-01-01

    Purpose: The purpose of this study is to report that commercially available eye shields (designed for orthovoltage x-rays) are inadequate to protect the ocular structures from penetrating electrons for electron beam energies equal to or greater than 6 MeV. Therefore, a prototype medium size tungsten eye shield was designed and fabricated. The advantages of the tungsten eye shield over lead are discussed. Methods and Materials: Electron beams (6-9 MeV) are often used to irradiate eyelid tumors to curative doses. Eye shields can be placed under the eyelids to protect the globe. Film and thermoluminescent dosimeters (TLDs) were used within a specially constructed polystyrene eye phantom to determine the effectiveness of various commercially available internal eye shields (designed for orthovoltage x-rays). The same procedures were used to evaluate a prototype medium size tungsten eye shield (2.8 mm thick), which was designed and fabricated for protection of the globe from penetrating electrons for electron beam energy equal to 9 MeV. A mini-TLD was used to measure the dose enhancement due to electrons backscattered off the tungsten eye shield, both with or without a dental acrylic coating that is required to reduce discomfort, permit sterilization of the shield, and reduce the dose contribution from backscattered electrons. Results: Transmission of a 6 MeV electron beam through a 1.7 mm thick lead eye shield was found to be 50% on the surface (cornea) of the phantom and 27% at a depth of 6 mm (lens). The thickness of lead required to stop 6-9 MeV electron beams is impractical. In place of lead, a prototype medium size tungsten eye shield was made. For 6 to 9 MeV electrons, the doses measured on the surface (cornea) and at 6 mm (lens) and 21 mm (retina) depths were all less than 5% of the maximum dose of the open field (4 x 4 cm). Electrons backscattered off a tungsten eye shield without acrylic coating increased the lid dose from 85 to 123% at 6 MeV and 87 to 119% at

  3. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  4. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  5. Dosimetric evaluation of total marrow irradiation using 2 different planning systems

    International Nuclear Information System (INIS)

    Nalichowski, Adrian; Eagle, Don G.; Burmeister, Jay

    2016-01-01

    This study compared 2 different treatment planning systems (TPSs) for quality and efficiency of total marrow irradiation (TMI) plans. The TPSs used in this study were VOxel-Less Optimization (VoLO) (Accuray Inc, Sunnyvale, CA) using helical dose delivery on a Tomotherapy Hi-Art treatment unit and Eclipse (Varian Medical Systems Inc, Palo Alto, CA) using volumetric modulated arc therapy (VMAT) dose delivery on a Varian iX treatment unit. A total dose of 1200 cGy was prescribed to cover 95% of the planning target volume (PTV). The plans were optimized and calculated based on a single CT data and structure set using the Alderson Rando phantom (The Phantom Laboratory, Salem, NY) and physician contoured target and organ at risk (OAR) volumes. The OARs were lungs, heart, liver, kidneys, brain, and small bowel. The plans were evaluated based on plan quality, time to optimize the plan and calculate the dose, and beam on time. The resulting mean and maximum doses to the PTV were 1268 and 1465 cGy for VoLO and 1284 and 1541 cGy for Eclipse, respectively. For 5 of 6 OAR structures the VoLO system achieved lower mean and D10 doses ranging from 22% to 52% and 3% to 44%, respectively. Total computational time including only optimization and dose calculation were 0.9 hours for VoLO and 3.8 hours for Eclipse. These times do not include user-dependent target delineation and field setup. Both planning systems are capable of creating high-quality plans for total marrow irradiation. The VoLO planning system was able to achieve more uniform dose distribution throughout the target volume and steeper dose fall off, resulting in superior OAR sparing. VoLO's graphics processing unit (GPU)–based optimization and dose calculation algorithm also allowed much faster creation of TMI plans.

  6. Dosimetric evaluation of total marrow irradiation using 2 different planning systems

    Energy Technology Data Exchange (ETDEWEB)

    Nalichowski, Adrian, E-mail: nalichoa@karmanos.org [Karmanos Cancer Center, Detroit, MI (United States); Eagle, Don G. [Wayne State University School of Medicine, Detroit, MI (United States); Burmeister, Jay [Karmanos Cancer Center, Detroit, MI (United States); Wayne State University School of Medicine, Detroit, MI (United States)

    2016-10-01

    This study compared 2 different treatment planning systems (TPSs) for quality and efficiency of total marrow irradiation (TMI) plans. The TPSs used in this study were VOxel-Less Optimization (VoLO) (Accuray Inc, Sunnyvale, CA) using helical dose delivery on a Tomotherapy Hi-Art treatment unit and Eclipse (Varian Medical Systems Inc, Palo Alto, CA) using volumetric modulated arc therapy (VMAT) dose delivery on a Varian iX treatment unit. A total dose of 1200 cGy was prescribed to cover 95% of the planning target volume (PTV). The plans were optimized and calculated based on a single CT data and structure set using the Alderson Rando phantom (The Phantom Laboratory, Salem, NY) and physician contoured target and organ at risk (OAR) volumes. The OARs were lungs, heart, liver, kidneys, brain, and small bowel. The plans were evaluated based on plan quality, time to optimize the plan and calculate the dose, and beam on time. The resulting mean and maximum doses to the PTV were 1268 and 1465 cGy for VoLO and 1284 and 1541 cGy for Eclipse, respectively. For 5 of 6 OAR structures the VoLO system achieved lower mean and D10 doses ranging from 22% to 52% and 3% to 44%, respectively. Total computational time including only optimization and dose calculation were 0.9 hours for VoLO and 3.8 hours for Eclipse. These times do not include user-dependent target delineation and field setup. Both planning systems are capable of creating high-quality plans for total marrow irradiation. The VoLO planning system was able to achieve more uniform dose distribution throughout the target volume and steeper dose fall off, resulting in superior OAR sparing. VoLO's graphics processing unit (GPU)–based optimization and dose calculation algorithm also allowed much faster creation of TMI plans.

  7. SU-F-T-82: Dosimetric Evaluation of a Shield Used for Hemi-Body Skin Electron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rivers, C; Singh, A [Roswell Park Cancer Institute, Buffalo, NY (United States); AlDahlawi, I; Wang, I; Podgorsak, M [Roswell Park Cancer Institute, Buffalo, NY (United States); State University of New York at Buffalo, Buffalo, NY (United States)

    2016-06-15

    Purpose: We had several mycosis fungoides patients with a limited disease to about half of the skin surface. A custom-made plywood shield was used to protect the non-targeted skin region with our total skin electron irradiation (TSEI) technique. We report a dosimetric evaluation for our “hemi-body” skin electron irradiation technique. Methods: The technique is similar to our clinical total skin electron irradiation (TSEI), performed with a six-pair dual field (Stanford technique) at an extended source-to-skin distance (SSD) of 377 cm, with the addition of a plywood shield placed 50 cm from the patient. The shield is made of three layers of standard 5/8″ thick plywood (total thickness of 4.75 cm) that are clamped securely on an adjustable-height stand. Gafchromic EBT3 films were used in assessing the shield’s transmission factor and the extend of the dose penumbra region. To verify the dose delivered for hemi-body skin radiation in a real patient treatment, in-vivo dosimetry using Gafchromic EBT3 films were performed. Film pieces were taped on the patient skin to measure the dose received during the first two fractions, placed on the forehead and upper body (shielded region); and also at the level of pelvic area, left thigh, and left ankle. Results: The shield transmission factor was found to be 10%, and the width of the penumbra (80-to-20% dose fall-off) was about 12 cm. In-vivo dosimetry of a real case confirmed the expected shielded area dose. Conclusion: Hemi-Body skin electron irradiation at an extended SSD is feasible with the addition of a plywood shield at a distance from patient skin. The penumbra dose region and the shield’s transmission factor should be evaluated prior to clinical use. We have treated several hemi-body skin patients with our custom-made plywood shield, the current patient measurements are representative of these for other patients as well.

  8. Dosimetric evaluation of photon dose calculation under jaw and MLC shielding

    International Nuclear Information System (INIS)

    Fogliata, A.; Clivio, A.; Vanetti, E.; Nicolini, G.; Belosi, M. F.; Cozzi, L.

    2013-01-01

    Purpose: The accuracy of photon dose calculation algorithms in out-of-field regions is often neglected, despite its importance for organs at risk and peripheral dose evaluation. The present work has assessed this for the anisotropic analytical algorithm (AAA) and the Acuros-XB algorithms implemented in the Eclipse treatment planning system. Specifically, the regions shielded by the jaw, or the MLC, or both MLC and jaw for flattened and unflattened beams have been studied.Methods: The accuracy in out-of-field dose under different conditions was studied for two different algorithms. Measured depth doses out of the field, for different field sizes and various distances from the beam edge were compared with the corresponding AAA and Acuros-XB calculations in water. Four volumetric modulated arc therapy plans (in the RapidArc form) were optimized in a water equivalent phantom, PTW Octavius, to obtain a region always shielded by the MLC (or MLC and jaw) during the delivery. Doses to different points located in the shielded region and in a target-like structure were measured with an ion chamber, and results were compared with the AAA and Acuros-XB calculations. Photon beams of 6 and 10 MV, flattened and unflattened were used for the tests.Results: Good agreement between calculated and measured depth doses was found using both algorithms for all points measured at depth greater than 3 cm. The mean dose differences (±1SD) were −8%± 16%, −3%± 15%, −16%± 18%, and −9%± 16% for measurements vs AAA calculations and −10%± 14%, −5%± 12%, −19%± 17%, and −13%± 14% for Acuros-XB, for 6X, 6 flattening-filter free (FFF), 10X, and 10FFF beams, respectively. The same figures for dose differences relative to the open beam central axis dose were: −0.1%± 0.3%, 0.0%± 0.4%, −0.3%± 0.3%, and −0.1%± 0.3% for AAA and −0.2%± 0.4%, −0.1%± 0.4%, −0.5%± 0.5%, and −0.3%± 0.4% for Acuros-XB. Buildup dose was overestimated with AAA, while Acuros-XB gave

  9. 3-D dosimetric evaluation of 2.5 mm HD120 multileaf system for intensity modulated stereotactic radiosurgery using optical CT based polymer gel dosimetry

    International Nuclear Information System (INIS)

    Wuu, C-S; Kessel, Jack; Xu, Y

    2009-01-01

    A Trilogy TX equipped with a 2.5 mm HD120 multileaf collimator system is available for the treatment of radiosurgery and IMRT. In this study, we evaluated the 3-D dosimetric impact of leaf width on an IMRT radiosurgery plan by comparing the target coverage and the dose gradient around the target, produced from both a 2.5 mm HD120 high-definition MLC system and a 5mm-leaf-width millennium 120 MLC system, using an optical CT based polymer gel dosimetry system. The 2.5 mm MLC improves target conformity and surrounding tissue sparing when compared to that of 5 mm MLC.

  10. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  11. Evaluation of physiological parameters and their influence on doses calculated from two alternative dosimetric models for the gastrointestinal tract

    International Nuclear Information System (INIS)

    Lessard, E.T.; Skrable, K.W.

    1981-01-01

    Two dosimetric models, the catenary compartmental model and the slug flow model are examined using three sets of physiological parameters. The impact of physiological parameters on the dosimetry of the tract is illustrated by comparing calculated maximum permissible daily activity ingestion rates for single, unabsorbed, particle emitting radionuclides with an effective energy term of unity. The conclusions drawn from this intercomparison of six different cases are: (1) Current dosimetric models which use physiological parameters described in this article do not significantly disagree, and (2) For the determination of average dose equivalent rates to segments of the tract due to chronic, long term ingestion of any radionuclide, the catenary compartmental model is a mathematically simpler approach. The catenary model in addition has certain advantages for the calculation of the photon dose contribution to one segment from cumulated activity (disintegrations) in another segment

  12. Dosimetric evaluation of lithium carbonate (Li2CO3) as a dosemeter for gamma-radiation dose measurements.

    Science.gov (United States)

    Popoca, R; Ureña-Núñez, F

    2009-06-01

    This work reports the possibility of using lithium carbonate as a dosimetric material for gamma-radiation measurements. Carboxi-radical ions, CO(2)(-) and CO(3)(-), arise from the gamma irradiation of Li(2)CO(3), and these radical ions can be quantified by electron paramagnetic resonance (EPR) spectrometry. The EPR-signal response of gamma-irradiated lithium carbonate has been investigated to determine some dosimetric characteristics such as: peak-to-peak signal intensity versus gamma dose received, zero-dose response, signal fading, signal repeatability, batch homogeneity, dose rate effect and stability at different environmental conditions. Using the conventional peak-to-peak method of stable ion radicals, it is concluded that lithium carbonate could be used as a gamma dosemeter in the range of 3-100 Gy.

  13. Modeling atmospheric dispersion for reactor accident consequence evaluation

    International Nuclear Information System (INIS)

    Alpert, D.J.; Gudiksen, P.H.; Woodard, K.

    1982-01-01

    Atmospheric dispersion models are a central part of computer codes for the evaluation of potential reactor accident consequences. A variety of ways of treating to varying degrees the many physical processes that can have an impact on the predicted consequences exists. The currently available models are reviewed and their capabilities and limitations, as applied to reactor accident consequence analyses, are discussed

  14. An evaluation on environment radiation impact of pulsed reactor

    International Nuclear Information System (INIS)

    Gao Yingwei; Pu Gongxu; Li Jian

    1991-01-01

    The dose regulation, assessment scope and assessment method adopted by the environment impact evaluation for the pulsed reactor are discussed. The compute model, the compute programme and the compute result of the dose adopted for the model pulsed reactor are introduced. The probable environment radiation impact under normal status and accident status are also appraised

  15. A computer code for Tokamak reactor concepts evaluation

    International Nuclear Information System (INIS)

    Rosatelli, F.; Raia, G.

    1985-01-01

    A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

  16. Dosimetric Evaluation of Metal Artefact Reduction using Metal Artefact Reduction (MAR) Algorithm and Dual-energy Computed Tomography (CT) Method

    Science.gov (United States)

    Laguda, Edcer Jerecho

    Purpose: Computed Tomography (CT) is one of the standard diagnostic imaging modalities for the evaluation of a patient's medical condition. In comparison to other imaging modalities such as Magnetic Resonance Imaging (MRI), CT is a fast acquisition imaging device with higher spatial resolution and higher contrast-to-noise ratio (CNR) for bony structures. CT images are presented through a gray scale of independent values in Hounsfield units (HU). High HU-valued materials represent higher density. High density materials, such as metal, tend to erroneously increase the HU values around it due to reconstruction software limitations. This problem of increased HU values due to metal presence is referred to as metal artefacts. Hip prostheses, dental fillings, aneurysm clips, and spinal clips are a few examples of metal objects that are of clinical relevance. These implants create artefacts such as beam hardening and photon starvation that distort CT images and degrade image quality. This is of great significance because the distortions may cause improper evaluation of images and inaccurate dose calculation in the treatment planning system. Different algorithms are being developed to reduce these artefacts for better image quality for both diagnostic and therapeutic purposes. However, very limited information is available about the effect of artefact correction on dose calculation accuracy. This research study evaluates the dosimetric effect of metal artefact reduction algorithms on severe artefacts on CT images. This study uses Gemstone Spectral Imaging (GSI)-based MAR algorithm, projection-based Metal Artefact Reduction (MAR) algorithm, and the Dual-Energy method. Materials and Methods: The Gemstone Spectral Imaging (GSI)-based and SMART Metal Artefact Reduction (MAR) algorithms are metal artefact reduction protocols embedded in two different CT scanner models by General Electric (GE), and the Dual-Energy Imaging Method was developed at Duke University. All three

  17. SU-F-T-224: Importance of Timely Review of Daily Cone-Beam CTs: Dosimetric Evaluation of Rejected CBCTs for Head and Neck Patients

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, M; Yu, N; Joshi, N; Koyfman, S; Xia, P [Cleveland Clinic, Cleveland, OH (United States); Lin, S [Cleveland State University, Cleveland, OH (United States)

    2016-06-15

    Purpose: To dosimetrically evaluate the importance of timely reviewing daily CBCTs for patients with head and neck cancer. Methods: After each fraction daily cone-beam CT (CBCT) for head and neck patients are reviewed by physicians prior to next treatment. Physician rejected image registrations of CBCT were identified and analyzed for 17 patients. These CBCT images were rigidly fused with planning CT images and the contours from the planning CT were transferred to CBCTs. Because of limited extension in the superior-inferior dimension contours with partial volumes in CBCTs were discarded. The treatment isocenter was placed by applying the clinically recorded shifts to the volume isocenter of the CBCT. Dose was recalculated at the shifted isocenter using a homogeneous dose calculation algorithm. Dosimetrically relevant changes defined as greater than 5% deviation from the clinically accepted plans but with homogeneous dose calculation were evaluated for the high dose (HD), intermediate dose (ID), and low dose (LD) CTVs, spinal cord, larynx, oropharynx, parotids, and submandibular glands. Results: Among seventeen rejected CBCTS, HD-CTVs, ID-CTVs, and LD-CTVs were completely included in the CBCTs for 17, 1, and 15 patients, respectively. The prescription doses to the HD-CTV, ID-CTV, and LD-CTV were received by < 95% of the CTV volumes in 5/17, 1/1, and 5/15 patients respectively. For the spinal cord, the maximum doses (D0.03cc) were increased > 5% in 13 of 17 patients. For the oropharynx, larynx, parotid, and submandibular glands, the mean dose of these organs at risk was increased > 5% in 7/17, 8/12, 11/16 and 6/16 patients, respectively. Conclusion: Timely review daily CBCTs for head and neck patients under daily CBCT guidance is important, and uncorrected setup errors can translate to dosimetrically relevant dose increases in organsat- risk and dose decreases in the clinical target volumes.

  18. SU-F-T-461: Dosimetric Evaluation of Indigenous Farmer Type Chamber FAR65- GB for Reference Dosimetry of FFF MV Photon Beam

    Energy Technology Data Exchange (ETDEWEB)

    Patwe, P; Mhatre, V; Dandekar, P [Sir HN RF Hospital, Mumbai, Maharashtra (India)

    2016-06-15

    Purpose: Indigenous Farmer type chamber FAR 65 GB is a reference class 0.6 cc ion chamber. It can be used for dosimetric evaluation of photon and high energy electron beams. We studied dosimetric characteristics of the chamber for 6MV and 10MV Flattening filter free FFF photon beams available on trueBEAM STx Linac. Methods: The study was carried out on trueBEAM STx Linac having 6 and 10 MV FFF photon beam with maximum dose rate 1400 and 2400 MU per min respectively. The dosimetric device to be evaluated is Rosalina Instruments FAR 65-GB Ion Chamber with active volume 0.65 cc, total active length 23.1cm, inner diameter of cylinder 6.2mm, wall thickness 0.4mm, inner electrode diameter 1mm. Inner and outer electrodes are made from Aluminium 2.7 gm per cc and graphite 1.82 gm per cc respectively. The ion chamber was placed along central axis of beam at 10cm depth and irradiated for 10cm × 10cm field size at SAD of 100 cm in plastic phantom. We studied Precision, Dose Linearity, Dose Rate dependence, directional dependence, Recombination effect. Recombination effect was determined using standard two-voltage method. Results: 1. Measurements were reproducible std deviation of 0.0105 and type A uncertainty 0.003265 under same set of reference conditions 2. Chamber exhibit dose linearity over a wider dose range. 3. Chamber shows dose rate independence for all available dose rate range. 4. Response of chamber with the angle of incidence of radiation is constant. 5. Recombination correction factors were 1.01848 and 1.02537 for dose rate 1400 and 2400 MU per min resp. Conclusion: Our study reveals that the chamber is prone to saturation effect at dose rate of 2400 MU per min. FAR 65-GB can be used for reference dosimetry of FFF MV photon beam with proper calculation of recombination effect.

  19. Exposure mode study to xenon-133 in a reactor building

    International Nuclear Information System (INIS)

    Perier, Aurelien

    2014-01-01

    The work described in this thesis focuses on the external and internal dose assessment to xenon-133. During the nuclear reactor operation, fission products and radioactive inert gases, as 133 Xe, are generated and might be responsible for the exposure of workers in case of clad defect. Particle Monte Carlo transport code is adapted in radioprotection to quantify dosimetric quantities. The study of exposure to xenon-133 is conducted by using Monte-Carlo simulations based on GEANT4, an anthropomorphic phantom, a realistic geometry of the reactor building, and compartmental models. The external exposure inside a reactor building is conducted with a realistic and conservative exposure scenario. The effective dose rate and the eye lens equivalent dose rate are determined by Monte-Carlo simulations. Due to the particular emission spectrum of xenon-133, the equivalent dose rate to the lens of eyes is discussed in the light of expected new eye dose limits. The internal exposure occurs while xenon-133 is inhaled. The lungs are firstly exposed by inhalation, and their equivalent dose rate is obtained by Monte-Carlo simulations. A biokinetic model is used to evaluate the internal exposure to xenon-133. This thesis gives us a better understanding to the dosimetric quantities related to external and internal exposure to xenon-133. Moreover the impacts of the dosimetric changes are studied on the current and future dosimetric limits. The dosimetric quantities are lower than the current and future dosimetric limits. (author)

  20. Liquid metal systems development: reactor vessel support structure evaluation

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration

  1. Human factors engineering evaluation of the UTR-10 Reactor

    International Nuclear Information System (INIS)

    Lahti, D.; Nilius, D.; Heithoff, D.; Roche, G.; Sage, S.

    1982-01-01

    This paper is a description of a student design team's review and evaluation of Iowa State University's University Test Reactor (UTR-10). The review was based on how well the control room of the UTR-10 measured up to selected portions of NUREG-0800, chapter 18, Human Factor Engineering/Standard Review Plan Development. The review was conducted by inspecting the reactor and interviewing reactor operators. The control room workspace, instrumentation controls and other equipment were evaluated from a human factors engineering point of view that takes into account both system demands and operator capabilities. Identification, assessment, and suggestion for control room design modifications that correct inadequate or unsuitable items was made

  2. Decision model for evaluating reactor disposition of excess plutonium

    International Nuclear Information System (INIS)

    Edmunds, T.

    1995-02-01

    The US Department of Energy is currently considering a range of technologies for disposition of excess weapon plutonium. Use of plutonium fuel in fission reactors to generate spent fuel is one class of technology options. This report describes the inputs and results of decision analyses conducted to evaluate four evolutionary/advanced and three existing fission reactor designs for plutonium disposition. The evaluation incorporates multiple objectives or decision criteria, and accounts for uncertainty. The purpose of the study is to identify important and discriminating decision criteria, and to identify combinations of value judgments and assumptions that tend to favor one reactor design over another

  3. Dosimetric fundamentals

    International Nuclear Information System (INIS)

    Nahum, A.E.

    2004-01-01

    This text covers some important concepts in radiation dosimetry with an emphasis on cavity theory, i.e. the theoretical evaluation of D med /D det , for two important classes of detector, 'large' and Bragg-Gray. Monte Carlo simulation continues to play a major role in evaluating this expression through its ability to compute the fluence spectra of electrons and photons as a function of their position in a medium. The key results in the paper can be summarised thus: - Fluence Φ = dN/da Σds/dV and is a scalar quantity. - Kerma K = dE tr /dm, i.e. kinetic energy (k.e.) transferred per unit mass; collision kerma K c excludes charged-particle k.e. converted to Bremsstrahlung. - Kerma and fluence are related by K med = Φ (μ tr /ρ) med for photons of energy E; for collision kerma, K c , the mass energy-absorption coefficient μ en replaces μ tr . - D med = (K c ) med under conditions of charged particle equilibrium (CPE), for a medium med irradiated by photons. - For a fluence Φ of charged particles, e.g. electrons, in medium med, the absorbed dose D med = Φ (S col /ρ) med provided there is δ-ray equilibrium. - For large detectors under photon irradiation (i.e. in which there is CPE as e - ranges - detector size), D med /D det is given by (μ en /ρ) med /(μ en /ρ) det which is evaluated over the photon spectrum at the detector position: e.g. TLD (e.g. LiF) in kV X-ray beams are large. - For 'small' or Bragg-Gray detectors under photon or electron irradiation (e - ranges - detector dimensions), D med /D det is given by (S col /ρ) med /(S col /ρ) det , the (mass) stopping-power ratio, usually written S med.det : e.g. (air-filled) ionisation chambers behave as Bragg -Gray detectors in megavoltage photon and electron beams, but not in kV X-ray beams. - Bragg-Gray theory was extended by Spencer and Attix to take into account the finite range of δ-rays. - General cavity theory provides an approximate treatment of detectors which are neither 'large' nor 'small

  4. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  5. About the contribution of occupational health's services for risk factors evaluation, medical and dosimetric follow-up in the workers monitoring exposed to ionising radiations in France

    International Nuclear Information System (INIS)

    Bailloeuil, C.; Gonin, M.; Gerondal, M.

    2006-01-01

    Full text of publication follows: French national regulation (31/03/2003) indicates principles of a global approach about the medical and dosimetric follow-up in the workers monitoring. Legislator insists on risks and expositions trace ability along all professional career and after. The aim of this French specific system is to institute medical clinic aspects in accordance with dosimetry and professional risks. The occupational practitioners are approved practitioners who have followed a specific training. The organisation guarantees that a worker will be followed by one specific practitioner in order to reinforce the quality and the traceability of follow up. Medical supervision is done at taking on and at least once a year. It means to identify and take care of risks and expositions at work stations. If necessary, biological measurements and recommendations about collective and individual protection equipments complete the estimation of risks. On the subject of emergency, first aid is delivered on sites by occupational health personnel, either for classic medical problem or for radiological accident. Furthermore, occupational health personnel assist outside emergency services with whom we have specific conventions. External dosimetric follow-up is done with radiation protection qualified expert of the company. The internal contamination supervision and internal dose evaluation are done by the occupational health services. Measurements either whole body counts or radio-toxicologic analysis are submitted to technical quality process. Beyond the respect of regulatory dose limits, the aim of the dosimetric follow-up is the contribution to the preparation of work places with strong dosimetric focus. Informations at workers are dispensed about every risks and every kinds of risks: ionising radiation health effects, ionising radiation and pregnancy, high exposition, chemical risks, work at heat, asbestos. All data are conserved 50 years after the exposure These data

  6. Dosimetric evaluation of Radiotherapy units wit {sup 60}Co; Evaluacion dosimetrica de unidades de radioterapia con {sup 60}Co

    Energy Technology Data Exchange (ETDEWEB)

    Leon, B. Salinas de; Tovar M, V.; Becerril V, A. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2000-07-01

    The SSDL network of the IAEA performs, every year, quality audit tests for radiotherapy services ({sup 60} Co units and linear accelerators), and for national SSDL as well. Because of the SSDL-Mexico results in these tests and due to our enthusiasm and confidence in our work, a parallel test has been done , which is described in this talk as well as the results. Nowadays, a second parallel test goes up, which could confirm our optimism and open the possibility to our country to start a national dosimetric audit of {sup 60} Co radiotherapy units. (Author)

  7. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marta, V; Manuel, P J [Instituto de Fusion Nuclear (DENIM)/ETSII, Universidad Politecnica Madrid (UPM) (Spain); Sedano Luis, A [Ministerio de Educacion y Ciencia, Ciemat (Spain)], E-mail: marta@denim.upm.es

    2008-05-15

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed {approx} 1 gr-T/a ({approx}20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimization in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  8. An evaluation of fast reactor blankets

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1974-01-01

    A comparative study of different types of fast reactor radial blankets is presented. Included are blankets of fertile material UO 2 , THO 2 and Th-metal blankets of pure reflectors C, BeO, Ni and combinations of reflecting and fertile blankets. The results for 1000MWe cores indicate that there is no incentive to use other than fertile blankets. The most favorable fertile material is thorium due to the prospective higher price of U-233

  9. Evaluation of physiological parameters and their influence on doses calculated from two alternative dosimetric models for the gastrointestinal tract

    International Nuclear Information System (INIS)

    Lessard, E.T.; Skrable, K.W.

    1981-01-01

    Two dosimetric models, the catenary compartmental model (Be70) and the slug flow model (Sk75), are examined using three sets of physiological parameters: those proposed by Eve, those proposed by ICRP, and those obtained from the Textbook of Physiology and Biochemistry by Bell et al. The impact of physiological parameters on the dosimetry of the tract is illustrated by comparing calculated maximum permissible daily activity ingestion rates for single, unabsorbed, particle emitting radionuclides with an effective energy term of unity. The conclusions drawn from this intercomparison of six different cases are: Current dosimetric models which use physiological parameters described in this article do not significantly disagree, and for the determination of average dose equivalent rates to segments of the tract due to chronic, long term ingestion of any radionuclide, the catenary compartmental model is a mathematically simpler approach. The catenary model in addition has certain advantages for the calculation of the photon dose contribution to one segment from cumulated activity (disintegrations) in another segment

  10. Utility industry evaluation of the Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; DelGeorge, L.O.; Tramm, T.R.; Gibbons, J.P.; High, M.D.; Neils, G.H.; Pilmer, D.F.; Tomonto, J.R.; Wells, J.T.

    1990-02-01

    A team of utility industry representatives evaluated the Sodium Advanced Fast Reactor plant design, a current liquid metal reactor design created by an industrial team led by Rockwell International under Department of Energy sponsorship. The utility industry team concluded that the plant design offers several attractive characteristics, especially in the safety arena, as well as preserving the traditional attraction of liquid metal reactors, very high fuel utilization. Specific comments and recommendations are provided as a contribution towards improving an already attractive plant design. 18 refs

  11. Evaluation of neutron exposure conditions for the Buffalo Reactor

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Kellogg, L.S.; McElroy, W.N.; Baldwin, C.A.

    1984-04-01

    The light water test reactor at the Nuclear Science and Technology Facility of the State University of New York at Buffalo is currently being used to irradiate specimens in in-core positions for NRC-sponsored metallurgical tests. It is important that the neutron exposures for these Buffalo tests be consistent with those determined for related irradiations in the BSR and ORR reactor at ORNL. Therefore, HEDL National Reactor Dosimetry Center dosimetry procedures and ORNL calculational procedures were used for an evaluation of typical test conditions

  12. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  13. SU-D-17A-01: Geometric and Dosimetric Evaluation of a 4D-CBCT Reconstruction Technique Using Prior Knowledge

    International Nuclear Information System (INIS)

    Zhang, Y; Yin, F; Ren, L

    2014-01-01

    Purpose: To evaluate a 4D-CBCT reconstruction technique both geometrically and dosimetrically Methods: A prior-knowledge guided 4DC-BCT reconstruction method named the motion-modeling and free-form deformation (MM-FD) has been developed. MM-FD views each phase of the 4D-CBCT as a deformation of a prior CT volume. The deformation field is first solved by principal component analysis based motion modeling, followed by constrained free-form deformation.The 4D digital extended-cardiac- torso (XCAT) phantom was used for comprehensive evaluation. Based on a simulated 4D planning CT of a lung patient, 8 different scenarios were simulated to cover the typical on-board anatomical and respiratory variations: (1) synchronized and (2) unsynchronized motion amplitude change for body and tumor; tumor (3) shrinkage and (4) expansion; tumor average position shift in (5) superior-inferior (SI) direction, (6) anterior-posterior (AP) direction and (7) SI, AP and lateral directions altogether; and (8) tumor phase shift relative to the respiratory cycle of the body. Orthogonal-view 30° projections were simulated based on the eight patient scenarios to reconstruct on-board 4D-CBCTs. For geometric evaluation, the volume-percentage-difference (VPD) was calculated to assess the volumetric differences between the reconstructed and the ground-truth tumor.For dosimetric evaluation, a gated treatment plan was designed for the prior 4D-CT. The dose distributions were calculated on the reconstructed 4D-CBCTs and the ground-truth images for comparison. The MM-FD technique was compared with MM-only and FD-only techniques. Results: The average (±s.d.) VPD values of reconstructed tumors for MM-only, FDonly and MM-FD methods were 59.16%(± 26.66%), 75.98%(± 27.21%) and 5.22%(± 2.12%), respectively. The average min/max/mean dose (normalized to prescription) of the reconstructed tumors by MM-only, FD-only, MM-FD methods and ground-truth tumors were 78.0%/122.2%/108.2%, 13%/117.7%/86%, 58

  14. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  15. Evaluation of the positional accuracy and dosimetric properties of a three-dimensional printed device for head and neck immobilization

    International Nuclear Information System (INIS)

    Sato, Kiyokazu; Yanagawa, Isao; Takeda, Ken; Dobashi, Suguru; Kadoya, Noriyuki; Ito, Kengo; Chiba, Mizuki; Jingu, Keiichi; Kishi, Kazuma

    2017-01-01

    Our aim was to investigate the feasibility of a three-dimensional (3D)-printed head-and-neck (HN) immobilization device by comparing its positional accuracy and dosimetric properties with those of a conventional immobilization device (CID). We prepared a 3D-printed immobilization device (3DID) consisting of a mask and headrest with acrylonitrile-butadiene-styrene resin developed from the computed tomography data obtained by imaging a HN phantom. For comparison, a CID comprising a thermoplastic mask and headrest was prepared using the same HN phantom. We measured the setup error using the ExacTrac X-ray image system. Furthermore, using the ionization chamber and the water-equivalent phantom, we measured the changes in the dose due to the difference in the immobilization device material from the photon of 4 MV and 6 MV. The positional accuracy of the two devices were almost similar in each direction except in the vertical, lateral, and pitch directions (t-test, p<0.0001), and the maximum difference was 1 mm, and 1deg. The standard deviations were not statistically different in each direction except in the longitudinal (F-test, p=0.034) and roll directions (F-test, p<0.0001). When the thickness was the same, the dose difference was almost similar at a 50 mm depth. At a 1 mm depth, the 3DID-plate had a 2.9-4.2% lower dose than the CID-plate. This study suggested that the positional accuracy and dosimetric properties of 3DID were almost similar to those of CID. (author)

  16. SU-E-T-368: Evaluating Dosimetric Outcome of Modulated Photon Radiotherapy (XMRT) Optimization for Head and Neck Patients

    Energy Technology Data Exchange (ETDEWEB)

    McGeachy, P; Villarreal-Barajas, JE; Khan, R [University of Calgary, Calgary, AB (Canada); Tom Baker Cancer Centre, Calgary, AB (Canada); Zinchenko, Y [University of Calgary, Calgary, AB (Canada)

    2015-06-15

    Purpose: The dosimetric outcome of optimized treatment plans obtained by modulating the photon beamlet energy and fluence on a small cohort of four Head and Neck (H and N) patients was investigated. This novel optimization technique is denoted XMRT for modulated photon radiotherapy. The dosimetric plans from XMRT for H and N treatment were compared to conventional, 6 MV intensity modulated radiotherapy (IMRT) optimization plans. Methods: An arrangement of two non-coplanar and five coplanar beams was used for all four H and N patients. Both XMRT and IMRT were subject to the same optimization algorithm, with XMRT optimization allowing both 6 and 18 MV beamlets while IMRT was restricted to 6 MV only. The optimization algorithm was based on a linear programming approach with partial-volume constraints implemented via the conditional value-at-risk method. H and N constraints were based off of those mentioned in the Radiation Therapy Oncology Group 1016 protocol. XMRT and IMRT solutions were assessed using metrics suggested by International Commission on Radiation Units and Measurements report 83. The Gurobi solver was used in conjunction with the CVX package to solve each optimization problem. Dose calculations and analysis were done in CERR using Monte Carlo dose calculation with VMC{sub ++}. Results: Both XMRT and IMRT solutions met all clinical criteria. Trade-offs were observed between improved dose uniformity to the primary target volume (PTV1) and increased dose to some of the surrounding healthy organs for XMRT compared to IMRT. On average, IMRT improved dose to the contralateral parotid gland and spinal cord while XMRT improved dose to the brainstem and mandible. Conclusion: Bi-energy XMRT optimization for H and N patients provides benefits in terms of improved dose uniformity to the primary target and reduced dose to some healthy structures, at the expense of increased dose to other healthy structures when compared with IMRT.

  17. SU-F-T-649: Dosimetric Evaluation of Non-Coplanar Arc Therapy Using a Novel Rotating Gamma Ray System

    Energy Technology Data Exchange (ETDEWEB)

    Eldib, A; Chibani, O; Jin, L; Fan, J; Veltchev, I; Ma, C [Fox Chase Cancer Center, Philadelphia, PA (United States); Mora, G [Universidade de Lisboa, Codex, Lisboa (Portugal); Li, J [Cyber Medical Inc, Xian, Shaanxi (China)

    2016-06-15

    Purpose: Stereotactic intra and extra-cranial body radiation therapy has evolved with advances in treatment accuracy, effective radiation dose, and parameters necessary to maximize machine capabilities. Novel gamma systems with a ring type gantry were developed having the ability to perform oblique arcs. The aim of this study is to explore the dosimetric advantages of this new system. Methods: The rotating Gamma system is named CybeRay (Cyber Medical Corp., Xian, China). It has a treatment head of 16 cobalt-60 sources focused to the isocenter, which can rotate 360° on the ring gantry and swing 35° in the superior direction. Treatment plans were generated utilizing our in-house Monte Carlo treatment planning system. A cylindrical phantom was modeled with 2mm voxel size. Dose inside the cylindrical phantom was calculated for coplanar and non-coplanar arcs. Dosimetric differences between CybeRay cobalt beams and CyberKnife 6MV beams were compared in a lung phantom and for previously treated SBRT patients. Results: The full width at half maxima of cross profiles in the S-I direction for the coplanar setup matched the cone sizes, while for the non-coplanar setup, FWHM was larger by 2mm for a 10mm cone and about 5mm for larger cones. In the coronal and sagittal view, coplanar beams showed elliptical shaped isodose lines, while non-coplanar beams showed circular isodose lines. Thus proper selection of the oblique angle and cone size can aid optimal dose matching to the target volume. Comparing a single 5mm cone from CybeRay to that from CyberKnife showed similar penumbra in a lung phantom but CybeRay had significant lower doses beyond lung tissues. Comparable treatment plans were obtained with CybeRay as that from CyberKnife.ConclusionThe noncoplanar multiple source arrangement of CybeRay will be of great clinical benefits for stereotactic intra and extra-cranial radiation therapy.

  18. Nonproliferation and safeguard considerations: Pebble Bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, conpare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  19. Technical evaluation of corium cooling at the reactor cavity

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Chan, Eun Sun; Lee, Jae Hun; Lee, Jong In

    1998-01-01

    To terminate the progression of the severe accident and mitigate the accident consequences, corium cooling has been suggested as one of most important design features considered in the severe accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially, the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the corium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study, technical descriptions are briefly described for the important issues related to the corium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluations for those items have been performed considering present technical levels including the experiment and analytical works

  20. Evolution of dosimetric phantoms

    International Nuclear Information System (INIS)

    Reddy, A.R.

    2010-01-01

    In this oration evolution of the dosimetric phantoms for radiation protection and for medical use is briefly reviewed. Some details of the development of Indian Reference Phantom for internal dose estimation are also presented

  1. An Overview of the International Reactor Physics Experiment Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  2. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

    International Nuclear Information System (INIS)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

  3. Dosimetric evaluation of the feasibility of stereotactic body radiotherapy for primary lung cancer with lobe-specific selective elective nodal irradiation.

    Science.gov (United States)

    Komatsu, Tetsuya; Kunieda, Etsuo; Kitahara, Tadashi; Akiba, Takeshi; Nagao, Ryuta; Fukuzawa, Tsuyoshi

    2016-01-01

    More than 10% of all patients treated with stereotactic body radiotherapy (SBRT) for primary lung cancer develop regional lymph node recurrence. We evaluated the dosimetric feasibility of SBRT with lobe-specific selective elective nodal irradiation (ENI) on dose-volume histograms. A total of 21 patients were treated with SBRT for Stage I primary lung cancer between January 2010 and June 2012 at our institution. The extents of lobe-specific selective ENI fields were determined with reference to prior surgical reports. The ENI fields included lymph node stations (LNS) 3 + 4 + 11 for the right upper lobe tumors, LNS 7 + 11 for the right middle or lower lobe tumors, LNS 5 + 11 for the left upper lobe tumors, and LNS 7 + 11 for the left lower lobe tumors. A composite plan was generated by combining the ENI plan and the SBRT plan and recalculating for biologically equivalent doses of 2 Gy per fraction, using a linear quadratic model. The V20 of the lung, D(1cm3) of the spinal cord, D(1cm3) and D(10cm3) of the esophagus and D(10cm3) of the tracheobronchial wall were evaluated. Of the 21 patients, nine patients (43%) could not fulfill the dose constraints. In all these patients, the distance between the planning target volume (PTV) of ENI (PTVeni) and the PTV of SBRT (PTVsrt) was ≤2.0 cm. Of the three patients who developed regional metastasis, two patients had isolated lymph node failure, and the lymph node metastasis was included within the ENI field. When the distance between the PTVeni and PTVsrt is >2.0 cm, SBRT with selective ENI may therefore dosimetrically feasible. © The Author 2015. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology.

  4. SU-F-T-545: Dosimetric and Radiobiological Evaluation of Dose Calculation Algorithms On Prostate Stereotactic Body Radiotherapy Using Conventional Flattened and Flattening-Filter-Free Beam

    International Nuclear Information System (INIS)

    Kang, S; Suh, T; Chung, J; Eom, K; Lee, J

    2016-01-01

    Purpose: The purpose of this study is to evaluate the dosimetric and radiobiological impact of Acuros XB (AXB) and Anisotropic Analytic Algorithm (AAA) dose calculation algorithms on prostate stereotactic body radiation therapy plans with both conventional flattened (FF) and flattening-filter free (FFF) modes. Methods: For thirteen patients with prostate cancer, SBRT planning was performed using 10-MV photon beam with FF and FFF modes. The total dose prescribed to the PTV was 42.7 Gy in 7 fractions. All plans were initially calculated using AAA algorithm in Eclipse treatment planning system (11.0.34), and then were re-calculated using AXB with the same MUs and MLC files. The four types of plans for different algorithms and beam energies were compared in terms of homogeneity and conformity. To evaluate the radiobiological impact, the tumor control probability (TCP) and normal tissue complication probability (NTCP) calculations were performed. Results: For PTV, both calculation algorithms and beam modes lead to comparable homogeneity and conformity. However, the averaged TCP values in AXB plans were always lower than in AAA plans with an average difference of 5.3% and 6.1% for 10-MV FFF and FF beam, respectively. In addition, the averaged NTCP values for organs at risk (OARs) were comparable. Conclusion: This study showed that prostate SBRT plan were comparable dosimetric results with different dose calculation algorithms as well as delivery beam modes. For biological results, even though NTCP values for both calculation algorithms and beam modes were similar, AXB plans produced slightly lower TCP compared to the AAA plans.

  5. SU-F-T-545: Dosimetric and Radiobiological Evaluation of Dose Calculation Algorithms On Prostate Stereotactic Body Radiotherapy Using Conventional Flattened and Flattening-Filter-Free Beam

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S; Suh, T [The catholic university of Korea, Seoul (Korea, Republic of); Chung, J; Eom, K [Seoul National University Bundang Hospital (Korea, Republic of); Lee, J [Konkuk University Medical Center (Korea, Republic of)

    2016-06-15

    Purpose: The purpose of this study is to evaluate the dosimetric and radiobiological impact of Acuros XB (AXB) and Anisotropic Analytic Algorithm (AAA) dose calculation algorithms on prostate stereotactic body radiation therapy plans with both conventional flattened (FF) and flattening-filter free (FFF) modes. Methods: For thirteen patients with prostate cancer, SBRT planning was performed using 10-MV photon beam with FF and FFF modes. The total dose prescribed to the PTV was 42.7 Gy in 7 fractions. All plans were initially calculated using AAA algorithm in Eclipse treatment planning system (11.0.34), and then were re-calculated using AXB with the same MUs and MLC files. The four types of plans for different algorithms and beam energies were compared in terms of homogeneity and conformity. To evaluate the radiobiological impact, the tumor control probability (TCP) and normal tissue complication probability (NTCP) calculations were performed. Results: For PTV, both calculation algorithms and beam modes lead to comparable homogeneity and conformity. However, the averaged TCP values in AXB plans were always lower than in AAA plans with an average difference of 5.3% and 6.1% for 10-MV FFF and FF beam, respectively. In addition, the averaged NTCP values for organs at risk (OARs) were comparable. Conclusion: This study showed that prostate SBRT plan were comparable dosimetric results with different dose calculation algorithms as well as delivery beam modes. For biological results, even though NTCP values for both calculation algorithms and beam modes were similar, AXB plans produced slightly lower TCP compared to the AAA plans.

  6. Human factors evaluation of the engineering test reactor control room

    International Nuclear Information System (INIS)

    Banks, W.W.; Boone, M.P.

    1981-03-01

    The Reactor and Process Control Rooms at the Engineering Test Reactor were evaluated by a team of human factors engineers using available human factors design criteria. During the evaluation, ETR, equipment and facilities were compared with MIL-STD-1472-B, Human Engineering design Criteria for Military Systems. The focus of recommendations centered on: (a) displays and controls; placing displays and controls in functional groups; (b) establishing a consistent color coding (in compliance with a standard if possible); (c) systematizing annunciator alarms and reducing their number; (d) organizing equipment in functional groups; and (e) modifying labeling and lines of demarcation

  7. Consequence evaluation of hypothetical reactor pressure vessel support failure

    International Nuclear Information System (INIS)

    Lu, S.C.; Holman, G.S.; Lambert, H.E.

    1991-01-01

    This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

  8. Cardiac dosimetric evaluation of deep inspiration breath-hold level variances using computed tomography scans generated from deformable image registration displacement vectors

    International Nuclear Information System (INIS)

    Harry, Taylor; Rahn, Doug; Semenov, Denis; Gu, Xuejun; Yashar, Catheryn; Einck, John; Jiang, Steve; Cerviño, Laura

    2016-01-01

    There is a reduction in cardiac dose for left-sided breast radiotherapy during treatment with deep inspiration breath-hold (DIBH) when compared with treatment with free breathing (FB). Various levels of DIBH may occur for different treatment fractions. Dosimetric effects due to this and other motions are a major component of uncertainty in radiotherapy in this setting. Recent developments in deformable registration techniques allow displacement vectors between various temporal and spatial patient representations to be digitally quantified. We propose a method to evaluate the dosimetric effect to the heart from variable reproducibility of DIBH by using deformable registration to create new anatomical computed tomography (CT) scans. From deformable registration, 3-dimensional deformation vectors are generated with FB and DIBH. The obtained deformation vectors are scaled to 75%, 90%, and 110% and are applied to the reference image to create new CT scans at these inspirational levels. The scans are then imported into the treatment planning system and dose calculations are performed. The average mean dose to the heart was 2.5 Gy (0.7 to 9.6 Gy) at FB, 1.2 Gy (0.6 to 3.8 Gy, p < 0.001) at 75% inspiration, 1.1 Gy (0.6 to 3.1 Gy, p = 0.004) at 90% inspiration, 1.0 Gy (0.6 to 3.0 Gy) at 100% inspiration or DIBH, and 1.0 Gy (0.6 to 2.8 Gy, p = 0.019) at 110% inspiration. The average mean dose to the left anterior descending artery (LAD) was 19.9 Gy (2.4 to 46.4 Gy), 8.6 Gy (2.0 to 43.8 Gy, p < 0.001), 7.2 Gy (1.9 to 40.1 Gy, p = 0.035), 6.5 Gy (1.8 to 34.7 Gy), and 5.3 Gy (1.5 to 31.5 Gy, p < 0.001), correspondingly. This novel method enables numerous anatomical situations to be mimicked and quantifies the dosimetric effect they have on a treatment plan.

  9. Cardiac dosimetric evaluation of deep inspiration breath-hold level variances using computed tomography scans generated from deformable image registration displacement vectors

    Energy Technology Data Exchange (ETDEWEB)

    Harry, Taylor [Department of Radiation Medicine and Applied Sciences, University of California San Diego, La Jolla, CA (United States); Department of Radiation Medicine, Oregon Health and Science University, Portland, OR (United States); Department of Nuclear Engineering and Radiation Health Physics, Oregon State University, Corvallis, OR (United States); Rahn, Doug; Semenov, Denis [Department of Radiation Medicine and Applied Sciences, University of California San Diego, La Jolla, CA (United States); Gu, Xuejun [Department of Radiation Oncology, University of Texas Southwestern Medical Center, Dallas, TX (United States); Yashar, Catheryn; Einck, John [Department of Radiation Medicine and Applied Sciences, University of California San Diego, La Jolla, CA (United States); Jiang, Steve [Department of Radiation Oncology, University of Texas Southwestern Medical Center, Dallas, TX (United States); Cerviño, Laura, E-mail: lcervino@ucsd.edu [Department of Radiation Medicine and Applied Sciences, University of California San Diego, La Jolla, CA (United States)

    2016-04-01

    There is a reduction in cardiac dose for left-sided breast radiotherapy during treatment with deep inspiration breath-hold (DIBH) when compared with treatment with free breathing (FB). Various levels of DIBH may occur for different treatment fractions. Dosimetric effects due to this and other motions are a major component of uncertainty in radiotherapy in this setting. Recent developments in deformable registration techniques allow displacement vectors between various temporal and spatial patient representations to be digitally quantified. We propose a method to evaluate the dosimetric effect to the heart from variable reproducibility of DIBH by using deformable registration to create new anatomical computed tomography (CT) scans. From deformable registration, 3-dimensional deformation vectors are generated with FB and DIBH. The obtained deformation vectors are scaled to 75%, 90%, and 110% and are applied to the reference image to create new CT scans at these inspirational levels. The scans are then imported into the treatment planning system and dose calculations are performed. The average mean dose to the heart was 2.5 Gy (0.7 to 9.6 Gy) at FB, 1.2 Gy (0.6 to 3.8 Gy, p < 0.001) at 75% inspiration, 1.1 Gy (0.6 to 3.1 Gy, p = 0.004) at 90% inspiration, 1.0 Gy (0.6 to 3.0 Gy) at 100% inspiration or DIBH, and 1.0 Gy (0.6 to 2.8 Gy, p = 0.019) at 110% inspiration. The average mean dose to the left anterior descending artery (LAD) was 19.9 Gy (2.4 to 46.4 Gy), 8.6 Gy (2.0 to 43.8 Gy, p < 0.001), 7.2 Gy (1.9 to 40.1 Gy, p = 0.035), 6.5 Gy (1.8 to 34.7 Gy), and 5.3 Gy (1.5 to 31.5 Gy, p < 0.001), correspondingly. This novel method enables numerous anatomical situations to be mimicked and quantifies the dosimetric effect they have on a treatment plan.

  10. Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

    International Nuclear Information System (INIS)

    Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

    2003-01-01

    The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

  11. Shadow corrosion evaluation in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Sanders, Ch.; Lysell, G.

    2000-01-01

    Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

  12. SU-F-J-121: Dosimetric Evaluation of Active Breathing Coordinator-Response Gating System Linked to Linear Accelerator in Volumetric Modulated Arc Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S; Zheng, Y; Albani, D; Colussi, V; Dorth, J; Sohn, J [Case Western University, Cleveland, OH (United States)

    2016-06-15

    Purpose: To reduce internal target volume (ITV), respiratory management is a must in imaging and treatment for lung, liver, and breast cancers. We investigated the dosimetric accuracy of VMAT treatment delivery with a Response™ gating system linked to linear accelerator. Methods: The Response™ gating module designed to directly control radiation beam by breath-holding with a ABC system (Elekta AB, Stockholm, Sweden) was tested for VMAT treatments. Seven VMAT plans including three conventional and four stereotactic body radiotherapy (SBRT) cases were evaluated. Each plan was composed of two or four arcs of 6MV radiation beam with prescribed dose ranged from 1.8 to 9 Gy per fraction. Each plan was delivered continuously without gating and delivered with multiple interruptions by the ResponseTM gating module with a 20 or 30 second breath-holding period. MapCheck2 and Gafchromic EBT3 films sandwiched in MapPHAN were used to measure the delivered dose with and without gating. Films were scanned on a flatbed color scanner, and red channel was extracted for film dosimetry. Gamma analysis was performed to analyze the dosimetrical accuracy of the radiation delivery with gating. Results: The measured doses with gating remarkably agree with the planned dose distributions in the results of gamma index passing rate (within 20% isodose; >98% for 3%/3mm and >92% for 2%/2mm in MapCheck2, and >91% for 3%/3mm criteria in EBT3 film except one case which was for large target and highly modulated). No significant difference (student t-test: p-value < 0.0005) was shown between the doses delivered with and without gating. There was no indication of radiation gap or overlapping during deliver interruption in film dosimetry. Conclusion: The Response™ gating system can be safely used during VMAT treatment. The accurate performance of the gating system linked to ABC can contribute to ITV reduction for SBRT using VMAT.

  13. Experimental evaluation of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1984-10-01

    The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

  14. A method for evaluation the activity of the reactor components

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Roth, Cs.

    2003-01-01

    The ability to predict the radioactivity levels of the reactor components is an important aspect from waste management point of view, as well as from radioprotection purposes. A special case is represented by the research reactors where, one of the major contributions to the radioactivity inventory is due to the experimental devices involved in various research works during reactor life. Generally, aluminum and aluminum alloys are used in manufacturing these devices; as a result, the work presented in this paper is focused on the qualitative and quantitative analysis of the radioactive isotopes contained in these materials. A device used for silicon doping by neutron transmutation that was placed near TRIGA reactor core is investigated. The isotopic content of various samplings drawn from various points of the device was analyzed by gamma spectrometry using a HPGe detector. Computations, using the MCNP5 code, are also performed in order to evaluate the reaction rates for all the isotopes and their reactions. The Monte Carlo simulations are performed for a detailed geometry and material composition of the reactor core and the device. The Origen-S code is also used in order to evaluate the isotopic inventory and the activity values. A detailed analysis regarding the possibility to estimate by computations and/or by gamma spectrometry the activity values of the isotopes which are of interest for decommissioning is presented in the paper. (authors)

  15. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  16. Evaluation of the influence of the TH-GEM detector components in dosimetric measurements of standard mammography beams

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Natália F.; Castro, Maysa C.; Caldas, Linda V.E., E-mail: nsilva@ipen.br, E-mail: maysadecastro@gmail.com, E-mail: fbelonsi@gmail.com, E-mail: lcaldas@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Silva, Tiago F.; Cintra, Felipe B.; Luz, Hugo N. da, E-mail: tfsilva@if.usp.br, E-mail: hugonluz@if.usp.br [Universidade de São Paulo (IF/USP), São Paulo, SP (Brazil). Instituto de Física

    2017-07-01

    GEM detectors have found applications in many areas due to their simplicity of construction, low cost, ruggedness and diversity of shape. A dosimeter with these qualities presents utility in several applications, as for example in diagnostic and therapeutic medicine, industrial radiography and nuclear meters. Furthermore, the high sensitivity provided by GEM detectors may extend their applications in low dose dosimetry. Based on these facts, it may be interesting to produce a prototype of a portable TH-GEM type detector with characteristics suitable for dosimetric use in X-rays with low and medium energies. The precise determination of the dosimeter characteristics is very important for laboratories of instrument calibration, as well as to determine how the various components of the detector may influence on the energy deposited in the sensitive volume. In this work, the results obtained about the influence of each one of the components present in this type of detector in standard mammography beams is presented. The code MCNP5 was used. The results allowed the adaptation of the detector to the desired conditions. (author)

  17. Dosimetric evaluation of the OneDoseTM MOSFET for measuring kilovoltage imaging dose from image-guided radiotherapy procedures.

    Science.gov (United States)

    Ding, George X; Coffey, Charles W

    2010-09-01

    The purpose of this study is to investigate the feasibility of using a single-use dosimeter, OneDose MOSFET designed for in vivo patient dosimetry, for measuring the radiation dose from kilovoltage (kV) x rays resulting from image-guided procedures. The OneDose MOSFET dosimeters were precalibrated by the manufacturer using Co-60 beams. Their energy response and characteristics for kV x rays were investigated by using an ionization chamber, in which the air-kerma calibration factors were obtained from an Accredited Dosimetry Calibration Laboratory (ADCL). The dosimetric properties have been tested for typical kV beams used in image-guided radiation therapy (IGRT). The direct dose reading from the OneDose system needs to be multiplied by a correction factor ranging from 0.30 to 0.35 for kilovoltage x rays ranging from 50 to 125 kVp, respectively. In addition to energy response, the OneDose dosimeter has up to a 20% reduced sensitivity for beams (70-125 kVp) incident from the back of the OneDose detector. The uncertainty in measuring dose resulting from a kilovoltage beam used in IGRT is approximately 20%; this uncertainty is mainly due to the sensitivity dependence of the incident beam direction relative to the OneDose detector. The ease of use may allow the dosimeter to be suitable for estimating the dose resulting from image-guided procedures.

  18. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1991-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author).

  19. Nuclear data evaluation and group constant generation for reactor analysis

    International Nuclear Information System (INIS)

    Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup

    1991-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author)

  20. The impact of WASH-1400 on reactor safety evaluation

    International Nuclear Information System (INIS)

    Tanguy, P.Y.

    1976-01-01

    Trends in reactor safety evaluation in France following the publication of WASH-1400 (the Rasmussen Report) are presented. What is called 'the meteorite case' is first schematically presented as follows: WASH-1400 shows nuclear risk equivalent to meteorite risk and reasonable corrections cannot make many orders of magnitude, consequently present safety rules are adequate. The very impact of WASH-1400 on safety approach is then discussed as for: assistance to deterministic safety analysis, introduction of probabilistic safety criteria, acceptable level of risk, and the use of results in research and reactor operating experience

  1. The evaluation of operator reliability factors on power reactor

    International Nuclear Information System (INIS)

    Karlina, Itjeu; Supriatna, Piping; W, Suharyo; Santosa, Kussigit; Darlis; S, Bambang; Y, Sasongko

    1999-01-01

    The sophisticated technology system was not assured the reliability system itself because it has contained a part of human dependence affected successfully of reactor operation either how work smoothly and safe or failure ac cured and then accident appears promptly. The evaluation of operator reliability factor on ABWR power reactor has been carried out which consist of criterion skill and workload according to NUREG/CR-2254, NUREG/CR-4016 and NUREG-0835 the reactor operation reliability emphasize to the operator are synergic between skill and workload themselves. The employee's skill will affect to the type and level of their tasks. The operator's skill depend on education and experiences, position or responsibility of tasks, physical conditions (age uninvalid of physic/mental

  2. Study, design and evaluation of nuclear reactor computer control system

    International Nuclear Information System (INIS)

    Menacer, S.

    1988-01-01

    Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

  3. Perturbation measurements in reactor LR-0 and their evaluation

    International Nuclear Information System (INIS)

    Rypar, W.; Faehrmann, K.H.

    1988-07-01

    To investigate space-dependent kinetic effects in reactors of the WWER-1000 type, two central and one eccentric perturbation measurements were performed in the zero power reactor LR-0 of the UJV Rez (CSSR) by trapeze-form movements of an absorber cluster. The measurements were based ona computer aided CAMAC system for the simultaneous data acquisition of 20 spatially distributed neutron detectors and for cluster movement control. The measurements were followed by a detailed evaluation in the ZfK Rossendorf (GDR) with respect to the calculation results of flux response obtained by nodal code HEXDYN3D, the aim of which was to demostrate the limits of the point reactor model and to account for space-dependent effects by approximative methods. A sensitive check of the calculation methods was made possible especially by the eccentric perturbation where the space dependent effects,due to a larger distance of cluster movement, were most significant. (author). 17 figs., 9 refs

  4. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  5. A Dosimetric Evaluation of Conventional Helmet Field Irradiation Versus Two-Field Intensity-Modulated Radiotherapy Technique

    International Nuclear Information System (INIS)

    Yu, James B.; Shiao, Stephen L.; Knisely, Jonathan

    2007-01-01

    Purpose: To compare dosimetric differences between conventional two-beam helmet field irradiation (external beam radiotherapy, EBRT) of the brain and a two-field intensity-modulated radiotherapy (IMRT) technique. Methods and Materials: Ten patients who received helmet field irradiation at our institution were selected for study. External beam radiotherapy portals were planned per usual practice. Intensity-modulated radiotherapy fields were created using the identical field angles as the EBRT portals. Each brain was fully contoured along with the spinal cord to the bottom of the C2 vertebral body. This volume was then expanded symmetrically by 0.5 cm to construct the planning target volume. An IMRT plan was constructed using uniform optimization constraints. For both techniques, the nominal prescribed dose was 3,000 cGy in 10 fractions of 300 cGy using 6-MV photons. Comparative dose-volume histograms were generated for each patient and analyzed. Results: Intensity-modulated radiotherapy improved dose uniformity over EBRT for whole brain radiotherapy. The mean percentage of brain receiving >105% of dose was reduced from 29.3% with EBRT to 0.03% with IMRT. The mean maximum dose was reduced from 3,378 cGy (113%) for EBRT to 3,162 cGy (105%) with IMRT. The mean percent volume receiving at least 98% of the prescribed dose was 99.5% for the conventional technique and 100% for IMRT. Conclusions: Intensity-modulated radiotherapy reduces dose inhomogeneity, particularly for the midline frontal lobe structures where hot spots occur with conventional two-field EBRT. More study needs to be done addressing the clinical implications of optimizing dose uniformity and its effect on long-term cognitive function in selected long-lived patients

  6. Dosimetric evaluation of 99mTc IgG as infection diagnostic agent for HIV positive patients

    International Nuclear Information System (INIS)

    Teran, Mariella; Paolino, Andrea; Vilar, Javier; Kapitan, Miguel; Andruskevicius, Patricia; Hermida, Juan C.; Gaudiano, Javier; Perez Sartori, Graciela; Savio Larriera, Eduardo

    2008-01-01

    A wide variety of radiopharmaceuticals are used as diagnostic or therapeutic agents. In this case 99m Tc-IgG was used to determine infection-inflammation processes in HIV patients, who sometimes are difficult to diagnose because of the presence of non specific signs and symptoms. The aim of this work was to estimate the hazard associated with the use of radiopharmaceuticals in nuclear medicine. In order to establish a proper design of kinetic studies and determine the radiation doses to individual human organs internal dosimetry methods were used. HIV positive patients with suspect of infection focus were administered via iv injection with 740 MBq (20 mCi) of 99m Tc-IgG. Anterior and posterior whole body images were acquired at 4 and 24 hours post injection in a gamma camera Mediso Medical Imaging, 1024 x 512 matrix. Geometric mean was calculated for different regions of interest taking into account decay, scattering and attenuation corrections. Blood and urine samples were collected at 1, 4, 8, 12 and 24 hours post injection. They were measured in a dose calibrator Capintec CR 5, corrections for geometry and decay were performed. For each patient, percentage of injected dose was calculated both for biological and image samples. The number of disintegrations was developed for those organs where higher concentration of activity was observed (liver, kidneys and spleen), the organs involved in the excretion (urinary bladder and intestines), red marrow and the reminder of the body. Total doses were estimated using OLINDA/EXM software. The code calculations showed that chosen organs as more compromised during the diagnostic procedure received very low effective doses. Correlation studies with calculations performed both for image and biological samples data were done. Despite the risk population under study the dosimetric estimations showed that 99m Tc-IgG is a safe radiopharmaceutical to be used in routine diagnostic procedures without hazardous effects. (author)

  7. Comparative evaluation of Map-Check and Arc-Check for dosimetric verification in patients treaties with IMRT

    International Nuclear Information System (INIS)

    Garcia, B.; Marquina, J.; Ramirez, J.; Gonzales, A.

    2014-08-01

    The dosimetric controls that are realized to the patients in the Intensity-Modulated Radiation Therapy (IMRT) and Volumetric Modulated Arc Therapy (VMAT) techniques; are indispensable since allows in real time to verify the quantity of imparted dose to the patient, these controls should be carried out every time that will begin a treatment, because these techniques impart dose dynamically modulating the dose intensity and movements of the Multi leaf Collimator (MLC), for they exist different diodes devices prepared in spiral (3-D) and planar form (2-D); that allows to estimate the dose fluence in a certain area. Treatment studies for head and neck with IMRT were compared regarding the reading average carried out by the diodes in the corresponding areas, using the criteria of the gamma index like dose difference 3% or 3m m of distance for both diode arrangements, in the IMRT case was found in Arc-Check a minor difference of 3/3 for an average of 99.37% of read diodes in a correct way contrary to the reading obtained with the Map-Check 3/3 an average of difference of 96.19%; in IMRT the difference was lower due to different factors like sensibility of the diodes reading, resolution, diodes disposition, as well as the average reading of entrance and exit of the radiation beams. Within the parameters delivered by the diodes arrangement is considered the positioning correction for both acceptance indexes like the gamma factor and the Distance-to-agreement (Dta), the existent difference of reading in factor gamma and Dta fundamentally is the way in like they compare the dose distribution; the Gamma uses dose averages of high and low gradients and Dta use only averages of areas of high gradients between the nearest points giving the distance as a result among the distribution point and the nearest point what makes stricter. (Author)

  8. Dosimetric evaluation of PLATO and Oncentra treatment planning systems for High Dose Rate (HDR) brachytherapy gynecological treatments

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Hardev; De La Fuente Herman, Tania; Showalter, Barry; Thompson, Spencer J.; Syzek, Elizabeth J.; Herman, Terence; Ahmad, Salahuddin [Department of Radiation Oncology, Peggy and Charles Stephenson Oklahoma Cancer Center, University of Oklahoma Health Sciences Center, Oklahoma City, OK 73104 (United States)

    2012-10-23

    This study compares the dosimetric differences in HDR brachytherapy treatment plans calculated with Nucletron's PLATO and Oncentra MasterPlan treatment planning systems (TPS). Ten patients (1 T1b, 1 T2a, 6 T2b, 2 T4) having cervical carcinoma, median age of 43.5 years (range, 34-79 years) treated with tandem and ring applicator in our institution were selected retrospectively for this study. For both Plato and Oncentra TPS, the same orthogonal films anterior-posterior (AP) and lateral were used to manually draw the prescription and anatomical points using definitions from the Manchester system and recommendations from the ICRU report 38. Data input for PLATO was done using a digitizer and Epson Expression 10000XL scanner was used for Oncentra where the points were selected on the images in the screen. The prescription doses for these patients were 30 Gy to points right A (RA) and left A (LA) delivered in 5 fractions with Ir-192 HDR source. Two arrangements: one dwell position and two dwell positions on the tandem were used for dose calculation. The doses to the patient points right B (RB) and left B (LB), and to the organs at risk (OAR), bladder and rectum for each patient were calculated. The mean dose and the mean percentage difference in dose calculated by the two treatment planning systems were compared. Paired t-tests were used for statistical analysis. No significant differences in mean RB, LB, bladder and rectum doses were found with p-values > 0.14. The mean percent difference of doses in RB, LB, bladder and rectum are found to be less than 2.2%, 1.8%, 1.3% and 2.2%, respectively. Dose calculations based on the two different treatment planning systems were found to be consistent and the treatment plans can be made with either system in our department without any concern.

  9. An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

    International Nuclear Information System (INIS)

    Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

    2006-01-01

    The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

  10. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  11. SU-E-J-94: Geometric and Dosimetric Evaluation of Deformation Image Registration Algorithms Using Virtual Phantoms Generated From Patients with Lung Cancer

    International Nuclear Information System (INIS)

    Shen, Z; Greskovich, J; Xia, P; Bzdusek, K

    2015-01-01

    Purpose: To generate virtual phantoms with clinically relevant deformation and use them to objectively evaluate geometric and dosimetric uncertainties of deformable image registration (DIR) algorithms. Methods: Ten lung cancer patients undergoing adaptive 3DCRT planning were selected. For each patient, a pair of planning CT (pCT) and replanning CT (rCT) were used as the basis for virtual phantom generation. Manually adjusted meshes were created for selected ROIs (e.g. PTV, lungs, spinal cord, esophagus, and heart) on pCT and rCT. The mesh vertices were input into a thin-plate spline algorithm to generate a reference displacement vector field (DVF). The reference DVF was used to deform pCT to generate a simulated replanning CT (srCT) that was closely matched to rCT. Three DIR algorithms (Demons, B-Spline, and intensity-based) were applied to these ten virtual phantoms. The images, ROIs, and doses were mapped from pCT to srCT using the DVFs computed by these three DIRs and compared to those mapped using the reference DVF. Results: The average Dice coefficients for selected ROIs were from 0.85 to 0.96 for Demons, from 0.86 to 0.97 for intensity-based, and from 0.76 to 0.95 for B-Spline. The average Hausdorff distances for selected ROIs were from 2.2 to 5.4 mm for Demons, from 2.3 to 6.8 mm for intensity-based, and from 2.4 to 11.4 mm for B-Spline. The average absolute dose errors for selected ROIs were from 0.2 to 0.6 Gy for Demons, from 0.1 to 0.5 Gy for intensity-based, and from 0.5 to 1.5 Gy for B-Spline. Conclusion: Virtual phantoms were modeled after patients with lung cancer and were clinically relevant for adaptive radiotherapy treatment replanning. Virtual phantoms with known DVFs serve as references and can provide a fair comparison when evaluating different DIRs. Demons and intensity-based DIRs were shown to have smaller geometric and dosimetric uncertainties than B-Spline. Z Shen: None; K Bzdusek: an employee of Philips Healthcare; J Greskovich: None; P Xia

  12. Postoperative Radiotherapy for Prostate Cancer: A Comparison of Four Consensus Guidelines and Dosimetric Evaluation of 3D-CRT Versus Tomotherapy IMRT

    International Nuclear Information System (INIS)

    Malone, Shawn; Croke, Jennifer; Roustan-Delatour, Nicolas; Belanger, Eric; Avruch, Leonard; Malone, Colin; Morash, Christopher; Kayser, Cathleen; Underhill, Kathryn; Li Yan; Malone, Kyle; Nyiri, Balazs; Spaans, Johanna

    2012-01-01

    Purpose: Despite the benefits of adjuvant radiotherapy after radical prostatectomy, approximately one-half of patients relapse. Four consensus guidelines have been published (European Organization for Research and Treatment of Cancer, Faculty of Radiation Oncology Genito-Urinary Group, Princess Margaret Hospital, Radiation Therapy Oncology Group) with the aim of standardizing the clinical target volume (CTV) delineation and improve outcomes. To date, no attempt has been made to compare these guidelines in terms of treatment volumes or organ at risk (OAR) irradiation. The extent to which the guideline-derived plans meet the dosimetric constraints of present trials or of the Quantitative Analysis of Normal Tissue Effects in the Clinic (QUANTEC) trial is also unknown. Our study also explored the dosimetric benefits of intensity-modulated radiotherapy (IMRT). Methods and Materials: A total of 20 patients treated with postoperative RT were included. The three-dimensional conformal radiotherapy (3D-CRT) plans were applied to cover the guideline-generated planning target volumes (66 Gy in 33 fractions). Dose–volume histograms (DVHs) were analyzed for CTV/planning target volume coverage and to evaluate OAR irradiation. The OAR DVHs were compared with the constraints proposed in the QUANTEC and Radiotherapy and Androgen Deprivation In Combination After Local Surgery (RADICALS) trials. 3D-CRT plans were compared with the tomotherapy plans for the Radiation Therapy Oncology Group planning target volume to evaluate the advantages of IMRT. Results: The CTV differed significantly between guidelines (p < 0.001). The European Organization for Research and Treatment of Cancer-CTVs were significantly smaller than the other CTVs (p < 0.001). Differences in prostate bed coverage superiorly accounted for the major volumetric differences between the guidelines. Using 3D-CRT, the DVHs rarely met the QUANTEC or RADICALS rectal constraints, independent of the guideline used. The RADICALS

  13. SU-F-T-37: Dosimetric Evaluation of Planned Versus Decay Corrected Treatment Plans for the Treatment of Tandem-Based Cervical HDR Brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Goyal, M [Texas Oncology, PA, Fort Worth, TX (United States); Shobhit University, Meerut, Uttar Pradesh (India); Manjhi, J; Rai, D [Shobhit University, Meerut, Uttar Pradesh (India); Kehwar, T [Pinnacle Health Cancer Center, Mechanicsburg, PA (United States); Barker, J; Heintz, B; Shide, K [Texas Oncology, PA, Fort Worth, TX (United States)

    2016-06-15

    Purpose: This study evaluated dosimetric parameters for actual treatment plans versus decay corrected treatment plans for cervical HDR brachytherapy. Methods: 125 plans of 25 patients, who received 5 fractions of HDR brachytherapy, were evaluated in this study. Dose was prescribed to point A (ICRU-38) and High risk clinical tumor volume (HR-CTV) and organs at risk (OAR) were, retrospectively, delineated on original CT images by treating physician. First HDR plan was considered as reference plan and decay correction was applied to calculate treatment time for subsequent fractions, and was applied, retrospectively, to determine point A, HR-CTV D90, and rectum and bladder doses. Results: The differences between mean point A reference doses and the point A doses of the plans computed using decay times were found to be 1.05%±0.74% (−2.26% to 3.26%) for second fraction; −0.25%±0.84% (−3.03% to 3.29%) for third fraction; 0.04%±0.70% (−2.68% to 2.56%) for fourth fraction and 0.30%±0.81% (−3.93% to 2.67%) for fifth fraction. Overall mean point A dose difference, for all fractions, was 0.29%±0.38% (within ± 5%). Mean rectum and bladder dose differences were calculated to be −3.46%±0.12% and −2.47%±0.09%, for points, respectively, and −1.72%±0.09% and −0.96%±0.06%, for D2cc, respectively. HR-CTV D90 mean dose difference was found to be −1.67% ± 0.11%. There was no statistically significant difference between the reference planned point A doses and that calculated using decay time to the subsequent fractions (p<0.05). Conclusion: This study reveals that a decay corrected treatment will provide comparable dosimetric results and can be utilized for subsequent fractions of cervical HDR brachytherapy instead of actual treatment planning. This approach will increase efficiency, decrease workload, reduce patient observation time between applicator insertion and treatment delivery. This would be particularly useful for institutions with limited

  14. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Min, Byung Joo; Lee, Jong Tai [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi-group constant library using the newly compiled data files and the code systems. As the results of this project, the latest version of NJOY nuclear data processing system, NJOY91.38 which is capable of processing data in ENDF-6 format, was compiled and installed in Cyber 960-31(OS : NOS/VE) and HP710 workstation. A 50-group constant library for fast reactor was generated with NJOY91.38 using evaluated data from JEF-1 and benchmark test of this library was performed. The newly generated library has been found to do an excellent job of calculating integral quantities for fast critical assemblies and is expected to be positively used to develop fast reactors. (Author).

  15. SU-E-T-319: Dosimetric Evaluation of IMRT with Mix-Energy Beam for Deep Seated Targets

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, S; Manigandan, D; Gandhi, A; Sharma, D; Subramani, V; Chander, S; Julkha, P [Fortis Hospital, Mohali, Punjab (India); Rath, G

    2015-06-15

    Purpose: IMRT is preferred in the range of 6–10MV X-rays. Partially adding high energy (>10MV) treatment fields, may provide advantage of both higher and lower energies. To study IMRT dose distribution obtained from treatment plans with single (6MV) and mixed-energy (6MV and 15MV) for deep seated targets (separation more than 30cm). Methods: Five patients of carcinoma of cervix were studied using eclipse planning system. Two different dynamic IMRT plans were generated for Varian CL2300C/D linear accelerator; one is by using 6MV X-ray with seven equally spaced coplanar beams. In second plan, 2 lateral oblique fields (gantry angle 102°, 255°) beam energy was modified to 15MV by keeping all other parameters and dose volume constraints constant. Dose prescription for the planning target volume (PTV) was (5040cGy/28f). For plan comparison, dose volume histogram (DVH) was used and PTV coverage index (CI=Target volume covered by prescription dose/Target volume), heterogeneity index (D5/D95), mean dose to organ at risk (OAR) and normal tissue integral dose (NTID, liter-Gray) was also noted. Total monitor unit (MU) required to deliver a plan was also noted. Results: Mixed-energy plan showed a better conformity and CI values were 0.942±0.032 and 0.960±0.040 for single-energy and mixed-energy plan, respectively. In addition, HI value of mixed energy beam is comparable to that of single energy and the values were within 1.084±0.034 and 1.082±0.032 for single energy and mixed-energy plan, respectively. Variation in mean dose to bladder, rectum and bowel were within 1.05%, 0.87% and 0.90%. NTID was lesser for mixed-energy beam due to use of two high-energy fields. NTID were 1573.40±214.60 and 1510.20±249.80 litre-Gray for single energy and mixed-energy plan. MU needed to deliver a plan was similar in both plans and MUs were 238±45 and 237±47. Conclusion: Partial use of 15MV treatment fields in IMRT plan for deep seated targets showed dosimetric advantage over 6MV

  16. SU-C-210-06: Quantitative Evaluation of Dosimetric Effects Resulting From Positional Variations of Pancreatic Tumor Volumes

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S; Sehgal, V; Wei, R; Lawrenson, L; Kuo, J; Hanna, N; Ramsinghani, N; Daroui, P; Al-Ghazi, M [University of California, Orange, CA (United States)

    2015-06-15

    Purpose: The aim of this study is to quantify dosimetric effects resulting from variation in pancreatic tumor position assessed by bony anatomy and implanted fiducial markers Methods: Twelve pancreatic cancer patients were retrospectively analyzed for this study. All patients received modulated arc therapy (VMAT) treatment using fiducial-based Image Guided Radiation Therapy (IGRT) to the intact pancreas. Using daily orthogonal kV and/or Cone beam CT images, the shift needed to co-register the daily pre-treatment images to reference CT from fiducial to bone (Fid-Bone) were recorded as Left-Right (LR), Anterior-Posterior (AP) and Superior-Inferior (SI). The original VMAT plan iso-center was shifted based on KV bone matching positions at 5 evenly spaced fractions. Dose coverage of the planning target volumes (PTVs) (V100%), mean dose to liver, kidney and stomach/duodenum were assessed in the modified plans. Results: A total of 306 fractions were analyzed. The absolute fiducial-bone positional shifts were greatest in the SI direction, (AP = 2.7 ± 3.0, LR = 2.8 ± 2.8, and SI 6.3 ± 7.9 mm, mean ± SD). The V100% was significantly reduced by 13.5%, (Fid-Bone = 95.3 ± 2.0 vs. 82.3 ± 11.8%, p=0.02). This varied widely among patients (Fid-Bone V100% Range = 2–60%), where 33% of patients had a reduction in V100% of more than 10%. The impact on OARs was greatest to the liver (Fid-Bone= 14.6 vs. 16.1 Gy, 10%), and stomach, (Fid-Bone = 23.9 vx. 25.5 Gy, 7%), however was not statistically significant (p=0.10 both). Conclusion: Compared to matching by fiducial markers, matching by bony anatomy would have substantially reduced the PTV coverage by 13.5%. This reinforces the importance of online position verification based on fiducial markers. Hence, implantation of fiducial markers is strongly recommended for pancreatic cancer patients undergoing intensity modulated radiation therapy treatments.

  17. Evaluation of the integrity of SEP reactor vessels

    International Nuclear Information System (INIS)

    Hoge, K.G.

    1979-12-01

    A documented review is presented of the integrity of the 11 reactor pressure vessels covered in the Systematic Evaluation Program. This review deals primarily with the design specifications and quality assurance programs used in the vessel construction and the status of material surveillance programs, pressure-temperature operating limits, and inservice inspection programs of the applicable plants. Several generic items such as PWR overpressurization protection and BWR nozzle and safe-end cracking also are evaluated. The 11 vessels evaluated include Dresden Units 1 and 2, Big Rock Point, Haddam Neck, Yankee Rowe, Oyster Creek, San Onofre 1, LaCrosse, Ginna, Millstone 1, and Palisades

  18. Safety re-evaluation of the HOR reactor

    International Nuclear Information System (INIS)

    Verkooijen, A.H.M.; Vries, J.W. de

    2001-01-01

    State. Requirement C16 in the new licence asks for a periodical integral safety re-evaluation of the HOR reactor every 10 years and starting after 2 years

  19. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1986-09-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data. 8 refs

  20. Joint evaluated file qualification for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; van der Gucht, C.; Vanuxeem, J.

    1986-01-01

    The neutron and nuclear data which are needed by reactor physicists to perform core calculations are brought together in the evaluated files. The files are processes to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data, which is sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. The authors show, in this paper, how the use of these integral experiments and the application of a tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Joint evaluated file. For this purpose, 56 buckling measurements and 42 isotopic analysis of irradiated fuel were used. Small modifications of the initial data are proposed. The final values are compared with recent recommended values or microscopic data

  1. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

    1991-12-01

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  2. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  3. Evaluation of the breed/burn fast reactor concept

    International Nuclear Information System (INIS)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH 16 ) as the moderator

  4. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  5. Evaluation of the Community's nuclear reactor safety research programme

    International Nuclear Information System (INIS)

    Brandstetter, A.; Goedkoop, J.A.; Jaumotte, A.; Malhouitre, G.; Tomkins, B.; Zorzoli, G.B.

    1986-01-01

    This report describes an evaluation of the 1980-85 CEC reactor safety programme prepared, at the invitation of the Commission, by a panel of six independent experts by means of examining the relevant document and by holding hearings with the responsible CEC staff. It contains the recommendations made by the panel on the following topics: the need for the JRC to continue to make its competence in the reactor safety field available to the Community; the importance of continuity in the JRC and shared-cost action programmes; the difficulty of developing reactor safety research programmes which satisfy the needs of users with diverse needs; the monitoring of the utilization of the research results; the maintenance of the JRC computer codes used by the Member States; the spin-off from research results being made available to other industrial sectors; the continued contact between the JRC researchers and the national experts; the coordination of LWR safety research with that of the Member States; and, the JRC work on fast breeders to be planned with regard to the R and D programmes of the Fast Reactor European Consortium

  6. Neutron streaming evaluation for the DREAM fusion power reactor

    International Nuclear Information System (INIS)

    Seki, Yasushi; Nishio, Satoshi; Ueda, Shuzo; Kurihara, Ryoichi

    2000-01-01

    Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

  7. SU-E-J-119: Head-And-Neck Digital Phantoms for Geometric and Dosimetric Uncertainty Evaluation of CT-CBCT Deformable Image Registration

    International Nuclear Information System (INIS)

    Shen, Z; Koyfman, S; Xia, P; Bzdusek, K

    2015-01-01

    Purpose: To evaluate geometric and dosimetric uncertainties of CT-CBCT deformable image registration (DIR) algorithms using digital phantoms generated from real patients. Methods: We selected ten H&N cancer patients with adaptive IMRT. For each patient, a planning CT (CT1), a replanning CT (CT2), and a pretreatment CBCT (CBCT1) were used as the basis for digital phantom creation. Manually adjusted meshes were created for selected ROIs (e.g. PTVs, brainstem, spinal cord, mandible, and parotids) on CT1 and CT2. The mesh vertices were input into a thin-plate spline algorithm to generate a reference displacement vector field (DVF). The reference DVF was applied to CBCT1 to create a simulated mid-treatment CBCT (CBCT2). The CT-CBCT digital phantom consisted of CT1 and CBCT2, which were linked by the reference DVF. Three DIR algorithms (Demons, B-Spline, and intensity-based) were applied to these ten digital phantoms. The images, ROIs, and volumetric doses were mapped from CT1 to CBCT2 using the DVFs computed by these three DIRs and compared to those mapped using the reference DVF. Results: The average Dice coefficients for selected ROIs were from 0.83 to 0.94 for Demons, from 0.82 to 0.95 for B-Spline, and from 0.67 to 0.89 for intensity-based DIR. The average Hausdorff distances for selected ROIs were from 2.4 to 6.2 mm for Demons, from 1.8 to 5.9 mm for B-Spline, and from 2.8 to 11.2 mm for intensity-based DIR. The average absolute dose errors for selected ROIs were from 0.7 to 2.1 Gy for Demons, from 0.7 to 2.9 Gy for B- Spline, and from 1.3 to 4.5 Gy for intensity-based DIR. Conclusion: Using clinically realistic CT-CBCT digital phantoms, Demons and B-Spline were shown to have similar geometric and dosimetric uncertainties while intensity-based DIR had the worst uncertainties. CT-CBCT DIR has the potential to provide accurate CBCT-based dose verification for H&N adaptive radiotherapy. Z Shen: None; K Bzdusek: an employee of Philips Healthcare; S Koyfman: None; P Xia

  8. Dosimetric evaluation of a commercial proton spot scanning Monte-Carlo dose algorithm: comparisons against measurements and simulations.

    Science.gov (United States)

    Saini, Jatinder; Maes, Dominic; Egan, Alexander; Bowen, Stephen R; St James, Sara; Janson, Martin; Wong, Tony; Bloch, Charles

    2017-09-12

     mm. In an anthropomorphic phantom, the gamma index (dose tolerance  =  3%, distance-to-agreement  =  3 mm) was greater than 90% for six out of seven planes using the RS-MC, and three out seven for the RS-PBA. The RS-MC algorithm demonstrated improved dosimetric accuracy over the RS-PBA in the presence of homogenous, heterogeneous and anthropomorphic phantoms. The computation performance of the RS-MC was similar to the RS-PBA algorithm. For complex disease sites like breast, head and neck, and lung cancer, the RS-MC algorithm will provide significantly more accurate treatment planning.

  9. Dosimetric evaluation of a commercial proton spot scanning Monte-Carlo dose algorithm: comparisons against measurements and simulations

    Science.gov (United States)

    Saini, Jatinder; Maes, Dominic; Egan, Alexander; Bowen, Stephen R.; St. James, Sara; Janson, Martin; Wong, Tony; Bloch, Charles

    2017-10-01

     mm. In an anthropomorphic phantom, the gamma index (dose tolerance  =  3%, distance-to-agreement  =  3 mm) was greater than 90% for six out of seven planes using the RS-MC, and three out seven for the RS-PBA. The RS-MC algorithm demonstrated improved dosimetric accuracy over the RS-PBA in the presence of homogenous, heterogeneous and anthropomorphic phantoms. The computation performance of the RS-MC was similar to the RS-PBA algorithm. For complex disease sites like breast, head and neck, and lung cancer, the RS-MC algorithm will provide significantly more accurate treatment planning.

  10. Dosimetric verification and evaluation of segmental multileaf collimator (SMLC)-IMRT for quality assurance. The second report. Absolute dose

    International Nuclear Information System (INIS)

    Tateoka, Kunihiko; Hareyama, Masato; Oouchi, Atsushi; Nakata, Kensei; Nagase, Daiki; Saikawa, Tsunehiko; Shimizume, Kazunari; Sugimoto, Harumi; Waka, Masaaki

    2003-01-01

    Intensity-modulated radiation therapy (IMRT) was developed to irradiate the target are more conformally, sparing organs at risk (OARs). Since the beams are sequentially delivered by many, small, irregular, and off-center fields in IMRT, dosimetric quality assurance (QA) is an extremely important issue. QA is performed by verifying both the dose distribution and doses at arbitrary points. In this work, we describe the verification of doses at arbitrary points in our hospital for Segmental multileaf collimator (SMLC)-IMRT. In general, verification of the absolute doses for IMRT is performed by comparison between the calculated doses using Radiation Treatment Planning Systems (RTP) and the measured doses using an ionization chamber with a small volume at arbitrary points in relatively flat regions of the dose gradients. However, no clear definitions of the dose gradients and the flat regions have yet been reported. We carried out verification by comparison of the measured doses with the average dose and the central point dose in a virtual Farmer type ionization chamber (V-F) and a virtual PinPoint ionization chamber (V-P) equal to the Farmer-type ionization chamber volume and PinPoint ionization chamber volumes using the RTP. Furthermore, we defined the dose gradients as the deviation of the maximum dose from the minimum dose in the virtual ionization chamber volume. In IMRT, the dose gradients may be as high as 80% or more in the virtual ionization chamber volume. Therefore, it is thought that the effective center of the ionization chamber varies by segment for IMRT fields (i.e., the variation of the ionization chamber replacement effect). Additionally, in regions with a higher dose gradient, uncertainty in the measured doses is influenced by the variations in the ionization chamber replacement effect and the ionization chamber positioning error. We more objectively examined the verification method for the absolute dose in IMRT using the virtual ionization chamber

  11. Dosimetric Verification and Evaluation of the 3-D Conformal Parotid Gland-Sparing Irradiation Technique for Bilateral Neck Treatment at University Hospital Centre Zagreb

    International Nuclear Information System (INIS)

    Kovacevic, N; Hrsak, H.; Bibic, J.

    2011-01-01

    3-D Conformal Parotid Gland-Sparing Irradiation Technique for Bilateral Neck (ConPas) is an alternative to Intensity-modulated radiotherapy (IMRT), and is in routine use at University Hospital Centre Rebro (KBC-Rebro), Zagreb. This technique includes highly asymmetric wedged conformal multi-leaf fields and demands very precise application. The aim of this paper is to present the dosimetric verification method of ConPas (and evaluation of ConPas applicability) as performed at KBC, taking into account the precision of the Treatment Planning System (TPS), possibilities of linear accelerator and patient set-up error. Results for two patients are shown in some details.ConPas is a rather sophisticated method and demands high precision in the whole radiotherapy process. Verification of ConPas using IMRT Verification Matrix Phantom shows good agreement between measured and predicted doses inside and outside PTV regions of the head and neck. Furthermore, a careful track of the positioning during the treatment shows that the overall set-up error is very small (practically negligible). When possible, one parotid gland may be partially spared, and therefore its function preserved at least to some extent. (author)

  12. Design and evaluation of materials for space reactors

    International Nuclear Information System (INIS)

    Tavassoli, A.A.; Vrillon, B.; Robert, G.

    1990-01-01

    The French programme envisages a 20 kWe reactor, project ERATO, with three technological options. The first option is a sodium cooled reactor, derived from the fast breeder reactor technology, (upper core outlet temperature of 700 0 C). The second option is based on the High Temperature Gas-cooled Reactor technology (outlet temperature range 700 0 C-900 0 C). The third option is the reference solution, lithium cooled and UN fuelled fast spectrum reactor, (outlet temperature as high as 1200 0 C). The choice is essentially dominated by material considerations, and more specifically by the problems related to the compatibility with the cooling medium and to the high temperature creep resistance. For the first system limited work will be needed as the technology used is well experimented and there is a wealth of information on the austenitic stainless steel Type 316L-SPH. For the second system, most of the work has been concentrated on characterization of existing commercial alloys. This has included the preselection and the testing of a number of superalloys irradiated or not. The results obtained from high temperature tensile and creep tests have allowed selection of Haynes 230 as the primary candidate material and have also permitted calculation of allowable design stresses for this alloy. For the very high temperature system the French R and D programme has focused on Mo-Re alloys. The results obtained to this date from microstructural examinations and mechanical tests performed on different alloy compositions have allowed selection of Mo-25%Re for future optimization work. They have also shown the need for evaluation of creep properties at low stresses where microstructural instabilities are likely to occur as a result of long exposure to high temperature

  13. Statistical evaluation of design-error related nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ott, K.O.; Marchaterre, J.F.

    1981-01-01

    In this paper, general methodology for the statistical evaluation of design-error related accidents is proposed that can be applied to a variety of systems that evolves during the development of large-scale technologies. The evaluation aims at an estimate of the combined ''residual'' frequency of yet unknown types of accidents ''lurking'' in a certain technological system. A special categorization in incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of U.S. nuclear power reactor technology, considering serious accidents (category 2 events) that involved, in the accident progression, a particular design inadequacy. 9 refs

  14. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  15. Power reactor embrittlement data base (PR-EDB): Uses in evaluating radiation embrittlement of reactor vessels

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1992-01-01

    Investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current Codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed, computerized data base. Also, such a data is essential for the evaluation of embrittlement prediction models by researchers. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The current compilation contains data from 92 reactors and consists of 175 data points for weld materials (79 different welds) and 395 data points for base materials (110 different base materials). The different types of data that are implemented or planned for this data base are discussed. ''User-friendly'' utility programs have been written to investigate a list of problems using this data base. The utility programs are also used to add and upgrade data, retrieve and select specific data, manipulate data, display data to the screen or printer, and to fit and plot Charpy impact data. The results of several studies investigated are presented in this paper

  16. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

    Maerker, R.E.; Broadhead, B.L.; Wagschal, J.J.

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  17. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current US innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery

  18. Cladding failure probability modeling for risk evaluations of fast reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Kramer, J.M.

    1987-01-01

    This paper develops the methodology to incorporate cladding failure data and associated modeling into risk evaluations of liquid metal-cooled fast reactors (LMRs). Current U.S. innovative designs for metal-fueled pool-type LMRs take advantage of inherent reactivity feedback mechanisms to limit reactor temperature increases in response to classic anticipated-transient-without-scram (ATWS) initiators. Final shutdown without reliance on engineered safety features can then be accomplished if sufficient time is available for operator intervention to terminate fission power production and/or provide auxiliary cooling prior to significant core disruption. Coherent cladding failure under the sustained elevated temperatures of ATWS events serves as one indicator of core disruption. In this paper we combine uncertainties in cladding failure data with uncertainties in calculations of ATWS cladding temperature conditions to calculate probabilities of cladding failure as a function of the time for accident recovery. (orig.)

  19. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  20. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-12-01

    In nuclear or shielding design analysis for reactors including nuclear facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multigroup constant library using the newly compiled data files and the code systems. As the results of this project, JEF-2.2 which is latest version of Joint Evaluated File developed at OECD/NEA was compiled and COMPLOT and EVALPLOT utility codes were installed in personal computer, which are able to draw ENDF/B-formatted nuclear data for comparison and check. Computer system (NJOY/ACER) for generating continuous energy Monte Carlo code MCNP library was established and the system was validated by analyzing a number of experimental data. (Author).

  1. On the use of advanced numerical models for the evaluation of dosimetric parameters and the verification of exposure limits at workplaces.

    Science.gov (United States)

    Catarinucci, L; Tarricone, L

    2009-12-01

    With the next transposition of the 2004/40/EC Directive, employers will become responsible for the electromagnetic field level at the workplace. To make this task easier, the scientific community is compiling practical guidelines to be followed. This work aims at enriching such guidelines, especially for the dosimetric issues. More specifically, some critical aspects related to the application of numerical dosimetric techniques for the verification of the safety limit compliance have been highlighted. In particular, three different aspects have been considered: the dosimetric parameter dependence on the shape and the inner characterisation of the exposed subject as well as on the numerical algorithm used, and the correlation between reference limits and basic restriction. Results and discussions demonstrate how, even by using sophisticated numerical techniques, in some cases a complex interpretation of the result is mandatory.

  2. On the use of advanced numerical models for the evaluation of dosimetric parameters and the verification of exposure limits at workplaces

    International Nuclear Information System (INIS)

    Catarinucci, L.; Tarricone, L.

    2009-01-01

    With the next transposition of the 2004/40/EC Directive, employers will become responsible for the electromagnetic field level at the workplace. To make this task easier, the scientific community is compiling practical guidelines to be followed. This work aims at enriching such guidelines, especially for the dosimetric issues. More specifically, some critical aspects related to the application of numerical dosimetric techniques for the verification of the safety limit compliance have been highlighted. In particular, three different aspects have been considered: the dosimetric parameter dependence on the shape and the inner characterisation of the exposed subject as well as on the numerical algorithm used, and the correlation between reference limits and basic restriction. Results and discussions demonstrate how, even by using sophisticated numerical techniques, in some cases a complex interpretation of the result is mandatory. (authors)

  3. Dosimetric characteristics of biological effect of sulfur-35

    International Nuclear Information System (INIS)

    Borisova, V.V.

    1990-01-01

    Experimental materials related to evaluation of dosimetric characteristics of sulfur-35 are presented. Hemogenic organs are subjected to greatest influence especially in the first hours after radionuclide entry into the organism. Comparison is made of absorbed doses in blood with observed blastomogen effect of hemogenic organs. It is noted, that quantitative evaluation of relative biological efficiency of low energy beta-emitters should be performed with account of dosimetric peculiarities of the nuclides mentioned above. 10 refs.; 3 tabs

  4. Evaluation for rigidity of box construction of nuclear reactor building

    International Nuclear Information System (INIS)

    Yamakawa, Tetsuo

    1979-01-01

    A huge box-shaped structure (hereafter, called box construction) of reinforced concrete is presently utilized as the reactor building structure in nuclear power plants. Evaluation of the rigidity of the huge box construction is required for making a vibration analysis model of nuclear reactor buildings. It is necessary to handle the box construction as the plates to which the force in plane is applied. This paper describes that the bending theory in elementary beam theory is equivalent to a peculiar, orthogonally anisotropic plate, the shearing rigidity and film rigidity in y direction of which are put to infinity and the Poisson's ratio is put to zero, viewed from the two-dimensional theory of elasticity. The form factor of 1.2 for shearing deformation in rectangular cross section was calculated from the parabolic distribution of shearing stress intensity, and it is the maximum value. The factor is equal to 1.2 for slender beams, but smaller than 1.2 for short and thick beams, having tendency to converge to 1.0. The non-conformity of boundary conditions regarding the shearing force at the both ends of cantilevers does not affect very seriously the evaluation of shearing rigidity. From the above results, it was found that the application of the theory to the box construction was able to give the rigidity evaluation with sufficient engineering accuracy. The theory can also be applied to the evaluation of tube type ultrahigh buildings. (Wakatsuki, Y.)

  5. Process Inherent Ultimate Safety (PIUS) reactor evaluation study: Final report

    International Nuclear Information System (INIS)

    1987-02-01

    This report presents the results of an independent study by United Engineers and Constructors (UNITED) of the SECURE-P Process Inherent Ultimate Safety (PIUS) Reactor Concept which is presently under development by the Swedish light water reactor vendor ASEA-ATOM of Vasteras, Sweden. This study was performed to investigate whether there is any realistic basis for believing that the PIUS reactor could be a viable competitor in the US energy market in the future. Assessments were limited to the technical, economic and licensing aspects of PIUS. Socio-political issues, while certainly important in answering this question, are so broad and elusive that it was considered that addressing them with the limited perspective of one small group from one company would be of questionable value and likely be misleading. Socio-political issues aside, the key issue is economics. For this reason, the specific objectives of this study were to determine if the estimated PIUS plant cost will be competitive in the US market and to identify and evaluate the technical and licensing risks that might make PIUS uneconomical or otherwise unacceptable

  6. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  7. Preliminary safety evaluation for a medical therapy reactor

    International Nuclear Information System (INIS)

    Jones, J.L.; Neuman, W.A.

    1989-01-01

    A conceptual design of a passively safe reactor facility for boron neutron capture therapy has been previously described. The medical therapy reactor (MTR) has a maximum power level of 10 MW(thermal) and utilizes 45 wt% uranium in UZrH, 20 wt% 235 U enriched hydride fuel matrix with 1 wt% erbium, which is a burnable poison and provides prompt negative reactivity feedback. The facility has five beam ports for patient treatment and advanced neutron beam research and is capable of 2,000 to 10,000 treatments per year, assuming single 8h/day, 5 day/week operation. The epithermal treatment flux from the beam ports is large, enabling single-session treatment of brain cancers of <10-min duration, with minimal fast neutron and gamma contaminants. The reactor core is designed with sufficient excess reactivity to yield a core lifetime equal to a facility lifetime of 30 yr. A preliminary safety evaluation was performed using the RELAP5 thermal-hydraulic code. The analysis addressed accidents in several major categories, including a pump coastdown, a loss of secondary heat sink, and a $0.5 step reactivity insertion

  8. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  9. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

  10. Evaluation of a hydrogen sensor for nuclear reactor containment monitoring

    International Nuclear Information System (INIS)

    Hoffheins, B.S.; McKnight, T.E.; Lauf, R.J.; Smith, R.R.; James, R.E.

    1997-01-01

    Measurement of hydrogen concentration in containment atmospheres in nuclear plants is a key safety capability. Current technologies require extensive sampling systems and subsequent maintenance and calibration costs can be very expensive. A new hydrogen sensor has been developed that is small and potentially inexpensive to install and maintain. Its size and low power requirement make it suitable in distributed systems for pinpointing hydrogen buildup. This paper will address the first phase of a testing program conducted to evaluate this sensor for operation in reactor containments

  11. Light water reactors fuel assembly mechanical design and evaluation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    This standard establishes a procedure for performing an evaluation of the mechanical design of fuel assemblies for light water-cooled commercial power reactors. It does not address the various aspects of neutronic or thermalhydraulic performance except where these factors impose loads or constraints on the mechanical design of the fuel assemblies. This standard also includes a set of specific requirements for design, various potential performance problems and criteria aimed specifically at averting them. This standard replaces ANSI/ANS-57.5-1978

  12. SU-F-19A-04: Dosimetric Evaluation of a Novel CT/MR Compatible Fletcher Applicator for Intracavitary Brachytherapy of the Cervix Uteri

    International Nuclear Information System (INIS)

    Gifford, K; Han, T; Mourtada, F; Eifel, P

    2014-01-01

    Purpose: To validate a Monte Carlo model and evaluate the dosimetric capabilities of a novel commercial CT/MR compatible Fletcher applicator for cervical cancer brachytherapy. Methods: MCNPX 2.7.0 was used to model the Fletcher CT/MR shielded applicator (FA) and 192Ir HDR source. Energy deposition was calculated with a track length estimator modified by an energy-dependent heating function. A high density polystyrene phantom was constructed with three film pockets for validation of the MCNPX model. Three planes of data were calculated with the MCNPX model corresponding to the three film planes in phantom. The planes were located 1 cm from the most anterior, posterior, and medial extents of the FA right ovoid. Unshielded distributions were calculated by modeling the shielded cells as air instead of the tungsten alloy. A third order polynomial fit to the OD to dose curve was used to convert OD of the three film planes to dose. Each film and MCNPX plane dose distribution was normalized to a point 2 cm from the center of the film plane and in a region of low dose gradient. MCNPX and film were overlaid and compared with a distance-to-agreement criterion of (±2%/±2mm). Shielded and unshielded distributions were overlaid and a percent shielded plot was created. Results: 85.2%, 97.1%, and 96.6% of the MCNPX points passed the (±2%/±2mm) criterion respectively for the anterior, lateral, and posterior film comparison planes. A majority of the points in the anterior plane that exceeded the DTA criterion were either along edges of where the film was cut or near the terminal edges of the film. The percent shielded matrices indicated that the maximum % shielding was 50%. Conclusion: These data confirm the validity of the FA Monte Carlo model. The FA ovoid can shield up to 50% of the dose in the anteroposterior direction

  13. Sci—Fri PM: Topics — 02: Evaluation of Dosimetric Variations in Partial Breast Seed Implant (PBSI) due to Patient Arm Position (Up vs. Down)

    International Nuclear Information System (INIS)

    Watt, E; Long, K; Husain, S; Meyer, T

    2014-01-01

    The planning for PBSI is done with the patient's ipsilateral arm raised, however, anatomical changes and variations are unavoidable as the patient resumes her daily activities, potentially resulting in significant deviations in implant geometry from the treatment plan. This study aims to quantify the impact of the ipsilateral arm position on the geometry and dosimetry of the implant at eight weeks, evaluated on post-plans using the MIM Symphony™ software (MIM Software, Cleveland, OH). The average dose metrics for the three patients treated at the TBCC thus far using rigid fusion and contour transfer for the arms up position were 76% for the CTV V100, 61% for the PTV V100, and 37% for the PTV V200; and for the arms down position 81% for the CTV V100, 64% for the PTV V100, and 42% for the PTV V200. Qualitative analysis of the post-implant CT for one of the three patients showed poor agreement between the seroma contour transferred from the pre-implant CT and the seroma visible on the post-implant CT. To obtain a clinically accurate plan for that patient, contour modifications were used, yielding improved dose metric averages for the arms-up position for all three patients of 87% for the CTV V100, 68% for the PTV V100, and 39% for the PTV V200. Overall, the data available shows that dosimetric parameters increase with the patient's arm down, both in terms of coverage and in terms of the hot spot, and accrual of more patients may confirm this in a larger population

  14. SU-F-19A-04: Dosimetric Evaluation of a Novel CT/MR Compatible Fletcher Applicator for Intracavitary Brachytherapy of the Cervix Uteri

    Energy Technology Data Exchange (ETDEWEB)

    Gifford, K; Han, T [UT MD Anderson Cancer Center, Houston, TX (United States); Mourtada, F [Christiana Care Hospital, Newark, DE (United States); Eifel, P [The UT MD Anderson Cancer Center, Houston, TX (United States)

    2014-06-15

    Purpose: To validate a Monte Carlo model and evaluate the dosimetric capabilities of a novel commercial CT/MR compatible Fletcher applicator for cervical cancer brachytherapy. Methods: MCNPX 2.7.0 was used to model the Fletcher CT/MR shielded applicator (FA) and 192Ir HDR source. Energy deposition was calculated with a track length estimator modified by an energy-dependent heating function. A high density polystyrene phantom was constructed with three film pockets for validation of the MCNPX model. Three planes of data were calculated with the MCNPX model corresponding to the three film planes in phantom. The planes were located 1 cm from the most anterior, posterior, and medial extents of the FA right ovoid. Unshielded distributions were calculated by modeling the shielded cells as air instead of the tungsten alloy. A third order polynomial fit to the OD to dose curve was used to convert OD of the three film planes to dose. Each film and MCNPX plane dose distribution was normalized to a point 2 cm from the center of the film plane and in a region of low dose gradient. MCNPX and film were overlaid and compared with a distance-to-agreement criterion of (±2%/±2mm). Shielded and unshielded distributions were overlaid and a percent shielded plot was created. Results: 85.2%, 97.1%, and 96.6% of the MCNPX points passed the (±2%/±2mm) criterion respectively for the anterior, lateral, and posterior film comparison planes. A majority of the points in the anterior plane that exceeded the DTA criterion were either along edges of where the film was cut or near the terminal edges of the film. The percent shielded matrices indicated that the maximum % shielding was 50%. Conclusion: These data confirm the validity of the FA Monte Carlo model. The FA ovoid can shield up to 50% of the dose in the anteroposterior direction.

  15. Economic evaluation of nuclear reactor operation utilizing power effect

    International Nuclear Information System (INIS)

    Budinsky, M.; Mydliar, J.

    1988-01-01

    The operation of a reactor at the so-called power effect may substantially increase the burnup of fuel to be removed. The aim of the evaluation of such reactor operation is the optimal determination of the time over which the yield of the higher use of fuel exceeds economic losses resulting from the increased share of constant expenditure of the price of generated kWh of electric power which ensues from such operation. A mathematical model is presented for such evaluation of reactor operation with regard to benefits for the national economy which is the basis of the ESTER 2 computer program. The calculations show that the prices of generated and delivered kWh are minimally 2% less than the prices of generated power without the power effect use. The minimum ranges in the interval of 30 to 50 days. The dependence of the price of generated and delivered kWh from the point of view of the operator of the power plant as well as the component of fuel price of generated kWh will not reach the minimum even after 50 days of operation. From the operating and physical points of view the duration of power effect is not expected to exceed 20 to 30 days which means that from the point of view of the national economy the price of generated and delivered kWh will be 1.6 to 2% less and the fuel component of the price of the generated kWh will be 3 to 4.5% lower. (Z.M.). 5 figs., 3 refs

  16. Dosimetric and geometric evaluation of an open low-field magnetic resonance simulator for radiotherapy treatment planning of brain tumours

    DEFF Research Database (Denmark)

    Kristensen, B.H.; Laursen, F.J.; Logager, V.

    2008-01-01

    Background and purpose: Magnetic resonance (MR) imaging is superior to computed tomography (CT) in radiotherapy of brain tumours. In this study an open low-field MR-simulator is evaluated in order to eliminate the cost of and time spent on additional CT scanning. Materials and methods: Eleven...

  17. A framework for comparative evaluation of dosimetric methods to triage a large population following a radiological event

    International Nuclear Information System (INIS)

    Flood, Ann Barry; Nicolalde, Roberto J.; Demidenko, Eugene; Williams, Benjamin B.; Shapiro, Alla; Wiley, Albert L.; Swartz, Harold M.

    2011-01-01

    Background: To prepare for a possible major radiation disaster involving large numbers of potentially exposed people, it is important to be able to rapidly and accurately triage people for treatment or not, factoring in the likely conditions and available resources. To date, planners have had to create guidelines for triage based on methods for estimating dose that are clinically available and which use evidence extrapolated from unrelated conditions. Current guidelines consequently focus on measuring clinical symptoms (e.g., time-to-vomiting), which may not be subject to the same verification of standard methods and validation processes required for governmental approval processes of new and modified procedures. Biodosimeters under development have not yet been formally approved for this use. Neither set of methods has been tested in settings involving large-scale populations at risk for exposure. Objective: To propose a framework for comparative evaluation of methods for such triage and to evaluate biodosimetric methods that are currently recommended and new methods as they are developed. Methods: We adapt the NIH model of scientific evaluations and sciences needed for effective translational research to apply to biodosimetry for triaging very large populations following a radiation event. We detail criteria for translating basic science about dosimetry into effective multi-stage triage of large populations and illustrate it by analyzing 3 current guidelines and 3 advanced methods for biodosimetry. Conclusions: This framework for evaluating dosimetry in large populations is a useful technique to compare the strengths and weaknesses of different dosimetry methods. It can help policy-makers and planners not only to compare the methods' strengths and weaknesses for their intended use but also to develop an integrated approach to maximize their effectiveness. It also reveals weaknesses in methods that would benefit from further research and evaluation.

  18. A framework for comparative evaluation of dosimetric methods to triage a large population following a radiological event

    Energy Technology Data Exchange (ETDEWEB)

    Flood, Ann Barry, E-mail: Ann.B.Flood@Dartmouth.Edu [Dartmouth Physically Based Biodosimetry Center for Medical Countermeasures Against Radiation (Dart-Dose CMCR), Dartmouth Medical School, Hanover, NH 03768 (United States); Nicolalde, Roberto J., E-mail: Roberto.J.Nicolalde@Dartmouth.Edu [Dartmouth Physically Based Biodosimetry Center for Medical Countermeasures Against Radiation (Dart-Dose CMCR), Dartmouth Medical School, Hanover, NH 03768 (United States); Demidenko, Eugene, E-mail: Eugene.Demidenko@Dartmouth.Edu [Dartmouth Physically Based Biodosimetry Center for Medical Countermeasures Against Radiation (Dart-Dose CMCR), Dartmouth Medical School, Hanover, NH 03768 (United States); Williams, Benjamin B., E-mail: Benjamin.B.Williams@Dartmouth.Edu [Dartmouth Physically Based Biodosimetry Center for Medical Countermeasures Against Radiation (Dart-Dose CMCR), Dartmouth Medical School, Hanover, NH 03768 (United States); Shapiro, Alla, E-mail: Alla.Shapiro@fda.hhs.gov [Food and Drug Administration (FDA), Rockville, MD (United States); Wiley, Albert L., E-mail: Albert.Wiley@orise.orau.gov [Oak Ridge Institute for Science and Education (ORISE), Oak Ridge, TN (United States); Swartz, Harold M., E-mail: Harold.M.Swartz@Dartmouth.Edu [Dartmouth Physically Based Biodosimetry Center for Medical Countermeasures Against Radiation (Dart-Dose CMCR), Dartmouth Medical School, Hanover, NH 03768 (United States)

    2011-09-15

    Background: To prepare for a possible major radiation disaster involving large numbers of potentially exposed people, it is important to be able to rapidly and accurately triage people for treatment or not, factoring in the likely conditions and available resources. To date, planners have had to create guidelines for triage based on methods for estimating dose that are clinically available and which use evidence extrapolated from unrelated conditions. Current guidelines consequently focus on measuring clinical symptoms (e.g., time-to-vomiting), which may not be subject to the same verification of standard methods and validation processes required for governmental approval processes of new and modified procedures. Biodosimeters under development have not yet been formally approved for this use. Neither set of methods has been tested in settings involving large-scale populations at risk for exposure. Objective: To propose a framework for comparative evaluation of methods for such triage and to evaluate biodosimetric methods that are currently recommended and new methods as they are developed. Methods: We adapt the NIH model of scientific evaluations and sciences needed for effective translational research to apply to biodosimetry for triaging very large populations following a radiation event. We detail criteria for translating basic science about dosimetry into effective multi-stage triage of large populations and illustrate it by analyzing 3 current guidelines and 3 advanced methods for biodosimetry. Conclusions: This framework for evaluating dosimetry in large populations is a useful technique to compare the strengths and weaknesses of different dosimetry methods. It can help policy-makers and planners not only to compare the methods' strengths and weaknesses for their intended use but also to develop an integrated approach to maximize their effectiveness. It also reveals weaknesses in methods that would benefit from further research and evaluation.

  19. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  20. Use of VAP3D software in the construction of pathological anthropomorphic phantoms for dosimetric evaluations; Uso do software VAP3D na construcao de fantomas antropomorficos patologicos para avaliacoes dosimetricas

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Lindeval Fernandes de [Universidade Federal de Pernambuco (DEM/UFPE), Recife, PE (Brazil). Dept. de Engenharia Mecanica; Vieira, Jose Wilson [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco, Recife, PE (Brazil); Lima, Fernando R.A., E-mail: falima@cnen.gov.b [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

    2011-10-26

    This paper performs a new type of dosimetric evaluation, where it was used a phantom of pathological voxels (representative phantom of sick person). The software VAP3D (Visualization and Analysis of Phantoms 3D) were used for, from a healthy phantom (phantom representative of healthy person), to introduce three dimensional regions to simulate tumors. It was used the Monte Carlo ESGnrc code to simulate the X ray photon transport, his interaction with matter and evaluation of absorbed dose in organs and tissues from thorax region of the healthy phantom and his pathological version. This is a computer model of typical exposure for programming the treatments in radiodiagnostic

  1. Clinical and dosimetric evaluation of RapidArc versus standard sliding window IMRT in the treatment of head and neck cancer

    Energy Technology Data Exchange (ETDEWEB)

    Smet, Stephanie; Lambrecht, Maarten; Vanstraelen, Bianca; Nuyts, Sandra [University Hospitals Leuven, Department of Radiation Oncology, Leuven (Belgium)

    2014-08-29

    Several planning studies have already proven the substantial dosimetric advantages of RapidArc (RA) over standard intensity-modulated radiotherapy. We retrospectively compared RapidArc and standard sliding window IMRT (swIMRT) in locally advanced head and neck cancer, looking both at dosimetrics as well as toxicity and outcome. CT datasets of 78 patients treated with swIMRT and 79 patients treated with RA were included. To compare the resulting dose distributions, the dose-volume parameters were evaluated for the planning target volumes (PTVs), clinical target volumes (CTVs), and organs at risk (OARs), and the number of MU were calculated. Acute toxicity was assessed by the Common Toxicity Criteria version 3.0. PTV coverage with the 95 % isodose was slightly better for RA. Dose distribution has proven to be significantly more homogenous with RA and led to a reduction of 62 % in MU with better OAR sparing. As for toxicity, more grade 3 mucositis and dysphagia was observed for swIMRT, though we observed more grade 3 dermatitis for RA. In our retrospective analysis, RA had better target coverage and better sparing of the OAR. Overall, the grade of acute toxicity was lower for RA than for swIMRT for the same types of tumor locations, except for the grade of dermatitis. (orig.) [German] Mehrere Studien haben die dosimetrische Ueberlegenheit der RapidArc (RA) gegenueber der intensitaetsmodulierten Standard-Radiotherapie (IMRT) bereits gezeigt. In unserer Studie verglichen wir retrospektiv die RapidArc und die dynamische (''standard sliding window'') IMRT (swIMRT) bei lokal fortgeschrittenen Kopf-Hals-Karzinomen sowohl hinsichtlich dosimetrischer Daten als auchEffektivitaet und Toxizitaet. Die CT-Datenanalysen von 78 Patienten, die mit swIMRT behandelt wurden, und von 79 Patienten, welche RA erhalten hatten, wurden in die Studie aufgenommen. Um die darauf resultierenden applizierten Dosen vergleichen zu koennen, wurden die Dosis-Volumen-Parameter fuer

  2. Evaluation of denatured thorium fuel cycles in pressurized water reactors

    International Nuclear Information System (INIS)

    Matzie, R.A.; Rec, J.R.; Terney, A.N.

    1977-01-01

    A developing national energy policy that is based in part on a substantial expansion of the LWR-based electrical generating capacity with deferment of the LMFBR has prompted a re-evaluation of our nuclear fuel resources and their utilization. The ancillary policy of minimizing nuclear weapons proliferation through diversion of bred fissile material has left in doubt the viability of fuel recycling as a means of extending these fuel resources. A substantial, government-sponsored effort is in progress to examine alternate fuel cycles and advanced reactor concepts which can lead to improved resource utilization while minimizing proliferation potential. This paper evaluates several improved fuel cycles for use in current design PWRs and develops selected scenarios for their use within the framework of the safeguarded Nuclear Energy Center (NEC) concept

  3. Dosimetric evaluation in panoramic and tele-radiography procedures; Avaliacao dosimetrica em procedimentos de radiografia panoramica e teleradiografias

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Georgge Gomes

    2004-07-01

    The present work had as an objective to evaluate the skin surface entrance dose in panoramic and tele radiography procedures in three clinics in Recife - Pernambuco - Brazil, and to contribute with data for the determination of reference levels for super cited extra oral procedures, for this purpose, operational conditions in 3 clinics were evaluated in Recife, aiming to evaluate the existence and integrity of the radioprotection equipment, manner and conditions of image processing; and radiographic equipment parameters such as the dimension of the irradiation filed, the total filtration, the exposure time and the potential applied to the X ray tube. For an estimation of the skin entrance dose of the patient, the phantom Randon Alderson and thermoluminescence dosemeters were used. From these values and the conversion factors determined by the Monte Carlo technique, with the phantom MAX it was possible to estimate the dose absorbed in the organ due to the tele radiography procedures. Regarding panoramic radiography the study showed that the more elevated doses occurred in the parotid gland region which is near rotational venters. In the case of tele radiography the highest dose value occurred in the regions corresponding to the temporal lobe of the brain, followed by linfonodes, ears and parotid glands. The doses absorbed in the eyes and the thyroid gland were, 0.037 mGy and 0.002 mGy in Clinic A and 0.062 mGy and 0.003 mGy in Clinic C, respectively. Regarding equipment test, inadequacy was found in the beam collimation in Clinic A and in the reproducibility of the X ray exposure in Clinic C. The total filtration in both clinics was inadequate.(author)

  4. Dosimetric evaluation of multi-sided irradiation on HDPE pipes under 2 MeV electron beam

    Energy Technology Data Exchange (ETDEWEB)

    Benny, P.G., E-mail: bennypg@yahoo.com; Khader, S.A.; Sarma, K.S.S.

    2014-03-01

    The use of electron beam technology has enabled the production of heat resistant pipe for hot water circulation. One of the difficulties in the irradiation of pipe products is the uneven penetration of electrons. Quality of the radiation process depends on radiation dose and homogeneity of the dose distribution, which becomes a major concern when treatments of circular objects like pipes are performed. One method to achieve uniformity in the absorbed dose in the product is to use multi-sided irradiation. The paper discusses the importance of dosimetry mapping in industrial electron beam radiation processing and outlines the challenges in delivering a uniform dose to cylindrical objects. In this study, HDPE pipe of 5 mm thickness of homogeneous material (40 mm outer diameter and 30 mm inner diameter) has been chosen for multi-sided irradiation under 2 MeV scanned electron beam from the ILU-6 accelerator. - Highlights: • The paper outlines the challenges in delivering uniform dose to cylindrical objects at 2 MeV industrial electron beam facility. • HDPE pipe of 40 mm outer diameter and 30 mm inner diameter has been chosen for the study. • The circumferential dose distribution inside and outside of the pipes were evaluated by using calibrated CTA dosimeter strips. • Using stack of dosimeter strips, changes in circumferential dose distribution in the annular region of the pipe was evaluated. • Optimization of multi-sided irradiation on the HDPE pipes for better dose homogeneity is reported in the paper.

  5. Dosimetric evaluation of multi-sided irradiation on HDPE pipes under 2 MeV electron beam

    International Nuclear Information System (INIS)

    Benny, P.G.; Khader, S.A.; Sarma, K.S.S.

    2014-01-01

    The use of electron beam technology has enabled the production of heat resistant pipe for hot water circulation. One of the difficulties in the irradiation of pipe products is the uneven penetration of electrons. Quality of the radiation process depends on radiation dose and homogeneity of the dose distribution, which becomes a major concern when treatments of circular objects like pipes are performed. One method to achieve uniformity in the absorbed dose in the product is to use multi-sided irradiation. The paper discusses the importance of dosimetry mapping in industrial electron beam radiation processing and outlines the challenges in delivering a uniform dose to cylindrical objects. In this study, HDPE pipe of 5 mm thickness of homogeneous material (40 mm outer diameter and 30 mm inner diameter) has been chosen for multi-sided irradiation under 2 MeV scanned electron beam from the ILU-6 accelerator. - Highlights: • The paper outlines the challenges in delivering uniform dose to cylindrical objects at 2 MeV industrial electron beam facility. • HDPE pipe of 40 mm outer diameter and 30 mm inner diameter has been chosen for the study. • The circumferential dose distribution inside and outside of the pipes were evaluated by using calibrated CTA dosimeter strips. • Using stack of dosimeter strips, changes in circumferential dose distribution in the annular region of the pipe was evaluated. • Optimization of multi-sided irradiation on the HDPE pipes for better dose homogeneity is reported in the paper

  6. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  7. Dosimetric evaluation of a moving tumor target in intensity-modulated radiation therapy (IMRT) for lung cancer patients

    Science.gov (United States)

    Kim, Sung Kyu; Kang, Min Kyu; Yea, Ji Woon; Oh, Se An

    2013-07-01

    Immobilization plays an important role in intensity-modulated radiation therapy (IMRT). The application of IMRT in lung cancer patients is very difficult due to the movement of the tumor target. Patient setup in radiation treatment demands high accuracy because IMRT employs a treatment size of a 1mm pixel unit. Hence, quality assurance of the dose delivered to patients must be at its highest. The radiation dose was evaluated for breathing rates of 9, 14, and 18 breaths per minute (bpm) for tumor targets moving up and down by 1.0 cm and 1.5 cm. The dose of the moving planned target volume (PTV) was measured by using a thermo-luminescent dosimeter (TLD) and Gafchromic™ EBT film. The measurement points were 1.0 cm away from the top, the bottom and the left and the right sides of the PTV center. The evaluated dose differences ranged from 94.2 to 103.8%, from 94.4 to 105.4%, and from 90.7 to 108.5% for 9, 14 and 18 bpm, respectively, for a tumor movement of 1.0 cm. The mean values of the doses were 101.4, 99.9, and 99.5% for 9, 14 and 18 bpm, respectively, for a tumor movement of 1.0 cm. Meanwhile, the evaluated dose differences ranged from 93.6 to 105.8%, from 95.9 to 111.5%, and from 96.2 to 111.7% for 9, 14 and 18 bpm, respectively, for a tumor movement of 1.5 cm. The mean values of the doses were 102.3, 103.4, and 103.1% for 9, 14 and 18 bpm, respectively, for a tumor movement of 1.5 cm. Therefore, we suggest that IMRT can be used in the treatment of lung cancer patients with vertical target movements within the range of 1.0 to 1.5 cm.

  8. Evaluation of the Gas Turbine Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  9. Evaluation of the Gas Turbine Modular Helium Reactor

    International Nuclear Information System (INIS)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs

  10. SU-E-I-21: Dosimetric Characterization and Image Quality Evaluation of the AIRO Mobile CT Scanner

    Energy Technology Data Exchange (ETDEWEB)

    Weir, V; Zhang, J; Bruner, A [University of Kentucky, Lexington, KY (United States)

    2015-06-15

    Purpose: The AIRO Mobile CT system was recently introduced which overcomes the limitations from existing CT, CT fluoroscopy, and intraoperative O-arm. With an integrated table and a large diameter bore, the system is suitable for cranial, spine and trauma procedures, making it a highly versatile intraoperative imaging system. This study is to investigate radiation dose and image quality of the AIRO and compared with those from a routine CT scanner. Methods: Radiation dose was measured using a conventional 100mm pencil ionization chamber and CT polymethylmetacrylate (PMMA) body and head phantoms. Image quality was evaluated with a CATPHAN 500 phantom. Spatial resolution, low contrast resolution (CNR), Modulation Transfer Function (MTF), and Normalized Noise Power Spectrum (NNPS) were analyzed. Results: Under identical technique conditions, radiation dose (mGy/mAs) from the AIRO mobile CT system (AIRO) is higher than that from a 64 slice CT scanner. MTFs show that both Soft and Standard filters of the AIRO system lost resolution quickly compared to the Sensation 64 slice CT. With the Standard kernel, the spatial resolutions of the AIRO system are 3lp/cm and 4lp/cm for the body and head FOVs, respectively. NNPSs show low frequency noise due to ring-like artifacts. Due to a higher dose in terms of mGy/mAs at both head and body FOV, CNR of the AIRO system is higher than that of the Siemens scanner. However detectability of the low contrast objects is poorer in the AIRO due to the presence of ring artifacts in the location of the targets. Conclusion: For image guided surgery applications, the AIRO has some advantages over a routine CT scanner due to its versatility, large bore size, and acceptable image quality. Our evaluation of the physical performance helps its future improvements.

  11. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  12. TH-AB-BRA-04: Dosimetric Evaluation of MR-Guided HDR Brachytherapy Planning for Cervical Cancer

    Energy Technology Data Exchange (ETDEWEB)

    Kamio, Y; Barkati, M; Beliveau-Nadeau, D [CHUM Notre Dame Hospital, Montreal, QC, CA (Canada)

    2016-06-15

    Purpose: To perform a retrospective study on 16 patients that had both CT and T2-weighted MR scans done at first fraction using the Utrecht CT/MR applicator (Elekta Brachytherapy) in order to evaluate uncertainties associated with an MR-only planning workflow. Methods: MR-workflow uncertainties were classified in three categories: reconstruction, registration and contouring. A systematic comparison of the CT and MR contouring, manual reconstruction and optimization process was performed to evaluate the impact of these uncertainties on the recommended GEC ESTRO DVH parameters: D90% and V100% for HR-CTV as well as D2cc for bladder, rectum, sigmoid colon and small bowel. This comparison was done using the following four steps: 1. Catheter reconstruction done on MR images with original CT-plan contours and dwell times. 2. OAR contours adjusted on MR images with original CT-plan reconstruction and dwell times. 3. Both reconstruction and contours done on MR images with original CT-plan dwell times. 4. Entire MR-based workflow optimized dwell times reimported to the original CT-plan. Results: The MR-based reconstruction process showed average D2cc deviations of 4.5 ± 3.0%, 1.5 ± 2.0%, 2.5 ± 2.0% and 2.0 ± 1.0% for the bladder, rectum, sigmoid colon and small bowels respectively with a maximum of 10%, 6%, 6% and 4%. The HR-CTV’s D90% and V100% average deviations was found to be 4.0 ± 3.0%, and 2.0 ± 2.0% respectively with a maximum of 10% and 6%. Adjusting contours on MR-images was found to have a similar impact. Finally, the optimized MR-based workflow dwell times were found to still give acceptable plans when re-imported to the original CT-plan which validated the entire workflow. Conclusion: This work illustrates a systematic validation method for centers wanting to move towards an MR-only workflow. This work will be expanded to model based reconstruction, PD-weighted images and other types of applicators.

  13. TH-AB-BRA-04: Dosimetric Evaluation of MR-Guided HDR Brachytherapy Planning for Cervical Cancer

    International Nuclear Information System (INIS)

    Kamio, Y; Barkati, M; Beliveau-Nadeau, D

    2016-01-01

    Purpose: To perform a retrospective study on 16 patients that had both CT and T2-weighted MR scans done at first fraction using the Utrecht CT/MR applicator (Elekta Brachytherapy) in order to evaluate uncertainties associated with an MR-only planning workflow. Methods: MR-workflow uncertainties were classified in three categories: reconstruction, registration and contouring. A systematic comparison of the CT and MR contouring, manual reconstruction and optimization process was performed to evaluate the impact of these uncertainties on the recommended GEC ESTRO DVH parameters: D90% and V100% for HR-CTV as well as D2cc for bladder, rectum, sigmoid colon and small bowel. This comparison was done using the following four steps: 1. Catheter reconstruction done on MR images with original CT-plan contours and dwell times. 2. OAR contours adjusted on MR images with original CT-plan reconstruction and dwell times. 3. Both reconstruction and contours done on MR images with original CT-plan dwell times. 4. Entire MR-based workflow optimized dwell times reimported to the original CT-plan. Results: The MR-based reconstruction process showed average D2cc deviations of 4.5 ± 3.0%, 1.5 ± 2.0%, 2.5 ± 2.0% and 2.0 ± 1.0% for the bladder, rectum, sigmoid colon and small bowels respectively with a maximum of 10%, 6%, 6% and 4%. The HR-CTV’s D90% and V100% average deviations was found to be 4.0 ± 3.0%, and 2.0 ± 2.0% respectively with a maximum of 10% and 6%. Adjusting contours on MR-images was found to have a similar impact. Finally, the optimized MR-based workflow dwell times were found to still give acceptable plans when re-imported to the original CT-plan which validated the entire workflow. Conclusion: This work illustrates a systematic validation method for centers wanting to move towards an MR-only workflow. This work will be expanded to model based reconstruction, PD-weighted images and other types of applicators.

  14. Dosimetric evaluation of a combination of brachytherapy applicators for uterine cervix cancer with involvement of the distal vagina

    International Nuclear Information System (INIS)

    Guimaraes, Roger Guilherme Rodrigues

    2009-01-01

    Objective: To evaluate an alternative brachytherapy technique for uterine cervix cancer involving the distal vagina, without increasing the risk of toxicity. Materials And Methods: Theoretical study comparing three different high-dose rate intracavitary brachytherapy applicators: intrauterine tandem and vaginal cylinder (TC); tandem/ring applicator combined with vaginal cylinder (TR+C); and a virtual applicator combining both the tandem/ring and vaginal cylinder in a single device (TRC). Prescribed doses were 7 Gy at point A, and 5 Gy on the surface or at a 5 mm depth of the vaginal mucosa. Doses delivered to the rectum, bladder and sigmoid colon were kept below the tolerance limits. Volumes covered by the isodoses, respectively, 50% (V50), 100% (V100), 150% (V150) and 200% (V200) were compared. Results: Both the combined TR+C and TRC presented a better dose distribution as compared with the TC applicator. The TR+C dose distribution was similar to the TRC dose, with V150 and V200 being about 50% higher for TR+C (within the cylinder). Conclusion: Combined TR+C in a two-time single application may represent an alternative therapy technique for patients affected by uterine cervix cancer involving the distal vagina. (author)

  15. Dosimetric evaluation of the response of the TLD-100 dosemeters in the IMRT technique by 'Step and Shoot'

    International Nuclear Information System (INIS)

    Vasquez, J.; Benavides, S.O.

    2005-01-01

    We show the results of the dosimetry response of LiF thermoluminescent crystals: TLD-100, where they were radiated in a linear accelerator Siemens Primus Hl using the Intensity Modulated Radiation Therapy (IMRT) by step and shoot technique. Previous to the crystals calibration and response evaluation, the acceptation procedures recommended by the TG-53 protocol for validation of the technique were carried out. The planning system utilized was the Theraplan Plus 3.8, using the algorithm of Pencil Kernel. The register and verification system was Lantis 5.2. The response curve of dose versus charge was obtained from the readings of the TLD in a Harshaw 3500. The crystals were radiated in a Bench- Marck phantom with doses previously determined by using ionization chambers for square radiation fields, in a beam with a 0.68 TPR20,10 corresponding to 6 MV of energy. We compare the response of these through of radiation of segmented fields in a Anthropomorphic phantom and the calculated doses by the planning system. The results obtained in the crystals response show deviations less than 5 % between the measured dose and the calculated dose in the zones of low gradient. It allows its implementation like routine control of quality by IMRT. (Author)

  16. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  17. Evaluation and improvement of dynamic optimality in electrochemical reactors

    International Nuclear Information System (INIS)

    Vijayasekaran, B.; Basha, C. Ahmed

    2005-01-01

    A systematic approach for the dynamic optimization problem statement to improve the dynamic optimality in electrochemical reactors is presented in this paper. The formulation takes an account of the diffusion phenomenon in the electrode/electrolyte interface. To demonstrate the present methodology, the optimal time-varying electrode potential for a coupled chemical-electrochemical reaction scheme, that maximizes the production of the desired product in a batch electrochemical reactor with/without recirculation are determined. The dynamic optimization problem statement, based upon this approach, is a nonlinear differential algebraic system, and its solution provides information about the optimal policy. Optimal control policy at different conditions is evaluated using the best-known Pontryagin's maximum principle. The two-point boundary value problem resulting from the application of the maximum principle is then solved using the control vector iteration technique. These optimal time-varying profiles of electrode potential are then compared to the best uniform operation through the relative improvements of the performance index. The application of the proposed approach to two electrochemical systems, described by ordinary differential equations, shows that the existing electrochemical process control strategy could be improved considerably when the proposed method is incorporated

  18. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    Bess, John D.; Gullifor, Jim

    2015-03-01

    The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

  19. Dosimetric evaluation of using in-house BoS Frame Fixation Tool for the Head and Neck Cancer Patient

    International Nuclear Information System (INIS)

    Kim, Kwang Suk; Jo, Kwang Hyun; Choi, Byeon Ki

    2016-01-01

    BoS(Base of Skull) Frame, the fixation tool which is used for the proton of brain cancer increases the lateral penumbra by increasing the airgap (the distance between patient and beam jet), due to the collision of the beam of the posterior oblique direction. Thus, we manufactured the fixation tool per se for improving the limits of BoS frame, and we'd like to evaluate the utility of the manufactured fixation tool throughout this study. We've selected the 3 patients of brain cancer who have received the proton therapy from our hospital, and also selected the 6 beam angles; for this, we've selected the beam angle of the posterior oblique direction. We've measured the planned BoS frame and the distance of Snout for each beam which are planned for the treatment of the patient using the BoS frame. After this, we've proceeded with the set-up that is above the location which was recommended by the manufacturer of the BoS frame, at the same beam angle of the same patient, by using our in-house Bos frame fixation tool. The set-up was above 21 cm toward the superior direction, compared to the situation when the BoS frame was only used with the basic couch. After that, we've stacked the snout to the BoS frame as much as possible, and measured the distance of snout. We've also measured the airgap, based on the gap of that snout distance; and we've proceeded the normalization based on each dose (100% of each dose), after that, we've conducted the comparative analysis of lateral penumbra. Moreover, we've established the treatment plan according to the changed airgap which has been transformed to the Raystation 5.0 proton therapy planning system, and we've conducted the comparative analysis of DVH(Dose Volume Histogram). When comparing the result before using the in-house Bos frame fixation tool which was manufactured for each beam angle with the result after using the fixation tool, we could figure out that airgap than when not used in accordance with the use of the in-house Bos

  20. Dosimetric evaluation of using in-house BoS Frame Fixation Tool for the Head and Neck Cancer Patient

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwang Suk; Jo, Kwang Hyun; Choi, Byeon Ki [Dept. of Radiation Oncology, Samsung Seoul Hospital, Seoul (Korea, Republic of)

    2016-06-15

    BoS(Base of Skull) Frame, the fixation tool which is used for the proton of brain cancer increases the lateral penumbra by increasing the airgap (the distance between patient and beam jet), due to the collision of the beam of the posterior oblique direction. Thus, we manufactured the fixation tool per se for improving the limits of BoS frame, and we'd like to evaluate the utility of the manufactured fixation tool throughout this study. We've selected the 3 patients of brain cancer who have received the proton therapy from our hospital, and also selected the 6 beam angles; for this, we've selected the beam angle of the posterior oblique direction. We've measured the planned BoS frame and the distance of Snout for each beam which are planned for the treatment of the patient using the BoS frame. After this, we've proceeded with the set-up that is above the location which was recommended by the manufacturer of the BoS frame, at the same beam angle of the same patient, by using our in-house Bos frame fixation tool. The set-up was above 21 cm toward the superior direction, compared to the situation when the BoS frame was only used with the basic couch. After that, we've stacked the snout to the BoS frame as much as possible, and measured the distance of snout. We've also measured the airgap, based on the gap of that snout distance; and we've proceeded the normalization based on each dose (100% of each dose), after that, we've conducted the comparative analysis of lateral penumbra. Moreover, we've established the treatment plan according to the changed airgap which has been transformed to the Raystation 5.0 proton therapy planning system, and we've conducted the comparative analysis of DVH(Dose Volume Histogram). When comparing the result before using the in-house Bos frame fixation tool which was manufactured for each beam angle with the result after using the fixation tool, we could figure out that airgap than when

  1. Evaluation of environmental impact of radioactive waste from reactor operation

    International Nuclear Information System (INIS)

    Lombard, J.; Pages, P.

    1989-10-01

    This paper evaluates the environmental impact of radioactive wastes from reactors operation. We estimate a case of a plant of 20 GWe power operating for 30 years which is equivalent to 600 tons of uranium per year. According to the properties, the waste is stored on surface (Aube site). Starting from the year of storage, we have defined the maximum dose equivalent for an individual from the reference group. The calculation depends on water of outlet water in which some initially stored radionuclides have migrated. Under the most pessimistic estimation, maximum annual dose was of the order of magnitude 0.5 μ Sv (0.05 mrem) for the storage 400 years after opening the site, and after 4000 years. Compared to the values obtained for the radioactive waste storage, the value of this impact is five times higher than the respective surface storage, but two time less than values for underground storage [fr

  2. EVALUATING HYDROGEN PRODUCTION IN BIOGAS REFORMING IN A MEMBRANE REACTOR

    Directory of Open Access Journals (Sweden)

    F. S. A. Silva

    2015-03-01

    Full Text Available Abstract Syngas and hydrogen production by methane reforming of a biogas (CH4/CO2 = 2.85 using carbon dioxide was evaluated in a fixed bed reactor with a Pd-Ag membrane in the presence of a nickel catalyst (Ni 3.31% weight/γ-Al2O3 at 773 K, 823 K, and 873 K and 1.01×105 Pa. Operation with hydrogen permeation at 873 K increased the methane conversion to approximately 83% and doubled the hydrogen yield relative to operation without hydrogen permeation. A mathematical model was formulated to predict the evolution of the effluent concentrations. Predictions based on the model showed similar evolutions for yields of hydrogen and carbon monoxide at temperatures below 823 K for operations with and without the hydrogen permeation. The hydrogen yield reached approximately 21% at 823 K and 47% at 873 K under hydrogen permeation conditions.

  3. Dosimetric evaluation in organs of the Tc99m, I123 bio-kinetics to estimate dose in thyroid children 1 and 5 years

    International Nuclear Information System (INIS)

    Vasquez, A. M.; Quispe, R.; Vasquez, D. J.; Rocha, M. D.; Morales, N. R.; Marin, R. K.; Zelada, A. L.

    2012-10-01

    Using the formalism MIRD and the representation of Cristy-Eckerman for the thyroid in children of 1 and 5 years, is demonstrated that the dosimetric contribution of the organs of I 123 (iodure) bio-kinetics is not significant in the dose estimate. The total dose absorbed by the gland is its auto dose. The dosimetric contribution of the organs source of the Tc 99m (pertechnetate) bio-kinetics in the gland is significant in the dose estimate like to be ignored. The reported results for the iodure are not significantly different to the found for the Marinelli scheme (auto-dose) for thyroid represented by a sphere of 1,78 and 3,45 grams. (Author)

  4. Identification and evaluation of accident sequences in nuclear power reactors

    International Nuclear Information System (INIS)

    Amendola, A.; Capobianchi, S.; Mancini, G.; Olivi, L.; Volta, G.; Reina, G.

    1981-01-01

    Probabilistic analysis techniques are being more and more used for the evaluation of accident progression in nuclear power plants, especially after the issue of the Reactor Safety Study (Report WASH-1400). This study and subsequent discussions have indicated the necessity of better investigating some major items, namely: adequate data base for the probabilistic evaluations; completeness of the analysis with respect both to accident initiation and behaviour; adequate treatment of uncertainties on the physical and operational parameters governing the accident behaviour. Furthermore, recent occurrences have stressed the importance of the operational aspects of reactor safety, such as plant-specific identification of possible occurrences, their prompt recognition, on-line prediction of subsequent developments and actions to be taken. The paper reviews the contributions in progress at JRC-Ispra to all these aspects, and specifically reports on the following: (1) The set-up of a European Reliability Data System for the acquisition and organisation of operational data of LWRs in the European Community. (2) The development of more complete and realistic models of systems. This work includes multistate static models of components and systems with a view to automatic fault-tree construction and dynamic models for accident sequence identification. The dynamic modelling approach ESCS (Event Sequence and Consequences Spectrum), shown in detail with an example, represents a step forward with respect to event-tree technique and opens new possibilities in dealing with human factors and on-line diagnosis problems. (3) The development of RSM (Response Surface Methodology) for the analysis of uncertainty propagations in consequence and in probability of accident chains. (author)

  5. SU-G-JeP2-06: Dosimetric and Workflow Evaluation of First Commercial Synthetic CT Software for Clinical Use in Pelvis

    Energy Technology Data Exchange (ETDEWEB)

    Tyagi, N; Zhang, J; Happersett, L; Kadbi, M; Mechalakos, J; Deasy, J; Hunt, M [Memorial Sloan Kettering Cancer Center, New York, NY (United States)

    2016-06-15

    Purpose: evaluate a commercial synthetic CT (syn-CT) software for use in prostate radiotherapy Methods: Twenty prostate patients underwent CT and MR simulation scans in treatment position on a 3T Philips scanner. The MR protocol consisted of a T2w turbo spin-echo for soft tissue contrast, a 2D balanced-fast field echo (b-FFE) for fiducial identification, a dual-echo 3D FFE B0 map for distortion analysis and a 3D mDIXON FFE sequence to generate syn-CT. Two echoes are acquired during mDIXON scan, allowing water, fat, and in-phase images to be derived using the frequency shift of the fat and water protons. Tissues were classified as: air, adipose, water, trabecular/spongy bone and compact/cortical bone and assigned specific bulk HU values. Bone structures are segmented based on a pelvis bone atlas. Accuracy of syn-CT for patient treatment planning was analyzed by transferring the original plan and structures from the CT to syn-CT via rigid registration and recalculating dose. In addition, new IMRT plans were generated on the syn-CT using structures contoured on MR and transferred to the syn-CT. Accuracy of fiducial-based localization at the treatment machine performed using syn-CT or DRRs generated from syn-CT was assessed by comparing to orthogonal kV radiographs or CBCT. Results: Dosimetric comparison between CT and syn-CT was within 0.5% for all structures. The de-novo optimized plans generated on the syn-CT met our institutional clinical objectives for target and normal structures. Patient-induced susceptibility distortion based on B0 maps was within 1mm and 0.4 mm in the body and prostate. The rectal and bladder outlines on the syn-CT were deemed sufficient for assessing rectal and bladder filling on the CBCT at the time of treatment. CBCT localization showed a median error of < ±1 mm in LR, AP and SI direction. Conclusion: MRI derived syn-CT can be used clinically in MR-alone planning and treatment process for prostate. Drs. Deasy, Hunt and Tyagi have Master

  6. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  7. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  8. Contribution to uncertainties evaluation for fast reactors neutronic cross sections

    International Nuclear Information System (INIS)

    Privas, Edwin

    2015-01-01

    The thesis has been motivated by a wish to increase the uncertainty knowledge on nuclear data, for safety criteria. It aims the cross sections required by core calculation for sodium fast reactors (SFR), and new tools to evaluate its.The main objective of this work is to provide new tools in order to create coherent evaluated files, with reliable and mastered uncertainties. To answer those problematic, several methods have been implemented within the CONRAD code, which is developed at CEA of Cadarache. After a summary of all the elements required to understand the evaluation world, stochastic methods are presented in order to solve the Bayesian inference. They give the evaluator more information about probability density and they also can be used as validation tools. The algorithms have been successfully tested, despite long calculation time. Then, microscopic constraints have been implemented in CONRAD. They are defined as new information that should be taken into account during the evaluation process. An algorithm has been developed in order to solve, for example, continuity issues between two energy domains, with the Lagrange multiplier formalism. Another method is given by using a marginalization procedure, in order to either complete an existing evaluation with new covariance or add systematic uncertainty on an experiment described by two theories. The algorithms are well performed along examples, such the 238 U total cross section. The last parts focus on the integral data feedback, using methods of integral data assimilation to reduce the uncertainties on cross sections. This work ends with uncertainty reduction on key nuclear reactions, such the capture and fission cross sections of 238 U and 239 Pu, thanks to PROFIL and PROFIL-2 experiments in Phenix and the Jezebel benchmark. (author) [fr

  9. Improvement of methods to evaluate brittle failure resistance of the WWER reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Popov, A A; Parshutin, E V [Engineering Center of Nuclear Equipment Strength, Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Rogov, M F; Dragunov, U G [Experimenter` s and Designer` s Office ` ` Hydropress` ` (Russian Federation)

    1997-09-01

    At the next 10 years a number of Russian WWER nuclear power plants will complete its design lifetime. Normative methods to evaluate brittle failure resistance of the reactor pressure vessels used in Russia have been intended for design stage. The evaluation of reactor pressure vessel lifetime in operation stage demands to create new methods of calculation and new methods for experimental evaluation of brittle failure resistance degradation. The main objective of the study in this type of reactor is weldment number 4. In this report an analysis is made of methods to determine critical temperature of reactor materials including the results of instrumented Charpy testing. 12 figs.

  10. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

  11. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  12. Dosimetric investigations in mammography

    International Nuclear Information System (INIS)

    Metges, P.J.; Lorrain, S.

    1981-01-01

    The development film-screen detectors in radiological equipment has led us to study how to improve standard mammographic pictures (focus 0.3 x 0.3 mm, focus-film distance: 65) of thick and dense breasts by the use of an anti-scatter grid and by magnification. A dosimetric study was necessary to assess the doses delivered during mammographic examinations carried out according to various procedures. The results led to modify breast examination procedures and use an anti-scatter grid for breasts thicker than 4 cm or known as dense. The dose increase due to a better quality image is the lowest provided depth penetration is increased by 2 kV as compared to a standard picture. Absorbed doses on the X-ray axis, at 3 cm depth, are below 0.1 rad [fr

  13. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  14. Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area

    International Nuclear Information System (INIS)

    Yazid, M; Sutrisno; Sukarman-Aminjoyo; Zaenal-Abidin

    1996-01-01

    Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area has been done. The aim of this investigation is for tracing the possibility of radioactivity released in the environment during the operation of Kartini reactor. The data was evaluated were monthly monitored data taken from 1986 to 1994 period. The method of analysis was done by comparing the environmental radioactivity data before and after reactor commissioning, off side the reactor up to a radius of 5.000 meters and more than 5.000 meters from Kartini reactor and also compared to the maximum permissible radioactivity according to the current regulation. This evaluation showed that there was no indication of radioactivity release to the environment during this period of reactor operation

  15. The evaluation of research reactor TRIGA MARK II safety

    International Nuclear Information System (INIS)

    Jordan, R.; Kozuh, M.; Mavko, B.

    1994-01-01

    In the paper the Probabilistic Safety Analysis (PSA) of a research reactor is described. Five different initiating events were selected and analyzed with the use of event trees. Seven reactor systems were modeled with fault trees. Three groups of radiation releases were introduced - Success, Reactor-Hall, Environment - and their frequencies were estimated. The importance factors of initiating events, human errors and basic events were calculated regarding the consequence groups. (author)

  16. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  17. Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

    International Nuclear Information System (INIS)

    Kotake, Syoji; Nishikawa, Akira

    2005-01-01

    Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

  18. Reliability evaluation of the Savannah River reactor leak detection system

    International Nuclear Information System (INIS)

    Daugherty, W.L.; Sindelar, R.L.; Wallace, I.T.

    1991-01-01

    The Savannah River Reactors have been in operation since the mid-1950's. The primary degradation mode for the primary coolant loop piping is intergranular stress corrosion cracking. The leak-before-break (LBB) capability of the primary system piping has been demonstrated as part of an overall structural integrity evaluation. One element of the LBB analyses is a reliability evaluation of the leak detection system. The most sensitive element of the leak detection system is the airborne tritium monitors. The presence of small amounts of tritium in the heavy water coolant provide the basis for a very sensitive system of leak detection. The reliability of the tritium monitors to properly identify a crack leaking at a rate of either 50 or 300 lb/day (0.004 or 0.023 gpm, respectively) has been characterized. These leak rates correspond to action points for which specific operator actions are required. High reliability has been demonstrated using standard fault tree techniques. The probability of not detecting a leak within an assumed mission time of 24 hours is estimated to be approximately 5 x 10 -5 per demand. This result is obtained for both leak rates considered. The methodology and assumptions used to obtain this result are described in this paper. 3 refs., 1 fig., 1 tab

  19. Outlines of guidelines for the inspection and evaluation of reactor vessel internals

    International Nuclear Information System (INIS)

    Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

    2014-01-01

    'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

  20. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G.

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) [Berglund, 1987] and the Sodium Advanced Fast Reactor (SAFR) [Baumeister, 1987], were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the ''inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II [NED, 1986]. The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs

  1. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  2. Evaluation of carbon-14 life cycle in reactors VVER-1000

    International Nuclear Information System (INIS)

    Lysakova, Katerina; Neumann, Jan; Vonkova, Katerina

    2012-09-01

    This work is aimed at the evaluation of carbon-14 life cycle in light water reactors VVER-1000. Carbon-14 is generated as a side product in different systems of nuclear reactors and has been an issue not only in radioactive waste management but mainly in release into the environment in the form of gaseous effluents. The principal sources of this radionuclide are in primary cooling water and fuel. Considerable amount of C-14 is generated by neutron reactions with oxygen 17 O and nitrogen 14 N present in water coolant and fuel. The reaction likelihood and consequently volume of generated radioisotope depends on several factors, especially on the effective cross-section, concentrations of parent elements and conditions of power plant operating strategies. Due to its long half-life and high capability of integration into the environment and thus into the living species, it is very important to monitor the movement of carbon-14 in all systems of nuclear power plant and to manage its release out of NPP. The dominant forms of radioactive carbon-14 are the hydrocarbons owing to the combinations with hydrogen used for absorption of radiolytic oxygen. These organic compounds, such as formaldehyde, methyl alcohol, ethyl alcohol and formic acid can be mostly retained on ion exchange resins used in the system for purifying primary cooling water. The gaseous carbon compounds (CH 4 and CO 2 ) are released into the atmosphere via the ventilation systems of NPP. Based on the information and data obtained from different sources, it has been designed a balance model of possible carbon-14 pathways throughout the whole NPP. This model includes also mass balance model equations for each important node in system and available sampling points which will be the background for further calculations. This document is specifically not to intended to describe the best monitoring program attributes or technologies but rather to provide evaluation of obtained data and find the optimal way to

  3. Evaluation of an automated struvite reactor to recover phosphorus ...

    African Journals Online (AJOL)

    In the present study we attempted to develop a reactor system to recover phosphorus by struvite precipitation, and which can be installed anywhere in the field without access to a laboratory. A reactor was developed that can run fully automated and recover up to 93% of total phosphorus (total P). Turbidity and conductivity ...

  4. Evaluation of an automated struvite reactor to recover phosphorus ...

    African Journals Online (AJOL)

    2015-04-03

    Apr 3, 2015 ... A reactor was developed that can run fully automated and recover up to 93% ..... uncertainty. This technique will work best when the concentration of ... Ken Jack at the School of Chemical Engineering to help build the reactor ...

  5. SU-E-J-55: Dosimetric Evaluation of Centrally Located Lung Tumors: A Monte Carlo (MC) Study of Lung SBRT Planning

    Energy Technology Data Exchange (ETDEWEB)

    Pokhrel, D; Badkul, R; Jiang, H; Saleh, H; Estes, C; Park, J; Kumar, P; Wang, F [University Kansas Medical Center, Kansas City, KS (United States)

    2014-06-01

    Purpose: To compare dose distributions calculated using the iPlan XVMC algorithm and heterogeneities corrected/uncorrected Pencil Beam (PB-hete/PB-homo) algorithms for SBRT treatments of lung tumors. Methods: Ten patients with centrally located solitary lung tumors were treated using MC-based SBRT to 60Gy in 5 fractions for PTVV100%=95%. ITV was delineated on MIP-images based on 4D-CT scans. PTVs(ITV+5mm margins) ranged from 10.1–106.5cc(mean=48.6cc). MC-SBRT plans were generated with a combination of non-coplanar conformal arcs/beams using iPlan-XVMC-algorithm (BrainLABiPlan ver.4.1.2) for Novalis-TX consisting of HD-MLCs and 6MV-SRS(1000MU/min) mode, following RTOG 0813 dosimetric criteria. For comparison, PB-hete/PB-homo algorithms were used to re-calculate dose distributions using same beam configurations, MLCs/monitor units. Plans were evaluated with isocenter/maximal/mean doses to PTV. Normal lung doses were evaluated with V5/V10/V20 and mean-lung-dose(MLD), excluding PTV. Other OAR doses such as maximal spinal cord/2cc-esophagus/max bronchial tree (BT/maximal heart doses were tabulated. Results: Maximal/mean/isocenter doses to PTV calculated by PB-hete were uniformly larger than MC plans by a factors of 1.09/1.13/1.07, on average, whereas they were consistently lower by PB-homo by a factors of 0.9/0.84/0.9, respectively. The volume covered by 5Gy/10Gy/20Gy isodose-lines of the lung were comparable (average within±3%) when calculated by PB-hete compared to XVMC, but, consistently lower by PB-homo by a factors of 0.90/0.88/0.85, respectively. MLD was higher with PB-hete by 1.05, but, lower by PB-homo by 0.9, on average, compared to XVMC. XVMC max-cord/max-BT/max-heart and 2cc of esophagus doses were comparable to PB-hete; however, PB-homo underestimates by a factors of 0.82/0.89/0.88/0.86, on average, respectively. Conclusion: PB-hete significantly overestimates dose to PTV relative to XVMC -hence underdosing the target. MC is more complex and accurate with

  6. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.M.; Ketcham, D.R.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table testing which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its ''Generic Safety Evaluation Report'' approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the United States and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluating program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  7. Seismic evaluation of safety systems at the Savannah River reactors

    International Nuclear Information System (INIS)

    Hardy, G.S.; Johnson, J.J.; Eder, S.J.; Monahon, T.; Ketcham, D.

    1989-01-01

    A thorough review of all safety related systems in commercial nuclear power plants was prompted by the accident at the Three Mile Island Nuclear Power Plant. As a consequence of this review, the Nuclear Regulatory Commission (NRC) focused its attention on the environmental and seismic qualification of the industry's electrical and mechanical equipment. In 1980, the NRC issued Unresolved Safety Issue (USI) A-46 to verify the seismic adequacy of the equipment required to safely shut down a plant and maintain a stable condition for 72 hours. After extensive research by the NRC, it became apparent that traditional analysis and testing methods would not be a feasible mechanism to address this USI A-46 issue. The costs associated with utilizing the standard analytical and testing qualification approaches were exorbitant and could not be justified. In addition, the only equipment available to be shake table tested which is similar to the item being qualified is typically the nuclear plant component itself. After 8 years of studies and data collection, the NRC issued its Generic Safety Evaluation Report approving an alternate seismic qualification approach based on the use of seismic experience data. This experience-based seismic assessment approach will be the basis for evaluating each of the 70 pre-1972 commercial nuclear power units in the US and for an undetermined number of nuclear plants located in foreign countries. This same cost-effective approach developed for the commercial nuclear power industry is currently being applied to the Savannah River Production Reactors to address similar seismic adequacy issues. This paper documents the results of the Savannah River Plant seismic evaluation program. This effort marks the first complete (non-trial) application of this state-of-the-art USI A-46 resolution methodology

  8. Evaluation of strategies for end storage of high-level reactor fuel

    International Nuclear Information System (INIS)

    2001-01-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized

  9. Reanalysis and evaluation of seismic response of reactor building

    International Nuclear Information System (INIS)

    Li Zhongcheng; Li Zhongxian

    2005-01-01

    For the Ling Ao phase-I (LA-I) Nuclear Power Plant (NPP), its' seismic analysis of nuclear island was in accordance with the approaches in RCC-G standard for the model M310 in France, in which the Simplified impedance method was employed for the consideration of SSI. Thanks to the rapid progress being made in upgrading the evaluation technology and the capability of data processing systems, methods and software tools for the SSI analysis have experienced significant development all over the world. Focused on the model of reactor building of the LA-I NPP, in this paper the more sophisticated 3D half-space continuum impedance method based on the Green functions is used to analyze the functions of the soil, and then the seismic responses of the coupled SSI system are calculated and compared with the corresponding design values. It demonstrates that the design method provides a set of conservatively safe results. The conclusions from the study are hopefully to provide some important references to the assessment of seismic safety margin for LA-I NPP. (authors)

  10. Neutron evaluation of burnable poison insertion in pressurized water reactor

    International Nuclear Information System (INIS)

    Faria, Rochkhudson Batista de

    2013-01-01

    The development of this work was to match the 'Burn-up Credit Criticality Benchmark - Phase II-D - PWR-UO 2 Assembly Study of Control Rod Effects on Spent Fuel Composition' (case 15), which was modeled using the code MCNP5 and SCALE 6.0. The results of the infinite multiplication factor (k inf ) were compared with those obtained by international institutions. Later we performed in this same benchmark, a sensitivity analysis using SCALE 6.0. Thus, we tested several changes in case 15 of Benchmark, such as insertion of different percentages of burnable poison, changing the number and positions of the rods. In all cases were analyzed, comparisons and discussions about the results. The same methodology was applied to the reactor core of the Nuclear Plant in Brazil, Angra II, initially to evaluate its behavior when subjected to a variation in the percentage of burnable poison and then, introduce changes also in the enrichment of nuclear fuel, doing the appropriate comparisons of results. Considering results and experience gained, the Department of Nuclear Engineering, is prepared to control analysis of reactivity with the use of different types of burnable poisons under the code SCALE 6.0 through its various modules. (author)

  11. Methodology development for statistical evaluation of reactor safety analyses

    International Nuclear Information System (INIS)

    Mazumdar, M.; Marshall, J.A.; Chay, S.C.; Gay, R.

    1976-07-01

    In February 1975, Westinghouse Electric Corporation, under contract to Electric Power Research Institute, started a one-year program to develop methodology for statistical evaluation of nuclear-safety-related engineering analyses. The objectives of the program were to develop an understanding of the relative efficiencies of various computational methods which can be used to compute probability distributions of output variables due to input parameter uncertainties in analyses of design basis events for nuclear reactors and to develop methods for obtaining reasonably accurate estimates of these probability distributions at an economically feasible level. A series of tasks was set up to accomplish these objectives. Two of the tasks were to investigate the relative efficiencies and accuracies of various Monte Carlo and analytical techniques for obtaining such estimates for a simple thermal-hydraulic problem whose output variable of interest is given in a closed-form relationship of the input variables and to repeat the above study on a thermal-hydraulic problem in which the relationship between the predicted variable and the inputs is described by a short-running computer program. The purpose of the report presented is to document the results of the investigations completed under these tasks, giving the rationale for choices of techniques and problems, and to present interim conclusions

  12. Evaluation of the effectiveness of a thermal hygienization reactor

    Directory of Open Access Journals (Sweden)

    Daniel Borski

    2011-01-01

    Full Text Available For reasons of limiting the spread of serious transmissible diseases, with regard to the requirement for reducing landfill of biodegradable waste (which may or contains animal by-products and thus presents a potential risk to human and animal health and with a focus on supporting its separate collection, there has been created a legal framework for processing and hygienization of materials containing animal by-products. For the above reasons new technologies are being developed and implemented. These technologies are able to ensure the processing of biological waste containing animal by-products. As a practical result of the effort to ensure the hygienization of biowaste, a hygienization unit of own design, which uses the thermal way of hygienization, is presented in this work. The general part of the work defines a legislative framework for the assignment and gives technical parameters and minimum requirements for conversion that hygienization unit should be able to perform, including the limits for digestion residues and compost.In the experimental section there are described operational tests which document the technological process of hygienization depending on the aeration of the contents of the reactor. Experiment III outlines the validation process which uses contamination by indicator organisms, including subsequent checking of their occurrence as well as processing of the results of experiments and evaluation of the process of hygienization.

  13. Dosimetric comparison of stereotactic body radiotherapy using 4D CT and multiphase CT images for treatment planning of lung cancer: Evaluation of the impact on daily dose coverage

    International Nuclear Information System (INIS)

    Wang Lu; Hayes, Shelly; Paskalev, Kamen; Jin Lihui; Buyyounouski, Mark K.; Ma, Charlie C.-M.; Feigenberg, Steve

    2009-01-01

    Purpose: To investigate the dosimetric impact of using 4D CT and multiphase (helical) CT images for treatment planning target definition and the daily target coverage in hypofractionated stereotactic body radiotherapy (SBRT) of lung cancer. Materials and methods: For 10 consecutive patients treated with SBRT, a set of 4D CT images and three sets of multiphase helical CT scans, taken during free-breathing, end-inspiration and end-expiration breath-hold, were obtained. Three separate planning target volumes (PTVs) were created from these image sets. A PTV 4D was created from the maximum intensity projection (MIP) reconstructed 4D images by adding a 3 mm margin to the internal target volume (ITV). A PTV 3CT was created by generating ITV from gross target volumes (GTVs) contoured from the three multiphase images. Finally, a third conventional PTV (denoted PTV conv ) was created by adding 5 mm in the axial direction and 10 mm in the longitudinal direction to the GTV (in this work, GTV = CTV = clinical target volume) generated from free-breathing helical CT scans. Treatment planning was performed based on PTV 4D (denoted as Plan-1), and the plan was adopted for PTV 3CT and PTV conv to form Plan-2 and Plan-3, respectively, by superimposing 'Plan-1' onto the helical free-breathing CT data set using modified beam apertures that conformed to either PTV 3CT or PTV conv . We first studied the impact of PTV design on treatment planning by evaluating the dosimetry of the three PTVs under the three plans, respectively. Then we examined the effect of the PTV designs on the daily target coverage by utilizing pre-treatment localization CT (CT-on-rails) images for daily GTV contouring and dose recalculation. The changes in the dose parameters of D 95 and D 99 (the dose received by 95% and 99% of the target volume, respectively), and the V p (the volume receiving the prescription dose) of the daily GTVs were compared under the three plans before and after setup error correction

  14. TL and LOE dosimetric evaluation of diamond films exposed to beta and ultraviolet radiation; Evaluacion dosimetrica TL y LOE de peliculas de diamante expuestas a radiacion beta y ultravioleta

    Energy Technology Data Exchange (ETDEWEB)

    Preciado F, S.; Melendrez, R.; Chernov, V.; Barboza F, M. [Universidad de Sonora, A.P. 13 y A.P. 5-088, 83000 Hermosillo, Sonora (Mexico); Schreck, M. [Universitaet Augsburg, Institut fuer Physik D-86135 (Germany); Cruz Z, E. [ICN, UNAM, 04500 Mexico D.F. (Mexico)

    2005-07-01

    The diamond possesses a privileged position regarding other materials of great technological importance. Their applications go from the optics, microelectronics, metals industry, medicine and of course as dosemeter, in the registration and detection of ionizing and non ionizing radiation. In this work the results of TL/LOE obtained in two samples of diamond of 10 {mu}m thickness grown by the chemical vapor deposition method (CVD) assisted by microwave plasma. The films were deposited in a silicon substrate (001) starting from a mixture of gases composed of CH{sub 4}/H{sub 2} and 750 ppm of molecular nitrogen as dopant. The samples were exposed to beta radiation (Sr{sup 90}/ Y{sup 90}) and ultraviolet, being stimulated later on thermal (TL) and optically (LOE) to evaluate their dosimetric properties. The sample without doping presented high response TL/LOE to the ultraviolet and beta radiation. The TL glow curve of the sample without doping showed two TL peaks with second order kinetics in the range of 520 to 550 K, besides a peak with first order kinetics of more intensity around 607 K. The TL efficiency of the non doped sample is bigger than the doped with nitrogen; however the LOE efficiency is similar in both samples. The results indicate that the CVD diamond possesses excellent perspectives for dosimetric applications, with special importance in radiotherapy due to it is biologically compatible with the human tissue. (Author)

  15. Dosimetric studies in diagnostic radiology

    International Nuclear Information System (INIS)

    Mohamadain, K. E. M.

    2004-04-01

    A dosimetric study in pediatric radiology and adult patients was currently being carried out at the pediatrics units of two large hospitals in Rio de Janeiro city: IPPMG (Instituto de Pediatric e Puericultura Martagao Gesteira, University hospital of federal University of Rio de Janeiro), IFF (Instituto Fernandes Figueira, FIOCRUZ) and Hospital Geral de Bonsucesso, a large public hospital in Rio de Janeiro city (HGB) Brazil. The dosimetric study was also performed at three pediatrics units in Sudan, namely, Ahmed Gasim, Khartoum and Omdurman hospitals. For chest x-ray examination the entrance skin dose(ESD) for AP, PA and LAT projections of pediatric patients, and the scattered dose at the thyroid, ovary and gonads have been obtained with thermoluminescent dosimeters (TLD) and with use of a software package Dosecal in thr Brazilian hospitals, and with the software dosecal in the Sudanese hospitals.The aim of this work was to estimate the entrance skin dose (ESD), the effective dose (ED) and the body organ dose (BOD) for chest x-ray exposure in pediatric patients, and different exams for adults patients, and to compare the results obtained in the tow Countries Sudan and Brazil with the reference dose level. For ESD evaluation of the chest x-ray, three different TL dosimeters have been used, namely LiF: Mg, Ti (TLD 100) CaSo 4 : Dy and LiF:Mg, Cu,P (TLD 100 H). The age intervals considered were: 0-1 years, 1-5 years, 5-10 years and 10-15 years. The results obtained with all dosimeters were in good agreement with, those obtained by the dosecal software, especially for AP and PA projection. However, some discrepancies were found for the LAT projection. The results within Brazil were some what consistent while in Sudan, large difference were observed, it was also noted that the doses in Brazil hospitals were less than the reference dose levels while in Sudanese hospitals the doses were higher than the reference dose levels. For adult patients only the software dosecal

  16. Dosimetric properties of MOS transistors

    International Nuclear Information System (INIS)

    Frank, H.; Petr, I.

    1977-01-01

    The structure of MOS transistors is described and their characteristics given. The experiments performed and data in the literature show the following dosimetric properties of MOS transistors: while for low gamma doses the transistor response to exposure is linear, it shows saturation for higher doses (exceeding 10 3 Gy in tissue). The response is independent of the energy of radiation and of the dose rate (within 10 -2 to 10 5 Gy/s). The spontaneous reduction with time of the spatial charge captured by the oxide layer (fading) is small and acceptable from the point of view of dosimetry. Curves are given of isochronous annealing of the transistors following irradiation with 137 Cs and 18 MeV electrons for different voltages during irradiation. The curves show that in MOS transistors irradiated with high-energy electrons the effect of annealing is less than in transistors irradiated with 137 Cs. In view of the requirement of using higher temperatures (approx. 400 degC) for the complete ''erasing'' of the captured charge, unsealed systems must be used for dosimetric purposes. The effect was also studied of neutron radiation, proton radiation and electron radiation on the MOS transistor structure. For MOS transistor irradiation with 14 MeV neutrons from a neutron generator the response was 4% of that for gamma radiation at the same dose equivalent. The effect of proton radiation was studied as related to the changes in MOS transistor structure during space flights. The response curve shapes are similar to those of gamma radiation curves. The effect of electron radiation on the MOS structure was studied by many authors. The experiments show that for each thickness of the SiO 2 layer an electron energy exists at which the size of the charge captured in SiO 2 is the greatest. All data show that MOS transistors are promising for radiation dosimetry. The main advantage of MOS transistors as gamma dosemeters is the ease and speed of evaluation, low sensitivity to neutron

  17. Probabilistic method for evaluating reactivity margin of nuclear reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1984-01-01

    A probabilistic method is proposed that will permit in the design stage to estimate quantitatively the likelihood with which any or all design criteria applicable to a nuclear reactor are actually satisfied after its construction. The method is trially applied to the core reactivity balance problem of the experimental Very High Temperature Reactor, and calculations are performed on the probability with which a design study core will, upon construction, satisfy design criteria concerning (a) one rod stuck and (b) startup margin. The method should prove useful in making engineering judgments before approving reactor core design. (author)

  18. Elements for evaluation of fast breeder reactor's potential in Argentina

    International Nuclear Information System (INIS)

    Gho, C.J.

    1985-01-01

    Fast Breeder Reactors (FBR) main features are presented in a general form, including their physical principles, the history of their evolution, their relevant technological aspects and the basis for their comparison to other energy sources. This is completed with descriptions of typical reactors and a model of FBR penetration in the Argentine electrical network. It is recommended to form a multidisciplinary board to study which position should be taken with respect to this type of reactors. In the author's opinion a Research activity should be started and gradually increased for passing to Development activities after a short while. (Author) [es

  19. Assessment of very high-temperature reactors in process application. Appendix I. Evaluation of the reactor system

    International Nuclear Information System (INIS)

    Jones, J.E. Jr.; Spiewak, I.

    1976-12-01

    In April 1974, the U.S. Atomic Energy Commission [now the Energy Research and Development Administration (ERDA)] authorized General Atomic Company, General Electric Company, and Westinghouse Electric Corp., Astronuclear Laboratory, to assess the available technology for producing heat using very high-temperature nuclear reactors. An evaulation of these studies and of the technical and economic potential of very high-temperature reactors (VHTR) is presented. The VHTR is a helium-cooled graphite-moderated reactor. The concepts and technology are evaluated for producing process stream temperatures of 649, 760, 871, 982, and 1093 0 C (1200, 1400, 1600, 1800, and 2000 0 F). There are a number of large industrial process heat applications that could utilize the VHTR

  20. Dosimetric verification of IMRT plans

    International Nuclear Information System (INIS)

    Bulski, W.; Cheimicski, K.; Rostkowska, J.

    2012-01-01

    Intensity modulated radiotherapy (IMRT) is a complex procedure requiring proper dosimetric verification. IMRT dose distributions are characterized by steep dose gradients which enable to spare organs at risk and allow for an escalation of the dose to the tumor. They require large number of radiation beams (sometimes over 10). The fluence measurements for individual beams are not sufficient for evaluation of the total dose distribution and to assure patient safety. The methods used at the Centre of Oncology in Warsaw are presented. In order to measure dose distributions in various cross-sections the film dosimeters were used (radiographic Kodak EDR2 films and radiochromic Gafchromic EBT films). The film characteristics were carefully examined. Several types of tissue equivalent phantoms were developed. A methodology of comparing measured dose distributions against the distributions calculated by treatment planning systems (TPS) was developed and tested. The tolerance level for this comparison was set at 3% difference in dose and 3 mm in distance to agreement. The so called gamma formalism was used. The results of these comparisons for a group of over 600 patients are presented. Agreement was found in 87 % of cases. This film dosimetry methodology was used as a benchmark to test and validate the performance of commercially available 2D and 3D matrices of detectors (ionization chambers or diodes). The results of these validations are also presented. (authors)

  1. Dosimetric approaches: pregnancy and lactation

    International Nuclear Information System (INIS)

    Rojo, Ana M.

    2001-01-01

    The female nuclear medicine patient is of special concern to the evaluation of radiation dose since radiation protection point of view: a)- The females overall body size and organ sizes are generally smaller than those of her male counterpart (thus her radiation doses will be higher, given the same amounts of administered activity and similar biokinetics), the effective doses could be 25 per cent higher than a man; b)- Female gonads are inside the body instead of outside and are near several organs often important as source organs in internal dosimetry; female gonads doses could be up to 10 or 30 higher than male gonads (usually 3 order); c)- Risk of breast cancer is significantly higher among females than males; d)- During the pregnancy due to placental transfer of radiopharmaceuticals or radiation exposure from the urinary bladder the embryo/fetus could receive doses that must be avoid; e)- In the case of nursing infant is of special concern in such an analysis to determine the interruption period to avoid doses in the nursing infant. The dosimetric approaches to take account to assess internal doses in the pregnant woman and during the breast feeding are discussed. (author)

  2. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  3. Evaluation of the trial design studies for an advanced marine reactor, (2)

    International Nuclear Information System (INIS)

    Ambo, Noriaki; Yokomura, Takeyoshi.

    1988-03-01

    As for the CARAMEL fuel (plate-type fuel) that was the fuel of the integrated-type reactor which was one of the trial design studies for an Advanced Marine Reactor, its structure and its fuel specific characteristics were studied and compared with a fuel rod (cylindrical fuel), and the total characteristics of the caramel fuel was reviewed and evaluated. (author)

  4. Evaluation of the impact of a committed site on fusion reactor development

    International Nuclear Information System (INIS)

    Reid, R.L.; Nagy, A.

    1979-01-01

    The technical and economic merits of a committed fusion site for development of tokamak, mirror, and EBT reactor from ignition through demo phases were evaluated. Schedule compression resulting from evolving several reactor concepts and/or phases on a committed site as opposed to sequential use of independent sites was estimated. Land, water, and electrical power requirements for a committed fusion site were determined. A conceptual plot plan for siting three fusion reactors on a committed site was configured. Reactor support equipment common to the various concepts was identified as candidates for sharing. Licensing issues for fusion plants were briefly addressed

  5. Evaluation of Metal-Fueled Surface Reactor Concepts

    International Nuclear Information System (INIS)

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-01

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

  6. Evaluation of the BRV 10 diesel engine disruption of the multi purpose reactor G.A Siwabessy reactor

    International Nuclear Information System (INIS)

    Asep Saepuloh; Kiswanto; Muh Taufiq; Yuyut, S.

    2014-01-01

    Diesel generator is one of the important components of emergency electrical power supply when the main power supply is disrupted. Unable to operation of diesel engines will have a serious impact to the operation of the reactor. This paper aims to evaluate the cause of disruption of the diesel generator BRV10 at the Multi Purpose Reactor GA Siwabessy occurred in 2014. This event makes enough attention because its cause is deemed unusual. Evaluation is done by investigating the causes of the disorder, do the repair, test functions and anticipate that similar events do not recur in the future. From the results of the evaluation of the causes of disorders known that diesel is a diesel mixing with water and mud that had buried long estimated in the diesel engine fuel tank. Is believed to cause the fuel tank care is less than optimal. (author)

  7. Development of system design and seismic performance evaluation for reactor pool working platform of a research reactor

    International Nuclear Information System (INIS)

    Kwag, Shinyoung; Lee, Jong-Min; Oh, Jinho; Ryu, Jeong-Soo

    2014-01-01

    Highlights: • Design of reactor pool working platform (RPWP) is newly proposed for an open-tank-in-pool type research reactor. • Main concept of RPWP is to minimize the pool top radiation level. • Framework for seismic performance evaluation of nuclear SSCs in a deterministic and a probabilistic manner is proposed. • Structural integrity, serviceability, and seismic margin of the RPWP are evaluated during and after seismic events. -- Abstract: The reactor pool working platform (RPWP) has been newly designed for an open-tank-in-pool type research reactor, and its seismic response, structural integrity, serviceability, and seismic margin have been evaluated during and after seismic events in this paper. The main important concept of the RPWP is to minimize the pool top radiation level by physically covering the reactor pool of the open-tank-in-pool type research reactor and suppressing the rise of flow induced by the primary cooling system. It is also to provide easy handling of the irradiated objects under the pool water by providing guide tubes and refueling cover to make the radioisotopes irradiated and protect the reactor structure assembly. For this concept, the new three dimensional design model of the RPWP is established for manufacturing, installation and operation, and the analytical model is developed to analyze the seismic performance. Since it is submerged under and influenced by water, the hydrodynamic effect is taken into account by using the hydrodynamic added mass method. To investigate the dynamic characteristics of the RPWP, a modal analysis of the developed analytical model is performed. To evaluate the structural integrity and serviceability of the RPWP, the response spectrum analysis and response time history analysis have been performed under the static load and the seismic load of a safe shutdown earthquake (SSE). Their stresses are analyzed for the structural integrity. The possibility of an impact between the RPWP and the most

  8. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  9. Evaluating Russian space nuclear reactor technology for United States applications

    International Nuclear Information System (INIS)

    Polansky, G.F.; Schmidt, G.L.; Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Space nuclear power and nuclear electric propulsion are considered important technologies for planetary exploration, as well as selected earth orbit applications. The Nuclear Electric Propulsion Space Test Program (NEPSTP) was intended to provide an early flight demonstration of these technologies at relatively low cost through extensive use of existing Russian technology. The key element of Russian technology employed in the program was the Topaz II reactor. Refocusing of the activities of the Ballistic Missile Defense Organization (BMDO), combined with budgetary pressures, forced the cancellation of the NEPSTP at the end of the 1993 fiscal year. The NEPSTP was faced with many unique flight qualification issues. In general, the launch of a spacecraft employing a nuclear reactor power system complicates many spacecraft qualification activities. However, the NEPSTP activities were further complicated because the reactor power system was a Russian design. Therefore, this program considered not only the unique flight qualification issues associated with space nuclear power, but also with differences between Russian and United States flight qualification procedures. This paper presents an overview of the NEPSTP. The program goals, the proposed mission, the spacecraft, and the Topaz II space nuclear power system are described. The subject of flight qualification is examined and the inherent difficulties of qualifying a space reactor are described. The differences between United States and Russian flight qualification procedures are explored. A plan is then described that was developed to determine an appropriate flight qualification program for the Topaz II reactor to support a possible NEPSTP launch

  10. Dosimetric evaluation of multi-pattern spatially fractionated radiation therapy using a multi-leaf collimator and collapsed cone convolution superposition dose calculation algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Stathakis, Sotirios [Department of Radiation Oncology, University of Texas Health Science Center San Antonio, 7979 Wurzbach Rd, San Antonio, TX 78229 (United States)], E-mail: stathakis@uthscsa.edu; Esquivel, Carlos; Gutierrez, Alonso N.; Shi, ChengYu; Papanikolaou, Niko [Department of Radiation Oncology, University of Texas Health Science Center San Antonio, 7979 Wurzbach Rd, San Antonio, TX 78229 (United States)

    2009-10-15

    Purpose: In this paper, we present an alternative to the originally proposed technique for the delivery of spatially fractionated radiation therapy (GRID) using multi-leaf collimator (MLC) shaped fields. We employ the MLC to deliver various pattern GRID treatments to large solid tumors and dosimetrically characterize the GRID fields. Methods and materials: The GRID fields were created with different open to blocked area ratios and with variable separation between the openings using a MLC. GRID designs were introduced into the Pinnacle{sup 3} treatment planning system, and the dose was calculated in a water phantom. Ionization chamber and film measurements using both Kodak EDR2 and Gafchromic EBT film were performed in a SolidWater phantom to determine the relative output of each GRID design as well as its spatial dosimetric characteristics. Results: Agreement within 5.0% was observed between the Pinnacle{sup 3} predicted dose distributions and the measurements for the majority of experiments performed. A higher magnitude of discrepancy (15%) was observed using a high photon beam energy (18 MV) and small GRID opening. Skin dose at the GRID openings was higher than the corresponding open field by a factor as high as three for both photon energies and was found to be independent of the open-to-blocked area ratio. Conclusion: In summary, we reaffirm that the MLC can be used to deliver spatially fractionated GRID therapy and show that various GRID patterns may be generated. The Pinnacle{sup 3} TPS can accurately calculate the dose of the different GRID patterns in our study to within 5% for the majority of the cases based on film and ion chamber measurements. Disadvantages of MLC-based GRID therapy are longer treatment times and higher surface doses.

  11. SU-F-SPS-04: Dosimetric Evaluation of the Dose Calculation Accuracy of Different Algorithms for Two Different Treatment Techniques During Whole Breast Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Pacaci, P; Cebe, M; Mabhouti, H; Codel, G; Serin, E; Sanli, E; Kucukmorkoc, E; Doyuran, M; Kucuk, N; Canoglu, D; Altinok, A; Acar, H; Caglar Ozkok, H [Medipol University, Istanbul, Istanbul (Turkey)

    2016-06-15

    Purpose: In this study, dosimetric comparison of field in field (FIF) and intensity modulated radiation therapy (IMRT) techniques used for treatment of whole breast radiotherapy (WBRT) were made. The dosimetric accuracy of treatment planning system (TPS) for Anisotropic Analytical Algorithm (AAA) and Acuros XB (AXB) algorithms in predicting PTV and OAR doses was also investigated. Methods: Two different treatment planning techniques of left-sided breast cancer were generated for rando phantom. FIF and IMRT plans were compared for doses in PTV and OAR volumes including ipsilateral lung, heart, left ascending coronary artery, contralateral lung and the contralateral breast. PTV and OARs doses and homogeneity and conformality indexes were compared between two techniques. The accuracy of TPS dose calculation algorithms was tested by comparing PTV and OAR doses measured by thermoluminescent dosimetry with the dose calculated by the TPS using AAA and AXB for both techniques. Results: IMRT plans had better conformality and homogeneity indexes than FIF technique and it spared OARs better than FIF. While both algorithms overestimated PTV doses they underestimated all OAR doses. For IMRT plan, PTV doses, overestimation up to 2.5 % was seen with AAA algorithm but it decreased to 1.8 % when AXB algorithm was used. Based on the results of the anthropomorphic measurements for OAR doses, underestimation greater than 7 % is possible by the AAA. The results from the AXB are much better than the AAA algorithm. However, underestimations of 4.8 % were found in some of the points even for AXB. For FIF plan, similar trend was seen for PTV and OARs doses in both algorithm. Conclusion: When using the Eclipse TPS for breast cancer, AXB the should be used instead of the AAA algorithm, bearing in mind that the AXB may still underestimate all OAR doses.

  12. Acceptance criteria for the evaluation of nuclear power reactor security plans

    International Nuclear Information System (INIS)

    1982-08-01

    This guidance document contains acceptance criteria to be used in the NRC license review process. It contains specific criteria for use in evaluating the acceptability of nuclear power reactor security programs as detailed in security plans

  13. Guideline for examination concerning the evaluation of safety in light water power reactor installations

    International Nuclear Information System (INIS)

    1978-01-01

    This guideline was drawn up as the guide for examination when the safety evaluation of nuclear reactors is carried out at the time of approving the installation of light water power reactors. Accordingly in case of the examination of safety, it must be confirmed that the contents of application are in conformity with this guideline. If they are in conformity, it is judged that the safety evaluation of the policy in the basic design of a reactor facility is adequate, and also that the evaluation concerning the separation from the public in surroundings is adequate as the condition of location of the reactor facility. This guideline is concerned with light water power reactors now in use, but the basic concept may be the reference for the examination of the other types of reactors. If such a case occurs that the safety evaluation does not conform to this guideline, it is not excluded when the appropriate reason is clarified. The purpose of safety evaluation, the scope to be evaluated, the selection of the events to be evaluated, the criteria for judgement, the matters taken into consideration at the time of analysis, the concrete events of abnormal transient change and accident in operation, and the concrete events of serious accident and hypothetic accident are stipulated. The explanation and two appendices are attached. (Kako, I.)

  14. Dosimetric system for prolonged manned flights

    International Nuclear Information System (INIS)

    Akatov, Yu.A.; Kovalev, E.E.; Sakovich, V.A.; Deme, Sh.; Fekher, I.; Nguen, V.D.

    1991-01-01

    Comments for the All-Union state standard 25645.202-83 named Radiation safety of a spacecraft crew during space flight. Requirements for personnel dosimetric control, are given. Devices for the dosimetric control used in manned space flights nowadays are reviewed. The performance principle and structure of the FEDOR dosimetric complex under development are discussed

  15. ECORA - Evaluation of Computational Methods for Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Scheuerer, Martina

    2002-01-01

    There were three motivations behind the ECORA Project: - the shortcomings of 0-D system codes in the simulation of 3-D, local flow and heat transfer phenomena, - increased interest in the application of 3-D CFD software as supplement to system codes, - high safety requirements in the nuclear industry required consistent standards for the use and assessment of CFD software. The purpose of ECORA was therefore: - to establish performance criteria for the assessment of CFD software, - to establish Best Practice Guidelines for application and use of CFD software, with the following objectives: - assessment of CFD applications in reactor safety: flows in containment (PANDA experiments) and flows in primary system (UPTF experiments) - Best Practice Guidelines for reactor safety: starting point (ERCOFTAC Best Practice Guidelines), adaptation to CFD application for nuclear safety, extension to assessment of experimental data - recommendations for improvements of CFD software, - network of European 'Centres of Competence for CFD Applications in Reactor Safety'. Currently, there were twelve partners in the ECORA Project, representing nine European countries. The Project was scheduled to last until September 2004. Ms Scheuerer then described the work programme and project structure, the Best Practice Guidelines for CFD simulations, the procedures for quantifying errors, applications of Best Practice Guidelines, Best Practice Guidelines for experimental data, applications to primary system, UPTF and PANDA data. Her conclusions were the following: - the Project had led to the improvement of the quality of CFD calculations in reactor safety, through: the ECORA Best Practice Guidelines, the assessment of shortcomings and the improvement of mathematical models. - It had also led to higher acceptance of CFD in reactor safety. - The next step was the establishment of European 'Centres of Competence for CFD Applications in reactor Safety'

  16. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  17. Ageing evaluation model of nuclear reactors structural elements

    International Nuclear Information System (INIS)

    Ziliukas, A.; Jutas, A.; Leisis, V.

    2002-01-01

    In this article the estimation of non-failure probability by random faults on the structural elements of nuclear reactors is presented. Ageing is certainly a significant factor in determining the limits of nuclear plant lifetime or life extensions. Usually the non failure probability rates failure intensity, which is characteristic for structural elements ageing in nuclear reactors. In practice the reliability is increased incorrectly because not all failures are fixed and cumulated. Therefore, the methodology with using the fine parameter of the failures flow is described. The comparison of non failure probability and failures flow is carried out. The calculation of these parameters in the practical example is shown too. (author)

  18. Evaluation of the activity levels in fusion reactor blankets

    International Nuclear Information System (INIS)

    Gruber, J.

    1977-05-01

    The activation of a fusion reactor blanket (316 SS or V-10Cr-10Ti as structure) with a minimum lithium inventory has been calculated for 0.83 MW/m 2 wall load. The resulting radiation levels and waste problems are discussed. The dose rate near the steel structure will always be higher than 0.1 rem/h due to its niobium content. After 200 to 100,000 years of decay the potential biological hazard originating from this high level fusion reactor waste (with plutonium recyclation). (orig.) [de

  19. Dosimetric essay in dental radiology

    International Nuclear Information System (INIS)

    Lopez Salaberry, M.

    1998-01-01

    A neck study was observated in the tiroids glands,laryngeal zone, sensitive organs for the ionizing radiation for increase dental xray exams. Was selected 29th patients with radiography prescription complete (in the Odontology Faculty Clinics Uruguaian). It took radiographies with and without tiroids necklace and apron lead using dosemeters. Dosimetric studies had demonstrated good dose between patients. For measuring the radiation dose have been used TLD thermoluminescence dosimetric and Harshaw 6600 for read it. The thyroids necklace use and odontology postgrading for training course for dentistry was the two recommendations advised

  20. The dosimetric control in radiotherapy

    International Nuclear Information System (INIS)

    Veres, A.

    2009-01-01

    The author first presents the thermoluminescent dosimetry method developed by the Equal-Estro Laboratory to control radiotherapy systems, according to which dosimeters are mailed by the radiotherapy centres to the laboratory, and then analyzed with respect to the level of dose bias. In a second part, he discusses the different techniques used for the dosimetric control of new radiotherapy methods (intensity-modulated radiation therapy, tomo-therapy) for which film dosimetry is applied. He also evokes the development of new phantoms and the development of a method for the dosimetric control of proton beams

  1. Structure design and realization of advanced nuclear reactor expert evaluation system

    International Nuclear Information System (INIS)

    Gao Bin; Zhou Zhiwei; Gu Junyang

    2007-01-01

    Advanced nuclear reactor expert evaluation system is the initial practice of software on nuclear power plants evaluation system. The system was developed in C++ code under the Visual Studio Net environment, and it used Model-View-Control (MVC) pattern as its basic frame. The system was used to access the advanced nuclear reactor in China. Available results illustrate that the frame of the system is feasible and effective. (authors)

  2. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  3. Design and cost evaluation of generic magnetic fusion reactor using the D-D fuel cycle

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1988-01-01

    A fusion reactor systems code has been developed to evaluate the economic potential of power generation from a toroidal magnetic fusion reactor using deuterium-deuterium (D-D) fuel. A method similar to that developed by J. Sheffield, of the Oak Ridge National Laboratory, for deuterium-tritium (D-T) fuel was used to model the generic aspects of magnetic fusion reactors. The results of the systems study and cost evaluation show that the cost of electricity produced by a D-D reactor is two times higher than that produced by an equivalent D-T reactor design. The significant finding of the study is that the cost ratio between the D-D and D-T systems can potentially be reduced to 1.5 by improved engineering design and even lower by better physics performance. The absolute costs for both systems at this level are close to the costs for nuclear fission and fossil fuel plants. A design for a magnet reinforced with advanced composite materials is presented as an example of an engineering improvement that could reduce the cost of electricity produced by both reactors. However, since the magnets in the D-D reactor are much larger than in the K-T reactor, the cost ratio of the two systems is significantly reduced

  4. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

  5. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  6. Evaluation guide for the international reactor physics experiments evaluation project (IRPhEP)

    International Nuclear Information System (INIS)

    Yamaji, Akifumi

    2013-01-01

    At present, there is an urgent need to preserve integral reactor physics experimental data including separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. The International Reactor Physics Evaluation Project (IRPhEP) was initiated as a pilot activity in 1999 by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. This document outlines the general presentation guidelines that evaluators should follow for the description of the experiments and all relevant experimental data in order to ensure the consistency between the evaluations published in the final Handbook. Publication templates will be used to ensure this consistency and will follow the general scheme below: 1 - Experiment identification number; 2- Date; 3 - Name of experiment (Purpose of experiment, Phenomena measured and scope); 4 - Name or designation of experimental programme; 5 - Description of facility; 6 - Description of test or experiment (Experimental configuration, Core life cycle, Experimental limitations or shortcomings); 7 - Phenomena measured (Description of results and analysis, Special features and characteristics of experiment, Measurement systems/methods and uncertainties); 8 - Duplicate or complementary experiments / other related experiments; 9 - Status of completion of the evaluation; 10 - References (pointer to evaluation, archive if available, otherwise generic bibliographic reference); 11 - Authors/ organisers 12 - Material available

  7. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    Energy Technology Data Exchange (ETDEWEB)

    Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2013-07-01

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  8. Evaluation and improvement of nondestructive evaluation reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Bates, D.J.; Deffenbaugh, J.D.; Good, M.S.; Heasler, P.G.; Mart, G.A.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Van Fleet, L.G.

    1987-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection (ISI) of Light Water Reactors (NDE Reliability) Program at Pacific Northwest Laboratory (PNL) was established to determine the reliability of current ISI techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this NRC program are to: determine the reliability of ultrasonic ISI performed on commercial light-water reactor (LWR) primary systems, using probabilistic fracture mechanics analysis, determine the impact of NDE unreliability on system safety and determine the level of inspection reliability required to ensure a suitably low failure probability, evaluate the degree of reliability improvement that could be achieved using improved and advanced NDE techniques, based on material properties, service conditions, and NDE uncertainties, recommend revisions to ASME Code, Section XI, and Regulatory Requirements that will ensure suitably low failure probabilities. The scope of this program is limited to ISI of primary systems; the results and recommendations may also be applicable to Class II piping systems

  9. A performance evaluation of a microchannel reactor for the production of hydrogen from formic acid for electrochemical energy applications

    CSIR Research Space (South Africa)

    Ndlovu, IM

    2017-12-01

    Full Text Available An experimental evaluation of a microchannel reactor was completed to assess the reactor performance for the catalytic decomposition of vaporised formic acid (FA) for H2 production. Initially, X-ray powder diffraction (XRD), elemental mapping using...

  10. Dosimetric adaptive IMRT driven by fiducial points

    International Nuclear Information System (INIS)

    Crijns, Wouter; Van Herck, Hans; Defraene, Gilles; Van den Bergh, Laura; Haustermans, Karin; Slagmolen, Pieter; Maes, Frederik; Van den Heuvel, Frank

    2014-01-01

    Purpose: Intensity modulated radiotherapy (IMRT) and volumetric modulated arc therapy have become standard treatments but are more sensitive to anatomical variations than 3D conformal techniques. To correct for inter- and intrafraction anatomical variations, fast and easy to implement methods are needed. Here, the authors propose a full dosimetric IMRT correction that finds a compromise in-between basic repositioning (the current clinical practice) and full replanning. It simplifies replanning by avoiding a recontouring step and a full dose calculation. It surpasses repositioning by updating the preoptimized fluence and monitor units (MU) using a limited number of fiducial points and a pretreatment (CB)CT. To adapt the fluence the fiducial points were projected in the beam's eye view (BEV). To adapt the MUs, point dose calculation towards the same fiducial points were performed. The proposed method is intrinsically fast and robust, and simple to understand for operators, because of the use of only four fiducial points and the beam data based point dose calculations. Methods: To perform our dosimetric adaptation, two fluence corrections in the BEV are combined with two MU correction steps along the beam's path. (1) A transformation of the fluence map such that it is realigned with the current target geometry. (2) A correction for an unintended scaling of the penumbra margin when the treatment beams scale to the current target size. (3) A correction for the target depth relative to the body contour and (4) a correction for the target distance to the source. The impact of the correction strategy and its individual components was evaluated by simulations on a virtual prostate phantom. This heterogeneous reference phantom was systematically subjected to population based prostate transformations to simulate interfraction variations. Additionally, a patient example illustrated the clinical practice. The correction strategy was evaluated using both dosimetric

  11. Dosimetric adaptive IMRT driven by fiducial points

    Energy Technology Data Exchange (ETDEWEB)

    Crijns, Wouter, E-mail: wouter.crijns@uzleuven.be [Department of Oncology, Laboratory of Experimental Radiotherapy, KU Leuven, Herestraat 49, 3000 Leuven, Belgium and Medical Imaging Research Center, KU Leuven, Herestraat 49, 3000 Leuven (Belgium); Van Herck, Hans [Medical Imaging Research Center, KU Leuven, Herestraat 49, 3000 Leuven, Belgium and Department of Electrical Engineering (ESAT) – PSI, Center for the Processing of Speech and Images, KU Leuven, 3000 Leuven (Belgium); Defraene, Gilles; Van den Bergh, Laura; Haustermans, Karin [Department of Oncology, Laboratory of Experimental Radiotherapy, KU Leuven, Herestraat 49, 3000 Leuven (Belgium); Slagmolen, Pieter [Medical Imaging Research Center, KU Leuven, Herestraat 49, 3000 Leuven (Belgium); Department of Electrical Engineering (ESAT) – PSI, Center for the Processing of Speech and Images, KU Leuven, 3000 Leuven (Belgium); iMinds-KU Leuven Medical IT Department, KU Leuven, 3000 Leuven (Belgium); Maes, Frederik [Medical Imaging Research Center, KU Leuven, Herestraat 49, 3000 Leuven (Belgium); Department of Electrical Engineering (ESAT) – PSI, Center for the Processing of Speech and Images, KU Leuven and iMinds, 3000 Leuven (Belgium); Van den Heuvel, Frank [Department of Oncology, Laboratory of Experimental Radiotherapy, KU Leuven, Herestraat 49, 3000 Leuven, Belgium and Department of Oncology, MRC-CR-UK Gray Institute of Radiation Oncology and Biology, University of Oxford, Oxford OX1 2JD (United Kingdom)

    2014-06-15

    Purpose: Intensity modulated radiotherapy (IMRT) and volumetric modulated arc therapy have become standard treatments but are more sensitive to anatomical variations than 3D conformal techniques. To correct for inter- and intrafraction anatomical variations, fast and easy to implement methods are needed. Here, the authors propose a full dosimetric IMRT correction that finds a compromise in-between basic repositioning (the current clinical practice) and full replanning. It simplifies replanning by avoiding a recontouring step and a full dose calculation. It surpasses repositioning by updating the preoptimized fluence and monitor units (MU) using a limited number of fiducial points and a pretreatment (CB)CT. To adapt the fluence the fiducial points were projected in the beam's eye view (BEV). To adapt the MUs, point dose calculation towards the same fiducial points were performed. The proposed method is intrinsically fast and robust, and simple to understand for operators, because of the use of only four fiducial points and the beam data based point dose calculations. Methods: To perform our dosimetric adaptation, two fluence corrections in the BEV are combined with two MU correction steps along the beam's path. (1) A transformation of the fluence map such that it is realigned with the current target geometry. (2) A correction for an unintended scaling of the penumbra margin when the treatment beams scale to the current target size. (3) A correction for the target depth relative to the body contour and (4) a correction for the target distance to the source. The impact of the correction strategy and its individual components was evaluated by simulations on a virtual prostate phantom. This heterogeneous reference phantom was systematically subjected to population based prostate transformations to simulate interfraction variations. Additionally, a patient example illustrated the clinical practice. The correction strategy was evaluated using both dosimetric

  12. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  13. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    International Nuclear Information System (INIS)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies

  14. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  15. Evaluation of the qualification of SPERT [Special Power Excursion Reactor Test] fuel for use in non-power reactors

    International Nuclear Information System (INIS)

    1987-08-01

    This report summarizes the US Nuclear Regulatory Commission staff's evaluation of the qualification of the stainless-steel-clad uranium/oxide (UO 2 ) fuel pins for use in non-power reactors. The fuel pins were originally procured in the 1960's as part of the Special Power Excursion Reactor Test (SPERT) program. Argonne National Laboratory (ANL) examined 600 SPERT fuel pins to verify that the pins were produced according to specification and to assess their present condition. The pins were visually inspected under 6X magnification and by X-radiographic, destructive, and metallographic examinations. Spectrographic and chemical analyses were performed on the UO 2 fuel. The results of the qualification examinations indicated that the SPERT fuel pins meet the requirements of Phillips Specification No. F-1-SPT and have suffered no physical damage since fabrication. Therefore, the qualification results give reasonable assurance that the SPERT fuel rods are suitable for use in non-power reactors provided that the effects of thin-wall defects in the region of the upper end cap and low-density fuel pellets are evaluated for the intended operating conditions. 1 ref., 4 figs., 11 tabs

  16. SU-E-T-131: Dosimetric Impact and Evaluation of Different Heterogenity Algorithm in Volumetric Modulated Arc Therapy Plan for Stereotactic Ablative Radiotherapy Lung Treatment with the Flattening Filter Free Beam

    Energy Technology Data Exchange (ETDEWEB)

    Chung, J; Kim, J [Seoul National University Bundang Hospital, Seongnam, Kyeonggi-do (Korea, Republic of); Lee, J [Konkuk University Medical Center, Seoul, Seoul (Korea, Republic of); Kim, Y [Choonhae College of Health Sciences, Ulsan (Korea, Republic of)

    2014-06-01

    Purpose: The present study aimed to investigate the dosimetric impacts of the anisotropic analytic algorithm (AAA) and the Acuros XB (AXB) plan for lung stereotactic ablative radiation therapy using flattening filter-free (FFF) beam. We retrospectively analyzed 10 patients. Methods: We retrospectively analyzed 10 patients. The dosimetric parameters for the target and organs at risk (OARs) from the treatment plans calculated with these dose calculation algorithms were compared. The technical parameters, such as the computation times and the total monitor units (MUs), were also evaluated. Results: A comparison of DVHs from AXB and AAA showed that the AXB plan produced a high maximum PTV dose by average 4.40% with a statistical significance but slightly lower mean PTV dose by average 5.20% compared to the AAA plans. The maximum dose to the lung was slightly higher in the AXB compared to the AAA. For both algorithms, the values of V5, V10 and V20 for ipsilateral lung were higher in the AXB plan more than those of AAA. However, these parameters for contralateral lung were comparable. The differences of maximum dose for the spinal cord and heart were also small. The computation time of AXB was found fast with the relative difference of 13.7% than those of AAA. The average of monitor units (MUs) for all patients was higher in AXB plans than in the AAA plans. These results indicated that the difference between AXB and AAA are large in heterogeneous region with low density. Conclusion: The AXB provided the advantages such as the accuracy of calculations and the reduction of the computation time in lung stereotactic ablative radiotherapy (SABR) with using FFF beam, especially for VMAT planning. In dose calculation with the media of different density, therefore, the careful attention should be taken regarding the impacts of different heterogeneity correction algorithms. The authors report no conflicts of interest.

  17. SU-E-T-131: Dosimetric Impact and Evaluation of Different Heterogenity Algorithm in Volumetric Modulated Arc Therapy Plan for Stereotactic Ablative Radiotherapy Lung Treatment with the Flattening Filter Free Beam

    International Nuclear Information System (INIS)

    Chung, J; Kim, J; Lee, J; Kim, Y

    2014-01-01

    Purpose: The present study aimed to investigate the dosimetric impacts of the anisotropic analytic algorithm (AAA) and the Acuros XB (AXB) plan for lung stereotactic ablative radiation therapy using flattening filter-free (FFF) beam. We retrospectively analyzed 10 patients. Methods: We retrospectively analyzed 10 patients. The dosimetric parameters for the target and organs at risk (OARs) from the treatment plans calculated with these dose calculation algorithms were compared. The technical parameters, such as the computation times and the total monitor units (MUs), were also evaluated. Results: A comparison of DVHs from AXB and AAA showed that the AXB plan produced a high maximum PTV dose by average 4.40% with a statistical significance but slightly lower mean PTV dose by average 5.20% compared to the AAA plans. The maximum dose to the lung was slightly higher in the AXB compared to the AAA. For both algorithms, the values of V5, V10 and V20 for ipsilateral lung were higher in the AXB plan more than those of AAA. However, these parameters for contralateral lung were comparable. The differences of maximum dose for the spinal cord and heart were also small. The computation time of AXB was found fast with the relative difference of 13.7% than those of AAA. The average of monitor units (MUs) for all patients was higher in AXB plans than in the AAA plans. These results indicated that the difference between AXB and AAA are large in heterogeneous region with low density. Conclusion: The AXB provided the advantages such as the accuracy of calculations and the reduction of the computation time in lung stereotactic ablative radiotherapy (SABR) with using FFF beam, especially for VMAT planning. In dose calculation with the media of different density, therefore, the careful attention should be taken regarding the impacts of different heterogeneity correction algorithms. The authors report no conflicts of interest

  18. Evaluation of enzymatic reactors for large-scale panose production.

    Science.gov (United States)

    Fernandes, Fabiano A N; Rodrigues, Sueli

    2007-07-01

    Panose is a trisaccharide constituted by a maltose molecule bonded to a glucose molecule by an alpha-1,6-glycosidic bond. This trisaccharide has potential to be used in the food industry as a noncariogenic sweetener, as the oral flora does not ferment it. Panose can also be considered prebiotic for stimulating the growth of benefic microorganisms, such as lactobacillus and bifidobacteria, and for inhibiting the growth of undesired microorganisms such as E. coli and Salmonella. In this paper, the production of panose by enzymatic synthesis in a batch and a fed-batch reactor was optimized using a mathematical model developed to simulate the process. Results show that optimum production is obtained in a fed-batch process with an optimum production of 11.23 g/l h of panose, which is 51.5% higher than production with batch reactor.

  19. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  20. EXPERIMENTAL EVALUATION OF THE FULLY LOADED ELK RIVER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, J. R.; Diaz, A.

    1963-06-15

    The loading and testing program of the Elk River Reactor confirmed the predicted values. The measured cold, clean excess reactivity agrees to 2% and the control rod worths to 1% of the calculated values. The reactivity for various core loadings and rod positions is tabulated. The effects of spiked elements on the reactivity and radial peak-toaverage power ratio were studied. (D.L.C.)

  1. Economic evaluation of reprocessing and thermal reactor recycle

    International Nuclear Information System (INIS)

    Marshall, W.

    This paper provides a summing up of the discussions on economic aspects in WG4. These discussions also took account of the strategic, ecological and public acceptance factors intimately involved. Tentative conclusions are put forward as a basis for discussion. Reprocessing may take place for reasons other than just strictly economic ones. The decisions facing various countries are rationalized by considering their reactions to the range of possible uranium prices and fast reactor costs in the future

  2. Development of source term evaluation method for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Keon Jae; Cheong, Jae Hak; Park, Jin Baek; Kim, Guk Gee [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-10-15

    This project had investigate several design features of radioactive waste processing system and method to predict nuclide concentration at primary coolant basic concept of next generation reactor and safety goals at the former phase. In this project several prediction methods of source term are evaluated conglomerately and detailed contents of this project are : model evaluation of nuclide concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant(NPP), investigation of prediction parameter of source term evaluation, basic parameter of PWR, operational parameter, respectively, radionuclide removal system and adjustment values of reference NPP, suggestion of source term prediction method of next generation NPP.

  3. Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

    2015-10-15

    Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

  4. Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

    International Nuclear Information System (INIS)

    Utoh, Hiroyasu; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

    2015-01-01

    Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

  5. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    International Nuclear Information System (INIS)

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi

    2014-01-01

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided

  6. Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

    International Nuclear Information System (INIS)

    Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

    2007-01-01

    In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

  7. The accident of stereotaxic radiosurgery at the University hospital center of Toulouse. Expert report n.2. Dosimetric and clinical evaluation. Risk analysis

    International Nuclear Information System (INIS)

    2008-01-01

    The regional center of stereotaxic radiosurgery (C.R.R.S.) of the University hospital center (C.H.U.) of Toulouse is equipped since april 2006 with a Novalis accelerator (Brainlab) devoted to radiosurgery and intra skull stereotaxic radiotherapy.In april 2007, during an intercomparison of dosimetry files coming from various sites, the Brainlab society detects an anomaly. The analysis made by the society concludes to the use of an unsuited detector for the measurement of a dosimetry parameter during the accelerator initial calibration. Following this error, 145 patients (on 172 patients treated since the center opening) suffer of an overdose whom importance is variable. On the 26. june 2007 the Authority of nuclear safety (Asn) requires an expertise on the following points: checking of the experimental protocols of micro-beams calibration before and after correction of the dysfunction; analysis at the theoretical level of the neurological complications risk at long term for the exposed patients. The second point of this request is the subject of this report. It gives the synthesis of the whole of information, at the clinical and dosimetric level and outlines successively: the expertise methodology; the cohort of patients treated at the C.R.R.S.; the parameters of the risk analysis of neurological complications; the different risk analysis according the pathologies treated at the C.R.R.S.; the recommendations. (N.C.)

  8. Evaluation of tritiated water retention capacity of fusion reactor concrete building

    International Nuclear Information System (INIS)

    Numata, S.; Fujii, Y.; Okamoto, M.

    1992-01-01

    In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment

  9. Seismic and cask drop excitation evaluation of the tower shielding reactor

    International Nuclear Information System (INIS)

    Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs

  10. Seismic and cask drop excitation evaluation of the Tower Shielding Reactor

    International Nuclear Information System (INIS)

    Stover, R.L.; Harris, S.P.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations

  11. Evaluation of aluminum capsules according to ISO 9978 to irradiation of gaseous samples in nuclear reactor

    International Nuclear Information System (INIS)

    Costa, Osvaldo L. da.; Tiezzi, Rodrigo; Souza, Daiane C.B.; Feher, Anselmo; Moura, Joao A.; Souza, Carla D.; Moura, Eduardo S.; Oliveira, Henrique B.; Zeituni, Carlos A.; Rostelato, Maria Elisa C.M.

    2015-01-01

    Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic press head was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978. (author)

  12. Criticality design evaluation of the White Sands reactor building storage vault

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.

    1979-03-01

    This report describes the conceptual design and criticality evaluation of a storage vault for components of the fast pulse reactor at White Sands Missile Range. Criticality calculations were performed with the KENO-IV Monte Carlo code for various storage configurations in order to investigate the coupling between the portable reactor and storage arrays of spare reactor rings or other fissile components of similar mass. Abnormal conditions corresponding to pseudo--random arrays of the fuel components, as well as a number of flooded configurations, were also evaluated to assess criticality potential for highly unlikely situations. In a normal, dry configuration, the neutron self-multiplication factor, k/sub eff/, of the fully loaded 3 x 8 planar array plus the reactor is less than 0.87. A completely flooded vault was found to produce self-multiplication factors in excess of 1.2

  13. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  14. Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.; Kamal, S.M.

    1994-01-01

    The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concrete shielding. Multiattribute utility theory is selected to accommodate decision maker's preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Illmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy weight heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Illmenite Serpentine concrete. (Author)

  15. Molten-Salt Reactors: Report for 1960 Ten-Year-Plan Evaluation

    International Nuclear Information System (INIS)

    MacPherson, H. G.

    1960-01-01

    For purposes of this evaluation, the molten-salt reactor is considered as an advanced concept. It is considered not to have a status of a current technology adequate to allow the construction of large-scale power plants, since no power reactor has been built or even designed in detail. As a result there can be no estimate of present cost of power, and the projection of power costs to later years is necessarily based on general arguments rather than detailed considerations.

  16. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour

  17. Evaluation of damages of airplane crash in European Advanced Boiling Water Reactor (EU-ABWR)

    International Nuclear Information System (INIS)

    Kamei, Kazuhiro; Tanoue, Tetsuharu; Kataoka, Kazuyoshi; Jimbo, Masakazu

    2011-01-01

    European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash (APC), severe accident mitigation systems, N+2 principle in safety systems and a large output of 1600 MWe. Thanks to above mentioned features, EU-ABWR's design objectives and principles are consistent with safety requirements in an European market. In this paper, evaluation of damages induced by APC has been summarized. (author)

  18. Technical committee meeting on evaluation of radioactive materials release and sodium fires in fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The objectives of the Technical Committee Meeting was to review the activities of research on radioactive materials release and sodium fires in fast reactors in each of the participating countries. It covered: out-of-pile experiments and analysis codes on source term; in-pile experiments on source term; core disruptive accidents; sodium leak experience in liquid metal fast reactors; evaluation of sodium fire; and aerosol behaviour.

  19. Critical technical issues and evaluation and comparison studies for inertial fusion energy reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.; Ying, A.Y.; Tillack, M.S.; Ghoniem, N.M.; Waganer, L.M.; Driemeyer, D.E.; Linford, G.J.; Drake, D.J.

    1994-01-01

    The critical issues, evaluation and comparison of two inertial fusion energy (IFE) reactor design concepts developed in the Prometheus studies are presented. The objectives were (1) to identify and characterize the critical issues and the R and D required to solve them, and (2) to establish a sound basis for future IFE technical and programmatic decisions by evaluating and comparing the different design concepts. Quantitative evaluation and comparison of the two design options have been made with special focus on physics feasibility, engineering feasibility, economics, safety and environment, and research and development (R and D) requirements. Two key conclusions are made based on the overall evaluation analysis: (1) The heavy-ion driven reactors appear to have an overall advantage over laser-driven reactors; and: (2) The differences in scores are not large and future results of R and D could change the overall ranking of the two IFE concepts

  20. Evaluating advanced LMR [liquid metal reactor] reactivity feedbacks using SSC

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.; Cheng, H.S.

    1988-01-01

    Analyses of the PRISM and SAFR Liquid Metal Reactors with SSC are discussed from a safety and licensing perspective. The PRISM and SAFR reactors with metal fuel are designed for inherent shutdown responses to loss-of-flow and loss-of-heat-sink events. The demonstration of this technology was performed by EBR-II during experiments in April 1986 by ANL (Planchon, et al.). Response to postulated TOPs (control rod withdrawal) are made acceptable largely by reducing reactivity swings, and therefore minimizing the size of possible ractivity insertions. Analyses by DOE and the contractors GE, RI, and ANL take credit for several reactivity feedback mechanisms during transient calculations. These feedbacks include Doppler, sodium density, and thermal expansion of the grid plates, the load pads, the fuel (axial) and the control rod which are now factored into the BNL SSC analyses. The bowing feedback mechanism is not presently modeled in the SSC due to its complexity and subsequent large uncertainty. The analysis is conservative by not taking credit for this negative feedback mechanism. Comparisons of BNL predictions with DOE contractors are provided

  1. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24

  2. The comet assay as a dosimetric tool in evaluation of overexposure localized irradiation; El ensayo de cometa como herramienta de la dosimetria biologica en la evaluacion de sobreexposiciones fuertemente localizadas

    Energy Technology Data Exchange (ETDEWEB)

    Giorgio, Marina Di; Taja, Maria R.; Nasazzi, Nora [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina); Bustos, Norma; Cavalieri, Hernan; Bolgiani, Alberto [Fundacion Fortunato Benaim, Buenos Aires (Argentina)

    2001-07-01

    With inhomogeneous exposures, as is characteristic in accidents, the skin may be an important organ in determining clinical prognosis, being dosimetric assessment a necessary requirement. In order to get information to be applied on the evaluation of skin biopsies without culture for an early assessment of irradiation consequences in locally irradiated individuals, contributing with the biophysical techniques, we evaluate the alkaline comet assay (for doses < 5 Gy) as a method for the detection of DNA radiation induced damage in keratinocytes from primary and secondary cultures obtained from medium thickness skin biopsies and epidermal cells, without culture, derived from the same sample of skin biopsies of patients requiring grafts. To extend the dose range (> 5 Gy), neutral comet assay was applied to keratinocytes from primary and secondary cultures and to a suspension of epidermal cells obtained from biopsies irradiated in vitro an afterwards processed to obtain the mentioned cellular suspension, in order to reproduce the closest condition to in vivo overexposure. The correlation of the obtained data with factors of the patient and the corresponding skin graft response, were evaluated. (author)

  3. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  4. Application of expert system to evaluating reactor water cleanup system performance

    International Nuclear Information System (INIS)

    Maeda, Katsuji; Nakamura, Masahiro; Nagasawa, Katsumi; Fushiki, Sumiyuki.

    1991-01-01

    Expert systems employing artificial intelligence (AI) have been developed for finding and elucidating causes of anomalies and malfunctions, presenting pertinent recommendation for countermeasures and for making precautionary diagnosis. On the other hand, further improvements in reliabilities for chemical control are required to promote BWR plant reliability and advancement. Especially, it is necessary to maintain the reactor water purity in high quality to minimize stress corrosion cracking (SCC) in primary cooling system, fuel performance degradation and radiation buildup. The reactor water quality is controlled by the reactor water cleanup (RWCU) system. So, it is very important to maintain the RWCU performance, in order to keep good reactor water quality. This paper describes an expert system used for evaluating RWCU system performance in BWR plants. (author)

  5. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon

    2006-02-01

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation

  6. A comparative evaluation of fuel utilisation by different thermal reactor systems

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1992-10-01

    A comparative assessment of fuel utilization efficiency of pressurised water reactors, pressure tube type pressurised heavy water reactors, pressure vessel type pressurised heavy water reactors and high temperature gas cooled graphite reactors operating on a number of different fuel cycles has been carried out. The fuel utilization efficiency has been defined as the amount of natural uranium consumed for the generation of one unit of electricity averaged over the period covered in the analysis. The comparative evaluation has been done with different projected growth of installed nuclear capacity for a period of 50 years. One of the models used to predict the installed nuclear capacity growth is the Fisher-Pry model. (author). refs., figs., tabs

  7. Investigation of fuel lattice pitch changes influence on reactor performance through evaluate the neutronic parameters

    International Nuclear Information System (INIS)

    Zareian Ronizi, F.; Fadaei, A.H.; Setayeshi, S.; Shahidi, A.R.

    2015-01-01

    Highlights: • One of the most complex issues that Nu-engineers deal with is the design of NR core. • Numerous factors in nuclear core design depend on Fuel-to-Moderator volume ratio. • Aim of this research is to investigate RX performance for different lattice pitches. - Abstract: Nuclear reactor core design is one of the most complex issues that nuclear engineers deal with. The number and complexity of effective parameters and their impact on reactor design, which makes the problem difficult to solve, require precise knowledge of these parameters and their influence on the reactor operation. Numerous factors in a nuclear reactor core design depend on the Fuel-to-Moderator volume ratio, V F /V M , in a fuel cell. This ratio can be modified by changing the lattice pitch which is the thickness of water channels between fuels plates while keeping fuel slab dimensions fixed. Cooling and moderating properties of water are affected by such a change in a reactor core, and hence some parameters related to these properties might be changed. The aim of this research is to provide the suitable knowledge for nuclear core designing. To reach this goal, the first operating core of Tehran Research Reactor (TRR) with different lattice pitches is simulated, and the effect of different lattice pitches on some parameters such as effective multiplication factor (K eff ), reactor life time, distribution of neutron flux and power density in the core, as well as moderator temperature and density coefficient of reactivity are evaluated. The nuclear reactor analysis code, MTR-PC package is employed to carry out the considered calculation. Finally, the results are presented in some tables and graphs that provide useful information for nuclear engineers in the nuclear reactor core design

  8. Response tree evaluation: experimental assessment of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1985-09-01

    The United States Nuclear Regulatory Commission (USNRC) sponsored a project performed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) to evaluate different display concepts for use in nuclear reactor control rooms. Included in this project was the evaluation of the response tree computer-based decision aid and its associated displays. The response tree evaluation task was deisgned to (1) assess the merit of the response tree decision aid and (2) develop a technical basis for recommendations, guidelines, and criteria for the design and evaluation of computerized decision aids for use in reactor control rooms. Two major experiments have been conducted to evaluate the response tree system. This report emphasizes the conduct and results of the second experiment. An enhanced version of the response tree system, known as the automated response tree system, was used in a controlled experiment using trained reactor operators as test subjects. This report discusses the automated response tree system, the design of the evaluation experiment, and the quantitative results of the experiment. The results of the experiment are compared to the results of the previous experiment to provide an integrated perspective of the response tree evaluation project. In addition, a subjective assessment of the results addresses the implications for the use of advanced, ''intelligent'' decision aids in the reactor control room

  9. Neutron spectrometric evaluations in the Argentine research reactor RA-1

    International Nuclear Information System (INIS)

    Kunst, J.J.; Papadopulos, S.B.; Gregori, B.N.; Cruzate, J.A.

    1998-01-01

    Full text: The determination of the quantities dose equivalent H * (10) and personal dose equivalent Hp(10) in mixed field (n,γ) needs the knowledge of the related spectrum. In order to fulfill this aim spectrometer system has been built based on the combination of polyethylene spheres of different diameters (Bonner Spheres System-BSS) and a He 3 proportional counter detector sensitive to thermal neutrons. The detector is located in the geometrical centre of each of the spheres and has an associated electronics with a charge preamplifier, an amplifier and a multichannel system that allows the outgoing spectrum analysis. In order to determine the neutron spectrum a deconvolution method is applied based on the LOUHI82 code. In this work are shown the spectra and the related values of H * (10) that have been got in five places of the reactor and in the command room with the BSS. (author) [es

  10. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1990-12-01

    A one-group cross section data base used by the ORIGEN2 computer code to simulate the depletion, buildup and decay of radionuclides in research reactor was developed. For this, ENDF/B-IV or -V data was processed using the NJOY code system into 69-group data. The burn-up-dependent weighting spectra were calculated with the WIMS-KAERI code, and then the 69-group data were collapsed to one-group using the spectra. The ORIGEN2 depletion calculations for the KMRR fuel were performed using an old PWR and the new data base. By comparing these results to the WIMS-KAERI calculations, it is seen that the results of actinide composition calculated by the ORIGEN2 with the new data base turn out to be in an excellent agreement with the WIMS-KAERI results in the range up to 120 GWD/MTIHM burnup. (Author).

  11. Economical evaluation on spent fuel storage technology away from reactor

    International Nuclear Information System (INIS)

    Itoh, Chihiro; Nagano, Koji; Saegusa, Toshiari

    2000-01-01

    Concerning the spent fuel storage away from reactor, economical comparison was carried out between metal cask and water pool storage technology. The economic index was defined by levelized cost (Unit storage cost) calculated on the assumption that the storage cost is paid at the receipt of the spent fuel at the storage facility. It is found that the cask storage is economical for small and large storage capacity. Unit storage cost of pool storage, however, is getting close to that of cask storage in case of storage capacity of 10,000 ton. Then, the unit storage cost is converted to power generation cost using data of the burn up of the fuel, etc. The cost is obtained as yen 0.09/kWh and yen 0. 15/kWh for cask storage and pool storage, respectively in case of the capacity of 5,000 tonU and the cooling time of 5 years. (author)

  12. Radiographic inspection and densitometric evaluation of CP-5 reactor fuel

    International Nuclear Information System (INIS)

    Staroba, J.F.; Knoerzer, T.W.

    1978-02-01

    This report covers the radiographic and densitometric techniques used as part of a quality verification program for CP-5 reactor fuel by the Nondestructive Assay Section of the Special Materials Division. Other nondestructive tests used were ultrasonic and gamma-ray spectrometry. The main objectives were to perform a one-hundred percent radiographic inspection of the fuel tubes and to derive a quantitative relationship between fuel thickness and film density with the use of fabricated fuel step wedges. By the use of tangential x-ray techniques, measurements were made of fuel peaks or ''hot spots'' that protruded above the main fuel line. Other general problems in radiographic inspection and solutions for the upgrading of the total radiographic inspection program are also discussed

  13. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  14. An evaluation of control rod motion simulator of research reactor

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

  15. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator has been carried out testing of the reactor control rod using a servomotor. Reactor control rod motion at any point should be in the right position, one of the motors that can move in a precise and correct the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servomotor function test for motor work to ensure the performance of the appliance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage Vout nets at 24 V, 6.5 A with 12 Ω load deviation obtained V0 = V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125%, next to the breakdown voltage Vout nets at 12V, 4.2 A with a 6 Ω load deviation obtained V0 = V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on Vout 24 V, 4.5 A with 12 Ω load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%. (author)

  16. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    International Nuclear Information System (INIS)

    Marvin, M.D.

    1978-01-01

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated

  17. Evaluation Of Fire Safety And Protection At PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ahmad Nabil Ab Rahim; Alfred Sanggau Ligam; Nurhayati Ramli; Mohd Fazli Zakaria; Naim Syauqi Hamzah; Phongsakorn Prak; Mohammad Suhaimi Kassim; Zarina Masood

    2014-01-01

    Fire hazard is one of many risks that can affect the safety operation of PUSPATI TRIGA Reactor. Reactor building in Malaysian Nuclear Agency was built in 1980s and the fire system has been introduced since then. The evaluation of the fire safety system at this time is important to ensure the efficiency of fire prevention, fighting and mitigation task that probably occurs. This evaluation involves with the fire fighting system and equipment, integrity of the system from the perspective of management and equipment, fire fighting procedure and fire fighting response team. (author)

  18. Evaluation of research reactor fuel reliability in support of regulatory requirements

    International Nuclear Information System (INIS)

    Sokolov, Eugene N.

    2005-01-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  19. Evaluation of research reactor fuel reliability in support of regulatory requirements

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Eugene N [Chalk River Laboratories, AECL, Chalk River, ON, K0J 1J0 (Canada)

    2005-07-01

    This standards, codes and practices survey is devoted to the problem of reliability of R and D especially research reactor fuel (RRF) performance-related processes. Regulatory R and D evaluations were based on one standard and just few of them provide correlation to other relative standards whereas synthetic process approach reflects actual status of particular R and D practices. Fuel performance regulatory parameters are based on quality standards. A reliability process-based method similar to PSA/FMEA is proposed to evaluate RRF performance- related parameters in terms of reactor safety. (author)

  20. Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

    International Nuclear Information System (INIS)

    Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

    1990-10-01

    This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns

  1. Evaluation of dosimetric variance in whole breast forward-planned intensity-modulated radiotherapy based on 4DCT and 3DCT

    International Nuclear Information System (INIS)

    Wang Wei; Li Jianbin; Hu Hongguang; Sun Tao; Xu Min; Fan Tingyong; Shao Qian

    2013-01-01

    This study was performed to explore and compare the dosimetric variance caused by respiratory movement in the breast during forward-planned intensity-modulated radiotherapy (IMRT) after breast-conserving surgery. A total of 17 enrolled patients underwent the three-dimensional computed tomography (3DCT) simulation scans followed by four-dimensional computed tomography (4DCT) simulation scans during free breathing. The treatment planning constructed using the 3DCT images was copied and applied to the end expiration (EE) and end inspiration (EI) scans and the dose distributions were calculated separately. CTV volume variance amplitude was very small (11.93±28.64 cm 3 ), and the percentage change of CTV volumes receiving 50 Gy and 55 Gy between different scans were all less than 0.8%. There was no statistically significant difference between EI and EE scans (Z=-0.26, P=0.795). However, significant differences were found when comparing the D mean at 3DCT planning with the EI and EE planning (P=0.010 and 0.019, respectively). The homogeneity index at EI, EE and 3D plannings were 0.139, 0.141 and 0.127, respectively, and significant differences existed between 3D and EI, and between 3D and EE (P=0.001 and 0.006, respectively). The conformal index (CI) increased significantly in 3D treatment planning (0.74±0.07) compared with the EI and EE phase plannings (P=0.005 and 0.005, respectively). The V 30 , V 40 , V 50 and D mean of the ipsilateral lung for EE phase planning were significantly lower than for EI (P=0.001-0.042). There were no significant differences in all the dose-volume histogram (DVH) parameters for the heart among these plannings (P=0.128-0.866). The breast deformation during respiration can be disregarded in whole breast IMRT. 3D treatment planning is sufficient for whole breast forward-planned IMRT on the basis of our DVH analysis, but 4D treatment planning, breath-hold, or respiratory gate may ensure precise delivery of radiation dose. (author)

  2. SU-C-210-05: Evaluation of Robustness: Dosimetric Effects of Anatomical Changes During Fractionated Radiation Treatment of Pancreatic Cancer Patients

    Energy Technology Data Exchange (ETDEWEB)

    Horst, A van der; Houweling, A C; Bijveld, M M C; Visser, J; Bel, A [Academic Medical Center, Amsterdam, Noord-Holland (Netherlands)

    2015-06-15

    Purpose: Pancreatic tumors show large interfractional position variations. In addition, changes in gastrointestinal air volume and body contour take place during treatment. We aim to investigate the robustness of the clinical treatment plans by quantifying the dosimetric effects of these anatomical changes. Methods: Calculations were performed for up to now 3 pancreatic cancer patients who had intratumoral fiducials for daily CBCT-based positioning during their 3-week treatment. For each patient, deformable image registration of the planning CT was used to assign Hounsfield Units to each of the 13—15 CBCTs; air volumes and body contour were copied from CBCT. The clinical treatment plan was used (CTV-PTV margin = 10 mm; 36Gy; 10MV; 1 arc VMAT). Fraction dose distributions were calculated and accumulated. The V95% of the clinical target volume (CTV) and planning target volume (PTV) were analyzed, as well as the dose to stomach, duodenum and liver. Dose accumulation was done for patient positioning based on the fiducials (as clinically used) as well as for positioning based on bony anatomy. Results: For all three patients, the V95% of the CTV remained 100%, for both fiducial- and bony anatomy-based positioning. For fiducial-based positioning, dose to duodenum en stomach showed no discernable differences with planned dose. For bony anatomy-based positioning, the PTV V95% of the patient with the largest systematic difference in tumor position (patient 1) decreased to 85%; the liver Dmax increased from 33.5Gy (planned) to 35.5Gy. Conclusion: When using intratumoral fiducials, CTV dose coverage was only mildly affected by the daily anatomical changes. When using bony anatomy for patient positioning, we found a decline in PTV dose coverage due to the interfractional tumor position variations. Photon irradiation treatment plans for pancreatic tumors are robust to variations in body contour and gastrointestinal gas, but the use of fiducial-based daily position verification

  3. Application of the Voxeldose software for dosimetric evaluation on the thyroid during thorax-AP irradiation considering the peak voltages (k Vp) most used in diagnostic X-ray

    International Nuclear Information System (INIS)

    Vieira, I.F.; Vieira, J.W.; Leal Neto, V.

    2009-01-01

    The evaluation of the absorbed dose distribution can be obtained through a computational model of exposures (ECM), being one the main difficulties at the specific dosimetric evaluation such as the radiodiagnostic, coupling the Monte Carlo computer code, developed for general use, to a anthropomorphic model. This problem can be solved by the software used in this paper, the VoxelDose, and it consists of an algorithm for X-ray diagnostic sources with the Monte Carlo EGS4 code coupled to the voxel anthropomorphic phantoms MAX (Male Adult voXel) and FAX (Female Adult voXel). The graphic interface allows the user to insert the mos common exams parameters, and to execute the simulation, obtaining conversion coefficients and the estimative of the deposited energy on organs/tissues radio sensible during the routine procedures. The data obtained were organized into graphics showing the thyroid equivalent dose, which is a radio sensible with 20 g mass and a weight factor of 5 %, compared with the effective dose during an irradiation of thorax-AP

  4. Dosimetric evaluation of spectrophotometric response of alanine gel solution for gamma, photons, electrons and thermal neutrons radiations; Avaliacao dosimetrica da resposta espectrofotometrica da solucao gel de alanina para radiacao gama, de fotons, de eletrons e de neutrons termicos

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Cleber Feijo

    2009-07-01

    Alanine Gel Dosimeter is a new gel material developed at IPEN that presents significant improvement on Alanine system developed by Costa. The DL-Alanine (C{sub 3}H{sub 7}NO{sub 2}) is an amino acid tissue equivalent that improves the production of ferric ions in the solution. This work aims to analyse the main dosimetric characteristics this new gel material for future application to measure dose distribution. The performance of Alanine gel solution was evaluated to gamma, photons, electrons and thermal neutrons radiations using the spectrophotometry technique. According to the obtained results for the different studied radiation types, the reproducibility intra-batches and inter-batches is better than 4% and 5%, respectively. The dose response presents a linear behavior in the studied dose range. The response dependence as a function of dose rate and incident energy is better 2% and 3%, respectively. The lower detectable dose is 0.1 Gy. The obtained results indicate that the Alanine gel dosimeter presents good performance and can be useful as an alternative dosimeter in the radiotherapy area, using MRI technique for tridimensional dose distribution evaluation. (author)

  5. Laminated dosimetric card

    International Nuclear Information System (INIS)

    Cox, F.M.; Chamberlain, J.D.; Shrader, E.F.; Shoffner, B.M.; Szalanczy, A.

    1975-01-01

    A laminated card with one or more apertures, each adapted to peripherally seal an encapsulated dosimeter, is formed by bonding a foraminous, code-adaptable, rigid sheet of low-Z material with a codedly transparent sheet of low-Z material in light-transmitting registry with particular code-holes of the rigid sheet. The laminated card may be coded to identify the person carrying it, and/or the location or circumstances related to its exposure to radiation. This card is particularly adapted for use in an instrument capable of evaluating a multiplicity of cards, substantially continuously. The coded identification from the card may be displayed by an appropriate machine, and if desired an evaluation may be recorded because of a ''parity checking'' system incorporated in each card, which permits ''auto-correction.'' Alternatively, where means for effecting the correction automatically are available, the operation of the machine may be interrupted to permit visual examination of a rejected card. The card of this invention is also coded for identifying the type of card with respect to its specific function, and whether or not a card is correctly positioned at any predetermined location during its sequential progress through the instrument in which it is evaluated. Dosimeters are evaluated and the card identified in one pass through the instrument. (auth)

  6. Evaluation of non radiation dangerous in multipurpose reactor GAS

    International Nuclear Information System (INIS)

    Suwarto, S.

    1998-01-01

    Evaluation of the potential non irradiation dangerous in RSG-GAS included of : the fire dangerous, the chemical hazard and gas dangerous have been performed to evaluate its potential on causing the accident and evaluate the performance of the equipment to protect the accident. Evaluate the performance of the equipment to protect to accident . Evaluate to the fire dangerous performed by identified the potential dangerous of fire at each rooms and evaluate the performance of each equipment included of dry powder fire extinguishing system, hydrant system, fire detectors and alarm system. Evaluation to the chemical hazard and gas dangerous performed by identified the number and it's the management of chemical hazard in the chemical storage and laboratory. The result of evaluation included of data of the fire dangerous potential class and the performance of its equipment in each room in RSG-GAS and the data of the number and the management system of the chemical hazard and gas in chemical storage and chemical laboratory. From this evaluation it is concluded that the equipment of fire system are available to protect against the accident and the chemical hazard and gas potential are relating small, and has been managed properly

  7. Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

  8. Burnup influence on the WWER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of WWER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in ? depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (Authors)

  9. Evaluation of fuel performance for fresh and aged CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

  10. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    Allen, T.R.; Was, G.S.

    2008-01-01

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept. The objective of the proposed research was to investigate degradation of materials in the supercritical water environment (SCW). First, representative alloys from the important classes of candidate materials were studied for their corrosion and stress-corrosion cracking (SCC) resistance in supercritical water. These included ferritic/martensitic (F/M) steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests were conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies were applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and SCC of alloys in the as-received and modified/engineered conditions were examined by irradiating samples using high-energy protons and then exposing them to SCW

  11. Evaluation of A-1 reactor heavy-water calandria specimens

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1976-01-01

    Container chains with surveillance specimens were placed in two special channels of the core peripheral part to test changes in mechanical properties due to reactor operation of caisson tube material. The specimens were made from the caisson tube material and placed by eight pieces on the outer surface of the containers. The first removed specimens were tested for corrosion losses, tensile strength, and fractured surfaces were then assessed. The changes in strength properties were found to be similar in both base material and welded joints. The corrosion film on surveillance specimens did not practically affect strength properties nor ductility. It was found that the Al-Mg-Si alloy used for the heavy water vessel caisson tubes following stabilization annealing was fully stable at operating temperatures of up to 100 degC. Slio.ht changes in properties can be attributed to the effect of a high neutron dose. Thus, the high radiation and temperature stability of the alloy was confirmed. (O.K.)

  12. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor

    International Nuclear Information System (INIS)

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-01-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. - Highlights: • Dynamic tritium concentration in the air and human body evaluated on the time scale. • Different intake forms and relevant radiation dose assessed to find out the potential relationships. • HT and HTO release amount limits for arbitrary accidents proposed for the fusion reactor according to dose limit

  13. Dosimetric lung models

    International Nuclear Information System (INIS)

    James, A.C.; Roy, M.

    1986-01-01

    The anatomical and physiological factors that vary with age and influence the deposition of airborne radionuclides in the lung are reviewed. The efficiency with which aerosols deposit in the lung for a given exposure at various ages from birth to adulthood is evaluated. Deposition within the lung is considered in relation to the clearance mechanisms acting in different regions or compartments. The procedure for evaluating dose to sensitive tissues in lung and transfer to other organs that is being considered by the Task Group established by ICRP to review the Lung Model is outlined. Examples of the application of this modelling procedure to evaluate lung dose as a function of age are given, for exposure to radon daughters in dwellings, and for exposure to an insoluble 239 Pu aerosol. The former represents exposure to short-lived radionuclides that deliver relatively high doses to bronchial tissue. In this case, dose rates are marginally higher in children than in adults. Plutonium exposure represents the case where dose is predominantly delivered to respiratory tissue and lymph nodes. In this case, the life-time doses tend to be lower for exposure in childhood. Some of the uncertainties in this modelling procedure are noted

  14. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

    International Nuclear Information System (INIS)

    2017-01-01

    The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

  15. A new method for evaluation and correction of thermal reactor power and present operational applications

    International Nuclear Information System (INIS)

    Langenstein, M.; Streit, S.; Laipple, B.; Eitschberger, H.

    2005-01-01

    The determination of the thermal reactor power is traditionally be done by heat balance: 1) for a boiling water reactor (BWR) at the interface of reactor control volume and heat cycle. 2) for a pressurised-water reactor (PWR) at the interface of the steam generator control volume and turbine island on the secondary side. The uncertainty of these traditional methods is not easy to determine and can be in the range of several percent. Technical and legal regulations (e.g. 10CFR50) cover an estimated error of instrumentation up to 2% by increasing the design thermal reactor power for emergency analysis to 102 % of the licensed thermal reactor power. Basically the licensee has the duty to warrant at any time operation inside the analyzed region for thermal reactor power. This is normally done by keeping the indicated reactor power at the licensed 100% value. The better way is to use a method which allows a continuous warranty evaluation. The quantification of the level of fulfilment of this warranty is only achievable by a method which: 1) is independent of single measurements accuracies. 2) results in a certified quality of single process values and for the total heat cycle analysis. 3)leads to complete results including 2-sigma deviation especially for thermal reactor power. Here this method, which is called 'process data reconciliation based on VDI 2048 guideline', is presented [1, 2]. This method allows to determine the true process parameters with a statistical probability of 95%, by considering closed material, mass- and energy balances following the Gaussian correction principle. The amount of redundant process information and complexity of the process improves the final results. This represents the most probable state of the process with minimized uncertainty according to VDI 2048. Hence, calibration and control of the thermal reactor power are possible with low effort but high accuracy and independent of single measurement accuracies. Further more, VDI 2048

  16. Measurement and evaluation of the external radiation level at reactor Kartini

    International Nuclear Information System (INIS)

    Atok Suhartanto; Suparno

    2013-01-01

    Measurement and evaluation of external radiation level at reactor Kartini in 2012 has been done. The purpose of this activity is to know the external radiation level as a result of the radioactive or radiation source usage, toward the operational of limit condition. The measurement is using survey meter Inspector 11086, factor of calibration 0.991 mR/h, at 9 locations is: Control room area, Thermal column facilities, Demineralizer, Beamport radiography facilities, bulk shielding Deck, Subcritical facilities, Reactor hall, Deck reactor and on the surface of reactor water tank . The highest room average measurement result in 9 working areas for 12 months continuously are at the reactor tank location is between 13.05±1.09 (xlO -2 mSv/hour) to 16.80±1.40 (x10 -2 mSv/hour), and the lowest measurement result in 1 location (control room) is 0.02±0.005 (x10 -2 mSv/hour) to 0.035±0.009 (x10 -2 mSv/hour). The Kartini reactor is involved in the control area which has potentially contaminated and has radiation exposure at the level of 6 mSv/year. Radiation Protection Officer that work in interval will received radiation exposure dosage of 8.4 mSv/year. This dosage is still below the Below Dosage Value which is recommended by, BAPETEN decree No, 4, 2013 about Protection and Radiation Safety in Nuclear Energy Application at 20 mSv/year. The result of the evaluation above shows that the external radiation which occurred in each area is still below the operational of limit condition that is written on the Kartini reactor safety analysis report, on document number: C7/05/B2/LAK/2010, revision 7. So that the workplace is safe for work monitored. (author)

  17. Evaluation of prestress losses in nuclear reactor containments

    International Nuclear Information System (INIS)

    Lundqvist, Peter; Nilsson, Lars-Olof

    2011-01-01

    Research highlights: → Prestress losses in reactor containments were estimated using prediction models. → The predicted prestress losses were compared to long-term measurements. → The accuracy of the models was improved by considering actual drying conditions. → Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage

  18. Evaluation of prestress losses in nuclear reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Lundqvist, Peter, E-mail: peter.lundqvist@kstr.lth.s [Div. of Structural Engineering, Lund University, Lund (Sweden); Nilsson, Lars-Olof [Div. of Building Materials, Lund University, Lund (Sweden)

    2011-01-15

    Research highlights: Prestress losses in reactor containments were estimated using prediction models. The predicted prestress losses were compared to long-term measurements. The accuracy of the models was improved by considering actual drying conditions. Predictions by CEB/FIP MC 1999 and ACI 209 were closest to the measured losses. - Abstract: The most critical safety barrier in a nuclear power plant, the concrete containment, is prestressed by hundreds of tendons, both horizontally and vertically. The main purpose of the containment is to prevent radioactive discharge to the environment in the case of a serious internal accident. Due to creep and shrinkage of concrete and relaxation of the prestressing steel, tendon forces decrease with time. These forces are thus measured in Swedish containments with unbonded tendons at regular in-service inspections. In this paper, the prestress losses obtained from these in-service inspections are compared to losses estimated using several prediction models for creep, shrinkage and relaxation. In an attempt to increase the accuracy of these models, existing expressions for the development of shrinkage were modified using previous findings on the humidity and temperature inside two Swedish containments. The models which were used and modified for predicting creep and shrinkage were CEB-FIP Model Codes 1990 and 1999, ACI 209, Model B3 and GL2000. Eurocode 2 was used for the prediction of relaxation. The results show that the most accurate of the models were CEB/FIP MC 99 and ACI 209. Depending on the model, the accuracy of the prediction models was increased by 0.5-1.2 percentage points of prestress losses when using the modified development of shrinkage. Furthermore, it was found that the differences between the different models depend mainly on the prediction of creep. Possible explanations for the deviation between the calculated and measured models can be the influence of reinforcement on creep and shrinkage of concrete and

  19. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C.; Palma, Daniel A.P.

    2015-01-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  20. Flaw evaluation of pressure vessel in pressurized water reactor

    International Nuclear Information System (INIS)

    Park, Ki Sung; Kim, Min Geol; Jeon, Chae Hong; Rhim, Soon Hyung; Kim, Seung Tae

    1999-01-01

    Flaw evaluation should be performed to determine the acceptance of a surface or a subsurface flaw detected during the in-service inspection without any repair or replacement. In this paper, the evaluation methodology and procedure were established according to ASME code Sec. XI and the evaluation program was coded. Using this program, a field engineer who doesn't have enough knowledge on fracture mechanics may be able to perform prompt and accurate flaw evaluation on site and decide whether a detected flaw be allowable or not. Analysis results were compared with those obtained from Westinghouse program called KCAL and FCG. Both results made good agreement and accuracy of the program developed in this paper was verified.=20

  1. Contura Multi-Lumen Balloon Breast Brachytherapy Catheter: Comparative Dosimetric Findings of a Phase 4 Trial

    Energy Technology Data Exchange (ETDEWEB)

    Arthur, Douglas W., E-mail: darthur@mcvh-vcu.edu [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia (United States); Vicini, Frank A. [Michigan Healthcare Professionals/21st Century Oncology, Farmington Hills, Michigan (United States); Todor, Dorin A. [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia (United States); Julian, Thomas B. [Allegheny General Hospital, Temple University School of Medicine, Pittsburgh, Pennsylvania (United States); Cuttino, Laurie W.; Mukhopadhyay, Nitai D. [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia (United States)

    2013-06-01

    Purpose: Final dosimetric findings of a completed, multi-institutional phase 4 registry trial using the Contura Multi-Lumen Balloon (MLB) breast brachytherapy catheter to deliver accelerated partial breast irradiation (APBI) in patients with early-stage breast cancer are presented. Methods and Materials: Three dosimetric plans with identical target coverage were generated for each patient for comparison: multilumen multidwell (MLMD); central-lumen multidwell (CLMD); and central-lumen single-dwell (CLSD) loading of the Contura catheter. For this study, a successful treatment plan achieved ideal dosimetric goals and included the following: ≥95% of the prescribed dose (PD) covering ≥95% of the target volume (TV); maximum skin dose ≤125% of the PD; maximum rib dose ≤145% of the PD; and V150 ≤50 cc and V200 ≤10 cc. Results: Between January 2008 and February 2011, 23 institutions participated. A total of 318 patients were available for dosimetric review. Using the Contura MLB, all dosimetric criteria were met in 78.93% of cases planned with MLMD versus 55.38% with the CLMD versus 37.66% with the CLSD (P≤.0001). Evaluating all patients with the full range of skin to balloon distance represented, median maximum skin dose was reduced by 12% and median maximum rib dose by 13.9% when using MLMD-based dosimetric plans compared to CLSD. The dosimetric benefit of MLMD was further demonstrated in the subgroup of patients where skin thickness was <5 mm, where MLMD use allowed a 38% reduction in median maximum skin dose over CLSD. For patients with rib distance <5 mm, the median maximum rib dose reduction was 27%. Conclusions: Use of the Contura MLB catheter produced statistically significant improvements in dosimetric capabilities between CLSD and CLMD treatments. This device approach demonstrates the ability not only to overcome the barriers of limited skin thickness and close rib proximity, but to consistently achieve a higher standard of dosimetric planning goals.

  2. Contura Multi-Lumen Balloon breast brachytherapy catheter: comparative dosimetric findings of a phase 4 trial.

    Science.gov (United States)

    Arthur, Douglas W; Vicini, Frank A; Todor, Dorin A; Julian, Thomas B; Cuttino, Laurie W; Mukhopadhyay, Nitai D

    2013-06-01

    Final dosimetric findings of a completed, multi-institutional phase 4 registry trial using the Contura Multi-Lumen Balloon (MLB) breast brachytherapy catheter to deliver accelerated partial breast irradiation (APBI) in patients with early-stage breast cancer are presented. Three dosimetric plans with identical target coverage were generated for each patient for comparison: multilumen multidwell (MLMD); central-lumen multidwell (CLMD); and central-lumen single-dwell (CLSD) loading of the Contura catheter. For this study, a successful treatment plan achieved ideal dosimetric goals and included the following: ≥ 95% of the prescribed dose (PD) covering ≥ 95% of the target volume (TV); maximum skin dose ≤ 125% of the PD; maximum rib dose ≤ 145% of the PD; and V150 ≤50 cc and V200 ≤ 10 cc. Between January 2008 and February 2011, 23 institutions participated. A total of 318 patients were available for dosimetric review. Using the Contura MLB, all dosimetric criteria were met in 78.93% of cases planned with MLMD versus 55.38% with the CLMD versus 37.66% with the CLSD (P ≤.0001). Evaluating all patients with the full range of skin to balloon distance represented, median maximum skin dose was reduced by 12% and median maximum rib dose by 13.9% when using MLMD-based dosimetric plans compared to CLSD. The dosimetric benefit of MLMD was further demonstrated in the subgroup of patients where skin thickness was <5 mm, where MLMD use allowed a 38% reduction in median maximum skin dose over CLSD. For patients with rib distance <5 mm, the median maximum rib dose reduction was 27%. Use of the Contura MLB catheter produced statistically significant improvements in dosimetric capabilities between CLSD and CLMD treatments. This device approach demonstrates the ability not only to overcome the barriers of limited skin thickness and close rib proximity, but to consistently achieve a higher standard of dosimetric planning goals. Copyright © 2013 Elsevier Inc. All rights

  3. Helical Tomotherapy-Based STAT Stereotactic Body Radiation Therapy: Dosimetric Evaluation for a Real-Time SBRT Treatment Planning and Delivery Program

    International Nuclear Information System (INIS)

    Dunlap, Neal; McIntosh, Alyson; Sheng Ke; Yang Wensha; Turner, Benton; Shoushtari, Asal; Sheehan, Jason; Jones, David R.; Lu Weigo; Ruchala, Keneth; Olivera, Gustavo; Parnell, Donald; Larner, James L.; Benedict, Stanley H.; Read, Paul W.

    2010-01-01

    iterations or 135 sec for STAT RT liver and lung SBRT plans and 7 iterations or 315 sec for STAT RT spine SBRT plans. Helical TomoTherapy-based STAT RT treatment planning with the 'full scatter' algorithm provides levels of dosimetric conformality, heterogeneity, and OAR avoidance for SBRT treatments that are clinically equivalent to those generated with the Helical TomoTherapy 'beamlet' algorithm. STAT RT calculation times for simple SBRT treatments are fast enough to warrant further investigation into their potential incorporation into an SBRT program with daily real-time planning. Development of methods for accurate target and OAR determination on megavoltage computed tomography scans incorporating high-resolution diagnostic image co-registration software and CT detector-based exit dose measurement for quality assurance are necessary to build a real-time SBRT planning and delivery program.

  4. Validation and dosimetric evaluation employing the techniques of TL and OSL of thermoluminescent materials for application in the dosimetry of clinical beams of electrons used in total irradiation of the skin - TSI

    International Nuclear Information System (INIS)

    Almeida, Shirlane Barbosa de

    2017-01-01

    In vivo dosimetry has become an important role for the treatment of total skin irradiation within a rigorous quality assurance program that should be an integral part of the radiotherapy departments. TSI dosimetry is difficult because of the complexity of the treatment in assessing dose uniformity and measuring the dose absorbed at shallow depths throughout the skin surface extent, resulting in a wide variation in dose distribution. The TLDs have proven to be very useful for the distribution and verification of the dose prescribed for the patient as the dose may differ from place to place due to patient body geometry, overlapping of structures and asymmetries of the radiation field. The use of TLDs in vivo can identify variations in the prescribed dose because its measurement accuracy and great precision. Several types of dosimeters have been used in the radiotherapy sectors, the most commonly used are Lithium Fluride (TLD-100), where it obtains a long history in this type of application. New dosimetric materials have gained great importance in the dosimetry of clinical electron beams, such as Dysprosium-doped Calcium Sulphate (TL) and Carbon doped (OSL) based Aluminum Oxide, This work evaluates the performance of the respective thermoluminescent dosimeters and the optically stimulated luminescence in the dosimetry of clinical electron beams used in total irradiation of the skin. (author)

  5. Evaluation of upper limit of accident probability in a nuclear reactor in Brazil

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1979-01-01

    This work calls attention to the great probability of accident in a pessimist vision regarding optimist one. The author uses the upper limit presented in Ford Foundation report and applies it on brazilian case to an evaluation of risk of reactor accident in Brazil. (C.M.)

  6. Solar membrane natural gas steam-reforming process: evaluation of reactor performance

    NARCIS (Netherlands)

    de Falco, M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  7. Solar membrane natural gas steam-reforming process : evaluation of reactor performance

    NARCIS (Netherlands)

    Falco, de M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  8. An evaluation of a mesophilic reactor for treating wastewater from a ...

    African Journals Online (AJOL)

    An evaluation of anaerobic treatment of potato-processing wastewater using an up flow Anaerobic Sludge Bed (UASB) reactor at 37°C was conducted. Wastewater from a potato-processing plant in Harare, with an average of 6.8 g COD/l, (COD = chemical oxygen demand) a high concentration of total solids (up to 6725 ...

  9. Evaluation of a nonevaporable getter pump for tritium handling in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Singleton, M.F.; Griffith, C.M.

    1978-01-01

    Lawrence Livermore Laboratory has tested and evaluated a commercially available getter pump for use with tritium in the Tokamak Fusion Test Reactor (TFTR). The pump contains Zr(84%)--Al in cartridge form with a concentric heating unit. It performed well in all tests, except for frequent heater failures

  10. Critical technical issues and evaluation and comparison studies for inertial fusion energy reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Ying, A.Y. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Tillack, M.S. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Ghoniem, N.M. (Mechanical, Aerospace and Nuclear Engineering Dept., Univ. of California, Los Angeles, CA (United States)); Waganer, L.M. (McDonnell Douglas Aerospace, St. Louis, MI (United States)); Driemeyer, D.E. (McDonnell Douglas Aerospace, St. Louis, MI (United States)); Linford, G.J. (TRW Space and Electronics Div., Redondo Beach, CA (United States)); Drake, D.J.

    1994-01-01

    Two inertial fusion energy (IFE) reactor design concepts developed in the Prometheus studies were evaluated. Objectives were to identify and characterize critical issues and the R and D required to resolve them, and to establish a sound basis for future IFE technical and programmatic decisions. Each critical issue contains several key physics and engineering issues associated with major reactor components and impacts key aspects of feasibility, safety, and economic potential of IFE reactors. Generic critical issues center around: demonstration of moderate gain at low driver energy, feasibility of direct drive targets, feasibility of indirect drive targets for heavy ions, feasibility of indirect drive targets for lasers, cost reduction strategies for heavy ion drivers, demonstration of higher overall laser driver efficiency, tritium self-sufficiency in IFE reactors, cavity clearing at IFE pulse repetition rates, performance/reliability/lifetime of final laser optics, viability of liquid metal film for first wall protection, fabricability/reliability/lifetime of SiC composite structures, validation of radiation shielding requirements, design tools, and nuclear data, reliability and lifetime of laser and heavy ion drivers, demonstration of large-scale non-linear optical laser driver architecture, demonstration of cost effective KrF amplifiers, and demonstration of low cost, high volume target production techniques. Quantitative evaluation and comparison of the two design options have been made with special focus on physics feasibility, engineering feasibility, economics, safety and environment, and research and development (R and D) requirements. Two key conclusions are made based on the overall evaluation analysis. The heavy-ion driven reactors appear to have an overall advantage over laser-driven reactors.

  11. Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2011-01-01

    Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

  12. Dosimetric methodology of the ICRP

    International Nuclear Information System (INIS)

    Eckerman, K.F.

    1994-01-01

    Establishment of guidance for the protection of workers and members of the public from radiation exposures necessitates estimation of the radiation dose to tissues of the body at risk. The dosimetric methodology formulated by the International Commission on Radiological Protection (ICRP) is intended to be responsive to this need. While developed for radiation protection, elements of the methodology are often applied in addressing other radiation issues; e.g., risk assessment. This chapter provides an overview of the methodology, discusses its recent extension to age-dependent considerations, and illustrates specific aspects of the methodology through a number of numerical examples

  13. Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

    International Nuclear Information System (INIS)

    Stevens, P.F.

    1994-01-01

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment

  14. Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

  15. Evaluation of activation detectors for the SPHINX project at the LR-0 experimental reactor

    International Nuclear Information System (INIS)

    Lahodova, Zdena; Viererbl, Ladislav; Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech

    2008-01-01

    This article summarizes the measurements of neutron fluence distributions carried out at the LR-0 research reactor (Czech Republic) in the frame of the SPHINX project. The influence of fluoride-salts or graphite filling in the SR-0 modules on neutron spectrum was studied using activation detectors. The activation detectors (Mn, Ni, In and Au) were evaluated to determine the changes in neutron field. The In and Au detectors were also irradiated with a cadmium cover. Five different configurations of reactor core (EROS) were realized. (authors)

  16. Evaluation of activation detectors for the SPHINX project at the LR-0 experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, Zdena; Viererbl, Ladislav [Research Center Rez Ltd (Czech Republic); Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech [Nuclear Power and Safety Division, Nuclear Research Institute Rez plc (Czech Republic)

    2008-07-01

    This article summarizes the measurements of neutron fluence distributions carried out at the LR-0 research reactor (Czech Republic) in the frame of the SPHINX project. The influence of fluoride-salts or graphite filling in the SR-0 modules on neutron spectrum was studied using activation detectors. The activation detectors (Mn, Ni, In and Au) were evaluated to determine the changes in neutron field. The In and Au detectors were also irradiated with a cadmium cover. Five different configurations of reactor core (EROS) were realized. (authors)

  17. Evaluation of thermal ratcheting of reactor vessel wall near the sodium surface

    International Nuclear Information System (INIS)

    Take, Kohji; Fujioka, Terutaka; Yano, Kazutaka

    1989-01-01

    Plastic ratcheting of reactor vessels may occur by an axially moving thermal gradient without primary stress. So there is a need to establish a proper prediction method for the plastic ratcheting. In this study, inelastic FEM analyses of reactor vessel model by using an advanced constitutive equation were carried out in order to comprehend plastic ratcheting behaviour of cylinder which subject to an axially moving thermal gradient. As a result of analyses, a basic mechanism of this ratcheting was found. And it also indicated that cyclic hardening behaviour will became important for development of evaluation method. (author)

  18. DEVELOPMENT OF HUMAN FACTORS ENGINEERING GUIDANCE FOR SAFETY EVALUATIONS OF ADVANCED REACTORS

    International Nuclear Information System (INIS)

    O'HARA, J.; PERSENSKY, J.; SZABO, A.

    2006-01-01

    Advanced reactors are expected to be based on a concept of operations that is different from what is currently used in today's reactors. Therefore, regulatory staff may need new tools, developed from the best available technical bases, to support licensing evaluations. The areas in which new review guidance may be needed and the efforts underway to address the needs will be discussed. Our preliminary results focus on some of the technical issues to be addressed in three areas for which new guidance may be developed: automation and control, operations under degraded conditions, and new human factors engineering methods and tools

  19. Evaluation of flow-induced vibration prediction techniques for in-reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.; Turula, P.

    1975-05-01

    Selected in-reactor components of a hydraulic and structural dynamic scale model of the U. S. Energy Research and Development Administration experimental Fast Test Reactor have been studied in an effort to develop and evaluate techniques for predicting vibration behavior of elastic structures exposed to a moving fluid. Existing analysis methods are used to compute the natural frequencies and modal shapes of submerged beam and shell type components. Component response is calculated, assuming as fluid forcing mechanisms both vortex shedding and random excitations characterized by the available hydraulic data. The free and force vibration response predictions are compared with extensive model flow and shaker test data. (U.S.)

  20. Fast neutron fluence evaluation of the smart reactor pressure vessel by using the GEOSHIELD code

    International Nuclear Information System (INIS)

    Kim, K.Y.; Kim, K.S.; Kim, H.Y.; Lee, C.C.; Zee, S.Q.

    2007-01-01

    In Korea, the design of an advanced integral reactor system called SMART has been developed by KAERI to supply energy for seawater desalination as well as an electricity generation. A fast neutron fluence distribution at the SMART reactor pressure vessel was evaluated to confirm the integrity of the vessel by using the GEOSHIELD code. The GEOSHIELD code was developed by KAERI in order to prepare an input list including a geometry modeling of the DORT code and to process results from the DORT code output list. Results by a GEOSHIELD code processing and by a manual processing of the DORT show a good agreement. (author)

  1. Experimental evaluation of methane dry reforming process on a membrane reactor to hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fabiano S.A.; Benachour, Mohand; Abreu, Cesar A.M. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. of Chemical Engineering], Email: f.aruda@yahoo.com.br

    2010-07-01

    In a fixed bed membrane reactor evaluations of methane-carbon dioxide reforming over a Ni/{gamma}- Al{sub 2}O{sub 3} catalyst were performed at 773 K, 823 K and 873 K. A to convert natural gas into syngas a fixed-bed reactor associate with a selective membrane was employed, where the operating procedures allowed to shift the chemical equilibrium of the reaction in the direction of the products of the process. Operations under hydrogen permeation, at 873 K, promoted the increase of methane conversion, circa 83%, and doubled the yield of hydrogen production, when compared with operations where no hydrogen permeation occurred. (author)

  2. Benchmark test of evaluated nuclear data files for fast reactor neutronics application

    International Nuclear Information System (INIS)

    Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

    2007-07-01

    A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

  3. Fracture mechanics and fatigue evaluation of nuclear reactor components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Andrade, Arnaldo H.P. de; Maneschy, Eduardo

    1995-01-01

    This paper presents a theoretical study available in the available literature for evaluation the environmental effects on the lifetime of nuclear power plant components. The author's motivation is to provide some technical tools to identify what research development could be done in this area

  4. Economic evaluation of small modular nuclear reactors and the complications of regulatory fee structures

    International Nuclear Information System (INIS)

    Vegel, Benjamin; Quinn, Jason C.

    2017-01-01

    Carbon emission concerns and volatility in fossil fuel resources have renewed world-wide interest in nuclear energy as a solution to growing energy demands. Several large nuclear reactors are currently under construction in the United States, representing the first new construction in over 30 years. Small Modular Reactors (SMRs) have been in design for many years and offer potential technical and economic advantages compared with traditionally larger reactors. Current SMR capital and operational expenses have a wide range of uncertainty. This work evaluates the potential for SMRs in the US, develops a robust techno-economic assessment of SMRs, and leverages the model to evaluate US regulatory fees structures. Modeling includes capital expenses of a factory facility and capital and operational expenses with multiple scenarios explored through a component-level capital cost model. Policy regarding the licensing and regulation of SMRs is under development with proposed annual US regulatory fees evaluated through the developed techno-economic model. Results show regulatory fees are a potential barrier to the economic viability of SMRs with an alternate fee structure proposed and evaluated. The proposed fee structure is based on the re-distribution of fees for all nuclear reactors under a single structure based on reactor thermal power rating. - Highlights: • Potential demand for new small modular nuclear power in the US is established. • Capital costs are broken down on component level and include factory production. • US regulatory fees structures are evaluated, results show potential barrier. • An additional fee structure is proposed and compared with current US fee structures.

  5. Flaw evaluation of thermally aged cast stainless steel in light-water reactor applications

    International Nuclear Information System (INIS)

    Lee, S.; Kuo, P.T.; Wichman, K.; Chopra, O.

    1997-01-01

    Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is 290 o C (550 o F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation. (Author)

  6. External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

    International Nuclear Information System (INIS)

    Mihalina, M.; Spalj, S.; Glaser, B.

    2016-01-01

    The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

  7. Evaluation of transmutation performance of long-lived fission products with a super fast reactor

    International Nuclear Information System (INIS)

    Lu, Haoliang; Han, Chiyoung; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    The performance of the Super Fast Reactor for transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super Fast Reactor. First is in the blanket assembly due to the ZrH 1.7 layer which can slow down the fast neutrons. Second is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected of transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe·y and 2.79%/GWe.y can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the outputs from 11.8 and 6.2 1000MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained by the Super Fast Reactor. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super Fast Reactor. It turns out that the 135 Cs transmutation is a challenge not only for the Super Fast Reactor but also for other commercial fast reactors. (author)

  8. Pre evaluation for heat balance of prototype sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

    2012-01-01

    Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

  9. Aspects of the Iea-R1 research reactor seismic evaluation

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    1996-01-01

    Codes and standards for the seismic evaluation of the research reactor IEA-R1 are presented. An approach to define the design basis earthquake based on the local seismic map and on simplified analysis methods is proposed. The site seismic evaluation indicates that the design earthquake intensity is IV MM. Therefore, according to the used codes and standards, no buildings, systems, and components seismic analysis are required. (author)

  10. Proceedings of the Second international Workshop on Nuclear Data Evaluation for Reactor applications (Wonder 2009)

    International Nuclear Information System (INIS)

    2009-01-01

    The NEA (Nuclear Energy Agency) has collaborated with the CEA in the organization of the second international workshop on nuclear data evaluation for reactor applications: Wonder 2009. About 50 scientists have participated to the workshop and 38 presentations have been made, they have been organized around 4 sessions: 1) nuclear data measurements, 2) theory, modeling and evaluation of nuclear data, 3) uncertainties and covariance matrices, and 4) processing and validation of nuclear data

  11. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  12. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs

  13. Intensity modulated radiotherapy and 3D conformal radiotherapy for whole breast irradiation: a comparative dosimetric study and introduction of a novel qualitative index for plan evaluation, the normal tissue index

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Jackie; Suttie, Clare; Bromley, Regina; Morgia, Marita; Lamoury, Gillian [Department of Radiation Oncology, Royal North Shore Hospital, St Leonards, New South Wales (Australia)

    2015-09-15

    We report on a retrospective dosimetric study, comparing 3D conformal radiotherapy (3DCRT) and hybrid intensity modulated radiotherapy (hIMRT). We evaluated plans based on their planning target volume coverage, dose homogeneity, dose to organs at risk (OARs) and exposure of normal tissue to radiation. The Homogeneity Index (HI) was used to assess the dose homogeneity in the target region, and we describe a new index, the normal tissue index (NTI), to assess the dose in the normal tissue inside the tangent treatment portal. Plans were generated for 25 early-stage breast cancer patients, using a hIMRT technique. These were compared with the 3DCRT plans of the treatment previously received by the patients. Plan quality was evaluated using the HI, NTI and dose to OARs. The hIMRT technique was significantly more homogenous than the 3DCRT technique, while maintaining target coverage. The hIMRT technique was also superior at minimising the amount of tissue receiving D{sub 105%} and above (P < 0.0001). The ipsilateral lung and contralateral breast maximum were significantly lower in the hIMRT plans (P < 0.05 and P < 0.005), but the 3DCRT technique achieved a lower mean heart dose in left-sided breast cancer patients (P < 0.05). Hybrid intensity modulated radiotherapy plans achieved improved dose homogeneity compared to the 3DCRT plans and superior outcome with regard to dose to normal tissues. We propose that the addition of both HI and NTI in evaluating the quality of intensity modulated radiotherapy (IMRT) breast plans provides clinically relevant comparators which more accurately reflect the new paradigm of treatment goals and outcomes in the era of breast IMRT.

  14. Intensity modulated radiotherapy and 3D conformal radiotherapy for whole breast irradiation: a comparative dosimetric study and introduction of a novel qualitative index for plan evaluation, the normal tissue index

    International Nuclear Information System (INIS)

    Yim, Jackie; Suttie, Clare; Bromley, Regina; Morgia, Marita; Lamoury, Gillian

    2015-01-01

    We report on a retrospective dosimetric study, comparing 3D conformal radiotherapy (3DCRT) and hybrid intensity modulated radiotherapy (hIMRT). We evaluated plans based on their planning target volume coverage, dose homogeneity, dose to organs at risk (OARs) and exposure of normal tissue to radiation. The Homogeneity Index (HI) was used to assess the dose homogeneity in the target region, and we describe a new index, the normal tissue index (NTI), to assess the dose in the normal tissue inside the tangent treatment portal. Plans were generated for 25 early-stage breast cancer patients, using a hIMRT technique. These were compared with the 3DCRT plans of the treatment previously received by the patients. Plan quality was evaluated using the HI, NTI and dose to OARs. The hIMRT technique was significantly more homogenous than the 3DCRT technique, while maintaining target coverage. The hIMRT technique was also superior at minimising the amount of tissue receiving D 105% and above (P < 0.0001). The ipsilateral lung and contralateral breast maximum were significantly lower in the hIMRT plans (P < 0.05 and P < 0.005), but the 3DCRT technique achieved a lower mean heart dose in left-sided breast cancer patients (P < 0.05). Hybrid intensity modulated radiotherapy plans achieved improved dose homogeneity compared to the 3DCRT plans and superior outcome with regard to dose to normal tissues. We propose that the addition of both HI and NTI in evaluating the quality of intensity modulated radiotherapy (IMRT) breast plans provides clinically relevant comparators which more accurately reflect the new paradigm of treatment goals and outcomes in the era of breast IMRT

  15. Impact induced response spectrum for the safety evaluation of the high flux isotope reactor

    International Nuclear Information System (INIS)

    Chang, S.J.

    1997-01-01

    The dynamic impact to the nearby HFIR reactor vessel caused by heavy load drop is analyzed. The impact calculation is carried out by applying the ABAQUS computer code. An impact-induced response spectrum is constructed in order to evaluate whether the HFIR vessel and the shutdown mechanism may be disabled. For the frequency range less than 10 Hz, the maximum spectral velocity of impact is approximately equal to that of the HFIR seismic design-basis spectrum. For the frequency range greater than 10 Hz, the impact-induced response spectrum is shown to cause no effect to the control rod and the shutdown mechanism. An earlier seismic safety assessment for the HFIR control and shutdown mechanism was made by EQE. Based on EQE modal solution that is combined with the impact-induced spectrum, it is concluded that the impact will not cause any damage to the shutdown mechanism, even while the reactor is in operation. The present method suggests a general approach for evaluating the impact induced damage to the reactor by applying the existing finite element modal solution that has been carried out for the seismic evaluation of the reactor

  16. Performance Evaluation of Moving Bed Bio Film Reactor in Saline Wastewater Treatment

    Directory of Open Access Journals (Sweden)

    M Ahmadi

    2013-06-01

    Full Text Available Background and purpose:Moving Bed Biofilm Reactor is an aerobic attached growth with better biofilm thickness control, lack of plugging and lower head loss. Consequently, this system is greatly used by different wastewater treatment plants. High TDS wastewater produced petrochemical, leather tanning, sea food processing, cannery, pickling and dairy industries. The aim of this study was to evaluate the performance of MBBR in saline wastewater treatment. Materials and methods: In this study, 50 percent of a cylindrical reactor with 9.5 liter occupied media with 650 m2.m-3. In the first step, hydraulic regime was evaluated and startup reactor was done by sanitary sludge. Bio film was generated with glucose as the sole carbon source in synthetic wastewater. MBBR performance evaluation was performed in 6:30 and 8:45 with saline wastewater after bio film produced on media. Results: After 83 days of passing MBBR operation with saline wastewater containing 3000-12000 mg.L-1 TDS, organic loading rate of 2.2-3.5 kg/m3.d COD removal efficiency reached 80-92%. Conclusion: Moving bed biofilm reactor is effective in organic load elimination from saline wastewater.

  17. An evaluation of the concept of transuranic burning using liquid metal reactors

    International Nuclear Information System (INIS)

    Rodwell, E.; Shaw, R.A.; Williams, R.F.

    1991-03-01

    The evaluation investigates the potential benefits to radioactive waste disposal of separating the transuranic elements from spent reactor fuel before the spent fuel is disposed of in geologic repositories. The evaluation addresses the question whether the benefits to radioactive waste disposal would justify processing the fuel to separate the transuranics, plus a liquid metal reactor (LMR) deployment program to transmute the separated transuranics. The evaluation concludes that adoption of a process-before-disposal policy for all the spent fuel from the light water reactors (LWRs) would accrue only modest benefits with respect to the accumulation of uranium mill tailings, the national inventory of transuranics and the licensing of a geologic repository. It is likely that this process-before-disposal policy would incur a large cost penalty, encounter major institutional difficulties, multiply licensing hurdles, and amplify political and public opposition to the overall nuclear power program. However, plutonium from spent LWR fuel is projected to be substantially more economic than enriched uranium, as fissile material for startup of LMRs when LMR deployment becomes economically justified. At that time, the LMR would fulfill its role as a reactor system that would protect the nation from diminishing energy resources. Development tasks towards defining and developing the most cost-effective LMR and associated fuel cycle remain very important. 29 refs., 3 figs., 18 tabs

  18. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    International Nuclear Information System (INIS)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-01

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  19. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    International Nuclear Information System (INIS)

    Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

  20. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

    2000-09-01

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

  1. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  2. Status of fusion reactor blanket evaluation studies in France

    International Nuclear Information System (INIS)

    Carre, F.; Chevereau, G.; Gervaise, F.; Proust, E.

    1985-03-01

    In the frame of recent CEA studies aiming at the evaluation and at the comparison of various candidate blanket concepts in moderate power conditions (Psub(n) approximately 2 MW/m 2 ), the present work examines the neutronic and thermomechanical performances of a water cooled Li 17 Pb 83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium/LiAlO 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress

  3. Progress in fusion reactors blanket analysis and evaluation at CEA

    International Nuclear Information System (INIS)

    Proust, E.; Gervaise, F.; Carre, F.; Chevereau, G.; Doutriaux, D.

    1986-09-01

    In the frame of the recent CEA studies aiming at the development, evaluation and comparison of solid breeder blanket concepts in view of their adaptation to NET, the evaluation of specific questions related to the first wall design, the present paper examines first the performances of a helium cooled toroidal blanket design for NET, based on innovative Beryllium/Ceramics breeder rod elements. Neutronic and thermo-mechanical optimisation converges on a concept featured by a breeding capability in excess of 1.2, a reasonnable pumping power of 1% and a narrow breeder temperature range (470+-30 deg C of the breeder), the latter being largely independent of the power level. This design proves naturally adapted to ceramic breeder assigned to very strict working conditions, and provides for any change in the thermal and heat transfer characteristics over the blanket lifetime. The final section of the paper is devoted to the evaluation of the heat load poloidal distribution and to the irradiation effects on first wall structural materials

  4. Evaluation of incompressible hydrodynamic mass methods in reactor applications

    International Nuclear Information System (INIS)

    Takeuchi, K.

    1981-01-01

    The hydrodynamic (or virtual) mass approach is evaluated by comparison of structural responses computed by the hydrodynamic mass method with those computed by MULTIFLEX code for a fluid/structure interaction problem with fluid compression effects taken into account. A sample problem used in that evaluation is a simplified 1-D PWR model which is first subjected to a LOCA type transient. The time history of structural displacement computed with the hydrodynamic mass approach is compared with MULTIFLEX results. The frequencies of structural oscillation of these two computations agree. The amplitudes disagree by more than 50%, which is attributed to the effect of fluid compressibility. For the seismic study, sinusoidal forces are applied to the floor at the vessel support. The system responses are expressed by the response functions or the maximum values of the barrel/vessel relative displacements as the applied frequency is varied. The response functions are computed by the hydrodynamic mass method and by MULTIFLEX for evaluation of the virtual mass method. For the pump pulsation study, sinusoidal pressure oscillations are applied at the pump outlet and the response functions are computed as above. 12 refs

  5. Inverse modeling approach for evaluation of kinetic parameters of a biofilm reactor using tabu search.

    Science.gov (United States)

    Kumar, B Shiva; Venkateswarlu, Ch

    2014-08-01

    The complex nature of biological reactions in biofilm reactors often poses difficulties in analyzing such reactors experimentally. Mathematical models could be very useful for their design and analysis. However, application of biofilm reactor models to practical problems proves somewhat ineffective due to the lack of knowledge of accurate kinetic models and uncertainty in model parameters. In this work, we propose an inverse modeling approach based on tabu search (TS) to estimate the parameters of kinetic and film thickness models. TS is used to estimate these parameters as a consequence of the validation of the mathematical models of the process with the aid of measured data obtained from an experimental fixed-bed anaerobic biofilm reactor involving the treatment of pharmaceutical industry wastewater. The results evaluated for different modeling configurations of varying degrees of complexity illustrate the effectiveness of TS for accurate estimation of kinetic and film thickness model parameters of the biofilm process. The results show that the two-dimensional mathematical model with Edward kinetics (with its optimum parameters as mu(max)rho(s)/Y = 24.57, Ks = 1.352 and Ki = 102.36) and three-parameter film thickness expression (with its estimated parameters as a = 0.289 x 10(-5), b = 1.55 x 10(-4) and c = 15.2 x 10(-6)) better describes the biofilm reactor treating the industry wastewater.

  6. Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

    Science.gov (United States)

    Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

    2016-12-01

    This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent. Copyright © 2015 Elsevier Inc. All rights reserved.

  7. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  8. Evaluating the consequences of loss of flow accident for a typical VVER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mirvakili, S.M.; Safaei, S. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Dept. of Nuclear Engineering, School of Mechanical Engineering; Faghihi, F. [Shiraz Univ., Shiraz (Iran, Islamic Republic of). Safety Research Center

    2010-07-01

    The loss of coolant flow in a nuclear reactor can result from a mechanical or electrical failure of the coolant pump. If the reactor is not tripped promptly, the immediate effect is a rapid increase in coolant temperature, decrease in minimum departure from nucleate boiling ratio (DNBR) and fuel damage. This study evaluated the shaft seizure of a reactor coolant pump in a VVER-1000 nuclear reactor. The locked rotor results in rapid reduction of flow through the affected reactor coolant loop and in turn leads to an increase in the primary coolant temperature and pressure. The analysis was conducted with regard for superimposing loss of power to the power plant at the initial accident moment. The required transient functions of flow, pressure and power were obtained using system transient calculations applied in COBRA-EN computer code in order to calculate the overall core thermal-hydraulic parameters such as temperature, critical heat flux and DNBR. The study showed that the critical period for the locked rotor accident is the first few seconds during which the maximum values of pressure and temperature are reached. 10 refs., 1 tab., 3 figs.

  9. Evaluation of the radial design of fuel cells in an operation cycle of a BWR reactor

    International Nuclear Information System (INIS)

    Gonzalez C, J.; Martin del Campo M, C.

    2003-01-01

    This work is continuation of one previous in the one that the application of the optimization technique called Tabu search to the radial design of fuel cells of boiling water reactors (BWR, Boiling Water Reactor) is presented. The objective function used in the optimization process only include neutron parameters (k-infinite and peak of radial power) considering the cell at infinite media. It was obtained to reduce the cell average enrichment completing the characteristics of reactivity of an original cell. The objective of the present work is to validate the objective function that was used for the radial design of the fuel cell (test cell), analyzing the operation of a one cycle of the reactor in which fuels have been fresh recharged that contain an axial area with the nuclear database of the cell designed instead of the original cell. For it is simulated it with Cm-Presto the cycle 10 of the reactor operation of the Unit 1 of the Nuclear Power station of Laguna Verde (U1-CNLV). For the cycle evaluation its were applied so much the simulation with the Haling strategy, as the simulation of the one cycle with control rod patterns and they were evaluated the energy generation and several power limits and reactivity that are used as design parameters in fuel reloads of BWR reactors. The results at level of an operation cycle of the reactor, show that the objective function used in the optimization and radial design of the cell is adequate and that it can induce to one good use of the fuel. (Author)

  10. Development of inspection and evaluation guidelines for light water reactor internals

    International Nuclear Information System (INIS)

    Aoki, T.; Yamashita, H.; Sakai, K.

    2002-01-01

    Full text: In Japan, before a nuclear power plant reaches its 30 years of operation, the Japanese utilities carry out a 'study on plant life management'. Reflecting the results of that study into the maintenance, the utilities make efforts to maintain and improve the safety and reliability of their nuclear power plants. In this study, all safety related components are evaluated from the viewpoint of aging degradation, assuming a long-term operation. If a crack should be found at components such as reactor internals, which is deemed important for safety and are difficult to repair or replace may provide a serious impact on the plant operation and management. Reactor internals, for instance, made of austenitic stainless steel and nickel base alloy, are not completely free from aging degradation including stress corrosion cracking (SCC). Therefore, it is concluded in the study on plant life management that they are required continuous planned inspections to confirm their integrity while continuing plant operation. If an aging degradation such as SCC should be found at the reactor internals, a great amount of labor and time may be required for root cause investigation and analysis and subsequent repairs because it is very difficult to reach to the degraded portion due to structural, dimensional and environmental restrictions. Therefore, such situations may provide serious impact on plant operation and management. As to the reactor internals and other components with similar characteristics, it is strongly recommended that contingency plan should be prepared in advance. Thus, considering the significance of aging degradation, it is necessary to develop some standard rules for the inspection and evaluation of reactor internals. Such rules should specify when, where and how to inspect. They should also specify the evaluation method in case such degradation as a crack is found, and the repair method and extent if repairs are required. These standard rules must be

  11. Evaluation of upper-shelf toughness requirements for reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, R.M.; Zahoor, A. (NOVETECH Corp., Rockville, MD (USA)); Hiser, A. (Materials Engineering Associates, Inc., Lanham, MD (USA)); Ernst, H.A.; Pollitz, E.T. (Georgia Inst. of Tech., Atlanta, GA (USA))

    1990-04-01

    This work assesses and applies the criteria recommended by the ASME Subgroup on Evaluation Standards for the evaluation of reactor pressure vessel beltline materials having upper shelf Charpy energies less than 50 ft-lbs. The assessment included comparison of the upper shelf energies required by the criteria recommended for Service Level A and B conditions and criteria proposed for evaluation of postulated Service Level C and D events. The criteria recommended for Service Level A and B conditions was used to evaluate Linde 80 weld material. 9 refs., 4 figs.

  12. Evaluation of upper-shelf toughness requirements for reactor pressure vessels

    International Nuclear Information System (INIS)

    Gamble, R.M.; Zahoor, A.; Hiser, A.; Ernst, H.A.; Pollitz, E.T.

    1990-04-01

    This work assesses and applies the criteria recommended by the ASME Subgroup on Evaluation Standards for the evaluation of reactor pressure vessel beltline materials having upper shelf Charpy energies less than 50 ft-lbs. The assessment included comparison of the upper shelf energies required by the criteria recommended for Service Level A and B conditions and criteria proposed for evaluation of postulated Service Level C and D events. The criteria recommended for Service Level A and B conditions was used to evaluate Linde 80 weld material. 9 refs., 4 figs

  13. Development of source term evaluation method for Korean Next Generation Reactor(III)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geon Jae; Park, Jin Baek; Lee, Yeong Il; Song, Min Cheonl; Lee, Ho Jin [Korea Advanced Institue of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    This project had investigated irradiation characteristics of MOX fuel method to predict nuclide concentration at primary and secondary coolant using a core containing 100% of all MOX fuel and development of source term evaluation tool. In this study, several prediction methods of source term are evaluated. Detailed contents of this project are : an evaluation of model for nuclear concentration at Reactor Coolant System, evaluation of primary and secondary coolant concentration of reference Nuclear Power Plant using purely MOX fuel, suggestion of source term prediction method of NPP with a core using MOX fuel.

  14. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    International Nuclear Information System (INIS)

    1991-04-01

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig

  15. Restart of K-Reactor, Savannah River Site: Safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This Safety Evaluation Report (SER) focuses on those issues required to support the restart of the K-Reactor at the Savannah River Plant. This SER provides the safety criteria for restart and documents the results of the staff reviews of the DOE and operating contractor activities to meet these criteria. To develop the restart criteria for the issues discussed in this SER, the Savannah River Restart Office and Savannah River Special Projects Office staffs relied, when possible, on commercial industry codes and standards and on NRC requirements and guidelines for the commercial nuclear industry. However, because of the age and uniqueness of the Savannah River reactors, criteria for the commercial plants were not always applicable. In these cases, alternate criteria were developed. The restart criteria applicable to each of the issues are identified in the safety evaluations for each issue. The restart criteria identified in this report are intended to apply only to restart of the Savannah River reactors. Following the development of the acceptance criteria, the DOE staff and their support contractors evaluated the results of the DOE and operating contractor (WSRC) activities to meet these criteria. The results of those evaluations are documented in this report. Deviations or failures to meet the requirements are either justified in the report or carried as open or confirmatory items to be completed and evaluated in supplements to this report before restart. 62 refs., 1 fig.

  16. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  17. Evaluation of strategies for end storage of high-level reactor fuel; Vurdering av strategier for sluttlagring av hoeyaktivt reaktorbrensel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report evaluates a national strategy for end-storage of used high-level reactor fuel from the research reactors at Kjeller and in Halden. This strategy presupposes that all the important phases in handling the high-level material, including temporary storage and deposition, are covered. The quantity of spent fuel from Norwegian reactors is quite small. In addition to the technological issues, ethical, environmental, safety and economical requirements are emphasized.

  18. Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Smith, D.L.

    1988-01-01

    The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs

  19. The evaluation of equipment and Instrumentation Reliability Factors on Power Reactor

    International Nuclear Information System (INIS)

    Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Santosa, Kussigit; Darlis; Sudiyono, Bambang; Yuniyanta, Sasongko; Sudarmin

    1999-01-01

    Equipment and instrumentation reliability on type power reactor control room was determined by its pattern and design. the principle of ergonomy applied on equipment and instrumentation layout in this ABWR type reactor are geometric pattern appropriate with economic body motion, average anthropometry data of operator especially operator hand-reach, range of vision, angle of vision, lighting, color arrangement and harmony as will as operator case in operating the equipment system. Limitation criteria of the parameter mentioned above are based on EPRI NP-3659, NURG 0700, and NUREG/CR-3331 documents. Besides that, the (working) physical environment parameter factor of the control room must be designed in order to fulfil the standard criteria of ergonomic condition based on NUREG-0800. The reliability evaluation of equipment and instrumentation system also occurs observed from man machine interaction side which happen between operator and equipment and instrumentation in the ABWR type power reactor control room. From the MMI analysis can be known the working failure possibility which is caused by the operator. The evaluation result of equipment and instrumentation reliability on ABWR type power reactor control room showed that the design of this ABWR control room is good and fulfils the ergonomy standard criteria have been determined

  20. New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

    2017-02-15

    Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

  1. Stochastic estimation approach for the evaluation of thermal-hydraulic parameters in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, M.G.

    1986-01-01

    A method based on the extended Kalman filter is developed for the estimation of the core coolant mass flow rate in pressurized water reactors. The need for flow calibration can be avoided by a direct estimation of this parameter. A reduced-order neutronic and thermal-hydraulic model is developed for the Loss-of-Fluid Test (LOFT) reactor. The neutron detector and core-exit coolant temperature signals from the LOFT reactor are used as measurements in the parameter estimation algorithm. The estimation sensitivity to model uncertainties was evaluated using the ambiguity function analysis. This also provides a lower bound on the measurement sample size necessary to achieve a certain estimation accuracy. A sequential technique was developed to minimize the computational effort needed to discretize the continuous time equations, and thus achieve faster convergence to the true parameter value. The performance of the stochastic approximation method was first evaluated using simulated random data, and then applied to the estimation of coolant flow rate using the operational data from the LOFT reactor at 100 and 65% flow rate conditions

  2. NEA Activities in Preserving, Evaluating and Applying Data from Fast Reactor Experiments

    International Nuclear Information System (INIS)

    Gulliford, N.T.; Cornet, S.M.; Hill, I.; Yamaji, A.

    2015-01-01

    The goal of the OECD Nuclear Energy Agency (NEA) in the area of nuclear science is to help member countries identify, collate, develop and disseminate the basic scientific and technical knowledge required to ensure safe and reliable operation of current nuclear systems and to develop next generation technologies. Within these general goals, the current nuclear science programme has three key objectives: (i) to help advance the existing scientific knowledge needed to enhance the performance and safety of current nuclear systems, (ii) to contribute to building a solid scientific and technical basis for the development of future generation nuclear systems and (iii) to support the preservation of essential knowledge in the field of nuclear science. As part of the second and third of these objectives, an extensive programme of work to preserve and evaluate data from integral experiments has been established, including reactor physics, shielding and criticality safety experiments on fast reactor systems. Data from experimental facilities are reviewed and, if necessary, archives of information are made safe. This may typically involve the indexing and scanning of key documents and archiving of logbooks, for example. Selected experiments go through a detailed evaluation process and where deemed appropriate, a benchmark description is created in a standardized format for inclusion in one of the NEA Data Bank international databases. This information is used extensively by the international nuclear science community to validate their modelling and simulation tools. The process can be viewed as part of a broader knowledge management function, where information is gathered, evaluated, linked and made accessible to a wide range of users. The presentation gives details of the main databases maintained and developed by the NEA, focusing on those related to fast reactor applications. The status of recent preservation activities for fast reactor archives in the United Kingdom is

  3. Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tarchalski, M.; Pytel, K.; Wroblewska, M.; Marcinkowska, Z.; Boettcher, A.; Prokopowicz, R. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Sireta, P.; Gonnier, C.; Bignan, G. [CEA, DEN, Reactor Studies Department, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Fourmentel, D.; Barbot, L.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C.; Brun, J. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Jagielski, J. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Institute of Electronic Materials Technolgy, Wolczynska 133, 01-919 Warszawa (Poland); Luks, A. [Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland)

    2015-07-01

    Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from

  4. Evaluation of plutonium, uranium, and thorium use in power reactor fuel cycles

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.

    1977-01-01

    The increased cost of uranium and separative work has increased the attractiveness of plutonium use in both uranium and thorium fuel cycles in thermal reactors. A technology, fuel utilization, and economic evaluation is given for uranium and thorium fuel cycles in various reactor types, along with the use of plutonium and 238 U. Reactors considered are LWRs, HWRs, LWBRs, HTGRs, and FBRs. Key technology factors are fuel irradiation performance and associated physical property values. Key economic factors are unit costs for fuel fabrication and reprocessing, and for refabrication of recycle fuels; consistent cost estimates are utilized. In thermal reactors, the irradiation performance of ceramic fuels appears to be satisfactory. At present costs for uranium ore and separative work, recycle of plutonium with thorium rather than uranium is preferable from fuel utilization and economic viewpoints. Further, the unit recovery cost of plutonium is lower from LWR fuels than from natural-uranium HWR fuels; use of LWR product permits plutonium/thorium fueling to compete with uranium cycles. Converting uranium cycles to thorium cycles increases the energy which can be extracted from a given uranium resource. Thus, additional fuel utilization improvement can be obtained by fueling all thermal reactors with thorium, but this requires use of highly enriched uranium; use of 235 U with thorium is most economic in HTGRs followed by HWRs and then LWRs. Marked improvement in long-term fuel utilization can be obtained through high thorium loadings and short fuel cycle irradiations as in the LWBR, but this imposes significant economic penalties. Similar operating modes are possible in HWRs and HTGRs. In fast reactors, use of the plutonium-uranium cycle gives advantageous fuel resource utilization in both LMFBRs and GCFRs; use of the thorium cycle provides more negative core reactivity coefficients and more flexibility relative to use of recycle fuels containing uranium of less than 20

  5. Comparison of dosimetric methods for virtual wedge analysis

    International Nuclear Information System (INIS)

    Bailey, M.; Nelson, V.; Collins, O.; West, M.; Holloway, L.; Rajapaske, S.; Arts, J.; Varas, J.; Cho, G.; Hill, R.

    2004-01-01

    Full text: The Siemens Virtual Wedge (Concord, USA) creates wedged beam profile by moving a single collimator jaw across the specified field size whilst varying the dose rate and jaw speed for use in the delivery of radiotherapy treatments. The measurement of the dosimetric characteristics of the Siemens Virtual Wedge poses significant challenges to medical physicists. This study investigates several different methods for measuring and analysing the virtual wedge for data collection for treatment planning systems and ongoing quality assurance. The beam profiles of the Virtual Wedge (VW) were compared using several different dosimetric methods. Open field profiles were measured with Kodak X-Omat V (Rochester, NY, USA) radiographic film and compared with measurements made using the Sun Nuclear Profiler with a Motorized Drive Assembly (MDA) (Melbourne, FL, USA) and the Scanditronix Wellhofer CC13 ionisation chamber and 24 ion Chamber Array (CA24) (Schwarzenbruck, Germany). The resolution of each dosimetric method for open field profiles was determined. The Virtual Wedge profiles were measured with radiographic film the Profiler and the Scanditronix Wellhofer CA 24 ion Chamber Array at 5 different depths. The ease of setup, time taken, analysis and accuracy of measurement were all evaluated to determine the method that would be both appropriate and practical for routine quality assurance of the Virtual Wedge. The open field profiles agreed within ±2% or 2mm for all dosimetric methods. The accuracy of the Profiler and CA24 are limited to half of the step size selected for each of these detectors. For the VW measurements a step size of 2mm was selected for the Profiler and the CA24. The VW profiles for all dosimetric methods agreed within ±2% or 2mm for the main wedged section of the profile. The toe and heel ends of the wedges showed the significant discrepancies dependent upon the dosimetry method used, up to 7% for the toe end with the CA24. The dosimetry of the

  6. N Reactor Production Assurance Program blance of plant evaluation: report of findings and conclusions

    International Nuclear Information System (INIS)

    Hurd, E.N.; Bitten, E.J.

    1985-03-01

    Fourteen tasks were identified by UNC Nuclear Industries for evaluating the life expectancy of N Reactor structures, systems and components in the Balance of Plant portion of the Production Assurance Program. A Westinghouse Hanford Company (WHC) evaluation team was assigned to each of these fourteen tasks. A uniform set of criteria was used by all teams in evaluating the problems that may be encountered during the extended plant operating lifetime. The overall conclusion is that extended life to those Balance of Plant components and systems studied can be achieved. Problems with the potential for compromising that conclusion are identified, and feasible solutions are provided

  7. Report and analysis on 'PR and PP evaluation. Example sodium fast reactor full system case study'

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Inoue, Naoko; Kawakubo, Yoko; Watahiki, Masaru

    2011-01-01

    The Generation IV (GEN IV) Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December 2002 in order to develop the PR and valuation methodology for GEN IV nuclear energy systems. In the final report of 'PR and PP Evaluation: Example Sodium Fast Reactor (ESFR) Full System Case Study,' issued in October 2009, the demonstration study of PR and PP evaluation with the qualitative approach are summarized using ESFR with four scenario threats. The present paper reviews and analyzes some results of the ESFR case study, and identifies the challenges and direction for the PR and PP evaluation methodology with quantitative approach. (author)

  8. IPIP: A new approach to inverse planning for HDR brachytherapy by directly optimizing dosimetric indices

    International Nuclear Information System (INIS)

    Siauw, Timmy; Cunha, Adam; Atamtuerk, Alper; Hsu, I-Chow; Pouliot, Jean; Goldberg, Ken

    2011-01-01

    Purpose: Many planning methods for high dose rate (HDR) brachytherapy require an iterative approach. A set of computational parameters are hypothesized that will give a dose plan that meets dosimetric criteria. A dose plan is computed using these parameters, and if any dosimetric criteria are not met, the process is iterated until a suitable dose plan is found. In this way, the dose distribution is controlled by abstract parameters. The purpose of this study is to develop a new approach for HDR brachytherapy by directly optimizing the dose distribution based on dosimetric criteria. Methods: The authors developed inverse planning by integer program (IPIP), an optimization model for computing HDR brachytherapy dose plans and a fast heuristic for it. They used their heuristic to compute dose plans for 20 anonymized prostate cancer image data sets from patients previously treated at their clinic database. Dosimetry was evaluated and compared to dosimetric criteria. Results: Dose plans computed from IPIP satisfied all given dosimetric criteria for the target and healthy tissue after a single iteration. The average target coverage was 95%. The average computation time for IPIP was 30.1 s on an Intel(R) Core TM 2 Duo CPU 1.67 GHz processor with 3 Gib RAM. Conclusions: IPIP is an HDR brachytherapy planning system that directly incorporates dosimetric criteria. The authors have demonstrated that IPIP has clinically acceptable performance for the prostate cases and dosimetric criteria used in this study, in both dosimetry and runtime. Further study is required to determine if IPIP performs well for a more general group of patients and dosimetric criteria, including other cancer sites such as GYN.

  9. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  10. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  11. Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1997-07-01

    The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

  12. Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

    1978-01-01

    An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

  13. Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1988-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactor (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987

  14. The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

    International Nuclear Information System (INIS)

    Dickson, T.L.; Simonen, F.A.

    1992-01-01

    Probabilistic fracture mechanics analysis is a major element of the comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-11 perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel

  15. Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

    International Nuclear Information System (INIS)

    Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

    1988-01-01

    The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987. (author)

  16. The application of probabilistic fracture analysis to residual life evaluation of embrittled reactor vessels

    International Nuclear Information System (INIS)

    Dickson, T.L.; Simonen, F.A.

    1992-01-01

    Probabilistic fracture mechanics analysis is a major element of comprehensive probabilistic methodology on which current NRC regulatory requirements for pressurized water reactor vessel integrity evaluation are based. Computer codes such as OCA-P and VISA-II perform probabilistic fracture analyses to estimate the increase in vessel failure probability that occurs as the vessel material accumulates radiation damage over the operating life of the vessel. The results of such analyses, when compared with limits of acceptable failure probabilities, provide an estimation of the residual life of a vessel. Such codes can be applied to evaluate the potential benefits of plant-specific mitigating actions designed to reduce the probability of failure of a reactor vessel. 10 refs

  17. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    International Nuclear Information System (INIS)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

  18. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  19. Evaluation of final vapor pressures in the loss of flow accident in an irradiation device of a pool reactor core

    International Nuclear Information System (INIS)

    Verri, A.

    1987-01-01

    The reliability feature, are described for a device containing samples, at a temperatures of 300 grade centigrades, in a reactor core for a long time. After an examination of the maximum accident event, the maximum vapour pressure originated by the inlet of reactor cooling water into the experimental device, is evaluated

  20. 9 CFR 147.16 - Procedure for the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment).

    Science.gov (United States)

    2010-01-01

    ... mycoplasma reactors by in vivo bio-assay (enrichment). 147.16 Section 147.16 Animals and Animal Products... the evaluation of mycoplasma reactors by in vivo bio-assay (enrichment). This procedure has been shown... publications: (a) Bigland, C. H. and A. J. DaMassa, “A Bio-Assay for Mycoplasma Gallisepticum.” In: United...

  1. Population Sensitivity Evaluation of Two Proposed Hampton Roads Area Sites for a Possible Small Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Belles, R. J. [ORNL; Omitaomu, O. A. [ORNL

    2014-08-01

    The overall objective of this research project is to use the OR-SAGE tool to support the US Department of Energy (DOE) Office of Nuclear Energy (NE) in evaluating future electrical generation deployment options for small modular reactors (SMRs) in areas with significant energy demand from the federal sector. Deployment of SMRs in zones with high federal energy use can provide a means of meeting federal clean energy goals.

  2. Evaluation of reactor induced (n,p) reactions for activation analysis of titanium in geological materials

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa Garcia, R; Cohen, I M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    1984-05-01

    The possibilities of reactor induced (n,p) reactions as a tool for neutron activation analysis of titanium in geological samples are discussed. The interference of calcium and scandium is experimentally evaluated. Results for Ti, Ca and Sc in GSP-1 and PCC-1 standard rocks are presented. Based on the experimental values, it is concluded that the /sup 47/Ti(n,p)/sup 47/Sc reaction is the most favourable for titanium determination. 11 refs.

  3. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    Energy Technology Data Exchange (ETDEWEB)

    Holmes, Forest Howard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  4. Evaluation of actinide partitioning and transmutation in light-water reactors

    International Nuclear Information System (INIS)

    Collins, Emory D.; Renier, John-Paul

    2004-01-01

    Advanced Fuel Cycle Initiative (AFCI) studies were made to evaluate the feasibility of multicycle transmutation of plutonium and the minor actinides (MAs) in light-water reactors (LWRs). Results showed that significant repository benefits, cost reductions, proliferation resistance, and effective use of facilities can be obtained. Key advantages are shown to be made possible by processing 30-year-decayed spent fuel rather than the more traditional 5-year-decayed fuel. (authors)

  5. Evaluation of 'period-generated' control laws for the time-optimal control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1988-01-01

    Time-Optimal control of neutronic power has recently been achieved by developing control laws that determine the actuator mechanism velocity necessary to produce a specified reactor period. These laws are designated as the 'MIT-SNL Period-Generated Minimum Time Control Laws'. Relative to time-optimal response, they function by altering the rate of change of reactivity so that the instantaneous period is stepped from infinity to its minimum allowed value, held at that value until the desired power level is attained, and then stepped back to infinity. The results of a systematic evaluation of these laws are presented. The behavior of each term in the control laws is shown and the capability of these laws to control properly the reactor power is demonstrated. Factors affecting the implementation of these laws, such as the prompt neutron lifetime and the differential reactivity worth of the actuators, are discussed. Finally, the results of an experimental study in which these laws were used to adjust the power of the 5 MWt MIT Research Reactor are shown. The information presented should be of interest to those designing high performance control systems for test, spacecraft, or, in certain instances, commercial reactors

  6. Evaluation of Continuous Stirred Tank Reactor Performance by Using Radioisotope Tracer

    International Nuclear Information System (INIS)

    Noor Anis Kundari; Djoko Marjanto; Ardhani Dyah W

    2009-01-01

    Research on performance evaluation of continuous stirred tank reactor (CSTR) using radioisotope tracer has been carried out. The aim of research is to assess a validity of assumption that stirring or mixing process in a CSTR is perfect. In order to follow the flow dynamics process of the fluid in the reactor, I-131 was used. The reactor was equipped with four baffles. The fluid/water leaving the reactor was sampled at 13 up to 1393 seconds and analysed its I-131 concentration. The performance of CSTR is expressed as dispersed number (D/uL) as function of retention time and Reynolds number under axial dispersed model. The experimental result show that the relation between the dispersion number and retention time is D/uL = 9X10 -4 (t s * ) 2 - 6.9X10 -1 (t s * ) + 148 and the dispersion number and Reynolds number is D/uL = 65.7 e 0.0003/Re . The dispersion number obtained were much higher than 0.01 that in between 11.08 up to 21.4. That mean the mixing process occurred in the CSTR can be assumed to be ideal. (author)

  7. Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Hess, J. D.; Gauld, I. C.; Gulliford, J.; Hill, I.; Okajima, S.

    2017-01-01

    Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

  8. Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2009-01-01

    This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

  9. Analysis and evaluation of the hydrogen risk in a thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Chaudron, V.; Arnould, F.; Latge, C.; Laurent, A.

    2001-01-01

    After a recall of the principle of controlled thermonuclear fusion, the ITER reactor project is briefly described. The integrity of the reactor must be preserved in the case of a potential explosion of the hydrogen generated inside the reactor, in order to avoid any dispersion radioactive, chemical or toxic materials in the environment. The fundamental principles of safety developed to fulfill these objectives, in particular the defense-in-depth concept, are presented. The main potential source of hydrogen production is the oxidation of beryllium, which is used as protection material in the first wall of the torus, and the accidental presence of water, as reported in several scenarios. The confinement strategy is then described with the qualification of the role of the different barriers. Finally, the hydrogen explosion risk is analyzed and evaluated with respect to the sources, to the reference envelope scenarios and to the location of hydrogen inside the ITER reactor. It appears, at the engineering stage, that the vacuum toric vessel, the discharge reservoir and the exchanger compartments are the most worrying parts. (J.S.)

  10. Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2005-01-01

    KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

  11. An evaluation of alternative reactor vessel cutting technologies for the decommissioning of the experimental boiling water reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

    1991-01-01

    This paper will detail (1) a brief overview of the current status of the EBWR D ampersand D Project, and (2) the results of a study performed to evaluate the metal cutting technologies available to size reduce the EBWR reactor vessel. The techniques evaluated were: Plasma arc, Arc saw, Oxyacetylene, Electric arc gouging, Mechanical cladding removal/flame cutting, Exothermic reaction, Diamond wire, Water jet, Laser, Mechanical milling, Controlled explosives, and Electrical discharge. After a detailed review of these 12 techniques, the decision was made by ANL that the most appropriate method for segmenting the EBWR reactor vessel would be to rift the vessel from the vessel cavity and use an abrasive water jet positioned on the main floor to perform the cutting of the reactor vessel

  12. Dosimetric system for measurement of radioactive contaminations

    International Nuclear Information System (INIS)

    Litynski, Z.; Pienkos, J.P.; Witkowski, J.; Zadrozny, S.

    1985-01-01

    A dosimetric system for personnel dosimetry and monitoring measuring a contamination without time delay and dead time is described. The system ensures many-point measurement and minimalization of background radiation influence. 1 fig. (A.S.)

  13. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

    2009-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the proces