Search Results

Advanced search parameters have been applied.
open access

Hot helium flow test facility summary report

Description: This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations.
Date: June 1, 1980
Partner: UNT Libraries Government Documents Department
open access

The multipulse Thomson scattering diagnostic on the DIII-D tokamak

Description: This paper describes the design and operation of a 40-spatial channel Thomson scattering system that uses multiple 20 Hz Nd:YAG lasers to measure the electron temperature and density profiles periodically throughout an entire plasma discharge. Interference filter polychromators disperse the scattered light which is detected by silicon avalanche photodiodes. The measurable temperature range from 10 eV to 20 keV and the minimum detectable density is about 2 {times} 10{sup 18} m{sup {minus}3}. Laser control and data acquisition are performed in real-time by a VME-based microcomputer. Data analysis is performed by a MicroVAX 3400. Unique features of this system include burst mode'' operation, where multiple lasers are fired in rapid succession (< 10 KHz), real-time analysis capability, and laser beam quality and alignment monitoring during plasma operation. Results of component testing, calibration, and plasma operation are presented. 8 refs. 6 figs.
Date: September 1, 1991
Creator: Carlstrom, T. N.; Campbell, G. L.; DeBoo, J. C.; Evanko, R. G.; Evans, J.; Greenfield, C. M. et al.
Partner: UNT Libraries Government Documents Department
open access

850/sup 0/C VHTR plant technical description

Description: This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 950/sup 0/C (1742/sup 0/F) core outlet temperature.
Date: June 1, 1980
Partner: UNT Libraries Government Documents Department
open access

Non-linear instability of DIII-D to error fields

Description: Otherwise stable DIII-D discharges can become nonlinearly unstable to locked modes and disrupt when subjected to resonant m = 2, n = 1 error field caused by irregular poloidal field coils, i.e. intrinsic field errors. Instability is observed in DIII-D when the magnitude of the radial component of the m = 2, n = 1 error field with respect to the toroidal field is B{sub r21}/B{sub T} of about 1.7 {times} 10{sup {minus}4}. The locked modes triggered by an external error field are aligned with the static error field and the plasma fluid rotation ceases as a result of the growth of the mode. The triggered locked modes are the precursors of the subsequent plasma disruption. The use of an n = 1 coil'' to partially cancel intrinsic errors, or to increase them, results in a significantly expanded, or reduced, stable operating parameter space. Precise error field measurements have allowed the design of an improved correction coil for DIII-D, the C-coil'', which could further cancel error fields and help to avoid disruptive locked modes. 6 refs., 4 figs.
Date: October 1, 1991
Creator: La Haye, R. J. & Scoville, J. T.
Partner: UNT Libraries Government Documents Department
open access

Progress report on the development of the General Atomic thermochemical water-splitting process

Description: The major accomplishments of the DOE funded part of the GA thermochemical water-splitting program are reported. They include: completion of installation of all bench-scale equipment; operation and preliminary data acquisition for bench-scale subunits I and II; design, installation and operation of a system for iodine removal from the low phase; review and modification of Section III of the engineering flowsheet resulting in an increase in process efficiency and decrease in capital cost; and completion of the Funk panel reivew. The results of the experimental work have demonstrated that flowsheet conditions can be achieved in all cases tested. Continued work on the flowsheet has increased our confidence in the economic viability of the sulfur-iodine process.
Date: August 1, 1980
Creator: Besenbruch, G.E.; Allen, C.L.; Brown, L.C.; McCorkle, K.; Rode, J.S.; Norman, J.H. et al.
Partner: UNT Libraries Government Documents Department
open access

HTGR safety philosophy

Description: The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.
Date: August 1, 1980
Creator: Joskimovic, V. & Fisher, C.R.
Partner: UNT Libraries Government Documents Department
open access

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

Description: The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium graphite and release to the coolant stream. The comparisons showed that the cesium profile shape in the graphite web and the partition coefficient between fuel rod matrix material and fuel element graphite were correctly modeled, although the overall release was significantly underpredicted. Uncertainties in the source term (fissile particle failure fraction) and total release to the coolant precluded an accurate appraisal of the validity of FIPERQ. However, several recommendations are presented to improve the applicability of future in-pile test data for the validation of fission metal release codes. The half-life dependence of fission gas release during irradiation was found to be in good agreement with the model used in the reference design materials, providing assurance that this aspect of the fission gas release predictions is properly modeled.
Date: February 1, 1978
Creator: Jensen, D. D.; Haire, M. J. & Ballagny, A.
Partner: UNT Libraries Government Documents Department
open access

Design configuration of GCFR core assemblies

Description: The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the same assemblies would be used in a commercial plant.
Date: May 1, 1980
Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.
Partner: UNT Libraries Government Documents Department
open access

HTGR-GT closed-cycle gas turbine: a plant concept with inherent cogeneration (power plus heat production) capability

Description: The high-grade sensible heat rejection characteristic of the high-temperature gas-cooled reactor-gas turbine (HTGR-GT) plant is ideally suited to cogeneration. Cogeneration in this nuclear closed-cycle plant could include (1) bottoming Rankine cycle, (2) hot water or process steam production, (3) desalination, and (4) urban and industrial district heating. This paper discusses the HTGR-GT plant thermodynamic cycles, design features, and potential applications for the cogeneration operation modes. This paper concludes that the HTGR-GT plant, which can potentially approach a 50% overall efficiency in a combined cycle mode, can significantly aid national energy goals, particularly resource conservation.
Date: April 1, 1980
Creator: McDonald, C. F.
Partner: UNT Libraries Government Documents Department
open access

Radiochemical analysis of the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor

Description: This report presents the analysis of radioactive elements on the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor. The plateout probe is a device which samples the primary coolant for condensible fission products. Circuit inventories of individual radionuclides are estimated from the probe analysis. The analysis shows that the radioactive contamination in the primary circuit is remarkable low, with activation product concentrations much greater than that of fission products. The analysis demonstrates that the concentrations of the key fission products I-131 and Sr-90 are far below the limits allowed by the technical specification.
Date: June 1, 1982
Creator: Burnette, R. D.
Partner: UNT Libraries Government Documents Department
open access

General Atomic reprocessing pilot plant: description and results of initial testing

Description: In June 1976 General Atomic completed the construction of a reprocessing head-end cold pilot plant. In the year since then, each system within the head end has been used for experiments which have qualified the designs. This report describes the equipment in the plant and summarizes the results of the initial phase of reprocessing testing.
Date: December 1, 1977
Partner: UNT Libraries Government Documents Department
open access

HTGR-GT systems optimization studies

Description: The compatibility of the inherent features of the high-temperature gas-cooled reactor (HTGR) and the closed-cycle gas turbine combined into a power conversion system results in a plant with characteristics consistent with projected utility needs and national energy goals. These characteristics are: (1) plant siting flexibility; (2) high resource utilization; (3) low safety risks; (4) proliferation resistance; and (5) low occupational exposure for operating and maintenance personnel. System design and evaluation studies on dry-cooled intercooled and nonintercooled commercial plants in the 800-MW(e) to 1200-MW(e) size range are described, with emphasis on the sensitivity of plant design objectives to variation of component and plant design parameters. The impact of these parameters on fuel cycle, fission product release, total plant economics, sensitivity to escalation rates, and plant capacity factors is examined.
Date: June 1, 1980
Creator: Kammerzell, L.L. & Read, J.W.
Partner: UNT Libraries Government Documents Department
open access

HTGR fuel recycle program. Quarterly progress report for the period ending November 30, 1977

Description: The work reported includes the development of unit processes and equipment for reprocessing of High-Temperature Gas-Cooled Reactor (HTGR) fuel, the design and development of an integrated pilot line to demonstrate the head end of HTGR reprocessing using unirradiated fuel materials, and design work in support of Hot Engineering Tests (HET). Work is also described on trade-off studies concerning the required design of facilities and equipment for the large-scale recycle of HTGR fuels in order to guide the development activities for HTGR fuel recycle.
Date: December 1, 1977
Partner: UNT Libraries Government Documents Department
open access

Negative ion-based neutral injection on DIII-D

Description: High energy negative ion-based neutral beam injection is a strong candidate for heating and non-inductive current drive in tokamaks. Many of the questions related to the physics and engineering of this technique remain unanswered. In this paper, we consider the possibility of negative ion-based neutral beam injection on DIII-D. We establish the desired parameter space by examining physics trades. This is combined with potential design constraints and a survey of component technology options to establish an injector concept. Injector performance is estimated assuming particular component technologies, and concept flexibility with respect to incorporating alternate technologies is described. 9 refs., 6 figs., 4 tabs.
Date: January 1, 1990
Creator: Stewart, L. D.; Bhadra, D. K.; Colleraine, A. P. & Kim, J.
Partner: UNT Libraries Government Documents Department
open access

GCFR core thermal-hydralic design

Description: The approach for developing the thermal-hydraulic core assembly designs for the gas-cooled fast reactor (GCFR) is reviewed, and key considerations for improving the core performance at all power and flow conditions are discussed. It is shown how the thermal-hydraulic core assembly designs evolve from evaluations of plant size, material limitations, safety criteria, and structural performance considerations.
Date: May 1, 1980
Creator: Schleuter, G.; Baxi, C.B. & Bennett, F.O.
Partner: UNT Libraries Government Documents Department
open access

Probabilistic risk assessment of HTGRs

Description: Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed.
Date: August 1980
Creator: Fleming, K. N.; Houghton, W. J.; Hannaman, G. W. & Joksimovic, V.
Partner: UNT Libraries Government Documents Department
open access

End of life fission product distributions in F-1 experiment fuel rods

Description: Fission product migration and end-of-life distributions were examined in the F-1 (X094) series of sealed, mixed-oxide fuel rods irradiated in the fast flux of EBR-II. Cesium, rubidium, iodine, and strontium data obtained from axial gamma scanning, mass spectrometry, and radiochemical analyses are presented. The results show significant migration of cesium, probably as both a volatile species and as the noble gas precursor. Cesium metal species leaving the fuel region accumulate predominately at the fuel-blanket interface. Volatile cesium reaching the fission product traps is readily sorbed by the charcoal.
Date: May 1, 1980
Creator: Goodin, D. T.; Langer, S. & Bell, W. E.
Partner: UNT Libraries Government Documents Department
open access

US GCFR demonstration plant design

Description: A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed.
Date: May 1, 1980
Creator: Hunt, P. S. & Snyder, H. J.
Partner: UNT Libraries Government Documents Department
open access

Personnel Radiation Exposure in HTGR Plants

Description: Occupational radiation exposures in high-temperature gas-cooled reactor (HTGR) plants were assessed. The expected rate of dose accumulations for a large HTGR steam cycle (HTGR-SC) unit is 0.07 man-rem/MW(e)y, while the design basis is 0.17 man-rem/MW(e)y. The comparable figure for actual light water reactor (LWR) experience is 1.3 man-rem/MW(e)y. The favorable HTGR occupational exposure is supported by results from the Peach Bottom Unit No. 1 HTGR and Fort St. Vrain HTGR plants and by operating experience at British gas-cooled reactor (GCR) stations.
Date: January 1, 1980
Creator: Su, S. & Engholm, B. A.
Partner: UNT Libraries Government Documents Department
open access

Design and analysis of PCRV core cavity closure

Description: Design requirements and considerations for a core cavity closure which led to the choice of a concrete closure with a toggle hold-down as the design for the Gas-Cooled Fast Breeder Reactor (GCFR) plant are discussed. A procedure for preliminary stress analysis of the closure by means of a three-dimensional finite element method is described. A limited parametric study using this procedure indicates the adequacy of the present closure design and the significance of radial compression developed as a result of inclined support reaction.
Date: May 1, 1980
Creator: Lee, T.T.; Schwartz, A.A. & Koopman, D.C.A.
Partner: UNT Libraries Government Documents Department
open access

Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents. [HTGR]

Description: A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors (HTGRs).
Date: June 1, 1980
Creator: Houghton, W.J.
Partner: UNT Libraries Government Documents Department
open access

Inherent design features of the GCFR

Description: This paper discusses several inherent design features of the GCFR that enhance its safety and presents analyses to demonstrate the degree of protection they provide. These features are a subset of a larger group of potential inherent features that form the third line of protection (LOP-3) for the GCFR. The function of LOP-3 is to demonstrate that the inherent response of the reactor system will limit core damage even if active cooling and shutdown systems in LOP-1 and LOP-2 fail. By providing this function with inherent features, which do not depend on active components and are self-controlling, an additional level of protection against common cause failure mechanisms is provided for both protected and unprotected events. The examples of LOP-3 discussed in this paper are natural circulation core cooling to the ultimate atmospheric heat sink and inherent reactor shutdown mechanisms.
Date: May 1, 1980
Creator: Medwid, W.; Breher, W.; Shenoy, A. & Elliott, R.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen