Open data from the third observing run of LIGO, Virgo, KAGRA, and GEO

The global network of gravitational-wave observatories now includes five detectors, namely LIGO Hanford, LIGO Livingston, Virgo, KAGRA, and GEO 600. These detectors collected data during their third observing run, O3, composed of three phases: O3a starting in 2019 April and lasting six months, O3b starting in 2019 November and lasting five months, and O3GK starting in 2020 April and lasting two weeks. In this paper we describe these data and various other science products that can be freely accessed through the Gravitational Wave Open Science Center at https://gwosc.org. The main data set, consisting of the gravitational-wave strain time series that contains the astrophysical signals, is released together with supporting data useful for their analysis and documentation, tutorials, as well as analysis software packages

REED BESLER BOILER HIGH PRESSURE STEAM SYSTEM AND THERMAL CYCLING FACILITY. Summary Report

A high pressure boiler has been installed at ORNL. This Besler boiler is capabie of producing from 150 to 2000 psi saturated steam at steaming rates up to 5000 lbs/hr. The boiler is part of a water-steam circuit whteh also includes two spray water pumps, a steam pressure control valve, a high pressure trapping station, and a low pressure deaerated feedwater system. The new boiler system is piped and instrumented to serve as a thermal cycling facility. Shakedown test thermal cycles to requirements set forth in HRT Specification 1113a have been conducted using the existing Dump Test Autoclave as a test piece. Fourty-four cycles have been run through mid February, 1958. The boiler has been operated a total of 142 hours. Cycles are run completely automatically. Better than three- fourths of the cycles as run fall within the specification prescribed limits. (auth

Half-bead weld repairs for in-service applications

Successful half- or temper-bead technique weld repairs performed to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code guidelines were made to two Heavy-Section Steel Technology Program vessels and a qualification prolongation. Intermediate sized vessels, equivalent in thickness to nuclear pressure vessels, were repair welded and subsequently flawed and pressure tested to approximately 2/sup 1///sub 4/ times design pressure before leakage occurred. Discussed are the standards and procedures used with half-bead repairs, resultant induced metallurgical and stress effects, flaw test criterion, pressure test details and results, and recommendations for further development work for a speedier application process

Repair weld induced residual stresses in thick-walled steel pressure vessels

If a flaw requiring corrective action were to be found in an operating nuclear pressure vessel, there would be considerable safety and economic implications. Should such a flaw be found, one possible corrective action would be an in situ repair weld. A repair of this type would presumably involve grinding away material in a region encompassing the flaw and then filling the resulting cavity with weld metal. Thermal stress relieving under those conditions could lead to serious difficulties associated with thermal expansion and warpage and would therefore most likely be avoided. Such a departure from normal procedure raises questions relating to residual stresses and material toughness levels which would have to be assessed before a repair could be recommended or approved. The residual stress measurements reported are intended to provide baseline information to aid in an assessment should such a repair ever have to be seriously considered

Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed

Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading. [BWR; PWR]

HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7