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in subjects: "Ruptures"

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Limits to Ductility Set by Plastic Flow Localization

The theory of strain localization is reviewed with reference both to local necking in sheet metal forming processes and to more general three dimensional shear band localizations that sometimes mark the onset of ductile rupture. Both bifurcation behavior and the growth of initial imperfections are considered. In addition to analyses based on classical Mises-like constitutive laws, approaches to localization based on constitutive models that may more accurately model processes of slip and progre… more

Date: November 1, 1977

Creator: Needleman, A. & Rice, J. R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Steam generator secondary pH during a steam generator tube rupture

The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL t… more

Date: December 1, 1991

Creator: Adams, J. P. & Peterson, E. S.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Some recent observations on the radiation behavior of uranium silicide dispersion fuel

Addition of B{sub 4}C burnable poison results in higher plate swelling in both U{sub 3}Si{sub 2} and U{sub 3}Si-Al dispersion fuel plates and also decreases the blister threshold temperature of these plates. Prolonged annealing of U{sub 3}Si{sub 2}-Al fuel plates produced no blister after 696 hours at 400{degrees}C. Blister formation started between 257 hours and 327 hours at 425{degrees}C and between 115 hours and 210 hours at 450{degrees}C. Operation with breached cladding resulted in pillowi… more

Date: January 1, 1988

Creator: Hofman, G.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

Date: January 1, 1991

Creator: Chen, N.C.J.; Yoder, G.L. (Oak Ridge National Lab., TN (USA)) & Wendel, M.W. (Martin Marietta Energy Systems, Inc., Oak Ridge, TN (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood tr… more

Date: October 1, 1977

Creator: Esparza, V. & Sackett, K.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Air flow into the LBL Bevatron

The Lawrence Berkeley Laboratory is currently installing an improved liner in its Bevatron. The new liner will be capable of producing a vacuum of 2 x 10/sup -8/ N/M/sup 2/ (1.5 x 10/sup -10/ Torr) and a temperature on the order of 12K. There has been concern for quite some time about possible damage to the liner in the event of a beam line window breaking allowing atmospheric air to rush into the vacuum. The installation of the new more fragile liner has heightened this concern. This effort is… more

Date: September 1, 1979

Creator: Williams, J.D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Finite element analysis for the initiation of lamellar tearing in welded joints

A numerical procedure using the finite element method is presented for predicting the initiation of lamellar tearing in fillet welded T-joints commonly employed in large structures. Starting with a prescribed geometry, the welding process is approximated by a known time-dependent volumetric heat source which simulates the arc heating and deposition of liquid metal. The transient nonlinear thermal and stress problems are then solved using finite element computer codes. Results of the elastic-pla… more

Date: January 1, 1980

Creator: Krieg, R. D. & Thomas, R. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Aging assessment of auxiliary feedwater systems

A study of Pressurized Water Reactor Auxiliary Feedwater (AFW) Systems has been conducted by Oak Ridge National Laboratory (ORNL) under the auspices of the Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. The study has reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results. 7 figs.

Date: January 1, 1989

Creator: Casada, D.A. (Oak Ridge National Lab., TN (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Probabilistic models of the stress-rupture of composite materials: Phase III. Progress report, June 15, 1978-June 14, 1980

The development of static and time dependent models for the failure of composite materials has been pursued. These models are based on the chain-of-bundles probability model for composite failure wherein local load sharing among non-failed fiber elements in a bundle plays a key role. At the outset, it was believed that the static model was well developed in the literature by others, and that work on the time dependent versions could proceed. In these versions, the viscoelasticity of the matrix … more

Date: March 1, 1980

Creator: Phoenix, S.L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood tr… more

Date: October 1, 1977

Creator: Collins, B.L.; Coppin, C.E. & Sackett, K.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Mechanical properties of types 304 and 316 stainless steel after long-term aging and exposure

Because designs for Liquid Metal Fast Breeder Reactor (LMFBR) power plants include plant lifetimes to 40 years, an understanding of the mechanical behavior of the structural alloys used is required for times of approx. 2 to 2.5 x 10/sup 5/ h. Most of the alloys used for LMFBR out-of-core structures and components are in a metastable state at the beginning of plant lifetime and evolve to a more stable state and, therefore, microstructure during plant operation. We reviewed mechanical properties … more

Date: January 1, 1983

Creator: Horak, J.A.; Sikka, V.K. & Raske, D.T.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Experiment data report for Semiscale Mod-1 Tests S-28-7, S-28-9, and S-28-12. [PWR]

Recorded test data are presented for Tests S-28-7, S-28-9, and S-28-12 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Tests S-28-7, S-28-9, and S-28-12 were conducted from initial conditions of 15 736 kPa and 557 K, 15 754 kPa and 556 K, and 15 704 kPa and 559 K, … more

Date: February 1, 1978

Creator: Esparza, V.; Collins, B.L.; Sackett, K.E. & Coppin, C.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, … more

Date: January 1, 1977

Creator: Lin, E.I.H.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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TRAC analysis and support of Oconee-1 PTS studies. [Pressurized thermal shock following overcooling transients]

This paper describes the overall pressurized thermal shock (PTS) program at Los Alamos with emphasis on TRAC-PF1 calculations of severe overcooling transients for the Oconee-1 pressurized water reactor (PWR). A summary of results for several calculations are presented for the Oconee-1 PWR along with detailed discussions of two of the most severe overcooling transients predicted (main steam-line break and turbine-bypass valve (TBV) failures). The calculations performed were plant specific in tha… more

Date: January 1, 1983

Creator: Ireland, J.R.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Combustion Engineering, Inc. [LMFBR]

Four (4) 3 '' O.D. x 0.470'' nominal wall thickness (NWT) hot rotary pierced/roll reduced modified AOD/ESR tube hollows were cold pilger reduced through one pass to 2'' O.D. x 0.250'' NWT tubing. Two (2) additional hollows of same size and process history were cold pilger reduced through one pass to 2 1/8'' O.D. x 0.200'' NWT. Six (6) 3 3/4'' O.D. x 0.600'' NWT hot extruded tube hollows were cold pilger reduced through two passes to 2'' O.D. x 0.250'' NWT tubing. Four of the extrusions represen… more

Date: January 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Status of RBCB testing of LMR oxide fuel in EBR-II

The status is given of the the American-Japanese collaborative program in Experimental Breeder Reactor 2 to determine the run-beyond-cladding-breach performance of (UPu)O{sub 2} fuel pins for liquid-metal cooled reactors. Phase 1 of the collaboration involved eighteen irradiation tests over 1981--86 with 5.84-mm pins in 316 or D9 stainless steel. Emphasis in Phase 2 tests from 1989 onwards is with larger diameter (7.5mm) pins in advanced claddings. Results include delayed neutron and fission ga… more

Date: January 1, 1991

Creator: Strain, R.V.; Bottcher, J.H.; Gross, K.C.; Lambert, J.D.B. (Argonne National Lab., IL (United States)); Ukai, S.; Nomura, S. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Progress toward analytical description of the creep strain-time behavior of engineering alloys

Elevated-temperature design methods in the United States often require a comprehensive description of the properties of the construction materials. These descriptions include representations for creep strain-time behavior as a function of stress, temperature, and material variability. Work conducted at this laboratory in the past five years toward the development of analytical techniques to derive such representations is summarized. Results for several common elevated-temperature structural mat… more

Date: January 1, 1980

Creator: Booker, M. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Prediction of long-term failure in Kevlar 49 composites

Creep rupture data in Kevlar 49 epoxy usually exhibit considerable scatter: the coefficient of variation (CV) about the mean failure time at a given stress exceeds 100%. Quasi-static strength data, in contrast, shows little scatter: <4% CV for pressure vessels and <10% for impregnated strands. In this paper analysis of existing creep rupture data on Kevlar epoxy vessels at four storage pressures has produced an interesting and useful result. It was found that a significant portion of the scatte… more

Date: January 1, 1982

Creator: Gerstle, F.P. Jr.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Experiment data report for Semiscale Mod-1 test S-28-6 (steam generator tube rupture test)

Recorded test data are presented for Test S-28-6 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-6 was conducted from initial conditions of 15,770 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood tr… more

Date: November 1, 1977

Creator: Patton, M. L.; Sackett, K. E. & Coppin, C. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Reliability of fast reactor mixed-oxide fuel during operational transients

Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 f… more

Date: July 1, 1991

Creator: Boltax, A.; Neimark, L.A.; Tsai, Hanchung (Argonne National Lab., IL (United States)); Katsuragawa, M. & Shikakura, S. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Deformation and Fracture Map Methodology for Predicting Cladding Behavior During Dry Storage

The licensing of interim dry storage of light-water reactor spent fuel requires assurance that release limits of radioactive materials are not exceeded. The extent to which Zircaloy cladding can be relied upon as a barrier to prevent release of radioactive spent fuel and fission products depends upon its integrity. The internal pressure from helium and fission gases could become a source of hoop stress for creep rupture if pressures and temperatures were sufficiently high. Consequently, it is o… more

Date: September 1, 1986

Creator: Chin, B. A.; Khan, M. A. & Tarn, J. C. L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Feasibility of methods and systems for reducng LNG tanker fire hazards

In this program concepts for reducing fire hazards that may result from LNG tanker collisions are identified and their technical feasibility evaluated. Concepts considered include modifications to the shipborne LNG containers so that in the event of a container rupture less of the contents would spill and/or the contents would spill at a reduced rate. Changes in the cargo itself, including making the LNG into a gel, solidifying it, converting it to methanol, and adding flame suppressants are al… more

Date: August 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; condi… more

Date: January 1, 1985

Creator: Nassersharif, B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Failure criteria used in a probabilistic fracture mechanics code

Two criteria are implemented in a piping reliability analysis code to assess the stability of crack growth in pipes. One is the critical net section stress criterion. It is simple and convenient but its application is limited to very ductile materials. The other is the tearing modulus stability criterion. This criterion has a solid technical base. However, calculating the J-integral, J, and the associated tearing modulus, T, usually requires a complicated finite element method (FEM). In this pi… more

Date: January 1, 1985

Creator: Lo, T.Y.

Item Type: Article

Partner: UNT Libraries Government Documents Department