90. ARAIII. GCRE reactor building (ARA608) mechanical loop pit. Shows ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

90. ARA-III. GCRE reactor building (ARA-608) mechanical loop pit. Shows nitrogen gas compressor in foreground, piping installations on walls of pit, and other details. February 24, 1959. Ineel photo no. 59-880. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

51. ARAII. Camera looking southeast at foundation piers for SL1 ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

51. ARA-II. Camera looking southeast at foundation piers for SL-1 reactor building support. August 22, 1957. Ineel photo no. 57-4212. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

86. ARAIII. GCRE reactor building (ARA608) showing mechanical loop pit ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

86. ARA-III. GCRE reactor building (ARA-608) showing mechanical loop pit after building shell had been erected. Beyond pit are demineralized water surge tank and heat exchanger. Camera facing northeast. December 22, 1958. Ineel photo no. 58-6427. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

89. ARAIII. Petrochem oilfired gas heater installed in reactor building ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

89. ARA-III. Petro-chem oil-fired gas heater installed in reactor building (ARA-608). View is at floor level. Shows hand rails around heater pit and top of pit extending upwards through ceiling. January 20, 1959. Ineel photo no. 59-321. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

141. ARAIII Equipment location plan. Includes list of equipment and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

141. ARA-III Equipment location plan. Includes list of equipment and location in reactor, control, and other buildings. Aerojet-general 880-area/GCRE-101-U-1. Date: February 1958. Ineel index code no. 063-0101-65-013-192508. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

126. ARAII Plot plan showing location of SL1 power plant ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

126. ARA-II Plot plan showing location of SL-1 power plant (reactor) building, and planned location of administrative and technical support building. C.A. Sundberg and Associates 866-area/ALPR-606-U-1. Date: May 1958. Ineel index code no. 070-0100-00-822-102834. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

71. ARAII. Construction progress at SL1 site near end of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

71. ARA-II. Construction progress at SL-1 site near end of 1957. Buildings from right to left are guard house, support building, reactor building, water tank and pump house. Construction was 23 percent complete. December 20, 1957. Ineel photo no. 57-6224. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

88. ARAIII. "Petrochem" heater is hoisted over south exterior wall ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

88. ARA-III. "Petro-chem" heater is hoisted over south exterior wall of heater pit in GCRE reactor building (ARA-608). Printing on heater says, "Petro-chem iso-flow furnace; American industrial fabrications, inc." Camera facing north. January 7, 1959. Ineel photo no. 529-124. Photographer: Ken Mansfield. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

91. ARAIII. GCRE reactor building (ARA608) at 48 percent completion. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

91. ARA-III. GCRE reactor building (ARA-608) at 48 percent completion. Camera faces west end of building; shows roll-up door. High bay section on right view. Petro-chem heater stack extends above roof of low-bay section on left. Excavation for 13, 8 kv electrical conduit in foreground. January 20, 1959. Ineel photo no. 59-322. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

42. ARAIII Prototype assembly and evaluation building ARA630 interior. Typical ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

42. ARA-III Prototype assembly and evaluation building ARA-630 interior. Typical view room partitions. Ineel photo no. 3-28. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

74. ARAII. Dr. William Zinn of combustion engineering company and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

74. ARA-II. Dr. William Zinn of combustion engineering company and others at controls of SL-1. August 8, 1959. Ineel photo no. 59-4109. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

134. ARAII SL1 decontamination and lay down building (ARA614) erected ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

134. ARA-II SL-1 decontamination and lay down building (ARA-614) erected after accidental explosion of SL-1 reactor. Shows vicinity map, index of related drawings, plot plan and other detail. F.C. Torkelson Company 842-area/SL-1-101-U-2. Date: September 1962. Ineel index code no. 070-0101-65-851-150713. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

41. ARAIII Prototype assembly and evaluation building ARA630. West end ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

41. ARA-III Prototype assembly and evaluation building ARA-630. West end and south side of building. Camera facing northeast. Ineel photo no. 3-22. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

104. ARAIII. Interior view of room 110 in ARA607 used ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

104. ARA-III. Interior view of room 110 in ARA-607 used as data acquisition control room. Camera facing northeast. Ineel photo no. 81-103. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

92. ARAIII. Overall view of GCRE area in 1959. From ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

92. ARA-III. Overall view of GCRE area in 1959. From left to right: ARA-607 (control building), ARA-608 (with high-bay, reactor building), ARA-610 (service building), ARA-609 (guard house), ARA-709 (water storage tank) ARA-710 in front of ARA-709 (fuel oil tank), ARA-611 (well pumphouse), and the cooling tower. Note petro-chem stack and other stacks emerging from reactor building. Camera facing northeast. August 1959. Ineel photo no. 59-4444. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

8. ARAI Shop and maintenance building ARA627 interior view. Remains ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

8. ARA-I Shop and maintenance building ARA-627 interior view. Remains of cabinetry and electrical switch panel in one of rooms. Ineel photo no. 1-11. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

40. ARAIII Prototype assembly and evaluation building ARA630. East end ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

40. ARA-III Prototype assembly and evaluation building ARA-630. East end and south side of building. Camera facing west. Roof railing is part of demolition preparations. Building beyond ARA-622 is ARA-621. In left of view is reactor building. ARA-607 is low-roofed portion, while high-bay portion is ARA-608. Ineel photo no. 3-27. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

105. ARAIII. Interior view of ARA608 highbay pit in 1983 ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

105. ARA-III. Interior view of ARA-608 high-bay pit in 1983 modified to contain high-temperature, high-pressure autoclave and furnace test area. Ineel photo no. 81-109. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

146. ARAIII Control building (ARA607) Roof plan and details. Aerojetgeneral ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

146. ARA-III Control building (ARA-607) Roof plan and details. Aerojet-general 880-area/GCRE-607-A-3. Date: February 1958. Ineel index code no. 063-0607-00-013-102548. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

170. ARAIV Blast bunker installed after ML1 buildings were removed. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

170. ARA-IV Blast bunker installed after ML-1 buildings were removed. Isometric detail and section. EG&G Company. Date: June 1985. Ineel index code no. 066-0600-60-220-166261. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

166. ARAIII Fire hose houses (Probably numbered on site as ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

166. ARA-III Fire hose houses (Probably numbered on site as ARA-624). Aerojet-general 880-area/GCRE-701-S-4. Date: February 1958. Ineel index code no. 063-0624-00-013-102695. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

123. ARAI Substation (ARA726) plan, elevation, security fence details, and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

123. ARA-I Substation (ARA-726) plan, elevation, security fence details, and sections. Norman Engineering Company 961-area/SF-726-E-1. Date: January 1959. Ineel index code no. 068-0726-10-613-102778. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

144. ARAIII Control building (ARA607) Foundation plan. Aerojetgeneral 880area/GCRE607S1. Date: ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

144. ARA-III Control building (ARA-607) Foundation plan. Aerojet-general 880-area/GCRE-607-S-1. Date: February 1958. Ineel index code no. 063-0607-60-013-102568. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

2013-09-25

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in amore » remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.« less

SL-1 Accident Briefing Report - 1961 Nuclear Reactor Meltdown Educational Documentary

ScienceCinema

None

2018-01-16

U.S. Atomic Energy Commission (Idaho Operations Office) briefing about the SL-1 Nuclear Reactor Meltdown. The SL-1, or Stationary Low-Power Reactor Number One, was a United States Army experimental nuclear power reactor which underwent a steam explosion and meltdown on January 3, 1961, killing its three operators. The direct cause was the improper withdrawal of the central control rod, responsible for absorbing neutrons in the reactor core. The event is the only known fatal reactor accident in the United States. The accident released about 80 curies (3.0 TBq) of Iodine-131, which was not considered significant due to its location in a remote desert of Idaho. About 1,100 curies (41 TBq) of fission products were released into the atmosphere. The facility, located at the National Reactor Testing Station approximately 40 miles (64 km) west of Idaho Falls, Idaho, was part of the Army Nuclear Power Program and was known as the Argonne Low Power Reactor (ALPR) during its design and build phase. It was intended to provide electrical power and heat for small, remote military facilities, such as radar sites near the Arctic Circle, and those in the DEW Line. The design power was 3 MW (thermal). Operating power was 200 kW electrical and 400 kW thermal for space heating. In the accident, the core power level reached nearly 20 GW in just four milliseconds, precipitating the reactor accident and steam explosion.

132. ARAII Administration building (ARA613) elevations of north, south, east, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

132. ARA-II Administration building (ARA-613) elevations of north, south, east, and west sides. F. C. Torkelson Company 842-area/SL-1-613-A-2. Date: October 1959. Ineel index no. 070-0613-00-851-150054. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

119. ARAI Shop and maintenance (ARA627) building sections and details ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

119. ARA-I Shop and maintenance (ARA-627) building sections and details of interior mesh partitions. Norman Engineering Company 961-area/SF-627-A-3. Date: January 1959. Ineel index code no. 068-0627-00-613-102761. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

117. ARAI Shop and maintenance (ARA627) building roof and floor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

117. ARA-I Shop and maintenance (ARA-627) building roof and floor plan. Includes room finish and equipment schedule. Norman Engineering Company 961-area/SF-627-A-1. Date: January 1959. Ineel index code no. 068-0627-00-613-102759. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

140. ARAIII Grading and drainage plan showing plot plan, including ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

140. ARA-III Grading and drainage plan showing plot plan, including berms around waste storage tank and fuel oil storage tank. Aerojet-general 880-area-GCRE-101-1. Date: February 1958. Ineel index code no. 063-0101-00-013-102507. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

112. ARAI Hot cell (ARA626) Building roof plan and details ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

112. ARA-I Hot cell (ARA-626) Building roof plan and details of roof ventilating equipment and parapet. Norman Engineering Company: 961-area/SF-626-A-2. Date: January 1959. Ineel index code no. 068-0626-00-613-102722. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

168. ARAIV Index of drawings prepared by Norman Engineering Company ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

168. ARA-IV Index of drawings prepared by Norman Engineering Company in preparation for construction of ARA-IV. Norman Engineering Company 961-area/ML-1index. Date: March 1961. Ineel index code no. 066-9999-90-613-102731. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

147. ARAIII Control building (ARA607) Detail of instrument rack and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

147. ARA-III Control building (ARA-607) Detail of instrument rack and control console in control room. Aerojet-general 880-area/GCRE-607-E-4. Date: November 1958. Ineel index code no. 063-0607-10-013-102560. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

160. ARAIII Service building (ARA610). Includes floor plan, north and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

160. ARA-III Service building (ARA-610). Includes floor plan, north and west elevations, and section details. Aerojet-general 880-area/GCRE-610-A-1. Date: February 1958. Ineel index code no. 063-0610-00-013-102684. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

44. ARAIII Fuel oil tank ARA710. Camera facing west. Perimeter ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

44. ARA-III Fuel oil tank ARA-710. Camera facing west. Perimeter fence at left side of view. Gable-roofed building beyond tank on right is ARA-622. Gable-roofed building beyond tank on left is ARA-610. Ineel photo no. 3-16. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

114. ARAI Hot cell (ARA626) Building details of fuel storage ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

114. ARA-I Hot cell (ARA-626) Building details of fuel storage pit in plan and section. Spaces shown for 20 elements. Norman Engineering Company: 961-area/SF-626-S-4. Date: January 1959. Ineel index code no. 068-0626-60-613-102752. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

72. ARAII. Interior view in ARA602 support building showing oilfired ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

72. ARA-II. Interior view in ARA-602 support building showing oil-fired hot air furnace and hot water boiler in foreground; hot water tank and diesel generator in background. December 12, 1957. Ineel photo no. 57-6099. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

129. ARAII Administrative and technical support building (ARA606) sections showing ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

129. ARA-II Administrative and technical support building (ARA-606) sections showing roof and wall details and longitudinal section. C.A. Sundberg and Associates 866-area/ALPR-606-A-5. Date: May 1958. Ineel index code no. 070-0606-00-822-102828. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

142. ARAIII General plan of GCRE area, including electrical distribution ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

142. ARA-III General plan of GCRE area, including electrical distribution plan for power and lighting. Includes detail of floodlight and security lighting poles and fixtures. Aerojet-general 880-area/GCRE-406-1. Date: February 1958. Ineel index code no. 063-0406-00-013-102539. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

121. ARAI Guard house (ARA628). Drawing shows north, south, east, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

121. ARA-I Guard house (ARA-628). Drawing shows north, south, east, and west elevations, floor plan, counter details, and roof plan. Norman Engineering Corporation 961-area/SF-628-A-1. Date: January 1959. Ineel index code no. 063-0628-00-613-102772. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

145. ARAIII Control building (ARA607) Sections. Shows highbay section of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

145. ARA-III Control building (ARA-607) Sections. Shows high-bay section of building over crane rail and beam. Indicates materials of exterior siding. Aerojet-general 880-area/GCRE-607-A-11. Date: February 1958. Ineel index code no. 063-0607-00-013-102556. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

128. ARAII Administrative and technical support building (ARA606) elevations for ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

128. ARA-II Administrative and technical support building (ARA-606) elevations for northwest, southwest, northeast, and southeast sides. C.A. Sundberg and Associates 866-area/ALPR-606-A-3. Date: May 1958. Ineel index code no. 070-0606-00-822-102826. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

47. ARAI. Interior view of operating wall of hot cell ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

47. ARA-I. Interior view of operating wall of hot cell in ARA-626. Note stands for operators at viewing windows. Manipulators with hand grips extend cables and other controls into hot cell through ducts above windows. Ineel photo no. 81-27. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

77. ARAII. Room at northeast corner of ARA606 used for ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

77. ARA-II. Room at northeast corner of ARA-606 used for welding training and welding procedure qualification and performance testing. This was only building in use at ARA-II by 1983. Date: 1983. Ineel photo no. 83-476-3-5. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

138. ARAII Building ARA606 floor plan for remodel as Inel ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

138. ARA-II Building ARA-606 floor plan for remodel as Inel Welding Laboratory. Shows room divisions and welding stations to be installed. Aerojet Nuclear Company 1375-ARA-II-606-E-2. Date: June 1976. Ineel index code no. 070-0606-10-400-156552. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

110. ARAI support facilities. Index of drawings related to initial ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

110. ARA-I support facilities. Index of drawings related to initial construction of hot cell building ARA-626, shop and maintenance building ARA-627, and other buildings at ARA-I. Date: Circa January 1959. Norman Engineering Company. Ineel index code no. 068-9999-80-613-102703. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

116. ARAI Details of hot cell section of building ARA626. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

116. ARA-I Details of hot cell section of building ARA-626. Shows manipulator openings in operating face of hot cell, start/stop buttons, and other details. Norman Engineering Company 961/area/SF-626-E-6. Date: January 1959. Ineel index code no. 068-0626-10-613-102731. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

137. ARAII Building ARA602 floor plan as it appeared in ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

137. ARA-II Building ARA-602 floor plan as it appeared in 1980 when electrical modifications were being made. Shows partial layout of floor plan. EG&G Idaho, Inc. 1570-ARA-II-602-E-3. Date: April 1980. Ineel index code no. 070--0602-10-220-159761. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

111. ARAI Hot cell (ARA626) Building elevations of north, south, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

111. ARA-I Hot cell (ARA-626) Building elevations of north, south, east, and west sides. Includes details of personnel decontamination area, dark room, and other features. Norman Engineering Company: 961-area/SF-626-A-3. Date: January 1959. Ineel index code no. 068-0626-00-613-102723. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

118. ARAI Shop and maintenance (ARA627) building elevations of north, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

118. ARA-I Shop and maintenance (ARA-627) building elevations of north, south, east, and west sides and other details of door and window types. Norman Engineering Company 961-area/SF-627-A-2. Date: January 1959. Ineel index code no. 068-0627-00-613-102760. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

127. ARAII Administrative and technical support building (ARA606) ground floor ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

127. ARA-II Administrative and technical support building (ARA-606) ground floor plan. Indicates use of rooms for classrooms, offices, and lunch room. C.A. Sundberg and Associates 866-area-ALPR-606-A-2. Date: June 1958. Ineel index code no. 070-0606-00-822-102825. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

135. ARAII SLI decontamination and lay down building (ARA614) north, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

135. ARA-II SL-I decontamination and lay down building (ARA-614) north, south, east, and west elevations, floor plan, and detail of doors. F.C. Torkelson Company 842-area/SL-1-614-A-1. Date: September 1960. Ineel index code no. 070-0614-00-851-150061. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

125. ARAI Contaminated waste storage tank (ARA729). Shows location of ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

125. ARA-I Contaminated waste storage tank (ARA-729). Shows location of tank on the ARA-I site, section views, connecting pipeline, and other details. Norman Engineering Company 961-area/SF-301-3. Date: January 1959. Ineel index code no. 068-0301-00-613-102711. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

143. ARAIII Control building (ARA607) Floor plan. Shows control room, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

143. ARA-III Control building (ARA-607) Floor plan. Shows control room, contaminated work area, counting and computer room, health physics room, instrument repair room, offices, and other rooms. Aerojet-general 880-area/GCRE-607-A-1. Date: February 1958. Ineel index code no. 063-0607-00-013-102546. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

124. ARAI Reservoir (ARA727), later named water storage tank. Shows ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

124. ARA-I Reservoir (ARA-727), later named water storage tank. Shows plan of 100,000-gallon tank, elevation, image of "danger radiation hazard" sign, and other details. Norman Engineering Company 961-area/SF-727-S-1. Date: January 1959. Ineel index code no. 068-0727-60-613-102779. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

131. ARAII Administration building (ARA613) floor plans for first and ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

131. ARA-II Administration building (ARA-613) floor plans for first and second floors. Includes roof plan. Shows use of rooms as offices, laboratory, conference room. F.C. Torkelson Company 842-area/SL-1-613-A-1. Date: October 1958. Ineel index code no. 070-0613-00-851-150718. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

115. ARAI Details of hot cell section of building ARA626. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

115. ARA-I Details of hot cell section of building ARA-626. Shows location of high density concrete, viewing windows, filters, monorail crane, bridge crane, and other details. Norman Engineering Company 961-area/SF-626-MS-1. Date: January 1959. Ineel index code no. 068-0626-40-613-102737. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

133. ARAII SL1 burial ground. Shows gravel path from ARAII ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

133. ARA-II SL-1 burial ground. Shows gravel path from ARA-II compound to the burial ground, detail of security fence and entry gate, and sign "Danger radiation hazard." F. C. Torkelson Company 842-area-101-1. Date: October 1961. Ineel index code no. 059-0101-00-851-150723. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Fuel Processor Development for a Soldier-Portable Fuel Cell System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Palo, Daniel R.; Holladay, Jamie D.; Rozmiarek, Robert T.

2002-01-01

Battelle is currently developing a soldier-portable power system for the U.S. Army that will continuously provide 15 W (25 W peak) of base load electric power for weeks or months using a micro technology-based fuel processor. The fuel processing train consists of a combustor, two vaporizers, and a steam-reforming reactor. This paper describes the concept and experimental progress to date.

16. ARAII Administration building ARA613. South (front) and east sides. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

16. ARA-II Administration building ARA-613. South (front) and east sides. Camera facing northwest. Sign at left corner of building says, "Fuels and materials division, materials joining research and development laboratory." Part of south wall already has been demolished. Sign on roof railing says, "Danger--Abestos." Ineel photo no. 2-3. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

122. ARAI Pump House (ARA629). Drawing shows north, south, east, ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

122. ARA-I Pump House (ARA-629). Drawing shows north, south, east, and west elevations, floor plan, foundation plan, and other details. Note small enclosure at southwest corner of building to contain chlorination equipment. Norman Engineering Company 961-area/SF-629-A-1. Date: January 1959. Ineel index code no. 068-0629-00-613-102774. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

136. ARRII Plot plan as it appeared in 1980, when ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

136. ARR-II Plot plan as it appeared in 1980, when interior modifications were being prepared to remodel electrical apparatus in ARA-602 in connection with use as a research and development joining laboratory. EG&G, Idaho, Inc. 1570-ARA-II-100-1. Date: April 1980. Ineel index code no. 070-0199-00-220-159749. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

130. ARAII Administration building (ARA613) vicinity map and plot plan ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

130. ARA-II Administration building (ARA-613) vicinity map and plot plan showing relationship to other existing buildings on site and to ARA-602, to which this building was attached. F.C. Torkelson Comapny 842-area/SL-1-101-U-1. Date: October 1959. Ineel index code no. 070-0101-65-851-150053. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

113. ARAI Hot cell (ARA626) Building wall sections and details ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

113. ARA-I Hot cell (ARA-626) Building wall sections and details of radio chemistry lab. Shows high-bay roof over hot cells and isolation rooms below grade storage pit for fuel elements. Norman Engineering Company: 961-area/SF-626-A-4. Date: January 1959. Ineel index code no. 068-0626-00-613-102724. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

120. ARAI Expansion of ARA627 shop and maintenance building for ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

120. ARA-I Expansion of ARA-627 shop and maintenance building for new use as materials and metallurgy laboratory. Shows ground floor plan addition of gas analyzer room, fatigue testing room, microscope room, and offices. Idaho Nuclear Corporation 1230-ARA-627-A-5. Date: June 1970. Ineel index code no. 068-0627-00-400-154062. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

46. ARAI. Aerial view of ARAI buildings as they looked ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

46. ARA-I. Aerial view of ARA-I buildings as they looked in 1981. From left to right, buildings are tank (ARA-727), contaminated waste storage tank (ARA-629), trailer, hot cell building (ARA-626), fuel oil storage tank (ARA-728), guard house (ARA-628), shop and maintenance building (ARA-627), and two trailers. Ineel photo no. 81-297. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

Small Modular Reactors: The Army’s Secure Source of Energy?

DTIC Science & Technology

2012-03-21

significant advantages of SMRs is the minimal amount of carbon dioxide (greenhouse gases) that is released in conjunction with the lifecycle operations...moderator in these reactors as well as the cooling agent and the means by which heat is removed to produce steam for turning the turbines of the...separate water system to generate steam to turn a turbine which then produces electricity. In the second type of light water reactors, the boiling water

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Photocopy of drawing, February 1958, NUCLEAR REACTOR FACILITY, STRUCTURAL CROSS SECTION. Giffals & Vallet, Inc., L. Rosetti, Associated Architects and Engineers, Detroit, Michigan; and U.S. Army Engineer Division, New England Corps of Engineers, Boston, Massachusetts. Drawing Number 35-84-04. (Original: AMTL Engineering Division, Watertown). - Watertown Arsenal, Building No. 100, Wooley Avenue, Watertown, Middlesex County, MA

A laboratory treatability study on RDX-contaminated soil from the Iowa Army Ammunition Plant, Burlington, Iowa.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boopathy, R.; Manning, J. F.; Environmental Research

2000-03-01

Soil in certain areas of the Iowa Army Ammunition Plant in Burlington, Iowa, was contaminated with hexahydro-1,3,5-trinitro-1,3,5-triazine (RDX). A laboratory treatability study was conducted to examine the ability of native soil bacteria present in the contaminated site to degrade RDX. The results indicated that RDX can be removed effectively from the soil by native soil bacteria through a co-metabolic process. Molasses, identified as an effective cosubstrate, is inexpensive, and this factor makes the treatment system cost effective. The successful operation of aerobic-anoxic soil-slurry reactors in batch mode with RDX-contaminated soil showed that the technology can be scaled up for fieldmore » demonstration. The RDX concentration in the contaminated soil was decreased by 98% after 4 months of reactor operation. The advantage of the slurry reactor is the simplicity of its operation. The method needs only mixing and the addition of molasses as cosubstrate.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

Progress is reported on fundamental research in: crystal physics, reactions at metal surfaces, spectroscopy of ionic media, structure of metals, theory of alloying, physical properties, sintering, deformation of crystalline solids, x ray diffraction, metallurgy of superconducting materials, and electron microscope studies. Long-randge applied research studies were conducted for: zirconium metallurgy, materials compatibility, solid reactions, fuel element development, mechanical properties, non-destructive testing, and high-temperature materials. Reactor development support work was carried out for: gas-cooled reactor program, molten-salt reactor, high-flux isotope reactor, space-power program, thorium-utilization program, advanced-test reactor, Army Package Power Reactor, Enrico Fermi fast-breeder reactor, and water desalination program. Other programmore » activities, for which research was conducted, included: thermonuclear project, transuraniunn program, and post-irradiation examination laboratory. Separate abstracts were prepared for 30 sections of the report. (B.O.G.)« less

Army Institutional Training: Current Status and Future Research

DTIC Science & Technology

2010-03-01

one-group posttest 2. two-group posttest only with nonequivalent comparison group 3. two-group pretest - posttest with nonequivalent comparison group...multiple posttests ii. What problems have you encountered in conducting this type of experimental/quasi-experimental research ? 10. If you were...U.S. Army Research Institute for the Behavioral and Social Sciences Research Report 1921 Army Institutional Training: Current

Foreign Policy Benefits from Subsidization of Trade with Eastern Europe

DTIC Science & Technology

1989-02-01

AFUDC, the projected cost per kilowatt is $2440. A reactor containment for a 1000 MW pressur - ized water reactor costs about $100 million;96 let us ...diffprencpe in interests between the Soviet Union and its East European allies in the Warsaw Pact. It examines the use of economic policy by the West as a...instead to Soviet armies, fronts, or theaters of military operations (TVDs). The Groups of Soviet Forces are stationed in Eastern Europe in part in an

Laboratory Studies and Preliminary Evaluation of Destructive Technologies for the Removal of RDX from the Water Waste Stream of Holston Army Ammunition Plan

DTIC Science & Technology

2010-05-01

with electrode plates . ERDC/EL TR-10-4 29 Total Electrode Surface Area = 0.4311 m² Cathodic Surface Area = 0.2156 m² Reactor Volume = 1632 mL 35...27 Figure 15. Continuous flow electrochemical reactor packed with electrode plates . .......................... 28 Figure...Environmental Compliance (TDEC) is in the process of establishing a total maximum daily load (TMDL) that will regulate the mass of hexahydro- 1,3,5

Full Spectrum Operations: Is This the Science of Victory?

DTIC Science & Technology

2011-05-19

Secretary of the Army Louis Caldera and Army Chief of Staff General Eric Shinseki proposed yet another experimental operating concept, the Objective Force...19-22. 124 Adams, The Army After Next, 54. 125 Ibid., 68. 126 Louis Caldera and Eric K. Shinseki, “Army Vision: Soldiers On Point for the Nation...the United States of America. Washington, DC: Government Printing Office, 2006. Caldera , Louis and Eric K. Shinseki. “Army Vision: Soldiers On

NDI (Nondestructive Inspection) Oriented Corrosion Control for Army Aircraft. Phase 1. Inspection Methods

DTIC Science & Technology

1989-07-01

Appendices A and B and are provided as cover sheets from each item rather than completc packages. The Pamplet Series materials were furnished as camera-ready...34Stational Neutron Radiography System for Aircraft Reliability and Maintainability." G. A. Technologies Brochure , Triga Reactor Division, San Diego

Experimental Determination of Physical Properties of DNGU, TNBA, LLM-105, HK-56, and DNP

DTIC Science & Technology

2016-09-01

ARL-TN-0788 ● SEP 2016 US Army Research Laboratory Experimental Determination of Physical Properties of DNGU, TNBA, LLM-105, HK...NOTICES Disclaimers The findings in this report are not to be construed as an official Department of the Army position unless so designated by... Experimental Determination of Physical Properties of DNGU, TNBA, LLM-105, HK-56, and DNP by Rose A Pesce-Rodriguez Weapons and Materials

Work Program. Fiscal Year 1972 for The Department of the Army. Research and Development in Training, Motivation, and Leadership

DTIC Science & Technology

1971-07-01

itself. IlumRRO assistance was requested by the Infantry School both for design of experimental tests and for analysis and interpretation of the data from...of Research. To develop an experimental Army literacy training program designed to provide a level of functional literacy appropriate to present...assigned under the provisions of a long- range program (up to two years in duration) designed by tlumRRO. Specially designed experimental

An Operational Utility Assessment: Measuring the Effectiveness of the Experimental Forward Operating Base Program

DTIC Science & Technology

2014-06-01

SPACES product brochure . Retrieved from https://www.iristechnology.com/manuals/BR-Iris-SPACES.pdf Jameson, LLC. (2014, April 11). EMI hardened LED...Army. (2010). TRADOC generating force study (TRADOC Pamphlet 525-8-1). Retrieved from http://www.tradoc.army.mil/tpubs/pams/tp525-8-1.pdf U.S. Army

Development and Field Test of the Trial Battery for Project A. Improving the Selection, Classification and Utilization of Army Enlisted Personnel. Project A: Improving the Selection, Classification and Utilization of Army Enlisted Personnel. ARI Technical Report 739.

ERIC Educational Resources Information Center

Peterson, Norman G., Ed.

As part of the United States Army's Project A, research has been conducted to develop and field test a battery of experimental tests to complement the Armed Services Vocational Aptitude Battery in predicting soldiers' job performance. Project A is the United States Army's large-scale manpower effort to improve selection, classification, and…

In-Flight Performance Evaluation of Experimental Information Displays

DTIC Science & Technology

1979-05-01

Chemical Systems Laboratory Experimentation Command Aberden Proving Ground ,MD Technical Library 21010 (1) Box 22 Fort Ord, CA 93941 (1) 21 US Amy Materiel...US Army Missile R&D Command Library, Bldg 3071 Redstone Arsenal, AL 35809 (1) ATTN: ATSL-DOSL Aberdeen Proving Ground , MD US Army Yuma Proving Ground ...Systems Chief Analysis Agency Benet Weapons Laboratory ATTN: Reports Distribution LCWSL, USA ARRADCOH Aberdeen Proving Ground , MD ATTN: DRDAR-LCB-TL

Army Field-Oriented S&T Experimentation Venues: A Comparative Analysis

DTIC Science & Technology

2011-09-01

Microclimate Cooling Station (MCCS)). The Fort Benning AEWE provides the venue and the data collection and analysis. The costs to the S&T...forest, fields, etc.) and is designated as an Army experimental station with access to ground and an aerial fleet. Technology developers have optional...YTC), (2) tropical (the Tropic Regions Test Center, Panama Canal Zone), and (3) cold weather (CRTC, Bolio Lake Test Complex, AK. Special

Shoreline Erosion and Proposed Control at Experimental Facility 15-Spesutie Island

DTIC Science & Technology

2017-09-01

Island, it is made up of various facilities and ranges designed for weapons testing as well as automotive testing . These ranges belong to the...ARL-SR-0383 ● SEP 2017 US Army Research Laboratory Shoreline Erosion and Proposed Control at Experimental Facility 15–Spesutie...in this report are not to be construed as an official Department of the Army position unless so designated by other authorized documents. Citation

A History of U.S. Military Conceptual Solutions to the Uncertainty of Expeditions

DTIC Science & Technology

2016-06-10

War I, the Army appointed COL Samuel D. Rockenbach as commandant of the Tank School in Camp Meade , Maryland.4 In this role, COL Rockenbach intended to...France, led by Captain George S. Patton, Jr., which proved prolific both in the extent to which it influenced American thinking regarding tanks and...Tank In 1928, Secretary of War Dwight Davis directed the Army to establish an experimental armored force at Camp Meade , Maryland. This experimental

Annual Historical Report - AMEDD Activities, Calendar Year 1986

DTIC Science & Technology

1987-01-01

experimentation . A method was devised to determine the heat transfer properties of the head by use of a copper model which is unique and allows independent ...Church, VA 22041 Commander US Army Training and Doctrine Command ATTN: ATCD-S ATCD-ATMD Fort Monroe, VA 23651 Commander US Army Test and Experimentation ...4500 m). Soldiers with less than 20 torr increase test have only a 40-50% probability of acute mountain sickness. Therefore, CPT at sea level may be used

Robotic follower experimentation results: ready for FCS increment I

NASA Astrophysics Data System (ADS)

Jaczkowski, Jeffrey J.

2003-09-01

Robotics is a fundamental enabling technology required to meet the U.S. Army's vision to be a strategically responsive force capable of domination across the entire spectrum of conflict. The U. S. Army Research, Development and Engineering Command (RDECOM) Tank Automotive Research, Development & Engineering Center (TARDEC), in partnership with the U.S. Army Research Laboratory, is developing a leader-follower capability for Future Combat Systems. The Robotic Follower Advanced Technology Demonstration (ATD) utilizes a manned leader to provide a highlevel proofing of the follower's path, which operates with minimal user intervention. This paper will give a programmatic overview and discuss both the technical approach and operational experimentation results obtained during testing conducted at Ft. Bliss, New Mexico in February-March 2003.

BUILDING FOR THE EXPERIMENTAL SWIMMING POOL REACTOR OF 3Mw OF THE JUNTA DE ENERGIA NUCLEAR (in Spanish)

DOE Office of Scientific and Technical Information (OSTI.GOV)

de la Camara, S.N.

1958-10-01

The Spanish experimental swimming pool reactor is constructed on the grounds of the Ciudad Universitaria de Madrid. A general layout of the reactor building and its annexes is given, and the reactor building itself is described. The construction of the reactor building and the characteristics of the annex building are discussed. (J.S.R.)

Hardware Modifications to the US Army Research Laboratory’s Metalorganic Chemical Vapor Deposition (MOCVD) System for Optimization of Complex Oxide Thin Film Fabrication

DTIC Science & Technology

2015-04-01

studies on flow and thermal fields in MOCVD reactor. Chinese Science Bulletin. 2010;55:560–566. 36. Hampdensmith MJ, Kodas TT. Chemical vapor...Chemistry. 1995;19727–750. 47. Xu CY, Hampdensmith MJ, Kodas TT. Aerosol-assisted chemical-vapor- deposition (AACVD) of binary alloy (AGXPD1-X, CUXPD1-X

The Trickling Filter/Solids Contact Process: Application to Army Wastewater Plants

DTIC Science & Technology

1988-08-01

technology (activated sludge and rotating biological contactors [RBC]). 3 7 For the study, the plant was to be sized at 10 mgd. Electricity purchased from...Project Costs* Estimated Cost** ($K) Trickling Rotating Filter/Solids Activated Biological Item Contact Sludge Contactor Preliminary treatment 1100 1100...basins 4500 - Rotating biological contactor reactors - 4520 Flocculator clarifiers 2000 - - Conventional secondary clarifiers 1770 1500 Dual-media

Catalog of experimental projects for a fissioning plasma reactor

NASA Technical Reports Server (NTRS)

Lanzo, C. D.

1973-01-01

Experimental and theoretical investigations were carried out to determine the feasibility of using a small scale fissioning uranium plasma as the power source in a driver reactor. The driver system is a light water cooled and moderated reactor of the MTR type. The eight experiments and proposed configurations for the reactor are outlined.

Army gas-cooled reactor systems program. Preliminary design report off-normal scram system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bushnell, W.H.; Malmstrom, S.A.

1965-06-01

The maximum allowable ML-1 fuel element cladding (hot spot) temperature is established by ANTS 201 at 1750/sup 0/F. The existing ML-1 design makes no provision for automatic scram when this limit is reached. Operating experience has indicated a requirement for such an automatic system during plant startup and a revised hot spot envelope (generated during conceptual design of the scram system) established the desirability of extending this protection to operation at full power conditions. It was also determined that the scram system should include circuitry to initiate an automatic scram if reactor ..delta..T exceeded 450/sup 0/F (the limit established inmore » ANTS 201) and if reactor power exceeded 6 kw(t) without coolant flow in the main loop. The preliminary design of the scram system (designated off-normal scram system) which will provide the required protection is described.« less

An experimental predeployment training program improves self-reported patient treatment confidence and preparedness of Army combat medics.

PubMed

Gerhardt, Robert T; Hermstad, Erik L; Oakes, Michael; Wiegert, Richard S; Oliver, Jeffrey

2008-01-01

To develop and assess impact of a focused review of International Trauma Life Support (ITLS) and combat casualty care with hands-on procedure training for U.S. Army medics deploying to Iraq. The setting was a U.S. Army Medical Department Center and School and Camp Eagle, Iraq. Investigators developed and implemented a command-approved prospective educational intervention with a post hoc survey. Subjects completed a three-day course with simulator and live-tissue procedure laboratories. At deployment's end, medics were surveyed for experience, confidence, and preparedness in treating various casualty severity levels. Investigators used two-tailed t-test with unequal variance for continuous data and chi-square for categorical data. Twenty-nine medics deployed. Eight completed the experimental program. Twenty-one of 25 (84%) available medics completed the survey including six of the eight (75%) experimental medics. The experimental group reported significantly greater levels of preparedness and confidence treating "minimal," "delayed," and "immediate" casualties at arrival in Iraq. These differences dissipated progressively over the time course of the deployment. This experimental program increased combat medic confidence and perceived level of preparedness in treating several patient severity levels. Further research is warranted to determine if the experimental intervention objectively improves patient care quality and translates into lives saved early in deployment.

Comparisons of Theoretical Methods for Predicting Airfoil Aerodynamic Characteristics

DTIC Science & Technology

2010-08-01

Airfoil ,” Airfoils , U.S. Army Aviation Research, Development and Engineering Command, RDECOM TR 10-D-107, August 2010. [2] Somers, D.M. and...Maughmer, M.D., “Design and Experimental Results for the S407 Airfoil ,” U.S. Army Aviation Research, Development and Engineering Command, RDECOM TR 10-D...S414 Airfoil ,” U.S. Army Aviation Research, Development and Engineering Command, RDECOM TR 10-D-112, August 2010. [5] Somers, D.M. and Maughmer

Integrated voice and visual systems research topics

NASA Technical Reports Server (NTRS)

Williams, Douglas H.; Simpson, Carol A.

1986-01-01

A series of studies was performed to investigate factors of helicopter speech and visual system design and measure the effects of these factors on human performance, both for pilots and non-pilots. The findings and conclusions of these studies were applied by the U.S. Army to the design of the Army's next generation threat warning system for helicopters and to the linguistic functional requirements for a joint Army/NASA flightworthy, experimental speech generation and recognition system.

TSCA Experimental Release Application (TERA) for modified Gordonia terrae, R-13-0001 and modified Rhodococcus jostii, R-13-0002

EPA Pesticide Factsheets

TERAs submitted by the US Army Engineer Research and Development Center, Vicksburg, MS and US Army Corps of Engineers, Seattle, WA. These microorganisms will be used in a field demonstration of bioaugmentation to enhance RDX degradation.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tarchalski, M.; Pytel, K.; Wroblewska, M.

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 themore » 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 measurements is of the order of 2.5 W/g at half of the possible MARIA power - 15 MW. The approach and the detailed program for experimental verification of calculations will be presented. The following points will be discussed: - Development of a gamma heating model of MARIA reactor with TRIPOLI 4 (coupled neutron-photon mode) and APOLLO2 model taking into account the key parameters like: configuration of the core, experimental loading, control rod location, reactor power, fuel depletion); - Design of specific measurement tools for MARIA experiments including for instance a new single-cell calorimeter called KAROLINA calorimeter; - MARIA experimental program description and a preliminary analysis of results; - Comparison of calculations for JHR and MARIA cores with experimental verification analysis, calculation behavior and n-γ 'environments'. (authors)« less

Basic Nuclear Physics.

ERIC Educational Resources Information Center

Bureau of Naval Personnel, Washington, DC.

Basic concepts of nuclear structures, radiation, nuclear reactions, and health physics are presented in this text, prepared for naval officers. Applications to the area of nuclear power are described in connection with pressurized water reactors, experimental boiling water reactors, homogeneous reactor experiments, and experimental breeder…

One Door to the Corps: The U.S. Army Engineering and Support Center, Huntsville Historical Update, 1998-2007

DTIC Science & Technology

2009-01-01

with Transpiring Wall Reactor; and Gas Phase Chemical Reduction systems . After much study , ACWA selected three technologies for additional study ...detection systems , and hardware development. Importantly, these advancements allowed for a more effective and cost-efficient remediation process... grounds at the ASPs, the munitions contractors considered a variety of factors, including proximity to the local civilian population and potential

Combating WMD: Journal of the U.S. Army Nuclear and CWMD Agency. Issue 5, Spring/Summer 2010

DTIC Science & Technology

2010-06-01

reception . In the past, antennas were protected from unwanted signals with high capaci- tance metal oxide varistors (a type of surge suppressor) placed at...including a gas-cooled reactor design combined with a closed-cycle gas-turbine generator that could be transportable on semi- trailers , railroad...where else. Towns, schools, shopping areas, theatres, hospitals, residential areas with houses, trailers , hutments, and barracks went up by the

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

US Army Research Laboratory Visualization Framework Architecture Document

DTIC Science & Technology

2018-01-11

this report are not to be construed as an official Department of the Army position unless so designated by other authorized documents. Citation of...release; distribution is unlimited. 14. ABSTRACT Visualization of network science experimentation results is generally achieved using stovepipe...report documents the ARL Visualization Framework system design and specific details of its implementation. 15. SUBJECT TERMS visualization

Development, Implementation, and Evaluation of Leadership/Management Training Within Army Battalions: Volume I: Summary of Findings.

ERIC Educational Resources Information Center

Fry, John P.; Cliborn, Robert E.

The report describes the development and evaluation of an in-unit, leadership/management training program (based on experimental training methodology for providing Army leaders with behavioral skills and techniques) implemented within three battalion-sized combat units at Fort Bliss, Texas, in 1974. The program was organized around workshops…

Selected Topics in Experimental Statistics with Army Applications

DTIC Science & Technology

1983-12-01

could employ the sum indicated by ’t pjXij + f, as the model. 6-45 UM«OUIVI-K /Ub- IUd As is usual and for use in significance tests, we will...were taken from a study and classical example of Bortkiewicz (Ref. 40), which describes the number of deaths from kicks of horses in the Prussian Army

The Impact of a Visual Imagery Intervention on Army ROTC Cadets' Marksmanship Performance and Flow Experiences

ERIC Educational Resources Information Center

Rakes, Edward Lee

2012-01-01

This investigation used an experimental design to examine how a visual imagery intervention and two levels of challenge would affect the flow experiences and performance of cadets engaged in Army ROTC marksmanship training. I employed MANCOVA analyses, with gender and prior marksmanship training experience as covariates, to assess cadets' (n =…

Physical Fitness Index for Assess Fitness Speed Among Army Reserve Officer Training Unit Cadet in Malaysia

NASA Astrophysics Data System (ADS)

Rustam, Shahrulfadly; Kassim, Mohar

2018-05-01

The aim of this study is to develop physical fitness index (PFI) for physical fitness among Army Reserve Officer Training Unit Cadet Malaysia. This study use 30 meter speed run as a physical fitness test battery to develop physical fitness index (PFI) and to evaluate the subject fitness speed. 212 male respondent (N=212) was selected in this study including Army Reserve Officer Training Unit Cadet of National Defence University of Malaysia. 30 meter sprint was used as a instrument for this study. The methodology will be adopted for this study is quantitative research in the form of a quasi-experiment. Quasi-experimental methods is used to measure and evaluate the level of physical fitness and develop Physical Fitness Index especially in speed. The design of this study is quasi-experimental study design with pre-test and post-test. The study design is quasi-experimental research design in which the data is obtained through the practical test in the field. The data were analysed by using the SPSS software version 20 to calculate the mean, standard deviation and t-test for develop physical fitness index (PFI) and to evaluate the fitness speed level for Army Reserve Officers Training Unit Cadet Malaysia. The findings showed mean and standard deviation for develope physical fitness index is (M=4.84) and (SD=0.48). The t-test for evaluate fitness level in speed for pre-test and post-test is significantly difference (p ≤ 0.05). The implication at this study is that the develope of standard physical fitness index is able to identify the level of physical fitness among Army Reserve Officer Training Unit Cadet Malaysia.

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

A Comparison of Measured and Calculated Air-Transported Radiation from a Fast. Unshielded Nuclear Reactor.

DTIC Science & Technology

1980-12-01

UNSHIELDED NUCLEAR REACTOR by H. A. Robitaille and B. E. Hoffarth D T ICS ELECTE SEP 1 5 i9813 C- LA . DISTRIBUTION STATEMENT A O PROJECT NO. 1 DECEMBER 1980...et de neutrons b diff6rentes distances jusqu’ 1100 m~tres du r6acteur ncldaire & neutrons rapides de la Pulse Radiation Division de la U.S. Army...u)UOU 𔃽N~nl4 N0in0 -CuD z f- 000 LUU ’4 :1O 00 w4 w4 w (u3)uU3 𔃽N~n-A NO.Ln3 rc ino OQQ z ca co, 0 0- C ) LU c C. M CU m CD 0 LA (u3uOU 𔃽N~n1.4No

Cost Reductions for Wastewater Treatment Utilizing Water Management at Holston Army Ammunition Plant

DTIC Science & Technology

1976-05-01

says that the granular carbon used is made from bituminous coal. As the waste stream pass through a bed of carbon granules, com- pounds are adsorbed to...findings of laboratory-scale reactor studies conducted at Purdue University for * Clark, Dietz and Associates. The original recommendations and cost...Pretreatment Denitrification by Submerged Anaerbbic I ilters I ~ Trickling Filters S F ,2al Clarification "•’i Pump - ~ Sludge ,Treatment Dual Media Filh:ration

KINETICS OF TREAT USED AS A TEST REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickerman, C.E.; Johnson, R.D.; Gasidlo, J.

1962-05-01

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

Mission-Based Scenario Research: Experimental Design and Analysis

DTIC Science & Technology

2011-08-10

with Army Reserach Lab; DCS Corporation, Alexandria, Va and Warren, Mi 14. ABSTRACT In this paper , we discuss a neuroimaging experiment that...Development and Engineering Center Warren, MI Kelvin S. Oie, PhD Army Research Laboratory Aberdeen Proving Ground, MD ABSTRACT In this paper , we...experiment, this paper will emphasize analyses that employ a pattern classification analysis approach. These classification examples aim to identify

The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility (AEF)

DTIC Science & Technology

2015-10-01

ARL-TR-7506 ● OCT 2015 US Army Research Laboratory The Automation of the Transonic Experimental Facility (TEF) and the...Laboratory The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility (AEF) by Charith R Ranawake Weapons...To) 05/2015–08/2015 4. TITLE AND SUBTITLE The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility

Development of a Toolbox Using Chemical, Physical and Biological Technologies for Decontamination of Sediments to Support Strategic Army Response to Natural Disasters

DTIC Science & Technology

2006-11-01

disinfection) was tested using soil microcosms and respirometry to determine diesel range and total organic compound degradation. These tests were...grease) such as benzo(a)pyrene were detected above chronic (long term-measured in years) screening levels. Levels of diesel and oil range organics... bioremediation , and toxicity of liquid and solid samples. The Comput-OX 4R is a 4 reactor unit with no stirring modules or temperature controlled water bath

Evaluation of the Feasibility of Biodegrading Explosives-Contaminated Soils and Groundwater at the Newport Army Ammunition Plant (NAAP)

DTIC Science & Technology

1991-06-01

undamaged to its original location. 9 3 Biodegradation Studies The NAAP soils were used for both the basic microbiological studies and the bench scale...reactor studies. The microbiological studies were directed at measuring (1) the growth potential of bacteria present in the soil samples and (2) the...clear and odorless, and no TNT was detected in them. The detection limit for TNT in the water samples was 0.5 mg/L. Microbiological characterization

Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

NASA Astrophysics Data System (ADS)

Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

Performance Comparison between Stereausis and Incoherent Wideband Music for Localization of Ground Vehicles

DTIC Science & Technology

1999-09-01

PERFORMANCE COMPARISON BETWEEN STEREAUSIS AND INCOHERENT WIDEBAND MUSIC FOR LOCALIZATION OF GROUND VEHICLES September 1999 Tien Pham U.S. Army...present experimental results comparing the incoherent wideband MUSIC (IWM) algorithm developed by the Army Research Laboratory (ARL)1, 2 and the...Type N/A Dates Covered (from... to) ("DD MON YYYY") Title and Subtitle Performance Comparison Between Stereausis and Incoherent Wideband Music for

EBR-II Reactor Physics Benchmark Evaluation Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pope, Chad L.; Lum, Edward S; Stewart, Ryan

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.

Multivariate Analysis of High Through-Put Adhesively Bonded Single Lap Joints: Experimental and Workflow Protocols

DTIC Science & Technology

2016-06-01

unlimited. v List of Tables Table 1 Single-lap-joint experimental parameters ..............................................7 Table 2 Survey ...Joints: Experimental and Workflow Protocols by Robert E Jensen, Daniel C DeSchepper, and David P Flanagan Approved for...TR-7696 ● JUNE 2016 US Army Research Laboratory Multivariate Analysis of High Through-Put Adhesively Bonded Single Lap Joints: Experimental

Equipment for neutron measurements at VR-1 Sparrow training reactor.

PubMed

Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

2010-01-01

The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

PubMed

Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

2017-12-01

The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

Rotating Fluidized Bed Reactor for Space Nuclear Propulsion. Annual Report; Design Studies and Experimental Results

NASA Technical Reports Server (NTRS)

1971-01-01

The rotating fluidized bed reactor concept is being investigated for possible application in nuclear propulsion systems. Physics calculations show U-233 to be superior to U-235 as a fuel for a cavity reactor of this type. Preliminary estimates of the effect of hydrogen in the reactor, reflector material, and power peaking are given. A preliminary engineering analysis was made for U-235 and U-233 fueled systems. An evaluation of the parameters affecting the design of the system is given, along with the thrust-to-weight ratios. The experimental equipment is described, as are the special photographic techniques and procedures. Characteristics of the fluidized bed and experimental results are given, including photographic evidence of bed fluidization at high rotational velocities.

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental...

EXPERIMENTAL MOLTEN-SALT-FUELED 30-Mw POWER REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alexander, L.G.; Kinyon, B.W.; Lackey, M.E.

1960-03-24

A preliminary design study was made of an experimental molten-salt- fueled power reactor. The reactor considered is a single-region homogeneous burner coupled with a Loeffler steam-generating cycle. Conceptual plant layouts, basic information on the major fuel circuit components, a process flowsheet, and the nuclear characteristics of the core are presented. The design plant electrical output is 10 Mw, and the total construction cost is estimated to be approximately ,000,000. (auth)

A Study of the Effectiveness of the Army’s National Advertising Expenditures. Volume 2. Main Report.

DTIC Science & Technology

1981-08-31

effectiveness of the Army’s national recruitment advertising . N W Ayer’s Marketing Services Department undertook this task in September, 1979, with...Experimentation is clearly preferable whenever one can dramatically vary advertising in sensitive markets under controlled conditions; however, it...Such a feedback relationship is not uncommon, since many commercial and industrial markets routinely set advertising budgets as a percentage of their

Safe Operations of Unmanned Systems for Reconnaissance in Complex Environments-Army Technology Objective (SOURCE ATO) Field Experimentation Observations and Soldier Feedback

DTIC Science & Technology

2012-07-01

intent during mission activities. Autonomous assets will have interactions with humans in several different relationships that could benefit from...RDRL HRM CN R SPENCER DCSFDI HF HQ USASOC BLDG E2929 FORT BRAGG NC 28310-5000 1 ARMY RSCH LABORATORY – HRED HUMAN RSRCH AND ENGRNG...A. William Evans, III Human Research and Engineering Directorate, ARL Approved for public release

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« less

Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krass, A.W.

2005-12-19

This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. Themore » material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.« less

Low cost solar array project: Experimental process system development unit for producing semiconductor-grade silicon using the silane-to-silicon process

NASA Technical Reports Server (NTRS)

1981-01-01

The results of the free space reactor experimental work are summarized. Overall, the objectives were achieved and the unit can be confidently scaled to the EPSDU size based on the experimental work and supporting theoretical analyses. The piping and instrumentation of the fluidized bed reactor was completed.

An experimental study of ammonia borane based hydrogen storage systems

NASA Astrophysics Data System (ADS)

Deshpande, Kedaresh A.

2011-12-01

Hydrogen is a promising fuel for the future, capable of meeting the demands of energy storage and low pollutant emission. Chemical hydrides are potential candidates for chemical hydrogen storage, especially for automobile applications. Ammonia borane (AB) is a chemical hydride being investigated widely for its potential to realize the hydrogen economy. In this work, the yield of hydrogen obtained during neat AB thermolysis was quantified using two reactor systems. First, an oil bath heated glass reactor system was used with AB batches of 0.13 gram (+/- 0.001 gram). The rates of hydrogen generation were measured. Based on these experimental data, an electrically heated steel reactor system was designed and constructed to handle up to 2 grams of AB per batch. A majority of components were made of stainless-steel. The system consisted of an AB reservoir and feeder, a heated reactor, a gas processing unit and a system control and monitoring unit. An electronic data acquisition system was used to record experimental data. The performance of the steel reactor system was evaluated experimentally through batch reactions of 30 minutes each, for reaction temperatures in the range from 373 K to 430 K. The experimental data showed exothermic decomposition of AB accompanied by rapid generation of hydrogen during the initial period of the reaction. 90% of the hydrogen was generated during the initial 120 seconds after addition of AB to the reactor. At 430 K, the reaction produced 12 wt.% of hydrogen. The heat diffusion in the reactor system and the process of exothermic decomposition of AB were coupled in a two-dimensional model. Neat AB thermolysis was modeled as a global first order reactions based on Arrhenius theory. The values of equation constants were derived from curve fit of experimental data. The pre-exponential constant and the activation energy were estimated to be 4 s-1 (+/- 0.4 s-1) and 13000 J mol -1 s-1 (+/- 1050 J mol-1 s -1) respectively. The model was solved in COMSOL Multiphysics. The model was capable of simulating the transient response of the system and captured the observed trends such as the decrease in reactor temperature upon addition of AB and exothermic decomposition.

Engineering Design Handbook. Army Weapon Systems Analysis. Part 2

DTIC Science & Technology

1979-10-01

EXPERIMENTAL DESIGN ............................... ............ 41-3 41-5 RESULTS OF THE ASARS lIX SIMULATIONS ........................... 41-4 41-6 LATIN...sciences and human factors engineering fields utilizing experimental methodology and multi-variable statistical techniques drawn from experimental ...randomly to grenades for the test design . The nine experimental types of hand grenades (first’ nine in Table 33-2) had a "pip" on their spherical

Characterization of gamma field in the JSI TRIGA reactor

NASA Astrophysics Data System (ADS)

Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

2018-01-01

Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

Utilization of the Recycle Reactor in Determining Kinetics of Gas-Solid Catalytic Reactions.

ERIC Educational Resources Information Center

Paspek, Stephen C.; And Others

1980-01-01

Describes a laboratory scale reactor that determines the kinetics of a gas-solid catalytic reaction. The external recycle reactor construction is detailed with accompanying diagrams. Experimental details, application of the reactor to CO oxidation kinetics, interphase gradients, and intraphase gradients are discussed. (CS)

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor

PubMed Central

Zhang, Shubin; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance. PMID:29121067

High yields of hydrogen production from methanol steam reforming with a cross-U type reactor.

PubMed

Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Zhang, Xuelin; Liu, Xiaowei

2017-01-01

This paper presents a numerical and experimental study on the performance of a methanol steam reformer integrated with a hydrogen/air combustion reactor for hydrogen production. A CFD-based 3D model with mass and momentum transport and temperature characteristics is established. The simulation results show that better performance is achieved in the cross-U type reactor compared to either a tubular reactor or a parallel-U type reactor because of more effective heat transfer characteristics. Furthermore, Cu-based micro reformers of both cross-U and parallel-U type reactors are designed, fabricated and tested for experimental validation. Under the same condition for reforming and combustion, the results demonstrate that higher methanol conversion is achievable in cross-U type reactor. However, it is also found in cross-U type reactor that methanol reforming selectivity is the lowest due to the decreased water gas shift reaction under high temperature, thereby carbon monoxide concentration is increased. Furthermore, the reformed gas generated from the reactors is fed into a high temperature proton exchange membrane fuel cell (PEMFC). In the test of discharging for 4 h, the fuel cell fed by cross-U type reactor exhibits the most stable performance.

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

Determining the microwave coupling and operational efficiencies of a microwave plasma assisted chemical vapor deposition reactor under high pressure diamond synthesis operating conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nad, Shreya; Department of Physics and Astronomy, Michigan State University, East Lansing, Michigan 48824; Gu, Yajun

2015-07-15

The microwave coupling efficiency of the 2.45 GHz, microwave plasma assisted diamond synthesis process is investigated by experimentally measuring the performance of a specific single mode excited, internally tuned microwave plasma reactor. Plasma reactor coupling efficiencies (η) > 90% are achieved over the entire 100–260 Torr pressure range and 1.5–2.4 kW input power diamond synthesis regime. When operating at a specific experimental operating condition, small additional internal tuning adjustments can be made to achieve η > 98%. When the plasma reactor has low empty cavity losses, i.e., the empty cavity quality factor is >1500, then overall microwave discharge coupling efficienciesmore » (η{sub coup}) of >94% can be achieved. A large, safe, and efficient experimental operating regime is identified. Both substrate hot spots and the formation of microwave plasmoids are eliminated when operating within this regime. This investigation suggests that both the reactor design and the reactor process operation must be considered when attempting to lower diamond synthesis electrical energy costs while still enabling a very versatile and flexible operation performance.« less

REACTOR PHYSICS CONSTANTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

MIQSTURE: An Experimental Online Language for Army Tactical Intelligence Information Processing

DTIC Science & Technology

1978-07-01

algorithms. The most critical component of an active information processing model for Army tactical intelligence is the user interface, which must be based on...1976)** defined some preliminary notions of an active information model centered around a data base that can introspect about its contents and...34An Introspective Data Base for an Active Information Model." OSI Technical Note N76-017, 17 November 1976 1-4 L4 beyond optimistic expectations and

The Results of an Experimental Restructuring of EO Staffing Patterns in a Infantry Division

DTIC Science & Technology

1979-10-01

Research, Inc. ARI FIELD UNIT AT PRESIDIO OF MONTEREY, CALIFORNIA JI NV 8 1984 -U 1 S: U. S. Army Research Institute for the Behavioral and Social ...12. REPORT DATE Army Research Institute for the Behavioral and Social October 1979 Sciences, 5001 Eisenhower Ave., Alexandria, VA 22333 I. NUMBER OF...one of thiree prepared under this contract. The other reports are: Unit Equal Opportunity Training Diagnosis and Assessment System, and Development and

Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, L.K.; Mohr, D.; Planchon, H.P.

This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

ETRCF, TRA654, INTERIOR. REACTOR OPERATED IN WATERFILLED TANK. CAMERA LOOKS ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR-CF, TRA-654, INTERIOR. REACTOR OPERATED IN WATER-FILLED TANK. CAMERA LOOKS DOWN FROM ABOVE UPON LATER (NON-NUCLEAR) EXPERIMENTAL GEAR. INL NEGATIVE NO. HD24-1-1. Mike Crane, Photographer, ca. 2003 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The electrical characteristics of the dielectric barrier discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehia, Ashraf, E-mail: yehia30161@yahoo.com; Department of Physics, Faculty of Science, Assiut University, Assiut 71516

2016-06-15

The electrical characteristics of the dielectric barrier discharges have been studied in this paper under different operating conditions. The dielectric barrier discharges were formed inside two reactors composed of electrodes in the shape of two parallel plates. The dielectric layers inside these reactors were pasted on the surface of one electrode only in the first reactor and on the surfaces of the two electrodes in the second reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at the normal temperature and pressure, in parallel with applying a sinusoidal ac voltagemore » between the electrodes of the reactor. The amount of the electric charge that flows from the reactors to the external circuit has been studied experimentally versus the ac peak voltage applied to them. An analytical model has been obtained for calculating the electrical characteristics of the dielectric barrier discharges that were formed inside the reactors during a complete cycle of the ac voltage. The results that were calculated by using this model have agreed well with the experimental results under the different operating conditions.« less

Continuation of a Postdoctoral Research Associateship Program

DTIC Science & Technology

1998-12-01

ODRS and LPO can be alleviated by nitric oxide. Rsearch in Progress: The experimental part of the research has been completed. The experimental ... Rsearch in Progress: The experimental part of the research has been completed. The experimental results have been published, or prepared for...construed as an official Department of the Army position, policy or decision unless so designated by other documentation. •*•*«, 19990521 029 fPII

Experimental Flight Characterization of Spin Stabilized Projectiles at High Angle of Attack

DTIC Science & Technology

2017-08-07

ARL-TR-8082 ● AUG 2017 US Army Research Laboratory Experimental Flight Characterization of Spin- Stabilized Projectiles at High ...Experimental Flight Characterization of Spin- Stabilized Projectiles at High Angle of Attack by Frank Fresconi and Ilmars Celmins Weapons and Materials...June 2016–June 2017 4. TITLE AND SUBTITLE Experimental Flight Characterization of Spin-Stabilized Projectiles at High Angle of Attack 5a. CONTRACT

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

Summary and Findings of the ARL Dynamic Failure Forum

DTIC Science & Technology

2016-09-29

short beam shear, quasi -static indentation, depth of penetration, and V50 limit velocity. o Experimental technique suggestions for improvement included...art in experimental , theoretical, and computational studies of dynamic failure. The forum also focused on identifying technologies and approaches...Army-specific problems. Experimental exploration of material behavior and an improved ability to parameterize material models is essential to improving

Numerical and Experimental Thermal Responses of Single-cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brun, J.; Reynard-Carette, C.; Carette, M.

2015-07-01

The nuclear radiation energy deposition rate (usually expressed in W.g{sup -1}) is a key parameter for the thermal design of experiments, on materials and nuclear fuel, carried out in experimental channels of irradiation reactors such as the French OSIRIS reactor in Saclay or inside the Polish MARIA reactor. In particular the quantification of the nuclear heating allows to predicting the heat and thermal conditions induced in the irradiation devices or/and structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters also called in-pile calorimeters. Two main kinds of in-pile calorimeter exist with in particular specific designs:more » single-cell calorimeter and differential calorimeter. The present work focuses on these two calorimeter kinds from their out-of-pile calibration step (transient and steady experiments respectively) to comparison between numerical and experimental results obtained from two irradiation campaigns (MARIA reactor and OSIRIS reactor respectively). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions. (authors)« less

Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

NASA Astrophysics Data System (ADS)

Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

2017-08-01

The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

Validating a Finite Element Model of a Structure Subjected to Mine Blast with Experimental Modal Analysis

DTIC Science & Technology

2017-11-01

The Under-body Blast Methodology (UBM) for the Test and Evaluation (T&E) program was established to provide a capability for the US Army Test and... Evaluation Command to assess the vulnerability of vehicles to under-body blast. Finite element (FE) models are part of the current UBM for T&E methodology...Methodology (UBM) for the Test and Evaluation (T&E) program was established to provide a capability for the US Army Test and Evaluation Command

Research and Operational Support for the Study of Militarily Relevant Infectious Diseases of Interest to the United States Army and the Royal Thai Army

DTIC Science & Technology

2007-01-01

sporozoites from P. vivax specimens obtained from throughout the world . These sporozoites could then be used in a variety of experimental models (i.e., we...maintenance of a center for excellence focused on the basic biology and epidemiology of malaria. 31 2. To assess emerging febrile diseases along...classical and state- of -the-art technologies as possible to the above multi-faceted research. Clinical research included mobile epidemiology team able to

Acceleration of an Initially Moving Projectile: Velocity-Injected Railguns and Their Effect on Pulsed Power

DTIC Science & Technology

2009-07-01

section 2, the propellant gun is assessed from existing experimental data. In section 3, circuit analysis is used to model the railgun with a 2 launch...An Empirical Model for Plasma Armature Voltage. IEEE Transactions on Magnetics 1991, 27, 283–288. 11. Elder, D. The First Generation in the...1 US ARMY MISSLE COMMAND AMSRD AMR W W C MCCORKLE 5400 FOWLER RD REDSTONE ARSENAL AL 35898-5240 1 US ARMY TACOM TARDEC AMSTA TR

Army AL&T, April-June 2008

DTIC Science & Technology

2008-06-01

IR )/laser designator (LD)/laser range finder (LRF) sensor. The Class I UAS consists of a Class I UAV, a cen- tralized controller and a minimal set...utility of a backpackable, affordable, easy-to- operate and responsive reconnais- sance and surveillance system through experimentation. • Use EO/ IR ...ARMY AL&T 33APRIL - JUNE 2008 • “The IR sensor pinpointed the enemy even after the sun went down. We could have really used this in Iraq.” • “The UAV

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Military Personnel Attrition and Retention: Research in Progress

DTIC Science & Technology

1981-10-01

designed to assess the relative risks of attrition among non -h.gh. school graduate Army enlistees. As a group, these individuals historically...this methodology see the Majcýhrzak paper in this volume) Research in the PlanningStaaqe Stage ’TV: Administrative Experimentation In order to derive...different research designs : 1. True experimental designs ; 2. Correlational research designs ; 3. Quasi- experimental research designs . We rejected the idea

Experimental Flight Characterization of a Canard-Controlled, Subsonic Missile

DTIC Science & Technology

2017-08-01

ARL-TR-8086 ● AUG 2017 US Army Research Laboratory Experimental Flight Characterization of a Canard- Controlled , Subsonic Missile...Laboratory Experimental Flight Characterization of a Canard- Controlled , Subsonic Missile by Frank Fresconi, Ilmars Celmins, James Maley, and...valid OMB control number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) August 2017 2. REPORT TYPE Technical

Operational Evaluation of Self-Paced Instruction in U.S. Army Training.

DTIC Science & Technology

1979-01-01

one iteration of each course, and the on -going refinement and adjustment of managerial techniques. Research Approach A quasi - experimental approach was...research design employed experimental and control groups , posttest only with non-random groups . The design dealt with the six major areas identified as...course on Interpersonal Communications were conducted in the conventional, group -paced manner. Experimental course materials. Wherever possible, existing

Experimental studies of irradiated and hydrogen implantation damaged reactor steels

NASA Astrophysics Data System (ADS)

Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

2016-01-01

Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, J. D.; Briggs, J. B.; Gulliford, J.

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 Energymore » 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 is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

Experimental study of Siphon breaker about size effect in real scale reactor design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kang, S. H.; Ahn, H. S.; Kim, J. M.

2012-07-01

Rupture accident within the pipe of a nuclear reactor is one of the main causes of a loss of coolant accident (LOCA). Siphon-breaking is a passive method that can prevent a LOCA. In this study, either a line or a hole is used as a siphon-breaker, and the effect of various parameters, such as the siphon-breaker size, pipe rupture point, pipe rupture size, and the presence of an orifice, are investigated using an experimental facility similar in size to a full-scale reactor. (authors)

An update of engine system research at the Army Propulsion Directorate

NASA Technical Reports Server (NTRS)

Bobula, George A.

1990-01-01

The Small Turboshaft Engine Research (STER) program provides a vehicle for evaluating the application of emerging technologies to Army turboshaft engine systems and to investigate related phenomena. Capitalizing on the resources at hand, in the form of both the NASA facilities and the Army personnel, the program goal of developing a physical understanding of engine system dynamics and/or system interactions is being realized. STER entries investigate concepts and components developed both in-house and out-of-house. Emphasis is placed upon evaluations which evolved from on-going basic research and advanced development programs. Army aviation program managers are also encouraged to make use of STER resources, both people and facilities. The STER personnel have established their reputations as experts in the fields of engine system experimental evaluations and engine system related phenomena. The STER facility has STER program provides the Army aviation community the opportunity to perform system level investigations, and then to offer the findings to the entire engine community for their consideration in next generation propulsion systems. In this way results of the fundamental research being conducted to meet small turboshaft engine technology challenges expeditiously find their way into that next generation of propulsion systems.

Ammonia-oxidizing microbial communities in reactors with efficient nitrification at low-dissolved oxygen

PubMed Central

Fitzgerald, Colin M.; Camejo, Pamela; Oshlag, J. Zachary; Noguera, Daniel R.

2015-01-01

Ammonia-oxidizing microbial communities involved in ammonia oxidation under low dissolved oxygen (DO) conditions (<0.3 mg/L) were investigated using chemostat reactors. One lab-scale reactor (NS_LowDO) was seeded with sludge from a full-scale wastewater treatment plant (WWTP) not adapted to low-DO nitrification, while a second reactor (JP_LowDO) was seeded with sludge from a full-scale WWTP already achieving low-DO nitrifiaction. The experimental evidence from quantitative PCR, rDNA tag pyrosequencing, and fluorescence in situ hybridization (FISH) suggested that ammonia-oxidizing bacteria (AOB) in the Nitrosomonas genus were responsible for low-DO nitrification in the NS_LowDO reactor, whereas in the JP_LowDO reactor nitrification was not associated with any known ammonia-oxidizing prokaryote. Neither reactor had a significant population of ammonia-oxidizing archaea (AOA) or anaerobic ammonium oxidation (anammox) organisms. Organisms isolated from JP_LowDO were capable of autotrophic and heterotrophic ammonia utilization, albeit without stoichiometric accumulation of nitrite or nitrate. Based on the experimental evidence we propose that Pseudomonas, Xanthomonadaceae, Rhodococcus, and Sphingomonas are involved in nitrification under low-DO conditions. PMID:25506762

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

DOE Office of Scientific and Technical Information (OSTI.GOV)

John D. Bess; J. Blair Briggs; Jim Gulliford

2012-11-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set ofmore » 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. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.« less

Validation of the MCNP computational model for neutron flux distribution with the neutron activation analysis measurement

NASA Astrophysics Data System (ADS)

Tiyapun, K.; Chimtin, M.; Munsorn, S.; Somchit, S.

2015-05-01

The objective of this work is to demonstrate the method for validating the predication of the calculation methods for neutron flux distribution in the irradiation tubes of TRIGA research reactor (TRR-1/M1) using the MCNP computer code model. The reaction rate using in the experiment includes 27Al(n, α)24Na and 197Au(n, γ)198Au reactions. Aluminium (99.9 wt%) and gold (0.1 wt%) foils and the gold foils covered with cadmium were irradiated in 9 locations in the core referred to as CT, C8, C12, F3, F12, F22, F29, G5, and G33. The experimental results were compared to the calculations performed using MCNP which consisted of the detailed geometrical model of the reactor core. The results from the experimental and calculated normalized reaction rates in the reactor core are in good agreement for both reactions showing that the material and geometrical properties of the reactor core are modelled very well. The results indicated that the difference between the experimental measurements and the calculation of the reactor core using the MCNP geometrical model was below 10%. In conclusion the MCNP computational model which was used to calculate the neutron flux and reaction rate distribution in the reactor core can be used for others reactor core parameters including neutron spectra calculation, dose rate calculation, power peaking factors calculation and optimization of research reactor utilization in the future with the confidence in the accuracy and reliability of the calculation.

Computational Fluid Dynamics simulation of hydrothermal liquefaction of microalgae in a continuous plug-flow reactor.

PubMed

Ranganathan, Panneerselvam; Savithri, Sivaraman

2018-06-01

Computational Fluid Dynamics (CFD) technique is used in this work to simulate the hydrothermal liquefaction of Nannochloropsis sp. microalgae in a lab-scale continuous plug-flow reactor to understand the fluid dynamics, heat transfer, and reaction kinetics in a HTL reactor under hydrothermal condition. The temperature profile in the reactor and the yield of HTL products from the present simulation are obtained and they are validated with the experimental data available in the literature. Furthermore, the parametric study is carried out to study the effect of slurry flow rate, reactor temperature, and external heat transfer coefficient on the yield of products. Though the model predictions are satisfactory in comparison with the experimental results, it still needs to be improved for better prediction of the product yields. This improved model will be considered as a baseline for design and scale-up of large-scale HTL reactor. Copyright © 2018 Elsevier Ltd. All rights reserved.

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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 inmore » their evaluation process. 73 refs., 26 figs., 69 tabs.« less

Overview of Fuel Rod Simulator Usage at ORNL

NASA Astrophysics Data System (ADS)

Ott, Larry J.; McCulloch, Reg

2004-02-01

During the 1970s and early 1980s, the Oak Ridge National Laboratory (ORNL) operated large out-of-reactor experimental facilities to resolve thermal-hydraulic safety issues in nuclear reactors. The fundamental research ranged from material mechanical behavior of fuel cladding during the depressurization phase of a loss-of-coolant accident (LOCA) to basic heat transfer research in gas- or sodium-cooled cores. The largest facility simulated the initial phase (less than 1 min. of transient time) of a LOCA in a commercial pressurized-water reactor. The nonnuclear reactor cores of these facilities were mimicked via advanced, highly instrumented electric fuel rod simulators locally manufactured at ORNL. This paper provides an overview of these experimental facilities with an emphasis on the fuel rod simulators.

Special features of the inverse-beta-decay reaction proceeding on a proton in a reactor-antineutrino flux

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kopeikin, V. I., E-mail: kopeikin46@yandex.ru; Skorokhvatov, M. D., E-mail: skorokhvatov-md@nrcki.ru

2017-03-15

The evolution of the reactor-antineutrino spectrum and the evolution of the spectrum of positrons from the inverse-beta-decay reaction in the course of reactor operation and after reactor shutdown are considered. The present-day status in determining the initial reactor-antineutrino spectrum on the basis of spectra of beta particles from mixtures of products originating from uranium and plutonium fission is described. A local rise of the experimental spectrum of reactor antineutrinos with respect to the expected spectrum is studied.

Experimental and modeling study of a two-stage pilot scale high solid anaerobic digester system.

PubMed

Yu, Liang; Zhao, Quanbao; Ma, Jingwei; Frear, Craig; Chen, Shulin

2012-11-01

This study established a comprehensive model to configure a new two-stage high solid anaerobic digester (HSAD) system designed for highly degradable organic fraction of municipal solid wastes (OFMSW). The HSAD reactor as the first stage was naturally separated into two zones due to biogas floatation and low specific gravity of solid waste. The solid waste was retained in the upper zone while only the liquid leachate resided in the lower zone of the HSAD reactor. Continuous stirred-tank reactor (CSTR) and advective-diffusive reactor (ADR) models were constructed in series to describe the whole system. Anaerobic digestion model No. 1 (ADM1) was used as reaction kinetics and incorporated into each reactor module. Compared with the experimental data, the simulation results indicated that the model was able to well predict the pH, volatile fatty acid (VFA) and biogas production. Copyright © 2012 Elsevier Ltd. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nygaard, E. T.; Williams, M. M. R.; Angelo, P. L.

Babcock and Wilcox Technical Services Group (B and W) has identified aqueous homogeneous reactors (AHRs) as a technology well suited to produce the medical isotope molybdenum 99 (Mo-99). AHRs have never been specifically designed or built for this specialized purpose. However, AHRs have a proven history of being safe research reactors. In fact, in 1958, AHRs had 'a longer history of operation than any other type of research reactor using enriched fuel' and had 'experimentally demonstrated to be among the safest of all various type of research reactor now in use [1].' A 'Level 1' model representing B and W'smore » proposed Medical Isotope Production System (MIPS) reactor has been developed. The Level 1 model couples a series of differential equations representing neutronics, temperature, and voiding. Neutronics are represented by point reactor kinetics while temperature and voiding terms are axially varying (one-dimensional). While this model was developed specifically for the MIPS reactor, its applicability to the Japanese TRACY reactor was assessed. The results from the Level 1 model were in good agreement with TRACY experimental data and found to be conservative over most of the time domains considered. The Level 1 model was used to study the MIPS reactor. An analysis showed the Level 1 model agreed well with a more complex computational model of the MIPS reactor (a FETCH model). Finally, a significant reactivity insertion was simulated with the Level 1 model to study the MIPS reactor's time-dependent response. (authors)« less

Consumption of the electric power inside silent discharge reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehia, Ashraf, E-mail: yehia30161@yahoo.com

An experimental study was made in this paper to investigate the relation between the places of the dielectric barriers, which cover the surfaces of the electrodes in the coaxial cylindrical reactors, and the rate of change of the electric power that is consumed in forming silent discharges. Therefore, silent discharges have been formed inside three coaxial cylindrical reactors. The dielectric barriers in these reactors were pasted on both the internal surface of the outer electrode in the first reactor and the external surface of the inner electrode in the second reactor as well as the surfaces of the two electrodesmore » in the third reactor. The reactor under study has been fed by atmospheric air that flowed inside it with a constant rate at normal temperature and pressure, in parallel with the application of a sinusoidal ac voltage between the electrodes of the reactor. The electric power consumed in forming the silent discharges inside the three reactors was measured as a function of the ac peak voltage. The validity of the experimental results was investigated by applying Manley's equation on the same discharge conditions. The results have shown that the rate of consumption of the electric power relative to the ac peak voltage per unit width of the discharge gap improves by a ratio of either 26.8% or 80% or 128% depending on the places of the dielectric barriers that cover the surfaces of the electrodes inside the three reactors.« less

Combining SBR systems for chemical and biological treatment: the destruction of the nerve agent VX.

PubMed

Irvine, R L; Haraburda, S S; Galbis-Reig, C

2004-01-01

The US Army is pilot testing the neutralization of VX nerve agent stockpiled at Newport, Indiana using caustic hydrolysis in a Sequencing Batch Reactor (SBR). The resulting hydrolysate was tested at the bench-scale for treatment with activated sludge biodegradation in two distinct studies, one in the SBR and another, in the PACT process. The feed to both biological systems was pretreated to enhance the biodegradability of the hydrolysis products. Both biodegradation studies demonstrated that the hydrolysate could easily meet the Chemical Weapons Convention treaty and US environmental regulations following pretreatment.

On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chauvin, J.P.; Blaise, P.; Lyoussi, A.

2015-07-01

The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physicsmore » calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)« less

Neutrino scattering and the reactor antineutrino anomaly

NASA Astrophysics Data System (ADS)

Garcés, Estela; Cañas, Blanca; Miranda, Omar; Parada, Alexander

2017-12-01

Low energy threshold reactor experiments have the potential to give insight into the light sterile neutrino signal provided by the reactor antineutrino anomaly and the gallium anomaly. In this work we analyze short baseline reactor experiments that detect by elastic neutrino electron scattering in the context of a light sterile neutrino signal. We also analyze the sensitivity of experimental proposals of coherent elastic neutrino nucleus scattering (CENNS) detectors in order to exclude or confirm the sterile neutrino signal with reactor antineutrinos.

Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Woods, Brian; Gutowska, Izabela; Chiger, Howard

Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion tomore » ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with an electrical heat source. The peak core region temperature capability is 1400°C. As part of this project, an inventory of test facilities that could be used for these experimental programs was completed. Several of these facilities showed some promise, however, upon further investigation it became clear that only the OSU HTTF had the power and/or peak temperature limits that would allow for the experimental programs envisioned herein. Thus the conceptual design and feasibility study development focused on examining the feasibility of configuring the current HTTF to collect validation data for these experimental programs. In addition to the scaling analyses and conceptual design development, a test plan was developed for the envisioned modified test facility. This test plan included a discussion on an appropriate shakedown test program as well as the specific matrix tests. Finally, a feasibility study was completed to determine the cost and schedule considerations that would be important to any test program developed to investigate these designs and events.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucholz, J.A.

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

NASA Astrophysics Data System (ADS)

Rizzo, Axel; Vaglio-Gaudard, Claire; Martin, Julie-Fiona; Noguère, Gilles; Eschbach, Romain

2017-09-01

DARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN/PEPIN2 depletion code, each of them being developed by CEA/DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle.

MTR,TRA603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

MTR,TRA-603. EXPERIMENTERS' SPACE ALLOCATIONS IN BASEMENT AS OF 1963. SHIELDED CUBICLES WERE IDENTIFIED BY SPONSORING LABORATORY AND ITS TEST HOLE NUMBER IN THE REACTOR, IE, "KAPL HB-1" SIGNIFIED KNOLLS ATOMIC POWER LABORATORY, HORIZONTAL BEAM NO. 1. "WAPD" WAS WESTINGHOUSE ATOMIC POWER DIVISION. CATCH TANKS AND SAMPLE STATIONS FOR TEST LOOPS WERE ASSOCIATED WITH THESE CUBICLES. NOTE DESKS, STORAGE CABINETS, SWITCH GEAR, INSTRUMENT PANELS. PHILLIPS PETROLEUM COMPANY MTR-E-5205, 4/1963. INL INDEX NO. 531-0603-00-706-009757, REV. 5. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

DOE Office of Scientific and Technical Information (OSTI.GOV)

De Bruyn, D.; Engelen, J.; Ortega, A.

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of threemore » years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)« less

The Captive Helicopter as a Training Device: Experimental Evaluation of a Concept. Technical Report 68-9.

ERIC Educational Resources Information Center

Caro, Paul W., Jr.; And Others

As part of the Army's effort to use synthetic devices to improve training, researchers evaluated a captive helicopter attached to a ground effects machine. Experimental groups received varying amounts of pre-flight practice tasks designed to develop flight skills, while control groups received no device training. Student flight performance during…

Experimental Design and Analysis for the FIST (Fire Support Team) Force Development Testing and Experimentation II.

DTIC Science & Technology

1985-10-01

median service time for a FIST IHQ to service Copperhead missions while in review mode and for mission workload (FO + ARMOR + CPH) was only 6.0...07703 Uazhin.tou, DC 20036 2 Coui-,ander 1 Comwanaer US Aruy larry Diaiaond Labs. US Army Belvoir ATTN: AIILHD- TD , Dr. Scully Research & Development

Flow reversal power limit for the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, L.Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. Themore » experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.« less

Application of the shaped electrode technique to a large area rectangular capacitively coupled plasma reactor to suppress standing wave nonuniformity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sansonnens, L.; Schmidt, H.; Howling, A.A.

The electromagnetic standing wave effect can become the main source of nonuniformity limiting the use of very high frequency in large area reactors exceeding 1 m{sup 2} required for industrial applications. Recently, it has been proposed and shown experimentally in a cylindrical reactor that a shaped electrode in place of the conventional flat electrode can be used in order to suppress the electromagnetic standing wave nonuniformity. In this study, we show experimental measurements demonstrating that the shaped electrode technique can also be applied in large area rectangular reactors. We also present results of electromagnetic screening by a conducting substrate whichmore » has important consequences for industrial application of the shaped electrode technique.« less

Planned Destruction of Metal-Core Reactor: Simulation of Catastrophic Accidents and New Experimental Possibilities

NASA Astrophysics Data System (ADS)

Vorontsov, S. V.; Kuvshinov, M. I.; Narozhnyi, A. T.; Popov, V. A.; Solov'ev, V. P.; Yuferev, V. I.

2017-12-01

A reactor with a destructible core (RIR reactor) generating a pulse with an output of 1.5 × 1019 fissions and a full width at half maximum of 2.5 μs was developed and tested at VNIIEF. In the course of investigation, a computational-experimental method for laboratory calibration of the reactor was created and worked out. This method ensures a high accuracy of predicting the energy release in a real experiment with excess reactivity of 3βeff above prompt criticality. A transportable explosion-proof chamber was also developed, which ensures the safe localization of explosion products of the core of small-sized nuclear devices and charges of high explosives with equivalent mass of up to 100 kg of TNT.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The paper summarizes the results obtained in an exploratory evaluation of ceramics for automobile thermal reactors. Candidate ceramic materials were evaluated in several reactor designs using both engine dynamometer and vehicle road tests. Silicon carbide contained in a corrugated metal support structure exhibited the best performance, lasting 1100 hours in engine dynamometer tests and for more than 38,600 kilimeters (24,000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

Design and evaluation of experimental ceramic automobile thermal reactors

NASA Technical Reports Server (NTRS)

Stone, P. L.; Blankenship, C. P.

1974-01-01

The results obtained in an exploratory evaluation of ceramics for automobile thermal reactors are summarized. Candidate ceramic materials were evaluated in several reactor designs by using both engine-dynamometer and vehicle road tests. Silicon carbide contained in a corrugated-metal support structure exhibited the best performance, lasting 1100 hr in engine-dynamometer tests and more than 38,600 km (24000 miles) in vehicle road tests. Although reactors containing glass-ceramic components did not perform as well as those containing silicon carbide, the glass-ceramics still offer good potential for reactor use with improved reactor designs.

An update of engine system research at the Army Propulsion Directorate

NASA Technical Reports Server (NTRS)

Bobula, George A.

1990-01-01

The Small Turboshaft Engine Research (STER) program provides a vehicle for evaluating the application of emerging technologies to Army turboshaft engine systems and to investigate related phenomena. Capitalizing on the resources at hand, in the form of both the NASA facilities and the Army personnel, the program goal of developing a physical understanding of engine system dynamics and/or system interactions is being realized. STER entries investigate concepts and components developed both in-house and out-of-house. Emphasis is placed upon evaluations which have evolved from on-going basic research and advanced development programs. Army aviation program managers are also encouraged to make use of STER resources, both people and facilities. The STER personnel have established their reputations as experts in the fields of engine system experimental evaluations and engine system related phenomena. The STER facility has demonstrated its utility in both research and development programs. The STER program provides the Army aviation community the opportunity to perform system level investigations, and then to offer the findings to the entire engine community for their consideration in next generation propulsion systems. In this way results of the fundamental research being conducted to meet small turboshaft engine technology challenges expeditiously find their way into that next generation of propulsion systems.

Ozone generation by negative direct current corona discharges in dry air fed coaxial wire-cylinder reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehia, Ashraf; Mizuno, Akira

An analytical study was made in this paper for calculating the ozone generation by negative dc corona discharges. The corona discharges were formed in a coaxial wire-cylinder reactor. The reactor was fed by dry air flowing with constant rates at atmospheric pressure and room temperature, and stressed by a negative dc voltage. The current-voltage characteristics of the negative dc corona discharges formed inside the reactor were measured in parallel with concentration of the generated ozone under different operating conditions. An empirical equation was derived from the experimental results for calculating the ozone concentration generated inside the reactor. The results, thatmore » have been recalculated by using the derived equation, have agreed with the experimental results over the whole range of the investigated parameters, except in the saturation range for the ozone concentration. Therefore, the derived equation represents a suitable criterion for expecting the ozone concentration generated by negative dc corona discharges in dry air fed coaxial wire-cylinder reactors under any operating conditions in range of the investigated parameters.« less

Waste tyre pyrolysis: modelling of a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Grasa, G; Navarro, M V; García, T; Mastral, A M

2010-12-01

This paper describes the development of a new model for waste tyre pyrolysis in a moving bed reactor. This model comprises three different sub-models: a kinetic sub-model that predicts solid conversion in terms of reaction time and temperature, a heat transfer sub-model that calculates the temperature profile inside the particle and the energy flux from the surroundings to the tyre particles and, finally, a hydrodynamic model that predicts the solid flow pattern inside the reactor. These three sub-models have been integrated in order to develop a comprehensive reactor model. Experimental results were obtained in a continuous moving bed reactor and used to validate model predictions, with good approximation achieved between the experimental and simulated results. In addition, a parametric study of the model was carried out, which showed that tyre particle heating is clearly faster than average particle residence time inside the reactor. Therefore, this fast particle heating together with fast reaction kinetics enables total solid conversion to be achieved in this system in accordance with the predictive model. Copyright © 2010 Elsevier Ltd. All rights reserved.

Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

NASA Astrophysics Data System (ADS)

Viellieber, Mathias; Class, Andreas G.

2013-11-01

Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

Control rod calibration and reactivity effects at the IPEN/MB-01 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pinto, Letícia Negrão; Gonnelli, Eduardo; Santos, Adimir dos

2014-11-11

Researches that aim to improve the performance of neutron transport codes and quality of nuclear cross section databases are very important to increase the accuracy of simulations and the quality of the analysis and prediction of phenomena in the nuclear field. In this context, relevant experimental data such as reactivity worth measurements are needed. Control rods may be made of several neutron absorbing materials that are used to adjust the reactivity of the core. For the reactor operation, these experimental data are also extremely important: with them it is possible to estimate the reactivity worth by the movement of themore » control rod, understand the reactor response at each rod position and to operate the reactor safely. This work presents a temperature correction approach for the control rod calibration problem. It is shown the control rod calibration data of the IPEN/MB-01 reactor, the integral and differential reactivity curves and a theoretical analysis, performed by the MCNP-5 reactor physics code, developed and maintained by Los Alamos National Laboratory, using the ENDF/B-VII.0 nuclear data library.« less

The use of experimental data in an MTR-type nuclear reactor safety analysis

NASA Astrophysics Data System (ADS)

Day, Simon E.

Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Scaling and design analyses of a scaled-down, high-temperature test facility for experimental investigation of the initial stages of a VHTR air-ingress accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arcilesi, David J.; Ham, Tae Kyu; Kim, In Hun

2015-07-01

A critical event in the safety analysis of the very high-temperature gas-cooled reactor (VHTR) is an air-ingress accident. This accident is initiated, in its worst case scenario, by a double-ended guillotine break of the coaxial cross vessel, which leads to a rapid reactor vessel depressurization. In a VHTR, the reactor vessel is located within a reactor cavity that is filled with air during normal operating conditions. Following the vessel depressurization, the dominant mode of ingress of an air–helium mixture into the reactor vessel will either be molecular diffusion or density-driven stratified flow. The mode of ingress is hypothesized to dependmore » largely on the break conditions of the cross vessel. Since the time scales of these two ingress phenomena differ by orders of magnitude, it is imperative to understand under which conditions each of these mechanisms will dominate in the air ingress process. Computer models have been developed to analyze this type of accident scenario. There are, however, limited experimental data available to understand the phenomenology of the air-ingress accident and to validate these models. Therefore, there is a need to design and construct a scaled-down experimental test facility to simulate the air-ingress accident scenarios and to collect experimental data. The current paper focuses on the analyses performed for the design and operation of a 1/8th geometric scale (by height and diameter), high-temperature test facility. A geometric scaling analysis for the VHTR, a time scale analysis of the air-ingress phenomenon, a transient depressurization analysis of the reactor vessel, a hydraulic similarity analysis of the test facility, a heat transfer characterization of the hot plenum, a power scaling analysis for the reactor system, and a design analysis of the containment vessel are discussed.« less

ETR, TRA642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

ETR, TRA-642. BASEMENT SPACE ALLOCATION FOR EXPERIMENTERS CA. 1966, SOUTHEAST QUADRANT OF FLOOR. WESTINGHOUSE ATOMIC POWER DIVISION (WAPD) AND BETTIS ATOMIC POWER LABORATORY (BAPL) CONSUME MOST OF THE QUADRANT. PHILLIPS PETROLEUM COMPANY ETR-E-2256, 12/1966. INL INDEX NO. 532-0642-00-706-021256, REV. F. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gonnelli, Eduardo; Diniz, Ricardo

2013-05-06

The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

Novel Anaerobic Wastewater Treatment System for Energy Generation at Forward Operating Bases

DTIC Science & Technology

2016-08-01

AnMBR) technology with clinoptilolite ion exchange and GreenBox™ ammonia electrolysis. The system generates both methane and hydrogen fuels...experimental setup. ................................................ 21 Figure 10. Methane phase semi batch experimental setup, a total of three reactors were...set up for PS + solid, Bioc and ADS methane phase reactors. .................... 21 Figure 11. Dried PS solid for the control, Bioc blend for the

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Optimization of bio-ethanol autothermal reforming and carbon monoxide removal processes

NASA Astrophysics Data System (ADS)

Markova, D.; Bazbauers, G.; Valters, K.; Alhucema Arias, R.; Weuffen, C.; Rochlitz, L.

Experimental investigation of bio-ethanol autothermal reforming (ATR) and water-gas shift (WGS) processes for hydrogen production and regression analysis of the data is performed in the study. The main goal was to obtain regression relations between the most critical dependent variables such as hydrogen, carbon monoxide and methane content in the reformate gas and independent factors such as air-to-fuel ratio (λ), steam-to-carbon ratio (S/C), inlet temperature of reactants into reforming process (T ATRin), pressure (p) and temperature (T ATR) in the ATR reactor from the experimental data. Purpose of the regression models is to provide optimum values of the process factors that give the maximum amount of hydrogen. The experimental ATR system consisted of an evaporator, an ATR reactor and a one-stage WGS reactor. Empirical relations between hydrogen, carbon monoxide, methane content and the controlling parameters downstream of the ATR reactor are shown in the work. The optimization results show that within the considered range of the process factors the maximum hydrogen concentration of 42 dry vol. % and yield of 3.8 mol mol -1 of ethanol downstream of the ATR reactor can be achieved at S/C = 2.5, λ = 0.20-0.23, p = 0.4 bar, T ATRin = 230 °C, T ATR = 640 °C.

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

PubMed

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Evaluation of Selected Chemical Processes for Production of Low-cost Silicon, Phase 3. [using a fluidized bed reactor

NASA Technical Reports Server (NTRS)

Blocher, J. M., Jr.; Browning, M. F.

1979-01-01

The construction and operation of an experimental process system development unit (EPSDU) for the production of granular semiconductor grade silicon by the zinc vapor reduction of silicon tetrachloride in a fluidized bed of seed particles is presented. The construction of the process development unit (PDU) is reported. The PDU consists of four critical units of the EPSDU: the fluidized bed reactor, the reactor by product condenser, the zinc vaporizer, and the electrolytic cell. An experimental wetted wall condenser and its operation are described. Procedures are established for safe handling of SiCl4 leaks and spills from the EPSDU and PDU.

Army Communicator. Volume 37, Number 1, Spring 2012

DTIC Science & Technology

2012-01-01

consortium conducts a risk reduction experiment and lab-based assessment on each of them. A successful example of this risk reduction process was the...include pre-experimental assessments , risk mitigation experiments, and configuration testing. The Network Battle Lab is proud to serve as a...experimentation, assessment , rehearsals, SATCOM/network services and vendor mentorship support to the NIE team. COL Michael Brownfield

Solid Aerosol Program

DTIC Science & Technology

1993-04-01

not to be construed as an official Department of the Army position unless so designated by other authorizing documents. REPORT DOCUMENTATION PAGE...parameter sensitivity studies, and test procedure design . An experimental system providing reaL data on the parametters relevant to the calculations has been...experimental program was designed to exploit as much of the existing capabilities of the Ventilation Kinetics group as possible while keeping in mind

Experimental study on the instability of Pressure Balance Injection System (PBIS)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okamoto, Koji; Teshima, Hideyuki; Madarame, Haruki

1996-06-01

The Passive Safety Reactor has been developed to reduce the construction cost and to improve the safety. Japan Atomic Energy Research institute (JAERI) proposed the System-Integrated Pressurized Water Reactor (SPWR) as a Passive Safety Reactor. In the SPWR design, the Pressure Balanced Injection System (PBIS) was introduced for the passive safety concept. The water with boron in a containment vessel were passively injected into the core by the pressure difference between the containment vessel and reactor vessel at a severe accidental condition. However there are few studies on the thermo-hydraulic characteristics of the PBIS. In this study, the thermal hydraulicsmore » of the PBIS are experimentally investigated using the small scale model. The instability of the injected flow was observed in the adiabatic experiment. The instability was caused by the pressure balance between the two vessels. The mechanism of the instability are discussed, resulting in the good agreement with the experimental results. In the steam experiment, another instability was observed, which was caused by the heat balance in the main tank.« less

Optimization of the moving-bed biofilm sequencing batch reactor (MBSBR) to control aeration time by kinetic computational modeling: Simulated sugar-industry wastewater treatment.

PubMed

Faridnasr, Maryam; Ghanbari, Bastam; Sassani, Ardavan

2016-05-01

A novel approach was applied for optimization of a moving-bed biofilm sequencing batch reactor (MBSBR) to treat sugar-industry wastewater (BOD5=500-2500 and COD=750-3750 mg/L) at 2-4 h of cycle time (CT). Although the experimental data showed that MBSBR reached high BOD5 and COD removal performances, it failed to achieve the standard limits at the mentioned CTs. Thus, optimization of the reactor was rendered by kinetic computational modeling and using statistical error indicator normalized root mean square error (NRMSE). The results of NRMSE revealed that Stover-Kincannon (error=6.40%) and Grau (error=6.15%) models provide better fits to the experimental data and may be used for CT optimization in the reactor. The models predicted required CTs of 4.5, 6.5, 7 and 7.5 h for effluent standardization of 500, 1000, 1500 and 2500 mg/L influent BOD5 concentrations, respectively. Similar pattern of the experimental data also confirmed these findings. Copyright © 2016 Elsevier Ltd. All rights reserved.

Dynamic modeling of temperature change in outdoor operated tubular photobioreactors.

PubMed

Androga, Dominic Deo; Uyar, Basar; Koku, Harun; Eroglu, Inci

2017-07-01

In this study, a one-dimensional transient model was developed to analyze the temperature variation of tubular photobioreactors operated outdoors and the validity of the model was tested by comparing the predictions of the model with the experimental data. The model included the effects of convection and radiative heat exchange on the reactor temperature throughout the day. The temperatures in the reactors increased with increasing solar radiation and air temperatures, and the predicted reactor temperatures corresponded well to the measured experimental values. The heat transferred to the reactor was mainly through radiation: the radiative heat absorbed by the reactor medium, ground radiation, air radiation, and solar (direct and diffuse) radiation, while heat loss was mainly through the heat transfer to the cooling water and forced convection. The amount of heat transferred by reflected radiation and metabolic activities of the bacteria and pump work was negligible. Counter-current cooling was more effective in controlling reactor temperature than co-current cooling. The model developed identifies major heat transfer mechanisms in outdoor operated tubular photobioreactors, and accurately predicts temperature changes in these systems. This is useful in determining cooling duty under transient conditions and scaling up photobioreactors. The photobioreactor design and the thermal modeling were carried out and experimental results obtained for the case study of photofermentative hydrogen production by Rhodobacter capsulatus, but the approach is applicable to photobiological systems that are to be operated under outdoor conditions with significant cooling demands.

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

NASA Astrophysics Data System (ADS)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

Quantification of syntrophic acetate-oxidizing microbial communities in biogas processes

PubMed Central

Westerholm, Maria; Dolfing, Jan; Sherry, Angela; Gray, Neil D; Head, Ian M; Schnürer, Anna

2011-01-01

Changes in communities of syntrophic acetate-oxidizing bacteria (SAOB) and methanogens caused by elevated ammonia levels were quantified in laboratory-scale methanogenic biogas reactors operating at moderate temperature (37°C) using quantitative polymerase chain reaction (qPCR). The experimental reactor was subjected to gradually increasing ammonia levels (0.8–6.9 g NH4+-N l−1), whereas the level of ammonia in the control reactor was kept low (0.65–0.90 g NH4+-N l−1) during the entire period of operation (660 days). Acetate oxidation in the experimental reactor, indicated by increased production of 14CO2 from acetate labelled in the methyl carbon, occurred when ammonia levels reached 5.5 and 6.9 g NH4+-N l−1. Syntrophic acetate oxidizers targeted by newly designed qPCR primers were Thermacetogenium phaeum, Clostridium ultunense, Syntrophaceticus schinkii and Tepidanaerobacter acetatoxydans. The results showed a significant increase in abundance of all these bacteria except T. phaeum in the ammonia-stressed reactor, coincident with the shift to syntrophic acetate oxidation. As the abundance of the bacteria increased, a simultaneous decrease was observed in the abundance of aceticlastic methanogens from the families Methanosaetaceae and Methanosarcinaceae. qPCR analyses of sludge from two additional high ammonia processes, in which methane production from acetate proceeded through syntrophic acetate oxidation (reactor SB) or through aceticlastic degradation (reactor DVX), demonstrated that SAOB were significantly more abundant in the SB reactor than in the DVX reactor. PMID:23761313

THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Colley, J.R.

1962-12-01

The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

Station Blackout Analysis of HTGR-Type Experimental Power Reactor

NASA Astrophysics Data System (ADS)

Syarip; Zuhdi, Aliq; Falah, Sabilul

2018-01-01

The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.

First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set relatively well, with the (110) surface being more active at higher temperaures; in contrast, for the other experimental data set, our reactor simulations achieve surprisingly and perhaps fortuitously good agreement with the activity and phenomenological pressure dependence when it is assumed that the (111) facet is the only active facet present. Lastly, the active phase of catalytic CO oxidation over RuO 2 remains unsettled, but the present study presents proof of principle (and progress) toward more accurate multiscale modeling from electrons to reactors and new simulation results.« less

Electrons to Reactors Multiscale Modeling: Catalytic CO Oxidation over RuO 2

DOE PAGES

Sutton, Jonathan E.; Lorenzi, Juan M.; Krogel, Jaron T.; ...

2018-04-20

First-principles kinetic Monte Carlo (1p-kMC) simulations for CO oxidation on two RuO 2 facets, RuO 2(110) and RuO 2(111), were coupled to the computational fluid dynamics (CFD) simulations package MFIX, and reactor-scale simulations were then performed. 1p-kMC coupled with CFD has recently been shown as a feasible method for translating molecular scale mechanistic knowledge to the reactor scale, enabling comparisons to in situ and online experimental measurements. Only a few studies with such coupling have been published. This work incorporates multiple catalytic surface facets into the scale-coupled simulation, and three possibilities were investigated: the two possibilities of each facet individuallymore » being the dominant phase in the reactor, and also the possibility that both facets were present on the catalyst particles in the ratio predicted by an ab initio thermodynamics-based Wulff construction. When lateral interactions between adsorbates were included in the 1p-kMC simulations, the two surfaces, RuO 2(110) and RuO 2(111), were found to be of similar order-of-magnitude in activity for the pressure range of 1 × 10 –4 bar to 1 bar, with the RuO 2(110) surface-termination showing more simulated activity than the RuO 2(111) surface-termination. Coupling between the 1p-kMC and CFD was achieved with a lookup table generated by the error-based modified Shepard interpolation scheme. Isothermal reactor scale simulations were performed and compared to two separate experimental studies, conducted with reactant partial pressures of ≤0.1 bar. Simulations without an isothermality restriction were also conducted and showed that the simulated temperature gradient across the catalytic reactor bed is <0.5 K, which validated the use of the isothermality restriction for investigating the reactor-scale phenomenological temperature dependences. The approach with the Wulff construction based reactor simulations reproduced a trend similar to one experimental data set relatively well, with the (110) surface being more active at higher temperaures; in contrast, for the other experimental data set, our reactor simulations achieve surprisingly and perhaps fortuitously good agreement with the activity and phenomenological pressure dependence when it is assumed that the (111) facet is the only active facet present. Lastly, the active phase of catalytic CO oxidation over RuO 2 remains unsettled, but the present study presents proof of principle (and progress) toward more accurate multiscale modeling from electrons to reactors and new simulation results.« less

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

DOE Office of Scientific and Technical Information (OSTI.GOV)

Qu, Jianmin

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. Thesemore » materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.« less

The Experimental Breeder Reactor II seismic probabilistic risk assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roglans, J; Hill, D J

1994-02-01

The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

A study of the Coriolis effect on the fluid flow profile in a centrifugal bioreactor.

PubMed

Detzel, Christopher J; Thorson, Michael R; Van Wie, Bernard J; Ivory, Cornelius F

2009-01-01

Increasing demand for tissues, proteins, and antibodies derived from cell culture is necessitating the development and implementation of high cell density bioreactors. A system for studying high density culture is the centrifugal bioreactor (CCBR), which retains cells by increasing settling velocities through system rotation, thereby eliminating diffusional limitations associated with mechanical cell retention devices. This article focuses on the fluid mechanics of the CCBR system by considering Coriolis effects. Such considerations for centrifugal bioprocessing have heretofore been ignored; therefore, a simpler analysis of an empty chamber will be performed. Comparisons are made between numerical simulations and bromophenol blue dye injection experiments. For the non-rotating bioreactor with an inlet velocity of 4.3 cm/s, both the numerical and experimental results show the formation of a teardrop shaped plume of dye following streamlines through the reactor. However, as the reactor is rotated, the simulation predicts the development of vortices and a flow profile dominated by Coriolis forces resulting in the majority of flow up the leading wall of the reactor as dye initially enters the chamber, results are confirmed by experimental observations. As the reactor continues to fill with dye, the simulation predicts dye movement up both walls while experimental observations show the reactor fills with dye from the exit to the inlet. Differences between the simulation and experimental observations can be explained by excessive diffusion required for simulation convergence, and a slight density difference between dyed and un-dyed solutions. Implications of the results on practical bioreactor use are also discussed. (c) 2009 American Institute of Chemical Engineers Biotechnol. Prog., 2009.

A Study of the Coriolis Effect on the Fluid Flow Profile in a Centrifugal Bioreactor

PubMed Central

Detzel, Christopher J.; Thorson, Michael R.; Van Wie, Bernard J.; Ivory, Cornelius F.

2011-01-01

Increasing demand for tissues, proteins, and antibodies derived from cell culture is necessitating the development and implementation of high cell density bioreactors. A system for studying high density culture is the centrifugal bioreactor (CCBR) which retains cells by increasing settling velocities through system rotation, thereby eliminating diffusional limitations associated with mechanical cell retention devices. This paper focuses on the fluid mechanics of the CCBR system by considering Coriolis effects. Such considerations for centrifugal bioprocessing have heretofore been ignored; therefore a simpler analysis of an empty chamber will be performed. Comparisons are made between numerical simulations and bromophenol blue dye injection experiments. For the non-rotating bioreactor with an inlet velocity of 4.3 cm/s, both the numerical and experimental results show the formation of a teardrop shaped plume of dye following streamlines through the reactor. However, as the reactor is rotated the simulation predicts the development of vortices and a flow profile dominated by Coriolis forces resulting in the majority of flow up the leading wall of the reactor as dye initially enters the chamber, results confirmed by experimental observations. As the reactor continues to fill with dye, the simulation predicts dye movement up both walls while experimental observations show the reactor fills with dye from the exit to the inlet. Differences between the simulation and experimental observations can be explained by excessive diffusion required for simulation convergence, and a slight density difference between dyed and un-dyed solutions. Implications of the results on practical bioreactor use are also discussed. PMID:19455639

The Birth of Nuclear-Generated Electricity

DOE R&D Accomplishments Database

1999-09-01

The Experimental Breeder Reactor-I (EBR-I), built in Idaho in 1949, generated the first usable electricity from nuclear power on December 20, 1951. More importantly, the reactor was used to prove that it was possible to create more nuclear fuel in the reactor than it consumed during operation -- fuel breeding. The EBR-I facility is now a National Historic Landmark open to the public.

DESIGN AND HAZARDS SUMMARY REPORT, BOILING REACTOR EXPERIMENT V (BORAX V)

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1961-05-01

Design data for BORAX V are presented along with results of hazards evaluation studies. Considcration of the hazards associated with the operation of BORAX V was based on the following conditions: For normal steady-state power and experimental operation, the reactor and plant are adequately shielded and ventilated to allow personnel to be safely stationed in the turbine building and on the main floor of the reactor building. The control building is located one- half mile distant from the reactor building. For special, hazardous experiments, personnel are withdrawn from the reactor area. (M.C.G.)

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.

1983-06-01

During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

Criticality experiments and analysis of molybdenum reflected cylindrical uranyl fluoride water solution reactors

NASA Technical Reports Server (NTRS)

Fieno, D.; Fox, T.; Mueller, R.

1972-01-01

Clean criticality data were obtained from molybdenum-reflected cylindrical uranyl-fluoride-water solution reactors. Using ENDF/B molybdenum cross sections, a nine energy group two-dimensional transport calculation of a reflected reactor configuration predicted criticality to within 7 cents of the experimental value. For these reactors, it was necessary to compute the reflector resonance integral by a detailed transport calculation at the core-reflector interface volume in the energy region of the two dominant resonances of natural molybdenum.

The use of plasma technology for the treatment of noxious waste

NASA Astrophysics Data System (ADS)

Wilman, Jonathan James

This thesis begins by describing the common types of air pollution and the main effects of these pollutants. Natural and man-made sources are discussed as well as the current types of technology used for reduction of common pollutants. The use of atmospheric pressure non-thermal plasma reactors for the control of pollutants is introduced at this stage. The second chapter describes the different types of atmospheric pressure non-thermal reactor designs and their modes of operation. The fundamental processes behind the production of plasmas are discussed and the chemistry of some simple discharges is also presented. The third chapter begins the experimental and modelling work done at Manchester on the destruction of volatile organic compounds (VOCs) using packed bed reactors and pulsed corona reactors. This chapter is concerned with the destruction of toluene and its behaviour as the oxygen content of the carrier gas, flowing through the reactor, is changed. Work using a pulsed corona reactor is also presented showing the destruction of toluene as a function of the applied specific energy. A model is constructed using mainly atmospheric reactions and the predictions are compared with experimental values. The fourth chapter discusses the destruction of dichloromethane (DCM) as a function of the oxygen content of the carrier gas. A model is constructed, again from mainly atmospheric reactions, and the predictions compared with the experimental data obtained. Methane is chosen as a molecule to study in the fifth chapter. A model is constructed and compared with experimental findings, showing that the chemistry of non-thermal plasmas can be effectively represented using neutral gas phase chemistry. Finally chapter six is concerned with the use of a large scale pulsed corona system for the reduction of NO[x] in industrial flue gas. This system has been tested on a modem incinerator, showing encouraging results. The workings of a modem incinerator are described together with those of the corona facility and any instruments used in these tests. Some experimental results are discussed. The aim of this chapter is to show that plasma reactors can be scaled up for industrial use. This section also discusses the difficulty of analysing and working with industrial gases and large scale apparatus as opposed to laboratory scale experiments.

Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: simulations and experimental validations [ANIMMA--2015-IO-98

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jammes, C.; Filliatre, P.; De Izarra, G.

The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. (authors)

Ultrafiltration membrane reactors for enzymatic resolution of amino acids: design model and optimization.

PubMed

Bódalo, A; Gómez, J L.; Gómez, E; Bastida, J; Máximo, M F.; Montiel, M C.

2001-03-08

In this paper the possibility of continuous resolution of DL-phenylalanine, catalyzed by L-aminoacylase in a ultrafiltration membrane reactor (UFMR) is presented. A simple design model, based on previous kinetic studies, has been demonstrated to be capable of describing the behavior of the experimental system. The model has been used to determine the optimal experimental conditions to carry out the asymmetrical hydrolysis of N-acetyl-DL-phenylalanine.

Ballistic Characterization of the Scalability of Magnesium Alloy AMX602

DTIC Science & Technology

2015-07-01

Powder Metallurgy 4 5. Fabrication Procedure 4 6. Mechanical Property Analysis 5 7. Ballistic Experimental Procedures 6 8. Ballistic Experimental...compositions of noncombustive Mg alloy powders 4. Powder Metallurgy The powder was consolidated at room temperature using a 2,000-kN hydraulic press...evaluation of advanced powder metallurgy magnesium alloys for dynamic applications. Aberdeen Proving Ground (MD): Army Research Laboratory (US); 2009 May

Cooling molten salt reactors using "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

2014-08-01

This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

LBE water interaction in sub-critical reactors: First experimental and modelling results

NASA Astrophysics Data System (ADS)

Ciampichetti, A.; Agostini, P.; Benamati, G.; Bandini, G.; Pellini, D.; Forgione, N.; Oriolo, F.; Ambrosini, W.

2008-06-01

This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors.

Development of a polysilicon process based on chemical vapor deposition, phase 1 and phase 2

NASA Technical Reports Server (NTRS)

Plahutnik, F.; Arvidson, A.; Sawyer, D.; Sharp, K.

1982-01-01

High-purity polycrystalline silicon was produced in an experimental, intermediate and advanced CVD reactor. Data from the intermediate and advanced reactors confirmed earlier results obtained in the experimental reactor. Solar cells were fabricated by Westinghouse Electric and Applied Solar Research Corporation which met or exceeded baseline cell efficiencies. Feedstocks containing trichlorosilane or silicon tetrachloride are not viable as etch promoters to reduce silicon deposition on bell jars. Neither are they capable of meeting program goals for the 1000 MT/yr plant. Post-run CH1 etch was found to be a reasonably effective method of reducing silicon deposition on bell jars. Using dichlorosilane as feedstock met the low-cost solar array deposition goal (2.0 gh-1-cm-1), however, conversion efficiency was approximately 10% lower than the targeted value of 40 mole percent (32 to 36% achieved), and power consumption was approximately 20 kWh/kg over target at the reactor.

Simulation of the neutron flux in the irradiation facility at RA-3 reactor.

PubMed

Bortolussi, S; Pinto, J M; Thorp, S I; Farias, R O; Soto, M S; Sztejnberg, M; Pozzi, E C C; Gonzalez, S J; Gadan, M A; Bellino, A N; Quintana, J; Altieri, S; Miller, M

2011-12-01

A facility for the irradiation of a section of patients' explanted liver and lung was constructed at RA-3 reactor, Comisión Nacional de Energía Atómica, Argentina. The facility, located in the thermal column, is characterized by the possibility to insert and extract samples without the need to shutdown the reactor. In order to reach the best levels of security and efficacy of the treatment, it is necessary to perform an accurate dosimetry. The possibility to simulate neutron flux and absorbed dose in the explanted organs, together with the experimental dosimetry, allows setting more precise and effective treatment plans. To this end, a computational model of the entire reactor was set-up, and the simulations were validated with the experimental measurements performed in the facility. Copyright © 2011 Elsevier Ltd. All rights reserved.

Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

PubMed

Töre, Candan; Ortego, Pedro

2005-01-01

The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

Criticality Safety Evaluation of the LLNL Inherently Safe Subcritical Assembly (ISSA)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Percher, Catherine

2012-06-19

The LLNL Nuclear Criticality Safety Division has developed a training center to illustrate criticality safety and reactor physics concepts through hands-on experimental training. The experimental assembly, the Inherently Safe Subcritical Assembly (ISSA), uses surplus highly enriched research reactor fuel configured in a water tank. The training activities will be conducted by LLNL following the requirements of an Integration Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of LLNL instructors. This report provides the technical criticality safety basis for instructional operations with the ISSA experimental assembly.

Metal fires and their implications for advanced reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in thesemore » areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety analysis capabilities of the advanced-reactor community for directly relevant scenarios. Beyond the focus on the thermally-interacting and smoldering sodium pool fires, experimental and analysis capabilities for sodium spray fires have also been developed in this project.« less

Moving bed reactor setup to study complex gas-solid reactions.

PubMed

Gupta, Puneet; Velazquez-Vargas, Luis G; Valentine, Charles; Fan, Liang-Shih

2007-08-01

A moving bed scale reactor setup for studying complex gas-solid reactions has been designed in order to obtain kinetic data for scale-up purpose. In this bench scale reactor setup, gas and solid reactants can be contacted in a cocurrent and countercurrent manner at high temperatures. Gas and solid sampling can be performed through the reactor bed with their composition profiles determined at steady state. The reactor setup can be used to evaluate and corroborate model parameters accounting for intrinsic reaction rates in both simple and complex gas-solid reaction systems. The moving bed design allows experimentation over a variety of gas and solid compositions in a single experiment unlike differential bed reactors where the gas composition is usually fixed. The data obtained from the reactor can also be used for direct scale-up of designs for moving bed reactors.

[The SILENE reactor: a tool adapted for applied study of moderate and large doses].

PubMed

Verrey, B; Leo, Y; Fouillaud, P

2002-07-01

Designed in 1974 to study the phenomenology and consequences of a critical accident, the SILENE experimental reactor, an intense source of mixed neutron and gamma radiation, is also suited to radiobiological studies.

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

U.S. Army Medical Department Journal (October-December 2006)

DTIC Science & Technology

2006-12-01

prioritize Soldiers with medical conditions who need additional neurophysiological and/or neuroimaging evaluations. If local national facilities are...of Enlightenment . Focused experimentation and solid hard work by an increasingly diverse range of organizations leads to a true understanding of the

Unit mechanisms of fission gas release: Current understanding and future needs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel properties and, once the gas is released into the gap between the fuel and cladding, lowering gap thermal conductivity and increasing gap pressure. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are being applied to provide unprecedented understanding of the unit mechanisms that define the fission product behavior. In this article, existing research on the basic mechanisms behind the various stages of fission gas releasemore » during normal reactor operation are summarized and critical areas where experimental and simulation work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior during reactor operation and to design fuels that have improved fission product retention. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. Quarterly Progress Report, October 1-December 31, 1963

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1964-02-15

The ML-1 power plant did not operate during the report period; low power reactor physics and shielding experiments were conducted with the ML-1 reactor. Evaluation of moderate corrosion observed on aluminum parts exposed to the ML-1 shield solution indicated no loss of performance capability. Preliminary tests showed that the corrosion probably was caused by heavy metal ions or chlorides in the solution, Massive corrosion observed on the ML-1 fuel element lower spiders was attributed to sub-standard material; failure of some spiders was attributed to a combination of corrosion and sub-standard fabrication. Evaluation indicated that the upper spiders will perform satisfactorilymore » for the design lifetime. Modification, repair, and reassembly of the CSN-1A t-c set was completed. Operation demonstrated bearing stability, but showed that the turbine effective flow area was too large. A bypass flow path in the turbine was being corrected. The TCS-670 t-c set will be stored indefinitely. Since a commercial alternator will be used for the ML-1A, further development of the brushless alternator was postponed indefinitely. Evaluation revealed that the ML-1 improved precooler design was not compatible with ML-1A requirements. Operntion of the IB-17R-2 and -3 test elements in the GETR continued without incident. Preliminary design of the ML-1A power plant was initiated. Design of modifications to the GCRE facility to adapt it to testing the ML-1 reactor skid was initiated. (auth)« less

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

NASA Astrophysics Data System (ADS)

Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

2018-01-01

The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tan, Lizhen; Yang, Ying; Sridharan, K.

2015-12-01

The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computationalmore » tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe 2+) irradiation.« less

Analytic expressions for Atomic Layer Deposition: coverage, throughput, and materials utilization in cross-flow, particle coating, and spatial ALD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yanguas-Gil, Angel; Elam, Jeffrey W.

2014-05-01

In this work, the authors present analytic models for atomic layer deposition (ALD) in three common experimental configurations: cross-flow, particle coating, and spatial ALD. These models, based on the plug-flow and well-mixed approximations, allow us to determine the minimum dose times and materials utilization for all three configurations. A comparison between the three models shows that throughput and precursor utilization can each be expressed by universal equations, in which the particularity of the experimental system is contained in a single parameter related to the residence time of the precursor in the reactor. For the case of cross-flow reactors, the authorsmore » show how simple analytic expressions for the reactor saturation profiles agree well with experimental results. Consequently, the analytic model can be used to extract information about the ALD surface chemistry (e. g., the reaction probability) by comparing the analytic and experimental saturation profiles, providing a useful tool for characterizing new and existing ALD processes. (C) 2014 American Vacuum Society« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soldevilla, M.; Salmons, S.; Espinosa, B.

The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a uniquemore » repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)« less

Clinical Investigation Program, RCS MED-300 (RI).

DTIC Science & Technology

1984-10-01

temperature/dry. Technical Approach: 1) Experimental desion: a post-test only, equivalent - group experimental design will be used in this study. Random...5650 * It. CONTROLLING OFFICE NAME AND ADDRESS 12. REPORT DATE Commander October 1984 Dwight David Eisenhower Army Medical Center 1s. NUMBEROFPAGES...Fort Gordon, Georgia 30905-5650 210 14. MONITORING AGENCY NAME I ADORESS(11 different from Controlling Offce) IS. SECURITY CLASS. (of tAle MP*H

Assessing Sustainment Operations in a Decisive Action Training Environment

DTIC Science & Technology

2017-05-01

The findings in this Research Product are not to be construed as an official Department of the Army position, unless so designated by other authorized...Combat Training Center (CTC) rotations. The research design allowed for comparison of a control and experimental group. The experimental group received...CP Operations performance. The pocket-sized guide was designed as a quick reference for the proper planning, execution, and follow up of CP

Assessing Consequential Scenarios in a Complex Operational Environment Using Agent Based Simulation

DTIC Science & Technology

2017-03-16

RWISE) 93 5.1.5 Conflict Modeling, Planning, and Outcomes Experimentation Program (COMPOEX) 94 5.1.6 Joint Non -Kinetic Effects Model (JNEM)/Athena... experimental design and testing. 4.3.8 Types and Attributes of Agent-Based Model Design Patterns Using the aforementioned ABM flowchart design methodology ...speed, or flexibility during tactical US Army wargaming. The report considers methodologies to improve analysis of the human domain, identifies

Battle Lab Simulation Collaboration Environment (BLSCE): Multipurpose Platform for Simulation C2

DTIC Science & Technology

2006-06-01

encryption, low-probability of intercept and detection communications, and specialized intelligent agents will provide the brick an d mortar for our...echelons. It allows multi-celled experimentations among several locations that cover all of the United States. It has become a gateway for Joint...of exercises from remote locations , including live-force play. • Integration of combined arms experimentation in support of Army Transformation

A review and forecast of engine system research at the Army Propulsion Directorate

NASA Technical Reports Server (NTRS)

Bobula, George A.

1989-01-01

An account is given of the development status and achievements to date of the U.S. Army Propulsion Directorate's Small Turbine Engine Research (STER) programs, which are experimental investigations of the physics of entire engine systems from the viewpoints of component interactions and/or system dynamics. STER efforts are oriented toward the evaluation of complete turboshaft engine advanced concepts and are conducted at the ECRL-2 indoor, sea-level engine test facility. Attention is given to the results obtained by STER experiments concerned with IR-suppressing engine exhausts, a ceramic turbine-blade shroud, an active shaft-vibration control system, and a ceramic-matrix combustor liner.

Generating unstructured nuclear reactor core meshes in parallel

DOE PAGES

Jain, Rajeev; Tautges, Timothy J.

2014-10-24

Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns

2014-01-01

The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during themore » development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.« less

Influence of Natural Convection and Thermal Radiation Multi-Component Transport in MOCVD Reactors

NASA Technical Reports Server (NTRS)

Lowry, S.; Krishnan, A.; Clark, I.

1999-01-01

The influence of Grashof and Reynolds number in Metal Organic Chemical Vapor (MOCVD) reactors is being investigated under a combined empirical/numerical study. As part of that research, the deposition of Indium Phosphide in an MOCVD reactor is modeled using the computational code CFD-ACE. The model includes the effects of convection, conduction, and radiation as well as multi-component diffusion and multi-step surface/gas phase chemistry. The results of the prediction are compared with experimental data for a commercial reactor and analyzed with respect to the model accuracy.

Experimental heat transfer distribution on the SNAP 10A reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hopenfeld, J.; Toews, R.E.

1965-01-29

Heating distributions have been obtained for the SNAP 10A reactor by means of a thermal paint technique in the Rhodes and Bloxsom 60 in. hypersonic wind tunnel. Data and correlations are presented only for those reactor components where the ratio of the local heat transfer to that on the stagnation point of the calibration sphere was found to be independent of tunnel conditions. It is shown that these heating distributions can be applied directly to reentry conditions provided the thermally painted and the bare reactor surfaces are both catalytic to atom recombination.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

Nuclear reactors built, being built, or planned, 1991

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simpson, B.

1992-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor ismore » an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

NASA Astrophysics Data System (ADS)

Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

2017-12-01

The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

Helium-3 blankets for tritium breeding in fusion reactors

NASA Technical Reports Server (NTRS)

Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

1988-01-01

It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

Assessment of start-up mechanisms for anaerobic fluidized bed reactor in series based on mathematical simulation

NASA Astrophysics Data System (ADS)

Sudibyo, Hanifrahmawan; Guntama, Dody; Budhijanto, Wiratni

2017-05-01

Anaerobic digestion is associated with long hydraulic residence time and hence leads to huge reactor volume, especially for high rate input to the reactor. To overcome this major drawback, one of the possibilities is optimizing the schemes of reactor configuration and start-up mechanisms. This study aimed to determine the most promising start-up mechanism for anaerobic digestion reactors in series, with respect to the shortest hydraulic residence time to reach the highest biogas production rate. The reactor to be studied is anaerobic fluidized bed reactor (AFBR) which is known as the most efficient reactor for high organic loading rate. Case to be studied is landfill leachate digestion. Although reactor optimization can be conducted experimentally, it could be expensive and time consuming. This study proposed the utilization of mathematical modeling to screen the possibilities towards the best options to be verified experimentally. Kinetic study of landfill leachate anaerobic digestion was first conducted to depict the rate of microbial growth and the rate of substrate consumption. Kinetics constants obtained from this batch experiment were then used in the mathematical model representing AFBR. Several mechanisms were simulated in this study. In the first mechanism, all digesters were started simultaneously. In the second mechanism, each digester was started until it achieved steady-state condition before the next digester was started. The third mechanism was start-up scenario for single reactor as opposed to the previous two mechanisms. These all three mechanisms were simulated for either one-through stream and recycling a portion of the reactor effluent. The mathematical simulation result was used to evaluate each mechanism based on hydraulic residence time required for all digesters in series to reach the steady-state condition, the extent of pollutant removal, and the rate of biogas production. In the need of high sCOD removal, the second mechanism emerged as the best one.

Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

NASA Astrophysics Data System (ADS)

Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

2018-03-01

Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

Batch Tests To Determine Activity Distribution and Kinetic Parameters for Acetate Utilization in Expanded-Bed Anaerobic Reactors

PubMed Central

Fox, Peter; Suidan, Makram T.

1990-01-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (Ks) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for Ks. However, Ks was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of Ks on the effluent 3-ethylphenol concentration. A two-parameter search determined a Ks of 8.99 mg of acetate per liter and a Ki of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made. PMID:16348175

Batch tests to determine activity distribution and kinetic parameters for acetate utilization in expanded-bed anaerobic reactors.

PubMed

Fox, P; Suidan, M T

1990-04-01

Batch tests to measure maximum acetate utilization rates were used to determine the distribution of acetate utilizers in expanded-bed sand and expanded-bed granular activated carbon (GAC) reactors. The reactors were fed a mixture of acetate and 3-ethylphenol, and they contained the same predominant aceticlastic methanogen, Methanothrix sp. Batch tests were performed both on the entire reactor contents and with media removed from the reactors. Results indicated that activity was evenly distributed within the GAC reactors, whereas in the sand reactor a sludge blanket on top of the sand bed contained approximately 50% of the activity. The Monod half-velocity constant (K(s)) for the acetate-utilizing methanogens in two expanded-bed GAC reactors was searched for by combining steady-state results with batch test data. All parameters necessary to develop a model with Monod kinetics were experimentally determined except for K(s). However, K(s) was a function of the effluent 3-ethylphenol concentration, and batch test results demonstrated that maximum acetate utilization rates were not a function of the effluent 3-ethylphenol concentration. Addition of a competitive inhibition term into the Monod expression predicted the dependence of K(s) on the effluent 3-ethylphenol concentration. A two-parameter search determined a K(s) of 8.99 mg of acetate per liter and a K(i) of 2.41 mg of 3-ethylphenol per liter. Model predictions were in agreement with experimental observations for all effluent 3-ethylphenol concentrations. Batch tests measured the activity for a specific substrate and determined the distribution of activity in the reactor. The use of steady-state data in conjunction with batch test results reduced the number of unknown kinetic parameters and thereby reduced the uncertainty in the results and the assumptions made.

3D magnetohydrodynamic modelling of a dc low-current plasma arc batch reactor at very high pressure in helium

NASA Astrophysics Data System (ADS)

Lebouvier, A.; Iwarere, S. A.; Ramjugernath, D.; Fulcheri, L.

2013-04-01

This paper deals with a three-dimensional (3D) time-dependent magnetohydrodynamic (MHD) model under peculiar conditions of very high pressures (from 2 MPa up to 10 MPa) and low currents (<1 A). Studies on plasma arc working under these unusual conditions remain almost unexplored because of the technical and technological challenges to develop a reactor able to sustain a plasma at very high pressures. The combined effect of plasma reactivity and high pressure would probably open the way towards new promising applications in various fields: chemistry, lightning, materials or nanomaterial synthesis. A MHD model helps one to understand the complex and coupled phenomena surrounding the plasma which cannot be understood by simply experimentation. The model also provides data which are difficult to directly determine experimentally. The model simulates an experimental-based batch reactor working with helium. The particular reactor in question was used to investigate the Fischer-Tropsch application, fluorocarbon production and CO2 retro-conversion. However, as a first approach in terms of MHD, the model considers the case for helium as a non-reactive working gas. After a detailed presentation of the model, a reference case has been fully analysed (P = 8 MPa, I = 0.35 A) in terms of physical properties. The results show a bending of the arc and displacement of the anodic arc root towards the top of the reactor, due to the combined effects of convection, gravity and electromagnetic forces. A parametric study on the pressure (2-10 MPa) and current (0.25-0.4 A) was then investigated. The operating pressure does not show an influence on the contraction of the arc but higher pressures involve a higher natural convection in the reactor, driven by the density gradients between the cold and hot gas.

Study on bubbly flow behavior in natural circulation reactor by thermal-hydraulic simulation tests with SF6-Gas and ethanol liquid

NASA Astrophysics Data System (ADS)

Kondo, Yoshiyuki; Suga, Keishi; Hibi, Koki; Okazaki, Toshihiko; Komeno, Toshihiro; Kunugi, Tomoaki; Serizawa, Akimi; Yoneda, Kimitoshi; Arai, Takahiro

2009-02-01

An advanced experimental technique has been developed to simulate two-phase flow behavior in a light water reactor (LWR). The technique applies three kinds of methods; (1) use of sulfur-hexafluoride (SF6) gas and ethanol (C2H5OH) liquid at atmospheric temperature and a pressure less than 1.0MPa, where the fluid properties are similar to steam-water ones in the LWR, (2) generation of bubble with a sintering tube, which simulates bubble generation on heated surface in the LWR, (3) measurement of detailed bubble distribution data with a bi-optical probe (BOP), (4) and measurement of liquid velocities with the tracer liquid. This experimental technique provides easy visualization of flows by using a large scale experimental apparatus, which gives three-dimensional flows, and measurement of detailed spatial distributions of two-phase flow. With this technique, we have carried out experiments simulating two-phase flow behavior in a single-channel geometry, a multi-rod-bundle one, and a horizontal-tube-bundle one on a typical natural circulation reactor system. Those experiments have clarified a) a flow regime map in a rod bundle on the transient region between bubbly and churn flow, b) three-dimensional flow behaviour in rod-bundles where inter-subassembly cross-flow occurs, c) bubble-separation behavior with consideration of reactor internal structures. The data have given analysis models for the natural circulation reactor design with good extrapolation.

[Optimization Study on the Nitrogen and Phosphorus Removal of Modified Two- sludge System Under the Condition of Low Carbon Source].

PubMed

Yang, Wei-qiang; Wang, Dong-bo; Li, Xiao-ming; Yang, Qi; Xu, Qiu-xiang; Zhang, Zhi-bei; Li, Zhi-jun; Xiang, Hai-hong; Wang, Ya-li; Sun, Jian

2016-04-15

This paper explored the method of resolving insufficient carbon source in urban sewage by comparing and analyzing denitrification and phosphorus removal (NPR) effect between modified two-sludge system and traditional anaerobic-aerobic-anoxic process under the condition of low carbon source wastewater. The modified two-sludge system was the experimental reactor, which was optimized by adding two stages of micro-aeration (aeration rate 0.5 L · mm⁻¹) in the anoxic period of the original two-sludge system, and multi-stage anaerobic-aerobic-anoxic SBR was the control reactor. When the influent COD, ammonia nitrogen, SOP concentration were respectively 200, 35, 10 mg · L⁻¹, the NPR effect of the experimental reactor was hetter than that of thecontrol reactor with the removal efficiency of TN being 94.8% vs 60.9%, and TP removal being 96.5% vs 75%, respectively. The effluent SOP, ammonia, TN concentration of the experimental reactor were 0.35, 0.50, 1.82 mg · L⁻¹, respectively, which could fully meet the first class of A standard of the Pollutants Emission Standard of Urban Wastewater Treatment Firm (GB 18918-2002). Using the optimized treatment process, the largest amounts of nitrogen and phosphorus removal per unit carbon source (as COD) were 0.17 g · g⁻¹ and 0.048 g · g⁻¹ respectively, which could furthest solve the lower carbon concentration in current municipal wastewater.

Next generation fuel irradiation capability in the High Flux Reactor Petten

NASA Astrophysics Data System (ADS)

Fütterer, Michael A.; D'Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco; Raison, Philippe; Bakker, Klaas; de Groot, Sander; Klaassen, Frodo

2009-07-01

This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

Energy spectrum of 208Pb(n,x) reactions

NASA Astrophysics Data System (ADS)

Tel, E.; Kavun, Y.; Özdoǧan, H.; Kaplan, A.

2018-02-01

Fission and fusion reactor technologies have been investigated since 1950's on the world. For reactor technology, fission and fusion reaction investigations are play important role for improve new generation technologies. Especially, neutron reaction studies have an important place in the development of nuclear materials. So neutron effects on materials should study as theoretically and experimentally for improve reactor design. For this reason, Nuclear reaction codes are very useful tools when experimental data are unavailable. For such circumstances scientists created many nuclear reaction codes such as ALICE/ASH, CEM95, PCROSS, TALYS, GEANT, FLUKA. In this study we used ALICE/ASH, PCROSS and CEM95 codes for energy spectrum calculation of outgoing particles from Pb bombardment by neutron. While Weisskopf-Ewing model has been used for the equilibrium process in the calculations, full exciton, hybrid and geometry dependent hybrid nuclear reaction models have been used for the pre-equilibrium process. The calculated results have been discussed and compared with the experimental data taken from EXFOR.

Towards reducing impact-induced brain injury: lessons from a computational study of army and football helmet pads.

PubMed

Moss, William C; King, Michael J; Blackman, Eric G

2014-01-01

We use computational simulations to compare the impact response of different football and U.S. Army helmet pad materials. We conduct experiments to characterise the material response of different helmet pads. We simulate experimental helmet impact tests performed by the U.S. Army to validate our methods. We then simulate a cylindrical impactor striking different pads. The acceleration history of the impactor is used to calculate the head injury criterion for each pad. We conduct sensitivity studies exploring the effects of pad composition, geometry and material stiffness. We find that (1) the football pad materials do not outperform the currently used military pad material in militarily relevant impact scenarios; (2) optimal material properties for a pad depend on impact energy and (3) thicker pads perform better at all velocities. Although we considered only the isolated response of pad materials, not entire helmet systems, our analysis suggests that by using larger helmet shells with correspondingly thicker pads, impact-induced traumatic brain injury may be reduced.

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ballard, D.L.; Brown, W.W.; Harrison, C.W.

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

Schedule and status of irradiation experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

1998-09-01

The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has one irradiation experiment in reactor and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

Preliminary Tritium Management Design Activities at ORNL

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.

2016-09-01

Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritiummore » mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.« less

Measurement of the 23Na(n,2n) cross section in 235U and 252Cf fission neutron spectra

NASA Astrophysics Data System (ADS)

Košťál, Michal; Schulc, Martin; Rypar, Vojtěch; Losa, Evžen; Švadlenková, Marie; Baroň, Petr; Jánský, Bohumil; Novák, Evžen; Mareček, Martin; Uhlíř, Jan

2017-09-01

The presented paper aims to compare the calculated and experimental reaction rates of 23Na(n,2n)22Na in a well-defined reactor spectra and in the spontaneous fission spectrum of 252Cf. The experimentally determined reaction rate, derived using gamma spectroscopy of irradiated NaF sample, is used for average cross section determination.Estimation of this cross-section is important as it is included in International Reactor Dosimetry and Fusion File and is also relevant to the correct estimation of long-term activity of Na coolant in Sodium Fast Reactors. The calculations were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. In the case of reactor spectrum, reasonable agreement was not achieved with any library. However, in the case of 252Cf spectrum agreement was achieved with IRDFF, JEFF-3.1 and JENDL libraries.

An investigation of tritium transfer in reactor loops

NASA Astrophysics Data System (ADS)

Ilyasova, O. H.; Mosunova, N. A.

2017-09-01

The work is devoted to the important task of the numerical simulation and analysis of the tritium behaviour in the reactor loops. The simulation was carried out by HYDRA-IBRAE/LM code, which is being developed in Nuclear safety institute of the Russian Academy of Sciences. The code is intended for modeling of the liquid metal flow (sodium, lead and lead-bismuth) on the base of non-homogeneous and non-equilibrium two-fluid model. In order to simulate tritium transfer in the code, the special module has been developed. Module includes the models describing the main phenomena of tritium behaviour in reactor loops: transfer, permeation, leakage, etc. Because of shortage of the experimental data, a lot of analytical tests and comparative calculations were considered. Some of them are presented in this work. The comparison of estimation results and experimental and analytical data demonstrate not only qualitative but also good quantitative agreement. It is possible to confirm that HYDRA-IBRAE/LM code allows modeling tritium transfer in reactor loops.

Investigating the kinetics of the enzymatic depolymerization of polygalacturonic acid in continuous UF-membrane reactors.

PubMed

Gallifuoco, Alberto; Cantarella, Maria; Marucci, Mariagrazia

2007-01-01

A stirred tank membrane reactor is used to study the kinetics of polygalacturonic acid (PGA) enzymatic hydrolysis. The reactor operates in semicontinuous configuration: the native biopolymer is loaded at the initial time and the system is continuously fed with the buffer. The effect of retention time (from 101 to 142 min) and membrane molecular weight cutoff (from 1 to 30 kDa) on the rate of permeable oligomers production is investigated. Reaction products are clustered in two different classes, those sized below the membrane cutoff and those above. The reducing power measured in the permeate is used as an estimate of total product concentration. The characteristic breakdown times range from 40 to 100 min. The overall kinetics obeys a first-order law with a characteristic time estimated to 24 min. New mathematical data handling are developed and illustrated using the experimental data obtained. Finally, the body of the experimental results suggests useful indications (reactor productivity, breakdown induction period) for implementing the bioprocess at the industrial scale.

Detailed Measurements of the Aeroelastic Response of a Rigid Coaxial Rotor in Hover

DTIC Science & Technology

2017-08-11

included: hover testing of single and CCR rotors (Year 1), deformation measurement and modal identification of rotor blades in the non -rotating and...the rotor blades, as well as the detailed experimental data were shared with Dr. Rajneesh Singh and Dr. Hao Kang at Vehicle Technology Directorate...VTD), Aberdeen Proving Grounds, MD. In this way, the experimental data could be used to validate US Army comprehensive analysis tools, specifically

High Pressure Combustion Experimental Facility(HPCEF) for Studies on Combustion in Reactive Flows

DTIC Science & Technology

2017-12-13

SECURITY CLASSIFICATION OF: 1. REPORT DATE (DD-MM-YYYY) 4. TITLE AND SUBTITLE 13. SUPPLEMENTARY NOTES 12. DISTRIBUTION AVAILIBILITY STATEMENT 6...Report: High Pressure Combustion Experimental Facility (HPCEF) for Studies on Combustion in Reactive Flows The views, opinions and/or findings... contained in this report are those of the author(s) and should not contrued as an official Department of the Army position, policy or decision, unless so

Computer Simulations: A Tool to Predict Experimental Parameters with Cold Atoms

DTIC Science & Technology

2013-04-01

Department of the Army position unless so designated by other authorized documents. Citation of manufacturer’s or trade names does not constitute an...specifically designed to work with cold atom systems and atom chips, and is already able to compute their key properties. We simulate our experimental...also allows one to choose different physics and define the interdependencies between them. It is not specifically designed for cold atom systems or

Applications of computational modeling in ballistics

NASA Technical Reports Server (NTRS)

Sturek, Walter B.

1987-01-01

The development of the technology of ballistics as applied to gun launched Army weapon systems is the main objective of research at the U.S. Army Ballistic Research Laboratory (BRL). The primary research programs at the BRL consist of three major ballistic disciplines: exterior, interior, and terminal. The work done at the BRL in these areas was traditionally highly dependent on experimental testing. A considerable emphasis was placed on the development of computational modeling to augment the experimental testing in the development cycle; however, the impact of the computational modeling to this date is modest. With the availability of supercomputer computational resources recently installed at the BRL, a new emphasis on the application of computational modeling to ballistics technology is taking place. The major application areas are outlined which are receiving considerable attention at the BRL at present and to indicate the modeling approaches involved. An attempt was made to give some information as to the degree of success achieved and indicate the areas of greatest need.

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Experimental evidences of 95 mTc production in a nuclear reactor.

PubMed

Cohen, I M; Robles, A; Mendoza, P; Airas, R M; Montoya, E H

2018-05-01

95 m Tc has been identified as by-product in some solutions of 99 m Tc obtained by irradiation of molybdenum trioxide in a reactor neutron flux. The characterization was carried out using both measurements by gamma spectrometry and half-life determination. The possible ways that lead to the 95 m Tc production in a nuclear reactor are discussed. Copyright © 2018. Published by Elsevier Ltd.

THE ARMOUR DUST FUELED REACTOR (ADFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krucoff, D.

1958-01-01

The A-DFR is based on the use of a fissionable dust carried in a gas. This fuel ferm offers promise of a major economic advance through the use of 2,000 to 3,000 F operating temperatures and a low cost fuel cycle. The development program is described that was initiated to investigate experimentally the proposed fuel and study analytically other reactor characteristics. A brief review of the reactor concept is presented. (W.D.M.)

Integrated detoxification methodology of hazardous phenolic wastewaters in environmentally based trickle-bed reactors: Experimental investigation and CFD simulation.

PubMed

Lopes, Rodrigo J G; Almeida, Teresa S A; Quinta-Ferreira, Rosa M

2011-05-15

Centralized environmental regulations require the use of efficient detoxification technologies for the secure disposal of hazardous wastewaters. Guided by federal directives, existing plants need reengineering activities and careful analysis to improve their overall effectiveness and to become environmentally friendly. Here, we illustrate the application of an integrated methodology which encompasses the experimental investigation of catalytic wet air oxidation and CFD simulation of trickle-bed reactors. As long as trickle-bed reactors are determined by the flow environment coupled with chemical kinetics, first, on the optimization of prominent numerical solution parameters, the CFD model was validated with experimental data taken from a trickle bed pilot plant specifically designed for the catalytic wet oxidation of phenolic wastewaters. Second, several experimental and computational runs were carried out under unsteady-state operation to evaluate the dynamic performance addressing the TOC concentration and temperature profiles. CFD computations of total organic carbon conversion were found to agree better with experimental data at lower temperatures. Finally, the comparison of test data with simulation results demonstrated that this integrated framework was able to describe the mineralization of organic matter in trickle beds and the validated consequence model can be exploited to promote cleaner remediation technologies of contaminated waters. Copyright © 2011 Elsevier B.V. All rights reserved.

A Measurement of the Absolute Reactor Antineutrino Flux and Spectrum at Daya Bay

NASA Astrophysics Data System (ADS)

An, Fengpeng

2017-12-01

The Daya Bay Reactor Neutrino Experiment uses an array of eight underground detectors to study antineutrinos from six reactor cores with different baselines. Since the start of data-taking from late 2011, Daya Bay has collected the largest sample of reactor antineutrino events to date, and has made the most precise measurement of the neutrino oscillation parameters sin22θ13 and Δm2ee. Using the data from the four detectors in the near experimental halls, Daya Bay has made a high statistics measurement of the absolute reactor antineutrino flux and spectrum. In this paper we will present this measurement and its comparison to predictions based on different flux models.

Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

D.M. McEligot; K. G. Condie; G. E. McCreery

2005-10-01

Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generationmore » IV program.« less

Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard

2016-12-29

The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate themore » models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.« less

Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

NASA Astrophysics Data System (ADS)

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

The Simulator Development for RDE Reactor

NASA Astrophysics Data System (ADS)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Neutron flux and power in RTP core-15

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

Estimation of transient heat flux density during the heat supply of a catalytic wall steam methane reformer

NASA Astrophysics Data System (ADS)

Settar, Abdelhakim; Abboudi, Saïd; Madani, Brahim; Nebbali, Rachid

2018-02-01

Due to the endothermic nature of the steam methane reforming reaction, the process is often limited by the heat transfer behavior in the reactors. Poor thermal behavior sometimes leads to slow reaction kinetics, which is characterized by the presence of cold spots in the catalytic zones. Within this framework, the present work consists on a numerical investigation, in conjunction with an experimental one, on the one-dimensional heat transfer phenomenon during the heat supply of a catalytic-wall reactor, which is designed for hydrogen production. The studied reactor is inserted in an electric furnace where the heat requirement of the endothermic reaction is supplied by electric heating system. During the heat supply, an unknown heat flux density, received by the reactive flow, is estimated using inverse methods. In the basis of the catalytic-wall reactor model, an experimental setup is engineered in situ to measure the temperature distribution. Then after, the measurements are injected in the numerical heat flux estimation procedure, which is based on the Function Specification Method (FSM). The measured and estimated temperatures are confronted and the heat flux density which crosses the reactor wall is determined.

Flow optimization study of a batch microfluidics PET tracer synthesizing device

PubMed Central

Elizarov, Arkadij M.; Meinhart, Carl; van Dam, R. Michael; Huang, Jiang; Daridon, Antoine; Heath, James R.; Kolb, Hartmuth C.

2010-01-01

We present numerical modeling and experimental studies of flow optimization inside a batch microfluidic micro-reactor used for synthesis of human-scale doses of Positron Emission Tomography (PET) tracers. Novel techniques are used for mixing within, and eluting liquid out of, the coin-shaped reaction chamber. Numerical solutions of the general incompressible Navier Stokes equations along with time-dependent elution scalar field equation for the three dimensional coin-shaped geometry were obtained and validated using fluorescence imaging analysis techniques. Utilizing the approach presented in this work, we were able to identify optimized geometrical and operational conditions for the micro-reactor in the absence of radioactive material commonly used in PET related tracer production platforms as well as evaluate the designed and fabricated micro-reactor using numerical and experimental validations. PMID:21072595

Dynamic characteristics of a VK-50 reactor operating under conditions of the loss of a normal feedwater flow

NASA Astrophysics Data System (ADS)

Semidotskiy, I. I.; Kurskiy, A. S.

2013-12-01

The paper describes the conditions of the ATWS type with virtually complete cessation of the feed-water flow at the operating power level of a reactor of the VK-50 type. Under these conditions, the role of spatial kinetics in the system of feedback between thermohydraulic and nuclear processes with bulk boiling of the coolant in the reactor core is clearly seen. This feature determines the specific character of experimental data obtained and the suitability of their use for verification of the associated codes used for calculating water-water reactors.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sklenka, L.; Rataj, J.; Frybort, J.

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training programmore » are demonstrated. (authors)« less

Radiocarbon tracer measurements of atmospheric hydroxyl radical concentrations

NASA Technical Reports Server (NTRS)

Campbell, M. J.; Farmer, J. C.; Fitzner, C. A.; Henry, M. N.; Sheppard, J. C.

1986-01-01

The usefulness of the C-14 tracer in measurements of atmospheric hydroxyl radical concentration is discussed. The apparatus and the experimental conditions of three variations of a radiochemical method of atmosphere analysis are described and analyzed: the Teflon bag static reactor, the flow reactor (used in the Wallops Island tests), and the aircraft OH titration reactor. The procedure for reduction of the aircraft reactor instrument data is outlined. The problems connected with the measurement of hydroxyl radicals are discussed. It is suggested that the gas-phase radioisotope methods have considerable potential in measuring tropospheric impurities present in very low concentrations.

Influence of denitrification reactor retention time distribution (RTD) on dissolved oxygen control and nitrogen removal efficiency.

PubMed

Raboni, Massimo; Gavasci, Renato; Viotti, Paolo

2015-01-01

Low concentrations of dissolved oxygen (DO) are usually found in biological anoxic pre-denitrification reactors, causing a reduction in nitrogen removal efficiency. Therefore, the reduction of DO in such reactors is fundamental for achieving good nutrient removal. The article shows the results of an experimental study carried out to evaluate the effect of the anoxic reactor hydrodynamic model on both residual DO concentration and nitrogen removal efficiency. In particular, two hydrodynamic models were considered: the single completely mixed reactor and a series of four reactors that resemble plug-flow behaviour. The latter prove to be more effective in oxygen consumption, allowing a lower residual DO concentration than the former. The series of reactors also achieves better specific denitrification rates and higher denitrification efficiency. Moreover, the denitrification food to microrganism (F:M) ratio (F:MDEN) demonstrates a relevant synergic action in both controlling residual DO and improving the denitrification performance.

Preliminary Study of Realistic Blast Impact on Cultured Brain Slices

DTIC Science & Technology

2015-04-01

and/or multiple impacts in water. 3. Experimental Setup 3.1 The Aquarium Setup A 30.5-cm by 34.5- × 65-cm water-filled polymethylmethacrylate ...sodium bicarbonate PAGE polyacrylamide gel electrophoresis PMMA polymethylmethacrylate RDECOM U.S. Army Research Development and Engineering Command

Inverse modeling approach for evaluation of kinetic parameters of a biofilm reactor using tabu search.

PubMed

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.

KINETIC STUDY OF ADSORPTION AND TRANSFORMATION OF MERCURY ON FLY ASH PARTICLES IN AN ENTRAINED FLOW REACTOR

EPA Science Inventory

Experimental studies were performed to investigate the interactions of elemental mercury vapor with entrained fly ash particles from coal combustion in a flow reactor. The rate of transformation of elemental mercury on fly ash particles was evauated over the temperature range fro...

Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

NASA Technical Reports Server (NTRS)

Kieffer, A. W.

1972-01-01

Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

Universal Rate Model Selector: A Method to Quickly Find the Best-Fit Kinetic Rate Model for an Experimental Rate Profile

DTIC Science & Technology

2017-08-01

as an official Department of the Army position unless so designated by other authorizing documents. REPORT DOCUMENTATION PAGE Form Approved OMB...processes to find a kinetic rate model that provides a high degree of correlation with experimental data. Furthermore, the use of kinetic rate... correlation 16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF ABSTRACT 18. NUMBER OF PAGES 19a. NAME OF RESPONSIBLE PERSON Renu B

Experimental investigation of a new method for advanced fast reactor shutdown cooling

NASA Astrophysics Data System (ADS)

Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

2017-07-01

We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

Modeling phosphorus removal and recovery from anaerobic digester supernatant through struvite crystallization in a fluidized bed reactor.

PubMed

Rahaman, Md Saifur; Mavinic, Donald S; Meikleham, Alexandra; Ellis, Naoko

2014-03-15

The cost associated with the disposal of phosphate-rich sludge, the stringent regulations to limit phosphate discharge into aquatic environments, and resource shortages resulting from limited phosphorus rock reserves, have diverted attention to phosphorus recovery in the form of struvite (MAP: MgNH4PO4·6H2O) crystals, which can essentially be used as a slow release fertilizer. Fluidized-bed crystallization is one of the most efficient unit processes used in struvite crystallization from wastewater. In this study, a comprehensive mathematical model, incorporating solution thermodynamics, struvite precipitation kinetics and reactor hydrodynamics, was developed to illustrate phosphorus depletion through struvite crystal growth in a continuous, fluidized-bed crystallizer. A thermodynamic equilibrium model for struvite precipitation was linked to the fluidized-bed reactor model. While the equilibrium model provided information on supersaturation generation, the reactor model captured the dynamic behavior of the crystal growth processes, as well as the effect of the reactor hydrodynamics on the overall process performance. The model was then used for performance evaluation of the reactor, in terms of removal efficiencies of struvite constituent species (Mg, NH4 and PO4), and the average product crystal sizes. The model also determined the variation of species concentration of struvite within the crystal bed height. The species concentrations at two extreme ends (inlet and outlet) were used to evaluate the reactor performance. The model predictions provided a reasonably good fit with the experimental results for PO4-P, NH4-N and Mg removals. Predicated average crystal sizes also matched fairly well with the experimental observations. Therefore, this model can be used as a tool for performance evaluation and process optimization of struvite crystallization in a fluidized-bed reactor. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

NASA Astrophysics Data System (ADS)

Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

2016-04-01

Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first partmore » of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)« less

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Development of the cascade inertial-confinement-fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitts, J.H.

Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

Development of the cascade inertial-confinement-fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pitts, J.H.

Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

Surveillance application using patten recognition software at the EBR-II Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, D.L.

1992-01-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Surveillance application using patten recognition software at the EBR-II Reactor Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, D.L.

1992-05-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory`s Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

DE-NE0008277_PROTEUS final technical report 2018

DOE Office of Scientific and Technical Information (OSTI.GOV)

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

Comparison of Pictorial Techniques for Guiding Performance During Training.

ERIC Educational Resources Information Center

Miller, Elmo E.

An experimental program was conducted to develop effective methods for producing and utilizing filmed demonstrations and instructional manuals. Four variations on conventional filmed demonstrations were evaluated: 1) revising an Army film through repeated tryouts with novices, 2) stopping the projector after each step is demonstrated to allow…

New Concepts for the Development of Carbon Nanotube Materials for Army Related Applications

DTIC Science & Technology

2015-08-16

the microcavity exciton- polariton system, which started as a theoretical concept in the 1990s and has been a driving force for experimental physics of... polariton lasers, optical polarization switches, superfluid spintronic devices, etc. We, therefore, strongly believe that the quasi-1D exciton BEC effect

Peer Ratings as Predictors of Success in Military Aviation.

ERIC Educational Resources Information Center

Wahlberg, James L.; And Others

Three experimental peer rating forms were developed for use in research in prediction of the aviation training performance criterion--completion/attrition--from the training program for Aviation Warrant Officer Candidates at the U.S. Army Helicopter School. This paper describes the construction of the ratings, the "Potential Aviator…

Oxidative coupling of methane using inorganic membrane reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ma, Y.H.; Moser, W.R.; Dixon, A.G.

1995-12-31

The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gasmore » phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.« less

Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

NASA Astrophysics Data System (ADS)

Darmawan, R.

2018-01-01

Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

FOEHN: The critical experiment for the Franco-German High Flux Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Scharmer, K.; Eckert, H. G.

1991-01-01

A critical experiment for the Franco-German High Flux Reactor was carried out in the French reactor EOLE (CEN Cadarache). The purpose of the experiment was to check the calculation methods in a realistic geometry and to measure effects that can only be calculated imprecisely (e.g. beam hole effects). The structure of the experiment and the measurement and calculation methods are described. A detailed comparison between theoretical and experimental results was performed. 30 refs., 105 figs.

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

An atmospheric pressure high-temperature laminar flow reactor for investigation of combustion and related gas phase reaction systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oßwald, Patrick; Köhler, Markus

A new high-temperature flow reactor experiment utilizing the powerful molecular beam mass spectrometry (MBMS) technique for detailed observation of gas phase kinetics in reacting flows is presented. The reactor design provides a consequent extension of the experimental portfolio of validation experiments for combustion reaction kinetics. Temperatures up to 1800 K are applicable by three individually controlled temperature zones with this atmospheric pressure flow reactor. Detailed speciation data are obtained using the sensitive MBMS technique, providing in situ access to almost all chemical species involved in the combustion process, including highly reactive species such as radicals. Strategies for quantifying the experimentalmore » data are presented alongside a careful analysis of the characterization of the experimental boundary conditions to enable precise numeric reproduction of the experimental results. The general capabilities of this new analytical tool for the investigation of reacting flows are demonstrated for a selected range of conditions, fuels, and applications. A detailed dataset for the well-known gaseous fuels, methane and ethylene, is provided and used to verify the experimental approach. Furthermore, application for liquid fuels and fuel components important for technical combustors like gas turbines and engines is demonstrated. Besides the detailed investigation of novel fuels and fuel components, the wide range of operation conditions gives access to extended combustion topics, such as super rich conditions at high temperature important for gasification processes, or the peroxy chemistry governing the low temperature oxidation regime. These demonstrations are accompanied by a first kinetic modeling approach, examining the opportunities for model validation purposes.« less

Detecting Dark Photons with Reactor Neutrino Experiments

NASA Astrophysics Data System (ADS)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ <1.3 ×10-5 and ɛ <2.1 ×10-5, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

NASA Astrophysics Data System (ADS)

Koshelev, A. S.; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B.

2016-12-01

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

Evaluation of selected chemical processes for production of low-cost silicon, phase 2

NASA Technical Reports Server (NTRS)

Blocher, J. M., Jr.; Browning, M. F.; Wilson, W. J.; Carmichael, D. C.

1977-01-01

Potential designs for an integrated fluidized-bed reactor/zinc vaporizer/SiCl4 preheater unit are being considered and heat-transfer calculations have been initiated on versions of the zinc vaporizer section. Estimates of the cost of the silicon prepared in the experimental facility have been made for projected capacities of 25, 50, 75, and 100 metric ton of silicon. A 35 percent saving is obtained in going from 25 metric ton/year to the 50 metric ton/year level. This analysis, coupled with the recognition that use of two reactors in the 50 metric ton/year version allows for continued operation (at reduced capacity) with one reactor shut down, has resulted in a recommendation for adoption of an experimental facility capacity of 50 metric ton/year or greater. At this stage, the change to a larger size facility would not increase the design costs appreciably. In the experimental support program, the effects of seed bed particle size and depth were studied, and operation of the miniplant with a new zinc vaporizer was initiated, revealing the need for modification of the latter.

ATOMIC PHYSICS, AN AUTOINSTRUCTIONAL PROGRAM, VOLUME 4, SUPPLEMENT.

ERIC Educational Resources Information Center

DETERLINE, WILLIAM A.; KLAUS, DAVID J.

THE AUTOINSTRUCTIONAL MATERIALS IN THIS TEXT WERE PREPARED FOR USE IN AN EXPERIMENTAL STUDY, OFFERING SELF-TUTORING MATERIAL FOR LEARNING ATOMIC PHYSICS. THE TOPICS COVERED ARE (1) RADIATION USES AND NUCLEAR FISSION, (2) NUCLEAR REACTORS, (3) ENERGY FROM NUCLEAR REACTORS, (4) NUCLEAR EXPLOSIONS AND FUSION, (5) A COMPREHENSIVE REVIEW, AND (6) A…

Theoretical and Experimental Study of the Primary Current Distribution in Parallel-Plate Electrochemical Reactors

ERIC Educational Resources Information Center

Vazquez Aranda, Armando I.; Henquin, Eduardo R.; Torres, Israel Rodriguez; Bisang, Jose M.

2012-01-01

A laboratory experiment is described to determine the primary current distribution in parallel-plate electrochemical reactors. The electrolyte is simulated by conductive paper and the electrodes are segmented to measure the current distribution. Experiments are reported with the electrolyte confined to the interelectrode gap, where the current…

CIRFT Data Update and Data Analyses for Spent Nuclear Fuel Vibration Reliability Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Jy-An John; Wang, Hong

The objective of this research is to collect experimental data on spent nuclear fuel (SNF) from pressurized water reactors (PWRs), including the H. B. Robinson Nuclear Power Station (HBR), Catawba Nuclear Station, North Anna Nuclear Power Station (NA), and the Limerick Nuclear Power Station (LMK) boiling water reactor (BWR).

Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

ERIC Educational Resources Information Center

Malkawi, Salaheddin; Al-Araidah, Omar

2013-01-01

Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation.

PubMed

Jeong, K; Choo, Y S; Hong, H J; Yoon, Y S; Song, M H

2015-03-01

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, K.; Choo, Y. S.; Hong, H. J.

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 mlmore » and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.« less

Experimental Anomalies in Neutrino Physics

NASA Astrophysics Data System (ADS)

Palamara, Ornella

2014-03-01

In recent years, experimental anomalies ranging in significance (2.8-3.8 σ) have been reported from a variety of experiments studying neutrinos over baselines less than 1 km. Results from the LSND and MiniBooNE short-baseline νe /νe appearance experiments show anomalies which cannot be described by oscillations between the three standard model neutrinos (the ``LSND anomaly''). In addition, a re-analysis of the anti-neutrino flux produced by nuclear power reactors has led to an apparent deficit in νe event rates in a number of reactor experiments (the ``reactor anomaly''). Similarly, calibration runs using 51Cr and 37Ar radioactive sources in the Gallium solar neutrino experiments GALLEX and SAGE have shown an unexplained deficit in the electron neutrino event rate over very short distances (the ``Gallium anomaly''). The puzzling results from these experiments, which together may suggest the existence of physics beyond the Standard Model and hint at exciting new physics, including the possibility of additional low-mass sterile neutrino states, have raised the interest in the community for new experimental efforts that could eventually solve this puzzle. Definitive evidence for sterile neutrinos would be a revolutionary discovery, with implications for particle physics as well as cosmology. Proposals to address these signals by employing accelerator, reactor and radioactive source experiments are in the planning stages or underway worldwide. In this talk some of these will be reviewed, with emphasis on the accelerator programs.

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bucknor, Matthew; Hu, Rui; Lisowski, Darius

2016-04-17

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at themore » NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.« less

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perret, G.; Pattupara, R. M.; Girardin, G.

2012-07-01

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fastmore » Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)« less

Modeling a Packed Bed Reactor Utilizing the Sabatier Process

NASA Technical Reports Server (NTRS)

Shah, Malay G.; Meier, Anne J.; Hintze, Paul E.

2017-01-01

A numerical model is being developed using Python which characterizes the conversion and temperature profiles of a packed bed reactor (PBR) that utilizes the Sabatier process; the reaction produces methane and water from carbon dioxide and hydrogen. While the specific kinetics of the Sabatier reaction on the RuAl2O3 catalyst pellets are unknown, an empirical reaction rate equation1 is used for the overall reaction. As this reaction is highly exothermic, proper thermal control is of the utmost importance to ensure maximum conversion and to avoid reactor runaway. It is therefore necessary to determine what wall temperature profile will ensure safe and efficient operation of the reactor. This wall temperature will be maintained by active thermal controls on the outer surface of the reactor. Two cylindrical PBRs are currently being tested experimentally and will be used for validation of the Python model. They are similar in design except one of them is larger and incorporates a preheat loop by feeding the reactant gas through a pipe along the center of the catalyst bed. The further complexity of adding a preheat pipe to the model to mimic the larger reactor is yet to be implemented and validated; preliminary validation is done using the smaller PBR with no reactant preheating. When mapping experimental values of the wall temperature from the smaller PBR into the Python model, a good approximation of the total conversion and temperature profile has been achieved. A separate CFD model incorporates more complex three-dimensional effects by including the solid catalyst pellets within the domain. The goal is to improve the Python model to the point where the results of other reactor geometry can be reasonably predicted relatively quickly when compared to the much more computationally expensive CFD approach. Once a reactor size is narrowed down using the Python approach, CFD will be used to generate a more thorough prediction of the reactors performance.

2011 Nanoelectronic Devices for Defense & Security (NANO-DDS) Conference Held in Brooklyn, New York on August 29-September 1, 2011. Technical Program and Abstract Digest

DTIC Science & Technology

2011-08-01

challenges in new design methodologies . Particular examples involve an in-circuit functional timing testing of systems with millions of cores. I...TECHNIQUES Chair: Dwight Woolard, U.S. Army Research Office (ARO) 8:40-9:05 EXPERIMENTAL DESIGN OF SINGLE-CRYSTAL DNA FOR THZ SPECTROSCOPY...Detection Based Techniques EXPERIMENTAL DESIGN OF SINGLE-CRYSTAL DNA FOR THZ SPECTROSCOPY E. R. Brown, M.L. Norton, M. Rahman, W. Zhang Wright

Using a Second-Price Auction to set Military Retention Bonus Levels: An Application to the Australian Army

DTIC Science & Technology

2008-03-01

and do not reflect the official policy or position of the Department of Defense or the U.S. Government. 12a. DISTRIBUTION / AVAILABILITY STATEMENT...auction is used to set salaries levels in a generic labor market scenario. The experimental results support the literature: a second-price auction...analyses an experiment where a sealed-bid, second-price auction is used to set salaries levels in a generic labor market scenario. The experimental

Evaluation of Alternative Splicing Regulators as Targets for Selective Therapy of Triple Negative (Basal) Breast Carcinoma

DTIC Science & Technology

2017-10-01

inhibitor (LY294002) ? Figure 2. Signaling pathway and experimental design . Growth signal supporting tumor growth are transmitted through the PI3K...are those of the author(s) and should not be construed as an official Department of the Army position, policy or decision unless so designated by...are generating the experimental animals that combine the transgene with the knockout alleles. Under Task 4, we analyzed the expression of KHDRBS3 in

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

Nuclear reactors provide an excellent environment for studying neutrinos and continue to play a critical role in unveiling the secrets of these elusive particles. A rich experimental program with reactor antineutrinos is currently ongoing, and leads the way in precision measurements of several oscillation parameters and in searching for new physics, such as the existence of light sterile neutrinos. Ongoing experiments have also been able to measure the flux and spectral shape of reactor antineutrinos with unprecedented statistics and as a function of core fuel evolution, uncovering anomalies that will lead to new physics and/or to an improved understanding of antineutrino emission from nuclear reactors. The future looks bright, with an aggressive program of next generation reactor neutrino experiments that will go after some of the biggest open questions in the field. This includes the JUNO experiment, the largest liquid scintillator detector ever constructed which will push the limits of this detection technology.

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.« less

Artemisinins from folklore to modern medicine--transforming an herbal extract to life-saving drugs.

PubMed

Weina, P J

2008-06-01

The history of the artemisinins from Ge Hong in China during the 4th century, to the re-discovery of the qing hao derivatives in the 1970s, to the explosion of artemisinin derivatives and combinations throughout the world today is a fascinating story. The central and underappreciated role of the United States Army's 'drug company' known as the Division of Experimental Therapeutics at the Walter Reed Army Institute of Research is a story worth relating. From being the first group outside China to extract the active component of qing hao, to leading the work on neurotoxicity of the class in animals, to bringing a Good Manufacturing Practices intravenous formulation to the worldwide market is traced.

Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

2010-01-01

We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.« less

Generation and Reduction of NOx on Air-Fed Ozonizers

NASA Astrophysics Data System (ADS)

Ehara, Yoshiyasu; Amemiya, Yusuke; Yamamoto, Toshiaki

A generation and reduction of NOx on air-fed ozonizers using a ferroelectric packed bed reactor have been experimentally investigated. The reactors packed with CaTiO3, SrTiO3 and BaTiO3 pellets are examined for ozone generation. An ac voltage is applied to the reactor to generate partial discharge. Ozone concentration and the different nitrogen oxides at downstream of the packed bed reactor were measured with UV absorption ozone monitor and a Fourier transform infrared spectroscope respectively. The dielectric constant of packed ferroelectric pellets influences the discharge characteristic, ozone and NOx generations are varied by the dielectric constant value. Focusing on a discharge pulse current and maximum discharge magnitude, the ferroelectric packed bed plasma reactors have been evaluated on nitrogen oxide and ozone generated concentrations.

Microprocessor tester for the treat upgrade reactor trip system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

NASA Astrophysics Data System (ADS)

Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

2017-09-01

The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

Flight-testing and frequency-domain analysis for rotorcraft handling qualities

NASA Technical Reports Server (NTRS)

Ham, Johnnie A.; Gardner, Charles K.; Tischler, Mark B.

1995-01-01

A demonstration of frequency-domain flight-testing techniques and analysis was performed on a U.S. Army OH-58D helicopter in support of the OH-58D Airworthiness and Flight Characteristics Evaluation and of the Army's development and ongoing review of Aeronautical Design Standard 33C, Handling Qualities Requirements for Military Rotorcraft. Hover and forward flight (60 kn) tests were conducted in 1 flight hour by Army experimental test pilots. Further processing of the hover data generated a complete database of velocity, angular-rate, and acceleration-frequency responses to control inputs. A joint effort was then undertaken by the Airworthiness Qualification Test Dirtectorate and the U.S. Army Aeroflightdynamics Directorate to derive handling-quality information from the frequency-domain database using a variety of approaches. This report documents numerous results that have been obtained from the simple frequency-domain tests; in many areas, these results provide more insight into the aircraft dynmamics that affect handling qualities than do traditional flight tests. The handling-quality results include ADS-33C bandwidth and phase-delay calculations, vibration spectral determinations, transfer-function models to examine single-axis results, and a six-degree-of-freedom fully coupled state-space model. The ability of this model to accurately predict responses was verified using data from pulse inputs. This report also documents the frequency-sweep flight-test technique and data analysis used to support the tests.

Human perception testing methodology for evaluating EO/IR imaging systems

NASA Astrophysics Data System (ADS)

Graybeal, John J.; Monfort, Samuel S.; Du Bosq, Todd W.; Familoni, Babajide O.

2018-04-01

The U.S. Army's RDECOM CERDEC Night Vision and Electronic Sensors Directorate (NVESD) Perception Lab is tasked with supporting the development of sensor systems for the U.S. Army by evaluating human performance of emerging technologies. Typical research questions involve detection, recognition and identification as a function of range, blur, noise, spectral band, image processing techniques, image characteristics, and human factors. NVESD's Perception Lab provides an essential bridge between the physics of the imaging systems and the performance of the human operator. In addition to quantifying sensor performance, perception test results can also be used to generate models of human performance and to drive future sensor requirements. The Perception Lab seeks to develop and employ scientifically valid and efficient perception testing procedures within the practical constraints of Army research, including rapid development timelines for critical technologies, unique guidelines for ethical testing of Army personnel, and limited resources. The purpose of this paper is to describe NVESD Perception Lab capabilities, recent methodological improvements designed to align our methodology more closely with scientific best practice, and to discuss goals for future improvements and expanded capabilities. Specifically, we discuss modifying our methodology to improve training, to account for human fatigue, to improve assessments of human performance, and to increase experimental design consultation provided by research psychologists. Ultimately, this paper outlines a template for assessing human perception and overall system performance related to EO/IR imaging systems.

Antineutrino analysis for continuous monitoring of nuclear reactors: Sensitivity study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart, Christopher; Erickson, Anna

This paper explores the various contributors to uncertainty on predictions of the antineutrino source term which is used for reactor antineutrino experiments and is proposed as a safeguard mechanism for future reactor installations. The errors introduced during simulation of the reactor burnup cycle from variation in nuclear reaction cross sections, operating power, and other factors are combined with those from experimental and predicted antineutrino yields, resulting from fissions, evaluated, and compared. The most significant contributor to uncertainty on the reactor antineutrino source term when the reactor was modeled in 3D fidelity with assembly-level heterogeneity was found to be the uncertaintymore » on the antineutrino yields. Using the reactor simulation uncertainty data, the dedicated observation of a rigorously modeled small, fast reactor by a few-ton near-field detector was estimated to offer reduction of uncertainty on antineutrino yields in the 3.0–6.5 MeV range to a few percent for the primary power-producing fuel isotopes, even with zero prior knowledge of the yields.« less

Post Cultural Revolution Teaching Methods. Occasional Paper No. 77-7.

ERIC Educational Resources Information Center

Silvestri, Gary

This paper highlights the developments in teaching methods in China in the era of the post-Cultural Revolution, and explains how these methods grew out of educational experimentation during the Yenan period. The Yenan period followed the historic "long march, "when Mao Tse-tung and the People's Liberation Army (PLA) established their…

Metabolic Heal and Energy Expenditure Estimates of Border Patrol Personnel Obtained using the Factorial Method

DTIC Science & Technology

2010-08-01

of Environmental Medicine Building 42 - Kansas Street Natick, MA 01760 T10-03 U.S. Army Medical Research and Materiel Command Fort Detrick, MD 21702...training these individuals reported averaging 6.4 + 1.7 miles of running per workout , min = 3 miles, max = 8 miles. EXPERIMENTAL PROCEDURE

Dermal Sensitization Potential of DIGL-RP Solid Propellant in Guinea Pigs

DTIC Science & Technology

1989-10-01

y ’,c. ADM$$S (ft, SWOt , &Wd ZIP Cod 7b. ADDRESS (City, State, arid ZIP Code) Letterman Army Institute of Research Fort Detrick Presidio of San...for contact sensitization. Toxicol Appl Pharmacol 1969; Suppl 3:90-102. 7. Buehler EV, Griffith JF. Experimental skin sensitization in the guinea pig

Background and Situational Confidence; Their Relation to Performance Effectiveness. Professional Paper No. 22-68.

ERIC Educational Resources Information Center

Boyles, Wiley R.

Inventories designed to measure confidence in dangerous situations were administered to about 3,000 potential Army aviation warrant officers from January to December 1967. These paper-and-pencil inventories are based on a clinical-experimental fractional anticipatory response conceptualization of reactions to the psychological stresses of combat.…

CHANGES IN FLIGHT TRAINEE PERFORMANCE FOLLOWING SYNTHETIC HELICOPTER FLIGHT TRAINING.

ERIC Educational Resources Information Center

CARO, PAUL W., JR.; ISLEY, ROBERT N.

A STUDY WAS CONDUCTED AT THE U.S. ARMY PRIMARY HELICOPTER SCHOOL, FORT WOLTERS, TEXAS, TO DETERMINE WHETHER THE USE OF A HELICOPTER TRAINING DEVICE WOULD IMPROVE STUDENT PERFORMANCE DURING SUBSEQUENT HELICOPTER CONTACT FLIGHT TRAINING. SUBJECTS WERE TWO EXPERIMENTAL GROUPS AND TWO CONTROL GROUPS OF WARRANT OFFICER CANDIDATES ENROLLED FOR A…

Stabilized three-stage oxidation of DME/air mixture in a micro flow reactor with a controlled temperature profile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oshibe, Hiroshi; Nakamura, Hisashi; Tezuka, Takuya

Ignition and combustion characteristics of a stoichiometric dimethyl ether (DME)/air mixture in a micro flow reactor with a controlled temperature profile which was smoothly ramped from room temperature to ignition temperature were investigated. Special attention was paid to the multi-stage oxidation in low temperature condition. Normal stable flames in a mixture flow in the high velocity region, and non-stationary pulsating flames and/or repetitive extinction and ignition (FREI) in the medium velocity region were experimentally confirmed as expected from our previous study on a methane/air mixture. In addition, stable double weak flames were observed in the low velocity region for themore » present DME/air mixture case. It is the first observation of stable double flames by the present methodology. Gas sampling was conducted to obtain major species distributions in the flow reactor. The results indicated that existence of low-temperature oxidation was conjectured by the production of CH{sub 2}O occured in the upstream side of the experimental first luminous flame, while no chemiluminescence from it was seen. One-dimensional computation with detailed chemistry and transport was conducted. At low mixture velocities, three-stage oxidation was confirmed from profiles of the heat release rate and major chemical species, which was broadly in agreement with the experimental results. Since the present micro flow reactor with a controlled temperature profile successfully presented the multi-stage oxidations as spatially separated flames, it is shown that this flow reactor can be utilized as a methodology to separate sets of reactions, even for other practical fuels, at different temperature. (author)« less

Fischer-Tropsch Slurry Reactor modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soong, Y.; Gamwo, I.K.; Harke, F.W.

1995-12-31

This paper reports experimental and theoretical results on hydrodynamic studies. The experiments were conducted in a hot-pressurized Slurry-Bubble Column Reactor (SBCR). It includes experimental results of Drakeol-10 oil/nitrogen/glass beads hydrodynamic study and the development of an ultrasonic technique for measuring solids concentration. A model to describe the flow behavior in reactors was developed. The hydrodynamic properties in a 10.16 cm diameter bubble column with a perforated-plate gas distributor were studied at pressures ranging from 0.1 to 1.36 MPa, and at temperatures from 20 to 200{degrees}C, using a dual hot-wire probe with nitrogen, glass beads, and Drakeol-10 oil as the gas,more » solid, and liquid phase, respectively. It was found that the addition of 20 oil wt% glass beads in the system has a slight effect on the average gas holdup and bubble size. A well-posed three-dimensional model for bed dynamics was developed from an ill-posed model. The new model has computed solid holdup distributions consistent with experimental observations with no artificial {open_quotes}fountain{close_quotes} as predicted by the earlier model. The model can be applied to a variety of multiphase flows of practical interest. An ultrasonic technique is being developed to measure solids concentration in a three-phase slurry reactor. Preliminary measurements have been made on slurries consisting of molten paraffin wax, glass beads, and nitrogen bubbles at 180 {degrees}C and 0.1 MPa. The data show that both the sound speed and attenuation are well-defined functions of both the solid and gas concentrations in the slurries. The results suggest possibilities to directly measure solids concentration during the operation of an autoclave reactor containing molten wax.« less

Unit mechanisms of fission gas release: Current understanding and future needs

DOE PAGES

Tonks, Michael; Andersson, David; Devanathan, Ram; ...

2018-03-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tonks, Michael; Andersson, David; Devanathan, Ram

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. Here, this basic understanding of the fission gas behavior mechanisms has the potentialmore » to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.« less

Unit mechanisms of fission gas release: Current understanding and future needs

NASA Astrophysics Data System (ADS)

Tonks, Michael; Andersson, David; Devanathan, Ram; Dubourg, Roland; El-Azab, Anter; Freyss, Michel; Iglesias, Fernando; Kulacsy, Katalin; Pastore, Giovanni; Phillpot, Simon R.; Welland, Michael

2018-06-01

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

In-situ material-motion diagnostics and fuel radiography in experimental reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

DeVolpi, A.

1982-01-01

Material-motion monitoring has become a routine part of in-pile transient reactor-safety experiments. Diagnostic systems, such as the fast-neutron hodoscope, were developed for the purpose of providing direct time-resolved data on pre-failure fuel motion, cladding-breach time and location, and post-failure fuel relocation. Hodoscopes for this purpose have been installed at TREAT and CABRI; other types of imaging systems that have been tested are a coded-aperture at ACRR and a pinhole at TREAT. Diagnostic systems that use penetrating radiation emitted from the test section can non-invasively monitor fuel without damage to the measuring instrument during the radiographic images of test sections installedmore » in the reator. Studies have been made of applications of hodoscopes to other experimental reactors, including PBF, FARET, STF, ETR, EBR-II, SAREF-STF, and DMT.« less

Experimental power density distribution benchmark in the TRIGA Mark II reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Snoj, L.; Stancar, Z.; Radulovic, V.

2012-07-01

In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the fewmore » available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)« less

Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fourmentel, D.; Radulovic, V.; Barbot, L.

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed. (authors)« less

An experimental aluminum-fueled power plant

NASA Astrophysics Data System (ADS)

Vlaskin, M. S.; Shkolnikov, E. I.; Bersh, A. V.; Zhuk, A. Z.; Lisicyn, A. V.; Sorokovikov, A. I.; Pankina, Yu. V.

2011-10-01

An experimental co-generation power plant (CGPP-10) using aluminum micron powder (with average particle size up to 70 μm) as primary fuel and water as primary oxidant was developed and tested. Power plant can work in autonomous (unconnected from industrial network) nonstop regime producing hydrogen, electrical energy and heat. One of the key components of experimental plant is aluminum-water high-pressure reactor projected for hydrogen production rate of ∼10 nm3 h-1. Hydrogen from the reactor goes through condenser and dehumidifier and with -25 °C dew-point temperature enters into the air-hydrogen fuel cell 16 kW-battery. From 1 kg of aluminum the experimental plant produces 1 kWh of electrical energy and 5-7 kWh of heat. Power consumer gets about 10 kW of electrical power. Plant electrical and total efficiencies are 12% and 72%, respectively.

HORIZONTAL BEAM HOLE NO. 3. PLUG AND RADIATION DOOR HAVE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

HORIZONTAL BEAM HOLE NO. 3. PLUG AND RADIATION DOOR HAVE BEEN REMOVED. EXPERIMENTAL APPARATUS WAS INSERTED INTO THE HOLE. NOTE VALVE CUBICLES NEAR FLOOR ON EACH SIDE OF HB-3. INL NEGATIVE NO. 3471. Unknown Photographer, 10/12/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 2 For the Period May 21, 1958 to August 21, 1958

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loewe, W.E.; Krucoff, D.

1958-10-31

Study of the ADFR concept included experimental work on fuel dust suspension stability and redispersibility, erosion, and dust deposition using the fuel dust circulation loop. Some theoretical work was done in the areas of reactor safety and breeding. (For preceding period -see AECU-3827.) (auth)

The effect of mixing on fermentation of primary solids, glycerol, and biodiesel waste.

PubMed

Ghasemi, Marzieh; Randall, Andrew A

2018-03-01

In this study, the effect of mixing on volatile fatty acid (VFA) production and composition was investigated through running five identical bench-scale reactors that were filled with primary solid and dosed with either pure glycerol or biodiesel waste. Experimental results revealed that there was an inverse correlation between the mixing intensity and the VFA production. The total VFA production in the un-mixed reactor was 9,787 ± 3,601 mg COD/L, whereas in the reactor mixed at 100 rpm this dropped to 3,927 ± 1,175 mg COD/L, while both types of reactor were dosed with pure glycerol at the beginning of each cycle to reach the initial concentration of 1,000 mg/L (1,217 mg COD/L). Propionic acid was the dominant VFA in all the reactors except the reactor mixed at 30 rpm. It is hypothesized that low mixing facilitated hydrogen transfer between obligate hydrogen producing acetogens (OHPA) and hydrogen consuming acidogens in these non-methanogenic reactors. Also, in a narrower range of mixing (0 or 7 rpm), the total VFA production in biodiesel waste-fed reactors was considerably higher than that of pure glycerol-fed reactors.

Pyrolysis of furan in a microreactor

NASA Astrophysics Data System (ADS)

Urness, Kimberly N.; Guan, Qi; Golan, Amir; Daily, John W.; Nimlos, Mark R.; Stanton, John F.; Ahmed, Musahid; Ellison, G. Barney

2013-09-01

A silicon carbide microtubular reactor has been used to measure branching ratios in the thermal decomposition of furan, C4H4O. The pyrolysis experiments are carried out by passing a dilute mixture of furan (approximately 0.01%) entrained in a stream of helium through the heated reactor. The SiC reactor (0.66 mm i.d., 2 mm o.d., 2.5 cm long) operates with continuous flow. Experiments were performed with a reactor inlet pressure of 100-300 Torr and a wall temperature between 1200 and 1600 K; characteristic residence times in the reactor are 60-150 μs. The unimolecular decomposition pathway of furan is confirmed to be: furan (+ M) rightleftharpoons α-carbene or β-carbene. The α-carbene fragments to CH2=C=O + HC≡CH while the β-carbene isomerizes to CH2=C=CHCHO. The formyl allene can isomerize to CO + CH3C≡CH or it can fragment to H + CO + HCCCH2. Tunable synchrotron radiation photoionization mass spectrometry is used to monitor the products and to measure the branching ratio of the two carbenes as well as the ratio of [HCCCH2]/[CH3C≡CH]. The results of these pyrolysis experiments demonstrate a preference for 80%-90% of furan decomposition to occur via the β-carbene. For reactor temperatures of 1200-1400 K, no propargyl radicals are formed. As the temperature rises to 1500-1600 K, at most 10% of the decomposition of CH2=C=CHCHO produces H + CO + HCCCH2 radicals. Thermodynamic conditions in the reactor have been modeled by computational fluid dynamics and the experimental results are compared to the predictions of three furan pyrolysis mechanisms. Uncertainty in the pressure-dependency of the initiation reaction rates is a possible a source of discrepancy between experimental results and theoretical predictions.

Behaviour of a full-scale expanded bed reactor with overlaid anaerobic and aerobic zones for domestic wastewater treatment.

PubMed

Mendonça, N M; Siman, R R; Niciura, C L; Campos, J R

2006-01-01

This paper presents the behaviour of a full-scale expanded bed reactor (160 m3) with overlaid anaerobic and aerobic zones used for municipal wastewater treatment. The research was carried out in two experimental steps: anaerobic and anaerobic-aerobic conditions, and the experimental results presented in this paper refer to four months of reactor operation. In the anaerobic condition, after inoculation and 60 days of operation, the reactor treating 3.40 kg CODm(-3)d(-1) for thetaH of 2.69 h, reached mean removal efficiencies of 76% for BOD, 72% for COD, and 80% for TSS, when the effluent presented mean values of 225 mg.L(-1) of COD, 98 mg.L(-1) of BOD and 35 mg.L(-1) of TSS. Under these conditions, for nitrogen loading of 0.27 kgN.m(-3)d(-1), the reactor generated an effluent with mean N-org. of 8 mg.L(-1) and N-ammon. of 37 mg.L(-1), demonstrating high potential of ammonification. For the anaerobic-aerobic condition (118th day) the system was operated with thetaH of 5.38 h presented mean removal efficiencies of 84% for BOD, 79% for COD, 76% for TSS, and 30% for TKN. The reactor's operation time was less than two months, which was not long enough to reach nitrification. Regarding the obtained results, this research confirmed that this reactor is configured as a flexible and adequate alternative for the treatment of sewage, requiring relatively small area and only thetaH of 10 h that can be adjusted to the local circumstances.

Detecting Dark Photons with Reactor Neutrino Experiments.

PubMed

Park, H K

2017-08-25

We propose to search for light U(1) dark photons, A^{'}, produced via kinetically mixing with ordinary photons via the Compton-like process, γe^{-}→A^{'}e^{-}, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ε, the A^{'}-γ mixing parameter, ε, for dark-photon masses below 1 MeV of ε<1.3×10^{-5} and ε<2.1×10^{-5}, from NEOS and TEXONO experimental data, respectively. This study demonstrates the applicability of nuclear reactors as potential sources of intense fluxes of low-mass dark photons.

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Hydrogen Generation by Koh-Ethanol Plasma Electrolysis Using Double Compartement Reactor

NASA Astrophysics Data System (ADS)

Saksono, Nelson; Sasiang, Johannes; Dewi Rosalina, Chandra; Budikania, Trisutanti

2018-03-01

This study has successfully investigated the generation of hydrogen using double compartment reactor with plasma electrolysis process. Double compartment reactor is designed to achieve high discharged voltage, high concentration, and also reduce the energy consumption. The experimental results showed the use of double compartment reactor increased the productivity ratio 90 times higher compared to Faraday electrolysis process. The highest hydrogen production obtained is 26.50 mmol/min while the energy consumption can reach up 1.71 kJ/mmol H2 at 0.01 M KOH solution. It was shown that KOH concentration, addition of ethanol, cathode depth, and temperature have important effects on hydrogen production, energy consumption, and process efficiency.

Reducing acquisition risk through integrated systems of systems engineering

NASA Astrophysics Data System (ADS)

Gross, Andrew; Hobson, Brian; Bouwens, Christina

2016-05-01

In the fall of 2015, the Joint Staff J7 (JS J7) sponsored the Bold Quest (BQ) 15.2 event and conducted planning and coordination to combine this event into a joint event with the Army Warfighting Assessment (AWA) 16.1 sponsored by the U.S. Army. This multipurpose event combined a Joint/Coalition exercise (JS J7) with components of testing, training, and experimentation required by the Army. In support of Assistant Secretary of the Army for Acquisition, Logistics, and Technology (ASA(ALT)) System of Systems Engineering and Integration (SoSE&I), Always On-On Demand (AO-OD) used a system of systems (SoS) engineering approach to develop a live, virtual, constructive distributed environment (LVC-DE) to support risk mitigation utilizing this complex and challenging exercise environment for a system preparing to enter limited user test (LUT). AO-OD executed a requirements-based SoS engineering process starting with user needs and objectives from Army Integrated Air and Missile Defense (AIAMD), Patriot units, Coalition Intelligence, Surveillance and Reconnaissance (CISR), Focused End State 4 (FES4) Mission Command (MC) Interoperability with Unified Action Partners (UAP), and Mission Partner Environment (MPE) Integration and Training, Tactics and Procedures (TTP) assessment. The SoS engineering process decomposed the common operational, analytical, and technical requirements, while utilizing the Institute of Electrical and Electronics Engineers (IEEE) Distributed Simulation Engineering and Execution Process (DSEEP) to provide structured accountability for the integration and execution of the AO-OD LVC-DE. As a result of this process implementation, AO-OD successfully planned for, prepared, and executed a distributed simulation support environment that responsively satisfied user needs and objectives, demonstrating the viability of an LVC-DE environment to support multiple user objectives and support risk mitigation activities for systems in the acquisition process.

Validation of light water reactor calculation methods and JEF-1-based data libraries by TRX and BAPL critical experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paratte, J.M.; Pelloni, S.; Grimm, P.

1991-04-01

This paper analyzes the capability of various code systems and JEF-1-based nuclear data libraries to compute light water reactor lattices by comparing calculations with results from thermal reactor benchmark experiments TRX and BAPL and with previously published values. With the JEF-1 evaluation, eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and all methods give reasonable results for the measured reaction rate ratios within, or not too far from, the experimental uncertainty.

Modeling of Gallium Nitride Hydride Vapor Phase Epitaxy

NASA Technical Reports Server (NTRS)

Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

1997-01-01

A reactor model for the hydride vapor phase epitaxy of GaN is presented. The governing flow, energy, and species conservation equations are solved in two dimensions to examine the growth characteristics as a function of process variables and reactor geometry. The growth rate varies with GaCl composition but independent of NH3 and H2 flow rates. A change in carrier gas for Ga source from H2 to N2 affects the growth rate and uniformity for a fixed reactor configuration. The model predictions are in general agreement with observed experimental behavior.

Comprehensive Experiments on Subcritical Assemblies of Cascade Reactor Systems

NASA Astrophysics Data System (ADS)

Zavyalov, N. V.; Il'kaev, R. I.; Kolesov, V. F.; Ivanin, I. A.; Zhitnik, A. K.; Kuvshinov, M. I.; Nefedov, Yu. Ya.; Punin, V. T.; Tel'nov, A. V.; Khoruzhi, V. Kh.

2017-12-01

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003-2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 1 for the Period February 21, 1958 to May 21, 1958

DOE Office of Scientific and Technical Information (OSTI.GOV)

Loewe, W.E.; Krucoff, D.

1958-10-31

Work has begun on the ADFR, a reactor using a new fuel form -- fissionable dust carried in an inent gas. Temperatures in the range 2,000 to 3,000 deg F appear feasible in an all-ceramic system. Experimental study of the fuel form was initiated, and a loop to circulate the fuel dust was constructed. Initial operation is encouraging. Theoretical studies were carried on in the areas of reactor physics, heat transfer, and safety. (auth)

The effect of water presence on the photocatalytic oxidation of benzene, toluene, ethylbenzene and m-xylene in the gas-phase

NASA Astrophysics Data System (ADS)

Korologos, Christos A.; Philippopoulos, Constantine J.; Poulopoulos, Stavros G.

2011-12-01

In the present work, the gas-solid heterogeneous photocatalytic oxidation of benzene, toluene, ethylbenzene and m-xylene (BTEX) over UV-irradiated titanium dioxide was studied in an annular reactor operated in the CSTR (continuous stirred-tank reactor) mode. GC-FID and GC-MS were used for analysing reactor inlet and outlet streams. Initial BTEX concentrations were in the low parts per million (ppmv) range, whereas the water concentration was in the range of 0-35,230 ppmv and the residence time varied from 50 to 210 s. The effect of water addition on the photocatalytic process showed strong dependence on the type of the BTEX and the water vapour concentration. The increase in residence time resulted in a considerable increase in the conversion achieved for all compounds and experimental conditions. There was a clear interaction between residence time and water presence regarding the effect on conversions achieved. It was established that conversions over 95% could be achieved by adjusting appropriately the experimental conditions and especially the water concentration in the reactor. In all cases, no by-products were detected above the detection limit and carbon dioxide was the only compound detected. Finally, various Langmuir-Hinshelwood kinetic models have been tested in the analysis of the experimental data obtained. The kinetic data obtained confirmed that water had an active participation in the photocatalytic reactions of benzene, toluene, ethylbenzene and m-xylene since the model involving reaction of BTEX and water adsorbed on different active sites yielded the most successful fitting to the experimental results for the first three compounds, whereas the kinetic model based on the assumption that reaction between VOC and water dissociatively adsorbed on the photocatalyst takes place was the most appropriate in the case of m-xylene.

Experimental and Computational Study of Multiphase Flow Hydrodynamics in 2D Trickle Bed Reactors

NASA Astrophysics Data System (ADS)

Nadeem, H.; Ben Salem, I.; Kurnia, J. C.; Rabbani, S.; Shamim, T.; Sassi, M.

2014-12-01

Trickle bed reactors are largely used in the refining processes. Co-current heavy oil and hydrogen gas flow downward on catalytic particle bed. Fine particles in the heavy oil and/or soot formed by the exothermic catalytic reactions deposit on the bed and clog the flow channels. This work is funded by the refining company of Abu Dhabi and aims at mitigating pressure buildup due to fine deposition in the TBR. In this work, we focus on meso-scale experimental and computational investigations of the interplay between flow regimes and the various parameters that affect them. A 2D experimental apparatus has been built to investigate the flow regimes with an average pore diameter close to the values encountered in trickle beds. A parametric study is done for the development of flow regimes and the transition between them when the geometry and arrangement of the particles within the porous medium are varied. Liquid and gas flow velocities have also been varied to capture the different flow regimes. Real time images of the multiphase flow are captured using a high speed camera, which were then used to characterize the transition between the different flow regimes. A diffused light source was used behind the 2D Trickle Bed Reactor to enhance visualizations. Experimental data shows very good agreement with the published literature. The computational study focuses on the hydrodynamics of multiphase flow and to identify the flow regime developed inside TBRs using the ANSYS Fluent Software package. Multiphase flow inside TBRs is investigated using the "discrete particle" approach together with Volume of Fluid (VoF) multiphase flow modeling. The effect of the bed particle diameter, spacing, and arrangement are presented that may be used to provide guidelines for designing trickle bed reactors.

Kinetic Parameter Measurements in the MINERVE Reactor

NASA Astrophysics Data System (ADS)

Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas

2017-01-01

In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

Atomistic modeling of high temperature uranium-zirconium alloy structure and thermodynamics

NASA Astrophysics Data System (ADS)

Moore, A. P.; Beeler, B.; Deo, C.; Baskes, M. I.; Okuniewski, M. A.

2015-12-01

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium-zirconium alloy (γ-U-Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U-Zr alloys. Uranium-rich U-Zr alloys (e.g. U-10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U-Pu-Zr and U-Zr-Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U-Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U-Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U-Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

PubMed

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Evaluation of selected chemical processes for production of low-cost silicon, phase 3

NASA Technical Reports Server (NTRS)

Blocher, J. M., Jr.; Browning, M. F.; Seifert, D. A.

1981-01-01

A Process Development Unit (PDU), which consisted of the four major units of the process, was designed, installed, and experimentally operated. The PDU was sized to 50MT/Yr. The deposition took place in a fluidized bed reactor. As a consequences of the experiments, improvements in the design an operation of these units were undertaken and their experimental limitations were partially established. A parallel program of experimental work demonstrated that Zinc can be vaporized for introduction into the fluidized bed reactor, by direct induction-coupled r.f. energy. Residual zinc in the product can be removed by heat treatment below the melting point of silicon. Current efficiencies of 94 percent and above, and power efficiencies around 40 percent are achievable in the laboratory-scale electrolysis of ZnCl2.

A novel digital neutron flux monitor for international thermonuclear experimental reactor

NASA Astrophysics Data System (ADS)

Xiang, ZHOU; Zihao, LIU; Chao, CHEN; Renjie, ZHU; Li, ZHAO; Lingfeng, WEI; Zejie, YIN

2018-04-01

A novel full-digital real-time neutron flux monitor (NFM) has been developed for the International Thermonuclear Experimental Reactor. A measurement range of 109 counts per second is achieved with 3 different sensitive fission chambers. The Counting mode and Campbelling mode have been combined as a means to achieve higher measurement range. The system is based on high speed as well as parallel and pipeline processing of the field programmable gate array and has the ability to upload raw-data of analog-to-digital converter in real-time through the PXIe platform. With the advantages of the measurement range, real time performance and the ability of raw-data uploading, the digital NFM has been tested in HL-2A experiments and reflected good experimental performance.

Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

NASA Astrophysics Data System (ADS)

Damle, Ashok S.

One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and experimental results indicating that the underlying assumption of the simple model of conversion of hydrocarbons to CO and H 2 followed by equilibrium reconstitution to methane appears to be reasonable one.

PROSPECT - A precision oscillation and spectrum experiment

NASA Astrophysics Data System (ADS)

Langford, T. J.; PROSPECT Collaboration

2015-08-01

Segmented antineutrino detectors placed near a compact research reactor provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. Close proximity to a reactor combined with minimal overburden yield a high background environment that must be managed through shielding and detector technology. PROSPECT is a new experimental effort to detect reactor antineutrinos from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, managed by UT Battelle for the U.S. Department of Energy. The detector will use novel lithium-loaded liquid scintillator capable of neutron/gamma pulse shape discrimination and neutron capture tagging. These enhancements improve the ability to identify neutrino inverse-beta decays (IBD) and reject background events in analysis. Results from these efforts will be covered along with their implications for an oscillation search and a precision spectrum measurement.

Effect of small-scale biomass gasification at the state of refractory lining the fixed bed reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Janša, Jan, E-mail: jan.jansa@vsb.cz; Peer, Vaclav, E-mail: vaclav.peer@vsb.cz; Pavloková, Petra, E-mail: petra.pavlokova@vsb.cz

The article deals with the influence of biomass gasification on the condition of the refractory lining of a fixed bed reactor. The refractory lining of the gasifier is one part of the device, which significantly affects the operational reliability and durability. After removing the refractory lining of the gasifier from the experimental reactor, there was done an assessment how gasification of different kinds of biomass reflected on its condition in terms of the main factors affecting its life. Gasification of biomass is reflected on the lining, especially through sticking at the bottom of the reactor. Measures for prolonging the lifemore » of lining consist in the reduction of temperature in the reactor, in this case, in order to avoid ash fusion biomass which it is difficult for this type of gasifier.« less

Current and prospective safety issues at the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tichler, P.R.

The Brookhaven high-flux beam reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated, and reflected by heavy water and uses materials test reactor and engineering test reactor type of fuel elements containing enriched uranium. The reactor power when operation began in 1965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time, safety questions have been raised that resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper discusses the principalmore » safety issues and plans for their resolution and return to 60-MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its effect on the life of the facility are briefly discussed.« less

Analysis of the Browns Ferry Unit 3 irradiation experiments. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Simmons, G.L.

1984-11-01

The results of the analysis of two experiments performed at the Browns Ferry-3 reactor are presented. These calculations utilize state-of-the-art neutron transport techniques and a new neutron cross-section library that has been developed for LWR applications. The calculations agree well with the experimental data obtained in irradiations inside the reactor vessel. For the measurements performed in the reactor cavity, the calculations agree well at the reactor midplane. Accurate determination of the axial distribution of the neutron fluence in the reactor cavity depends on having a concise representation of the axial-void distribution in the core. Detailed data are presented describing themore » procedures used in the generation of the new cross-section library that has been named SAILOR. This library is available from the Radiation-Shielding Information Center.« less

Novel electrode structure in a DBD reactor applied to the degradation of phenol in aqueous solution

NASA Astrophysics Data System (ADS)

Mercado-Cabrera, Antonio; Peña-Eguiluz, Rosendo; López-Callejas, Régulo; Jaramillo-Sierra, Bethsabet; Valencia-Alvarado, Raúl; Rodríguez-Méndez, Benjamín; Muñoz-Castro, Arturo E.

2017-07-01

Phenol degradation experimental results are presented in a similar wastewater aqueous solution using a non-thermal plasma reactor in a coaxial dielectric barrier discharge. The novelty of the work is that one of the electrodes of the reactor has the shape of a hollow screw which shows an enhanced efficiency compared with a traditional smooth structure. The experimentation was carried out with gas mixtures of 90% Ar-10% O2, 80% Ar-20% O2 and 0% Ar-100% O2. After one hour of treatment the removal efficiency was 76%, 92%, and 97%, respectively, assessed with a gas chromatographic mass spectrometry technique. For both reactors used, the ozone concentration was measured. The screw electrode required less energy, for all gas mixtures, than the smooth electrode, to maintain the same ozone concentration. On the other hand, it was also observed that in both electrodes the electrical conductivity of the solution changed slightly from ˜0.0115 S m-1 up to ˜0.0430 S m-1 after one hour of treatment. The advantages of using the hollow screw electrode structure compared with the smooth electrode were: (1) lower typical power consumption, (2) the generation of a uniform plasma throughout the reactor benefiting the phenol degradation, (3) a relatively lower temperature of the aqueous solution during the process, and (4) the plasma generation length is larger.

Experimental investigation and CFD analysis on cross flow in the core of PMR200

DOE PAGES

Lee, Jeong -Hun; Yoon, Su -Jong; Cho, Hyoung -Kyu; ...

2015-04-16

The Prismatic Modular Reactor (PMR) is one of the major Very High Temperature Reactor (VHTR) concepts, which consists of hexagonal prismatic fuel blocks and reflector blocks made of nuclear gradegraphite. However, the shape of the graphite blocks could be easily changed by neutron damage duringthe reactor operation and the shape change can create gaps between the blocks inducing the bypass flow.In the VHTR core, two types of gaps, a vertical gap and a horizontal gap which are called bypass gap and cross gap, respectively, can be formed. The cross gap complicates the flow field in the reactor core by connectingmore » the coolant channel to the bypass gap and it could lead to a loss of effective coolant flow in the fuel blocks. Thus, a cross flow experimental facility was constructed to investigate the cross flow phenomena in the core of the VHTR and a series of experiments were carried out under varying flow rates and gap sizes. The results of the experiments were compared with CFD (Computational Fluid Dynamics) analysis results in order to verify its prediction capability for the cross flow phenomena. Fairly good agreement was seen between experimental results and CFD predictions and the local characteristics of the cross flow was discussed in detail. Based on the calculation results, pressure loss coefficient across the cross gap was evaluated, which is necessary for the thermo-fluid analysis of the VHTR core using a lumped parameter code.« less

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions.

PubMed

Geslot, B; Vermeeren, L; Filliatre, P; Lopez, A Legrand; Barbot, L; Jammes, C; Bréaud, S; Oriol, L; Villard, J-F

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 10(20) n∕cm(2). A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

NASA Astrophysics Data System (ADS)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

DOE Office of Scientific and Technical Information (OSTI.GOV)

REID, ROBERT S.; PEARSON, J. BOSIE; STEWART, ERIC T.

2007-01-16

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WSTmore » is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.« less

Simultaneous fermentation and separation in an immobilized cell trickle bed reactor: Acetone-butanol-ethane (ABE) and ethanol fermentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Park, C.H.

1989-01-01

A novel process employing immobilized cells and in-situ product removal was studied for acetone-butanol-ethanol (ABE) fermentation by Clostridium acetobutylicum and ethanol fermentation by Saccharomyces cerevisiae. Experimental studies of ABE fermentation in a trickle bed reactor without product separation showed that solvent production could be improved by one order of magnitude compared to conventional batch fermentation. Control of effluent pH near 4.3 and feed glucose concentrations higher than 10 g/L were the necessary conditions for cell growth and solvent production. A mathematical model using an equilibrium staged model predicted efficient separation of butanol from the fermentation broth. Activity coefficients of multicomponentmore » system were estimated by Wilson's equation or the ASOG method. Inhibition by butanol and organic acids was incorporated into the kinetic expression. Experimental performance of simultaneous fermentation and separation in an immobilized cell trickle bed reactor showed that glucose conversion was improved as predicted by mathematical modeling and analysis. The effect of pH and temperature on ethanol fermentation by Saccharomyces cerevisiae was studied in free and immobilized cell reactors. Conditions for the highest glucose conversion, cell viability and least glycerol yield were determined.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lisowski, Darius D.; Kraus, Adam R.; Bucknor, Matthew D.

A 1/2 scale test facility has been constructed at Argonne National Laboratory to study the heat removal performance and natural circulation flow patterns in a Reactor Cavity Cooling System (RCCS). Our test facility, the Natural convection Shutdown heat removal Test Facility (NSTF), supports the broader goal of developing an inherently safe and fully passive ex-vessel decay heat removal for advanced reactor designs. The project, initiated in 2010 to support the Advanced Reactor Technologies (ART), Small Modular Reactor (SMR), and Next Generation Nuclear Plant (NGNP) programs, has been conducting experimental operations since early 2014. The following paper provides a summary ofmore » some primary design features of the 26-m tall test facility along with a description of the data acquisition suite that guides our experimental practices. Specifics of the distributed fiber optic temperature measurements will be discussed, which introduces an unparalleled level of data density that has never before been implemented in a large scale natural circulation test facility. Results from our test series will then be presented, which provide insight into the thermal hydraulic behavior at steady-state and transient conditions for varying heat flux levels and exhaust chimney configuration states. (C) 2016 Elsevier B.V. All rights reserved.« less

Initial development and performance evaluation of a process for formation of dense carbon by pyrolysis of methane

NASA Technical Reports Server (NTRS)

Noyes, G. P.; Cusick, R. J.

1985-01-01

The three steps in pyrolytic carbon formation are: (1) gaseous hydrocarbon polymerization and aromatic formation; (2) gas-phase condensation and surface adsorption/impingement of polyaromatic hydrocarbon; and (3) final dehydration to carbon. The structure of the carbon in the various stages of formation is examined. The apparatuses and experimental procedures for the pyrolysis of methane in a 60 cm long quartz reactor tube at temperatures ranging from 1400-1600 K are described. The percentage of carbon converted and its density are calculated and tabularly presented. The results reveal that dense carbon formation is maximized and soot eliminated by this procedure. It is observed that conversion efficiency depends on the composition of the inlet gas and conversion increases with increasing temperature. Based on the experimental data a three-man carbon reactor subsystem (CRS) is developed; the functions of the Sabatier Methanation Reactor, two carbon formation reactors and fluid handling components of the CRS are analyzed. The CRS forms 16 kg of carbon at a rate of 0.8 kg/day for 20 days in a two percent volume density quartz wool packing at temperature of 1500-1600 K.

Neutron dose estimation in a zero power nuclear reactor

NASA Astrophysics Data System (ADS)

Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

2016-10-01

This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

Flooding Experiments and Modeling for Improved Reactor Safety

DOE Office of Scientific and Technical Information (OSTI.GOV)

Solmos, M.; Hogan, K. J.; Vierow, K.

2008-09-14

Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge linemore » can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.« less

New infrastructure for studies of transmutation and fast systems concepts

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-09-01

In this work we report initial studies on a low power Accelerator-Driven System as a possible experimental facility for the measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

A low power ADS for transmutation studies in fast systems

NASA Astrophysics Data System (ADS)

Panza, Fabio; Firpo, Gabriele; Lomonaco, Guglielmo; Osipenko, Mikhail; Ricco, Giovanni; Ripani, Marco; Saracco, Paolo; Viberti, Carlo Maria

2017-12-01

In this work, we report studies on a fast low power accelerator driven system model as a possible experimental facility, focusing on its capabilities in terms of measurement of relevant integral nuclear quantities. In particular, we performed Monte Carlo simulations of minor actinides and fission products irradiation and estimated the fission rate within fission chambers in the reactor core and the reflector, in order to evaluate the transmutation rates and the measurement sensitivity. We also performed a photo-peak analysis of available experimental data from a research reactor, in order to estimate the expected sensitivity of this analysis method on the irradiation of samples in the ADS considered.

EXPERIMENTAL STUDIES OF TRANSIENT EFFECTS IN FAST REACTOR FUELS. SERIES I. UO$sub 2$ IRRADIATIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Field, J.H.

1962-11-15

An experimental program to evaluate the performance of FCR and EFCR fuel during transient operation is outlined, and the initial series of tests are described in some detail. Test results from five experiments in the TREAT reactor, using 1-in. OD SS-clad UO/sub 2/ fuel specimens, are compared with regard to fuel temperatures, mechanical integrity, and post-irradiation appearance. Incipient fuel pin failure limits for transients are identified with maximum fuel temperatures in the range of 7000 deg F. Multiple transient damage to the cladding is likely for transients above the melting point of the fuel. (auth)

THE EXPERIENCE IN THE UNITED STATES WITH REACTOR OPERATION AND REACTOR SAFEGUARDS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCullough, C.R.

1958-10-31

Reactors are operating or planned at locations in the United States in cities, near cities, and at remote locations. There is a general pattern that the higher power reactors are not in, but fairly uear cities, and the testing reactors for more hazardous experiments are at remote locations. A great deal has been done on the theoretical and experimental study of importunt features of reactor design. The metal-water reaction is still a theoretical possibility but tests of fuel element burnout under conditions approaching reactor operation gave no reaction. It appears that nucleate boiling does not necessarily result in steam blanketingmore » and fuel melting. Much attention is being given to the calculation of core kinetics but it is being found that temperature, power, and void coefficients cannot be calculated with accuracy and experiments are required. Some surprises are found giving positive localized void coefficients. Possible oscillatory behavior of reactors is being given careful study. No dangerous oscillations have been found in operating reactors but osciliations hare appeared in experimeats. The design of control and safety systems varies wvith different constructors. The relation of control to the kinetic behavior of the reactor is being studied. The importance of sensing element locations in order to know actual local reactor power level is being recognized. The time constants of instrumentation as related to reactor kinetics are being studied. Pressure vessels for reactors are being designed and manufactured. Many of these are beyond any previous experience. The stress problem is being given careful study. The effect of radiation is being studied experimentally. The stress problems of piping and pressure vessels is a difficult design problem being met successfully in reactor plants. The proper organization and procedure for operation of reactors is being evolved for resourch, testing, and power reactors. The importance of written standards and instructions for both normal and abnormal operating conditions is recogmized. Corfinement of radioactive materials either by tight steel shells, tight buildings, or semi-tight structures vented through filters is considered necessary in the United States. A discussion will be given of specifications, construction, and testing of these structures. The need for emergency plans has been stressed by recent experiences in radioactive releases. The problems of such plans to cover all grades of accidents will be discussed. The theoretical consequences of releases of radioactive materials have been studied and these results will be compared with actual experience. The problem of exposures from normal and abnormal operetion of reactors is a problem of desiga and operation on one hand and the amount of damage to be expected on the other. The safeguard problem is closely related to the acceptable doses of radiouctivity which the ICRP recommend. The future of atomic energy depends upon adequate safeguards and economical design and operation. Accepted criteria are required to guide designers as to the proper balance of caution and boldness. (auth)« less

Investigation of sewage sludge treatment using air plasma assisted gasification.

PubMed

Striūgas, Nerijus; Valinčius, Vitas; Pedišius, Nerijus; Poškas, Robertas; Zakarauskas, Kęstutis

2017-06-01

This study presents an experimental investigation of downdraft gasification process coupled with a secondary thermal plasma reactor in order to perform experimental investigations of sewage sludge gasification, and compare process parameters running the system with and without the secondary thermal plasma reactor. The experimental investigation were performed with non-pelletized mixture of dried sewage sludge and wood pellets. To estimate the process performance, the composition of the producer gas, tars, particle matter, producer gas and char yield were measured at the exit of the gasification and plasma reactor. The research revealed the distribution of selected metals and chlorine in the process products and examined a possible formation of hexachlorobenzene. It determined that the plasma assisted processing of gaseous products changes the composition of the tars and the producer gas, mostly by destruction of hydrocarbon species, such as methane, acetylene, ethane or propane. Plasma processing of the producer gas reduces their calorific value but increases the gas yield and the total produced energy amount. The presented technology demonstrated capability both for applying to reduce the accumulation of the sewage sludge and production of substitute gas for drying of sewage sludge and electrical power. Copyright © 2017 Elsevier Ltd. All rights reserved.

On the possible use of the MASURCA reactor as a flexible, high-intensity, fast neutron beam facility

NASA Astrophysics Data System (ADS)

Dioni, Luca; Jacqmin, Robert; Sumini, Marco; Stout, Brian

2017-09-01

In recent work [1, 2], we have shown that the MASURCA research reactor could be used to deliver a fairly-intense continuous fast neutron beam to an experimental room located next to the reactor core. As a consequence of the MASURCA favorable characteristics and diverse material inventories, the neutron beam intensity and spectrum can be further tailored to meet the users' needs, which could be of interest for several applications. Monte Carlo simulations have been performed to characterize in detail the extracted neutron (and photon) beam entering the experimental room. These numerical simulations were done for two different bare cores: A uranium metallic core (˜30% 235U enriched) and a plutonium oxide core (˜25% Pu fraction, ˜78% 239Pu). The results show that the distinctive resonance energy structures of the two core leakage spectra are preserved at the channel exit. As the experimental room is large enough to house a dedicated set of neutron spectrometry instruments, we have investigated several candidate neutron spectrum measurement techniques, which could be implemented to guarantee well-defined, repeatable beam conditions to users. Our investigation also includes considerations regarding the gamma rays in the beams.

Army Posture Statement: A Statement on the Posture of the United States Army, 2009

DTIC Science & Technology

2009-05-07

Research Institute Army Physical Readiness Training (FM 3-22.02) Army Preparatory School Army Prepositioned Stocks (APS) Army Reserve Employer Relations...ARFORGEN Army Force Generation AFRICOM Africa Command AMAP Army Medical Action Plan AMC Army Material Command APA Army Prepositioned Stocks AR Army...Ordnance Disposal ES2 Every Soldier a Sensor ETF Enterprise Task Force FCS Future Combat Systems FM Field Manual FORSCOM Forces Command FY Fiscal Year

Phonological studies of the new gas-induced agitated reactor using computational fluid dynamics.

PubMed

Yang, T C; Hsu, Y C; Wang, S F

2001-06-01

An ozone-induced agitated reactor has been found to be very effective in degrading industrial wastewater. However, the cost of the ozone generation as well as its short residence time in reactors has restricted its application in a commercial scale. An innovated gas-induced draft tube installed inside a conventional agitated reactor was proved to effectively retain the ozone in a reactor. The setup was demonstrated to significantly promote the ozone utilization rate up to 96% from the conventional rate of 60% above the onset speed. This work investigates the mixing mechanism of an innovated gas-induced reactor for the future scale-up design by using the technique of computational fluid dynamics. A three-dimensional flow model was proposed to compute the liquid-gas free surface as well as the flow patterns inside the reactor. The turbulent effects generated by two 45 degrees pitch-blade turbines were considered and the two phases mixing phenomena were also manipulated by the Eulerian-Eulerian techniques. The consistency of the free surface profiles and the fluid flow patterns proved a good agreement between computational results and the experimental observation.

Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlisle, R.P.; Zenzen, J.M.

1994-01-01

This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story ofmore » fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.« less

Demonstration of catalytic combustion with residual fuel

NASA Technical Reports Server (NTRS)

Dodds, W. J.; Ekstedt, E. E.

1981-01-01

An experimental program was conducted to demonstrate catalytic combustion of a residual fuel oil. Three catalytic reactors, including a baseline configuration and two backup configurations based on baseline test results, were operated on No. 6 fuel oil. All reactors were multielement configurations consisting of ceramic honeycomb catalyzed with palladium on stabilized alumina. Stable operation on residual oil was demonstrated with the baseline configuration at a reactor inlet temperature of about 825 K (1025 F). At low inlet temperature, operation was precluded by apparent plugging of the catalytic reactor with residual oil. Reduced plugging tendency was demonstrated in the backup reactors by increasing the size of the catalyst channels at the reactor inlet, but plugging still occurred at inlet temperature below 725 K (845 F). Operation at the original design inlet temperature of 589 K (600 F) could not be demonstrated. Combustion efficiency above 99.5% was obtained with less than 5% reactor pressure drop. Thermally formed NO sub x levels were very low (less than 0.5 g NO2/kg fuel) but nearly 100% conversion of fuel-bound nitrogen to NO sub x was observed.

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

NASA Astrophysics Data System (ADS)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

Stimulation of methanogenesis in anaerobic digesters treating leachate from a municipal solid waste incineration plant with carbon cloth.

PubMed

Lei, Yuqing; Sun, Dezhi; Dang, Yan; Chen, Huimin; Zhao, Zhiqiang; Zhang, Yaobin; Holmes, Dawn E

2016-12-01

Bio-methanogenic digestion of incineration leachate is hindered by high OLRs, which can lead to build-up of VFAs, drops in pH and ultimately in reactor souring. It was hypothesized that incorporation of carbon cloth into reactors treating leachate would promote DIET and enhance reactor performance. To examine this possibility, carbon cloth was added to laboratory-scale UASB reactors that were fed incineration leachate. As expected, the carbon-cloth amended reactor could operate stably with a 34.2% higher OLR than the control (49.4 vs 36.8kgCOD/(m 3 d)). Microbial community analysis showed that bacteria capable of extracellular electron transfer and methanogens known to participate in DIET were enriched on the carbon cloth surface, and conductivity of sludge from the carbon cloth amended reactor was almost twofold higher than sludge from the control (9.77 vs 5.47μS/cm), suggesting that microorganisms in the experimental reactor may have been expressing electrically conductive filaments. Copyright © 2016 Elsevier Ltd. All rights reserved.

From the laboratory to the soldier: providing tactical behaviors for Army robots

NASA Astrophysics Data System (ADS)

Knichel, David G.; Bruemmer, David J.

2008-04-01

The Army Future Combat System (FCS) Operational Requirement Document has identified a number of advanced robot tactical behavior requirements to enable the Future Brigade Combat Team (FBCT). The FBCT advanced tactical behaviors include Sentinel Behavior, Obstacle Avoidance Behavior, and Scaled Levels of Human-Machine control Behavior. The U.S. Army Training and Doctrine Command, (TRADOC) Maneuver Support Center (MANSCEN) has also documented a number of robotic behavior requirements for the Army non FCS forces such as the Infantry Brigade Combat Team (IBCT), Stryker Brigade Combat Team (SBCT), and Heavy Brigade Combat Team (HBCT). The general categories of useful robot tactical behaviors include Ground/Air Mobility behaviors, Tactical Mission behaviors, Manned-Unmanned Teaming behaviors, and Soldier-Robot Interface behaviors. Many DoD research and development centers are achieving the necessary components necessary for artificial tactical behaviors for ground and air robots to include the Army Research Laboratory (ARL), U.S. Army Research, Development and Engineering Command (RDECOM), Space and Naval Warfare (SPAWAR) Systems Center, US Army Tank-Automotive Research, Development and Engineering Center (TARDEC) and non DoD labs such as Department of Energy (DOL). With the support of the Joint Ground Robotics Enterprise (JGRE) through DoD and non DoD labs the Army Maneuver Support Center has recently concluded successful field trails of ground and air robots with specialized tactical behaviors and sensors to enable semi autonomous detection, reporting, and marking of explosive hazards to include Improvised Explosive Devices (IED) and landmines. A specific goal of this effort was to assess how collaborative behaviors for multiple unmanned air and ground vehicles can reduce risks to Soldiers and increase efficiency for on and off route explosive hazard detection, reporting, and marking. This paper discusses experimental results achieved with a robotic countermine system that utilizes autonomous behaviors and a mixed-initiative control scheme to address the challenges of detecting and marking buried landmines. Emerging requirements for robotic countermine operations are outlined as are the technologies developed under this effort to address them. A first experiment shows that the resulting system was able to find and mark landmines with a very low level of human involvement. In addition, the data indicates that the robotic system is able to decrease the time to find mines and increase the detection accuracy and reliability. Finally, the paper presents current efforts to incorporate new countermine sensors and port the resulting behaviors to two fielded military systems for rigorous assessing.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zabelin, A.I.; Shmelev, V.E.

Radiolysis of the coolant proceeds at a higher rate in a boiling water reactor as compared to a water-moderated, water-cooled reactor. The radiolytic gases (hydrogen and oxygen) exiting the reactor together with steam can form a potentially explosive mixture. Special interest attaches to the results obtained under the codnitions of prolonged operation of the VK-50 reactor. Tests of various water-chemistry conditions which were performed in the experimental reactor showed their critical influence on the rate of progress of radiolytic processes. The entire period of operation of the reactor may be arbitrarily divided into three stages, each of which is characterizedmore » by its own peculiar conditions of water chemistry and range of thermal power. From stage to stage, there is a noticeable improvement in the coolant quality which to a limited extent is reflected in the exit of radiolytic gases with the steam. The concentration of radiolytic gases increases with decreased power and with an increased content of corrosion products and other contaminants in the coolant.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

Bioelectrochemical enhancement of methane production from highly concentrated food waste in a combined anaerobic digester and microbial electrolysis cell.

PubMed

Park, Jungyu; Lee, Beom; Tian, Donjie; Jun, Hangbae

2018-01-01

A microbial electrolysis cell (MEC) is a promising technology for enhancing biogas production from an anaerobic digestion (AD) reactor. In this study, the effects of the MEC on the rate of methane production from food waste were examined by comparing an AD reactor with an AD reactor combined with a MEC (AD+MEC). The use of the MEC accelerated methane production and stabilization via rapid organic oxidation and rapid methanogenesis. Over the total experimental period, the methane production rate and stabilization time of the AD+MEC reactor were approximately 1.7 and 4.0 times faster than those of the AD reactor. Interestingly however, at the final steady state, the methane yields of both the reactors were similar to the theoretical maximum methane yield. Based on these results, the MEC did not increase the methane yield over the theoretical value, but accelerated methane production and stabilization by bioelectrochemical reactions. Copyright © 2017 Elsevier Ltd. All rights reserved.

Operation of the WWR-S Reactor in Poland and in the Period 1961-1962; DOKLAD OB ISPOL'ZOVANII I EKSPLUATATSII REAKTORA TIPA VVR-C V POL'SHE ZA 1961-1962 GODA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aleksandrowicz, J.

1963-03-01

The experimental equipment used in the work at the horizontal reactor channels is listed. Diagrams of the utilization of the nominal reactor power and core loading are given, the reactivity fractions of the separate fuel assemblies are detonated, together with the diagram of reactivity versus burnup. Reactor channels and space used for sample irradiation and isotope production are described, and the total number of irradiations is given. Results of the measurements connected with the routine reactor operation are quoted, namely: analysis of water purity in the primary circuit, analysis of the work of the ion exchanger and mechanical filter, andmore » analysis of air activity in the special ventilation system. Data are given concerning radiation protection of personnel, including individual monitoring. Leak testing of the fuel elements is discussed. Damage of the reactor equipment and appearance of alarm signals are described. (auth)« less

Real-time LMR control parameter generation using advanced adaptive synthesis

DOE Office of Scientific and Technical Information (OSTI.GOV)

King, R.W.; Mott, J.E.

1990-01-01

The reactor delta T'', the difference between the average core inlet and outlet temperatures, for the liquid-sodium-cooled Experimental Breeder Reactor 2 is empirically synthesized in real time from, a multitude of examples of past reactor operation. The real-time empirical synthesis is based on reactor operation. The real-time empirical synthesis is based on system state analysis (SSA) technology embodied in software on the EBR 2 data acquisition computer. Before the real-time system is put into operation, a selection of reactor plant measurements is made which is predictable over long periods encompassing plant shutdowns, core reconfigurations, core load changes, and plant startups.more » A serial data link to a personal computer containing SSA software allows the rapid verification of the predictability of these plant measurements via graphical means. After the selection is made, the real-time synthesis provides a fault-tolerant estimate of the reactor delta T accurate to {plus}/{minus}1{percent}. 5 refs., 7 figs.« less

Development of a Polysilicon Process Based on Chemical Vapor Deposition of Dichlorosilane in an Advanced Siemen's Reactor

NASA Technical Reports Server (NTRS)

Arevidson, A. N.; Sawyer, D. H.; Muller, D. M.

1983-01-01

Dichlorosilane (DCS) was used as the feedstock for an advanced decomposition reactor for silicon production. The advanced reactor had a cool bell jar wall temperature, 300 C, when compared to Siemen's reactors previously used for DCS decomposition. Previous reactors had bell jar wall temperatures of approximately 750 C. The cooler wall temperature allows higher DCS flow rates and concentrations. A silicon deposition rate of 2.28 gm/hr-cm was achieved with power consumption of 59 kWh/kg. Interpretation of data suggests that a 2.8 gm/hr-cm deposition rate is possible. Screening of lower cost materials of construction was done as a separate program segment. Stainless Steel (304 and 316), Hastalloy B, Monel 400 and 1010-Carbon Steel were placed individually in an experimental scale reactor. Silicon was deposited from trichlorosilane feedstock. The resultant silicon was analyzed for electrically active and metallic impurities as well as carbon. No material contributed significant amounts of electrically active or metallic impurities, but all contributed carbon.

Evaluation of IKTS Transparent Polycrystalline Magnesium Aluminate Spinel (MgAl2O4) for Armor and Infrared Dome/Window Applications

DTIC Science & Technology

2013-03-01

interacted with (15). 4.3.3 Experimental Procedure Two MgAl2O4 spinel samples with nominal 0.6- and 1.6-μm mean grain sizes were tested using advanced...unable to make specific quantitative predictions at this time. Due to the nature of the experimental process, this technique is suitable only for...Information From Spherical Indentation; ARL-TR-4229; U.S. Army Research Laboratory: Aberdeen Proving Ground, MD, 2007. 24. ASTM E112. Standard Test

Experimental Measurements at the MASURCA Facility

NASA Astrophysics Data System (ADS)

Assal, W.; Bosq, J. C.; Mellier, F.

2012-12-01

Dedicated to the neutronics studies of fast and semi-fast reactor lattices, MASURCA (meaning “mock-up facility for fast breeder reactor studies at CADARACHE”) is an airflow cooled fast reactor operating at a maximum power of 5 kW playing an important role in the CEA research activities. At this facility, a lot of neutron integral experimental programs were undertaken. The purpose of this poster is to show a panorama of the facility from this experimental measurement point of view. A hint at the forthcoming refurbishment will be included. These programs include various experimental measurements (reactivity, distributions of fluxes, reaction rates), performed essentially with fission chambers, in accordance with different methods (noise methods, radial or axial traverses, rod drops) and involving several devices systems (monitors, fission chambers, amplifiers, power supplies, data acquisition systems ...). For this purpose are implemented electronics modules to shape the signals sent from the detectors in various mode (fluctuation, pulse, current). All the electric and electronic devices needed for these measurements and the relating wiring will be fully explained through comprehensive layouts. Data acquired during counting performed at the time of startup phase or rod drops are analyzed by the mean of a Neutronic Measurement Treatment (TMN in French) programmed on the basis of the MATLAB software. This toolbox gives the opportunity of data files management, reactivity valuation from neutronics measurements and transient or divergence simulation at zero power. Particular TMN using at MASURCA will be presented.

Polypropylene Production Optimization in Fluidized Bed Catalytic Reactor (FBCR): Statistical Modeling and Pilot Scale Experimental Validation

PubMed Central

Khan, Mohammad Jakir Hossain; Hussain, Mohd Azlan; Mujtaba, Iqbal Mohammed

2014-01-01

Propylene is one type of plastic that is widely used in our everyday life. This study focuses on the identification and justification of the optimum process parameters for polypropylene production in a novel pilot plant based fluidized bed reactor. This first-of-its-kind statistical modeling with experimental validation for the process parameters of polypropylene production was conducted by applying ANNOVA (Analysis of variance) method to Response Surface Methodology (RSM). Three important process variables i.e., reaction temperature, system pressure and hydrogen percentage were considered as the important input factors for the polypropylene production in the analysis performed. In order to examine the effect of process parameters and their interactions, the ANOVA method was utilized among a range of other statistical diagnostic tools such as the correlation between actual and predicted values, the residuals and predicted response, outlier t plot, 3D response surface and contour analysis plots. The statistical analysis showed that the proposed quadratic model had a good fit with the experimental results. At optimum conditions with temperature of 75°C, system pressure of 25 bar and hydrogen percentage of 2%, the highest polypropylene production obtained is 5.82% per pass. Hence it is concluded that the developed experimental design and proposed model can be successfully employed with over a 95% confidence level for optimum polypropylene production in a fluidized bed catalytic reactor (FBCR). PMID:28788576

Comprehensive kinetic model for the low-temperature oxidation of hydrocarbons

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gaffuri, P.; Faravelli, T.; Ranzi, E.

1997-05-01

The oxidation chemistry in the low- and intermediate-temperature regimes (600--900 K) is important and plays a significant role in the overall combustion process. Autoignition in diesel engines as well as end-gas autoignition and knock phenomena in s.i. engines are initiated at these low temperatures. The low-temperature oxidation chemistry of linear and branched alkanes is discussed with the aim of unifying their complex behavior in various experimental systems using a single detailed kinetic model. New experimental data, obtained in a pressurized flow reactor, as well as in batch- and jet-stirred reactors, are useful for a better definition of the region ofmore » cool flames and negative temperature coefficient (NTC) for pure hydrocarbons from propane up to isooctane. Thermochemical oscillations and the NTC region of the reaction rate of the low-temperature oxidation of n-heptane and isooctane in a jet-stirred flow reactor are reproduced quite well by the model, not only in a qualitative way but in terms of the experimental frequencies and intensities of cool flames. Very good agreement is also observed for fuel conversion and intermediate-species formation. Irrespective of the experimental system, the same critical reaction steps always control these phenomena. The results contribute to the definition of a limited set of fundamental kinetic parameters that should be easily extended to model heavier alkanes.« less

Experimental validation of the DARWIN2.3 package for fuel cycle applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

San-Felice, L.; Eschbach, R.; Bourdot, P.

2012-07-01

The DARWIN package, developed by the CEA and its French partners (AREVA and EDF) provides the required parameters for fuel cycle applications: fuel inventory, decay heat, activity, neutron, {gamma}, {alpha}, {beta} sources and spectrum, radiotoxicity. This paper presents the DARWIN2.3 experimental validation for fuel inventory and decay heat calculations on Pressurized Water Reactor (PWR). In order to validate this code system for spent fuel inventory a large program has been undertaken, based on spent fuel chemical assays. This paper deals with the experimental validation of DARWIN2.3 for the Pressurized Water Reactor (PWR) Uranium Oxide (UOX) and Mixed Oxide (MOX) fuelmore » inventory calculation, focused on the isotopes involved in Burn-Up Credit (BUC) applications and decay heat computations. The calculation - experiment (C/E-1) discrepancies are calculated with the latest European evaluation file JEFF-3.1.1 associated with the SHEM energy mesh. An overview of the tendencies is obtained on a complete range of burn-up from 10 to 85 GWd/t (10 to 60 GWcVt for MOX fuel). The experimental validation of the DARWIN2.3 package for decay heat calculation is performed using calorimetric measurements carried out at the Swedish Interim Spent Fuel Storage Facility for Pressurized Water Reactor (PWR) assemblies, covering a large burn-up (20 to 50 GWd/t) and cooling time range (10 to 30 years). (authors)« less

Biogasification of community-derived biomass and solid wastes in a pilot-scale SOLCON reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Srivastava, V.J.; Biljetina, R.; Isaacson, H.R.

1988-01-01

The Institute of Gas Technology has developed a novel, solids- concentrating (SOLCON) bioreactor to convert a variety of individual or mixed feedstocks (biomass and wastes) to methane at higher rates and efficiencies than those obtained from conventional high-rate anaerobic digesters. The biogasification studies are being conducted in a pilot-scale experimental test unit (ETU) located in the Walt Disney World Resort Complex, Orlando, Florida. This paper describes the ETU facility, the logistics of feedstock integration, the SOLCON reactor design and operating techniques, and the results obtained during 4 years of stable, uninterrupted operation with different feedstocks. The SOLCON reactor consistently outperformedmore » the conventional stirred-tank reactor by 20% to 50%.« less

Current Results of NEUTRINO-4 Experiment

NASA Astrophysics Data System (ADS)

Serebrov, A.; Ivochkin, V.; Samoilov, R.; Fomin, A.; Polyushkin, A.; Zinoviev, V.; Neustroev, P.; Golovtsov, V.; Chernyj, A.; Zherebtsov, O.; Martemyanov, V.; Tarasenkov, V.; Aleshin, V.; Petelin, A.; Izhutov, A.; Tuzov, A.; Sazontov, S.; Ryazanov, D.; Gromov, M.; Afanasiev, V.; Zaytsev, M.; Chaikovskii, M.

2017-12-01

The main goal of experiment “Neutrino-4” is to search for the oscillation of reactor antineutrino to a sterile state. Experiment is conducted on SM-3 research reactor (Dimitrovgrad, Russia). Data collection with full-scale detector with liquid scintillator volume of 3m3 was started in June 2016. We present the results of measurements of reactor antineutrino flux dependence on the distance in range 6- 12 meters from the center of the reactor. At that distance range, the fit of experimental dependence has good agreement with the law 1/L2. Which means, at achieved during the data collecting accuracy level oscillations to sterile state are not observed. In addition, the spectrum of prompt signals of neutrino-like events at different distances have been presented.

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Lisowski, Darius D.; Bucknor, Matthew

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performancemore » of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.« less

Integral Design Methodology of Photocatalytic Reactors for Air Pollution Remediation.

PubMed

Passalía, Claudio; Alfano, Orlando M; Brandi, Rodolfo J

2017-06-07

An integral reactor design methodology was developed to address the optimal design of photocatalytic wall reactors to be used in air pollution control. For a target pollutant to be eliminated from an air stream, the proposed methodology is initiated with a mechanistic derived reaction rate. The determination of intrinsic kinetic parameters is associated with the use of a simple geometry laboratory scale reactor, operation under kinetic control and a uniform incident radiation flux, which allows computing the local superficial rate of photon absorption. Thus, a simple model can describe the mass balance and a solution may be obtained. The kinetic parameters may be estimated by the combination of the mathematical model and the experimental results. The validated intrinsic kinetics obtained may be directly used in the scaling-up of any reactor configuration and size. The bench scale reactor may require the use of complex computational software to obtain the fields of velocity, radiation absorption and species concentration. The complete methodology was successfully applied to the elimination of airborne formaldehyde. The kinetic parameters were determined in a flat plate reactor, whilst a bench scale corrugated wall reactor was used to illustrate the scaling-up methodology. In addition, an optimal folding angle of the corrugated reactor was found using computational fluid dynamics tools.

Nuclear reactors built, being built, or planned, 1994

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1995-07-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation

NASA Astrophysics Data System (ADS)

Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François

2017-12-01

A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

Nuclear reactors built, being built, or planned: 1995

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-08-01

This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristicmore » and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).« less

Flight testing and frequency domain analysis for rotorcraft handling qualities characteristics

NASA Technical Reports Server (NTRS)

Ham, Johnnie A.; Gardner, Charles K.; Tischler, Mark B.

1993-01-01

A demonstration of frequency domain flight testing techniques and analyses was performed on a U.S. Army OH-58D helicopter in support of the OH-58D Airworthiness and Flight Characteristics Evaluation and the Army's development and ongoing review of Aeronautical Design Standard 33C, Handling Qualities Requirements for Military Rotorcraft. Hover and forward flight (60 knots) tests were conducted in 1 flight hour by Army experimental test pilots. Further processing of the hover data generated a complete database of velocity, angular rate, and acceleration frequency responses to control inputs. A joint effort was then undertaken by the Airworthiness Qualification Test Directorate (AQTD) and the U.S. Army Aeroflightdynamics Directorate (AFDD) to derive handling qualities information from the frequency response database. A significant amount of information could be extracted from the frequency domain database using a variety of approaches. This report documents numerous results that have been obtained from the simple frequency domain tests; in many areas, these results provide more insight into the aircraft dynamics that affect handling qualities than to traditional flight tests. The handling qualities results include ADS-33C bandwidth and phase delay calculations, vibration spectral determinations, transfer function models to examine single axis results, and a six degree of freedom fully coupled state space model. The ability of this model to accurately predict aircraft responses was verified using data from pulse inputs. This report also documents the frequency-sweep flight test technique and data analysis used to support the tests.

Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

NASA Technical Reports Server (NTRS)

Martin, James J.; Reid, Robert S.

2004-01-01

A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

Steam reforming of heptane in a fluidized bed membrane reactor

NASA Astrophysics Data System (ADS)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

A multi-physics analysis for the actuation of the SSS in opal reactor

NASA Astrophysics Data System (ADS)

Ferraro, Diego; Alberto, Patricio; Villarino, Eduardo; Doval, Alicia

2018-05-01

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

Experimental Constraints on Neutrino Spectra Following Fission

NASA Astrophysics Data System (ADS)

Napolitano, Jim; Daya Bay Collaboration

2016-09-01

We discuss new initiatives to constrain predictions of fission neutrino spectra from nuclear reactors. These predictions are germane to the understanding of reactor flux anomalies; are needed to reduce systematic uncertainty in neutrino oscillation spectra; and inform searches for the diffuse supernova neutrino background. The initiatives include a search for very high- Q beta decay components to the neutrino spectrum from the Daya Bay power plant; plans for a measurement of the β- spectrum from 252Cf fission products; and precision measurements of the 235U fission neutrino spectrum from PROSPECT and other very short baseline reactor experiments.

Core plasma design of the compact helical reactor with a consideration of the equipartition effect

NASA Astrophysics Data System (ADS)

Goto, T.; Miyazawa, J.; Yanagi, N.; Tamura, H.; Tanaka, T.; Sakamoto, R.; Suzuki, C.; Seki, R.; Satake, S.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

2018-07-01

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport.

2009 Center for Army Leadership Annual Survey of Army Leadership (CASAL): Army Education

DTIC Science & Technology

2010-06-11

right time, handling pre- education attitudes, and tracking performance gains and career advantages related to academics.  Developing current, relevant...Army Leadership Technical Report 2010-2 2009 CENTER FOR ARMY LEADERSHIP ANNUAL SURVEY OF ARMY LEADERSHIP (CASAL): ARMY EDUCATION ...Joshua Hatfield ICF International John P. Steele Center for Army Leadership June 2010 The Center for Army Leadership An

Student Collaboration in a Series of Integrated Experiments to Study Enzyme Reactor Modeling with Immobilized Cell-Based Invertase

ERIC Educational Resources Information Center

Taipa, M. A^ngela; Azevedo, Ana M.; Grilo, Anto´nio L.; Couto, Pedro T.; Ferreira, Filipe A. G.; Fortuna, Ana R. M.; Pinto, Ine^s F.; Santos, Rafael M.; Santos, Susana B.

2015-01-01

An integrative laboratory study addressing fundamentals of enzyme catalysis and their application to reactors operation and modeling is presented. Invertase, a ß-fructofuranosidase that catalyses the hydrolysis of sucrose, is used as the model enzyme at optimal conditions (pH 4.5 and 45 °C). The experimental work involves 3 h of laboratory time…

Operational performance of the three bean salad control algorithm on the ACRR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute.

FAST CHOPPER BUILDING, TRA665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKEDIN ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

FAST CHOPPER BUILDING, TRA-665, INTERIOR. LOWER (DETECTOR) LEVEL. NOTE BRICKED-IN WINDOW ON MTR SIDE. USED FOR STORAGE OF LEAD BRICKS AFTER EXPERIMENTAL NEUTRON INSTRUMENTS WERE REMOVED. SIGN SAYS "IN-PROCESS LEAD SOURCE STORAGE." INL NEGATIVE NO. HD-42-2. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Operational performance of the three bean salad control algorithm on the ACRR

NASA Astrophysics Data System (ADS)

Ball, Russell M.; Madaras, John J.; Trowbridge, F. Ray; Talley, Darren G.; Parma, Edward J.

1991-01-01

Experimental tests on the Annular Core Research Reactor have confirmed that the ``Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute.

Conceptual design of a thermalhydraulic loop for multiple test geometries at supercritical conditions named Supercritical Phenomena Experimental Test Apparatus (SPETA)

NASA Astrophysics Data System (ADS)

Adenariwo, Adepoju

The efficiency of nuclear reactors can be improved by increasing the operating pressure of current nuclear reactors. Current CANDU-type nuclear reactors use heavy water as coolant at an outlet pressure of up to 11.5 MPa. Conceptual SuperCritical Water Reactors (SCWRs) will operate at a higher coolant outlet pressure of 25 MPa. Supercritical water technology has been used in advanced coal plants and its application proves promising to be employed in nuclear reactors. To better understand how supercritical water technology can be applied in nuclear power plants, supercritical water loops are used to study the heat transfer phenomena as it applies to CANDU-type reactors. A conceptual design of a loop known as the Supercritical Phenomena Experimental Apparatus (SPETA) has been done. This loop has been designed to fit in a 9 m by 2 m by 2.8 m enclosure that will be installed at the University of Ontario Institute of Technology Energy Research Laboratory. The loop include components to safely start up and shut down various test sections, produce a heat source to the test section, and to remove reject heat. It is expected that loop will be able to investigate the behaviour of supercritical water in various geometries including bare tubes, annulus tubes, and multi-element-type bundles. The experimental geometries are designed to match the fluid properties of Canadian SCWR fuel channel designs so that they are representative of a practical application of supercritical water technology in nuclear plants. This loop will investigate various test section orientations which are the horizontal, vertical, and inclined to investigate buoyancy effects. Frictional pressure drop effects and satisfactory methods of estimating hydraulic resistances in supercritical fluid shall also be estimated with the loop. Operating limits for SPETA have been established to be able to capture the important heat transfer phenomena at supercritical conditions. Heat balance and flow calculations have been done to appropriately size components in the loop. Sensitivity analysis has been done to find the optimum design for the loop.

Henry Wade (1876-1955), pioneer of urological surgery, museum conservator and war veteran.

PubMed

Gardner, Dugald

2017-01-01

Henry Wade graduated in the Edinburgh Medical School in 1898 before spending two years with the British army during the Anglo-Boer war. Returning to this country, he joined Francis Caird, surgeon to the Edinburgh Royal Infirmary. Appointed Conservator of the museum of the Royal College of Surgeons of Edinburgh, Wade met young William Ford Robertson. In a study of experimental cancer they concluded that some neoplasms were caused by bacteria. Wade became increasingly recognised as an authority in urology. His growing practice was interrupted by the First World War. Already a member of the Royal Army Medical Corps, he served for five years in the Middle East, in Gallipoli and then with the army in an approach to Jerusalem. Resuming civilian life, Wade combined an extensive urological practice with membership of the Council of the RCSEd. He became President in 1935. Married in 1924, his wife died four years later after an operation by a colleague, David Wilkie. Director of Surgery to the Scottish Emergency Medical Service when the Second World War broke out, Wade was made a Knight Bachelor in 1946. He died in 1955.

The US Army Foreign Comparative Test fuel cell program

NASA Astrophysics Data System (ADS)

Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

(Boiling water reactor (BWR) CORA experiments)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ott, L.J.

To participate in the 1990 CORA Workshop at Kernforschungszentrum Karlsruhe (KfK) GmbH, Karlsruhe, FRG, on October 1--4, and to participate in detailed discussions on October 5 with the KfK CORA Boiling Water Reactor (BWR) experiments. The traveler attended the 1990 CORA Workshop at KfK, FRG. Participation included the presentation of a paper on work performed by the Boiling Water Reactor Core Melt Progression Phenomena Program at Oak Ridge National Laboratory (ORNL) on posttest analyses of CORA BWR experiments. The Statement of Work (November 1989) for the BWR Core Melt Progression Phenomena Program provides for pretest and posttest analyses of themore » BWR CORA experiments performed at KfK. Additionally, it is intended that ORNL personnel participate in the planning process for future CORA BWR experiments. For these purposes, meetings were held with KfK staff to discuss such topics as (1) experimental test schedule, (2) BWR test conduct, (3) perceived BWR experimental needs, and (4) KfK operational staff needs with respect to ORNL support. 19 refs.« less

Development and application of kinetic model on biological anoxic/aerobic filter.

PubMed

Kim, Youngnoh; Tanaka, Kazuhiro; Lee, Yong-Woo; Chung, Jinwook

2008-01-01

An up-flow biological anoxic filter (BANF) has been developed to achieve high removal performance of suspended solids and BOD removal as well as nitrogen. With a view to understand treatment mechanisms, we developed a filtration model that incorporates filtration, deposit scoring and biological reactions simultaneously. The biological reactions consist of four types of reaction; dissolution of organic particles; utilization of dissolved organic matter; denitrification; and self-degradation of bacteria. Whereas the reactor is generally assumed to be a plug flow reactor in the filtration model, it is assumed a continuous-flow stirred tank reactor (CSTR) in the model of biological reactions. The hydrodynamics is supposed that the filter bottom (the portion sludge settled) is a CSTR and the filter bed (the portion filled with filter media) consists of number of CSTR of equal size arranged in series. The model obtained in this study was verified and simulated using experimental results taken from a pilot-scale plant and predicted the experimental data well, applying to design and operate BANF.

Electrolytic ammonia removal and current efficiency by a vermiculite-packed electrochemical reactor

PubMed Central

Li, Liang; Yao, Ji; Fang, Xueyou; Huang, Yuanxing; Mu, Yan

2017-01-01

The ammonia removal as well as the current efficiency during electrolysis was investigated by using a vermiculite-packed electrochemical reactor under continuous mode. Experimental results showed that adsorption of ammonia by vermiculite and electrolytic desorption of ammonia simultaneously existed in the reactor, leading to 89% removal of initial 30 mg N/L ammonia and current efficiency of 25% under the condition of 2.0 A, 6.0 min hydraulic retention time with 300 mg Cl/L chloride as the catalyst. The ammonia removal capacity had a linear relationship with the products of hydraulic retention time, current and chloride concentration within experimental conditions. The treatment results of secondary effluent indicated that 29.9 mg N/L ammonia can be reduced to 4.6 mg N/L with 72% removal of total nitrogen and a current efficiency of 23%, which was 2% less than synthetic wastewater due to the reducing components in the real wastewater. PMID:28102340

Supercritical Water Experimental Setup for µSR

NASA Astrophysics Data System (ADS)

Liu, Guangdong; Chen, Yanggang; Morrison, Alexander H.; Koda, Akihiro; Percival, Paul W.; Ghandi, Khashayar

The Canadian design for Generation IV nuclear reactors uses supercritical water (SCW, water above its critical point of 374 °C, 221 bar (1 bar = 100 kPa)) as the coolant. Supercritical water-cooled reactors (SCWRs) are designed towards sustainability, economic benefits, improved safety, and longer lifespan. Despite the potential advantages of SCWRs, we know very little about the kinetics of radiolysis products that are formed in them because of the limitations of experimental instruments under the extreme conditions of SCW. The radiolysis products can accumulate over time and create a very corrosive environment. Our group has developed and tested an apparatus suitable for muon spin rotation (µSR) studies of water and aqueous solutions up to 550 °C and 250 bar, close to the conditions at the reactor outlet of the proposed Canadian SCWR design (625 °C and 250 bar). The reaction kinetics information obtained from our setup, together with computer simulations, will aid us in developing chemical control strategies to minimize corrosion in SCWRs.

Positive direct current corona discharges in single wire-duct electrostatic precipitators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yehia, Ashraf, E-mail: yehia30161@yahoo.com; Department of Physics, Faculty of Science, Assiut University, Assiut 71516, Arab Republic of Egypt; Abdel-Fattah, E.

This paper is aimed to study the characteristics of the positive dc corona discharges in single wire-duct electrostatic precipitators. Therefore, the corona discharges were formed inside dry air fed single wire-duct reactor under positive dc voltage at the normal atmospheric conditions. The corona current-voltage characteristics curves have been measured in parallel with the ozone concentration generated inside the reactor under different discharge conditions. The corona current-voltage characteristics curves have agreed with a semi empirical equation derived from the previous studies. The experimental results of the ozone concentration generated inside the reactor were formulated in the form of an empirical equationmore » included the different parameters that were studied experimentally. The obtained equations are valid to expect both the current-voltage characteristics curves and the corresponding ozone concentration that generates with the positive dc corona discharges inside single wire-duct electrostatic precipitators under any operating conditions in the same range of the present study.« less

Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, R.K.; Iskander, S.K.; Sokolov, M.A.

1997-02-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or {open_quotes}recovery,{close_quotes} of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices ofmore » Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed.« less

Combined Ceria Reduction and Methane Reforming in a Solar-Driven Particle-Transport Reactor.

PubMed

Welte, Michael; Warren, Kent; Scheffe, Jonathan R; Steinfeld, Aldo

2017-09-20

We report on the experimental performance of a solar aerosol reactor for carrying out the combined thermochemical reduction of CeO 2 and reforming of CH 4 using concentrated radiation as the source of process heat. The 2 kW th solar reactor prototype utilizes a cavity receiver enclosing a vertical Al 2 O 3 tube which contains a downward gravity-driven particle flow of ceria particles, either co-current or counter-current to a CH 4 flow. Experimentation under a peak radiative flux of 2264 suns yielded methane conversions up to 89% at 1300 °C for residence times under 1 s. The maximum extent of ceria reduction, given by the nonstoichiometry δ (CeO 2-δ ), was 0.25. The solar-to-fuel energy conversion efficiency reached 12%. The syngas produced had a H 2 :CO molar ratio of 2, and its calorific value was solar-upgraded by 24% over that of the CH 4 reformed.

Combined Ceria Reduction and Methane Reforming in a Solar-Driven Particle-Transport Reactor

PubMed Central

2017-01-01

We report on the experimental performance of a solar aerosol reactor for carrying out the combined thermochemical reduction of CeO2 and reforming of CH4 using concentrated radiation as the source of process heat. The 2 kWth solar reactor prototype utilizes a cavity receiver enclosing a vertical Al2O3 tube which contains a downward gravity-driven particle flow of ceria particles, either co-current or counter-current to a CH4 flow. Experimentation under a peak radiative flux of 2264 suns yielded methane conversions up to 89% at 1300 °C for residence times under 1 s. The maximum extent of ceria reduction, given by the nonstoichiometry δ (CeO2−δ), was 0.25. The solar-to-fuel energy conversion efficiency reached 12%. The syngas produced had a H2:CO molar ratio of 2, and its calorific value was solar-upgraded by 24% over that of the CH4 reformed. PMID:28966440

Fluidized-bed reactor modeling for production of silicon by silane pyrolysis

NASA Technical Reports Server (NTRS)

Dudukovic, M. P.; Ramachandran, P. A.; Lai, S.

1986-01-01

An ideal backmixed reactor model (CSTR) and a fluidized bed bubbling reactor model (FBBR) were developed for silane pyrolysis. Silane decomposition is assumed to occur via two pathways: homogeneous decomposition and heterogeneous chemical vapor deposition (CVD). Both models account for homogeneous and heterogeneous silane decomposition, homogeneous nucleation, coagulation and growth by diffusion of fines, scavenging of fines by large particles, elutriation of fines and CVD growth of large seed particles. At present the models do not account for attrition. The preliminary comparison of the model predictions with experimental results shows reasonable agreement. The CSTR model with no adjustable parameter yields a lower bound on fines formed and upper estimate on production rates. The FBBR model overpredicts the formation of fines but could be matched to experimental data by adjusting the unkown jet emulsion exchange efficients. The models clearly indicate that in order to suppress the formation of fines (smoke) good gas-solid contacting in the grid region must be achieved and the formation of the bubbles suppressed.

Experimental Plans for Subsystems of a Shock Wave Driven Gas Core Reactor

NASA Technical Reports Server (NTRS)

Kazeminezhad, F.; Anghai, S.

2008-01-01

This Contractor Report proposes a number of plans for experiments on subsystems of a shock wave driven pulsed magnetic induction gas core reactor (PMI-GCR, or PMD-GCR pulsed magnet driven gas core reactor). Computer models of shock generation and collision in a large-scale PMI-GCR shock tube have been performed. Based upon the simulation results a number of issues arose that can only be addressed adequately by capturing experimental data on high pressure (approx.1 atmosphere or greater) partial plasma shock wave effects in large bore shock tubes ( 10 cm radius). There are three main subsystems that are of immediate interest (for appraisal of the concept viability). These are (1) the shock generation in a high pressure gas using either a plasma thruster or pulsed high magnetic field, (2) collision of MHD or gas dynamic shocks, their interaction time, and collision pile-up region thickness, and (3) magnetic flux compression power generation (not included here).

Experimental validation of an 8 element EMAT phased array probe for longitudinal wave generation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Le Bourdais, Florian, E-mail: florian.lebourdais@cea.fr; Marchand, Benoit, E-mail: florian.lebourdais@cea.fr

2015-03-31

Sodium cooled Fast Reactors (SFR) use liquid sodium as a coolant. Liquid sodium being opaque, optical techniques cannot be applied to reactor vessel inspection. This makes it necessary to develop alternative ways of assessing the state of the structures immersed in the medium. Ultrasonic pressure waves are well suited for inspection tasks in this environment, especially using pulsed electromagnetic acoustic transducers (EMAT) that generate the ultrasound directly in the liquid sodium. The work carried out at CEA LIST is aimed at developing phased array EMAT probes conditioned for reactor use. The present work focuses on the experimental validation of amore » newly manufactured 8 element probe which was designed for beam forming imaging in a liquid sodium environment. A parametric study is carried out to determine the optimal setup of the magnetic assembly used in this probe. First laboratory tests on an aluminium block show that the probe has the required beam steering capabilities.« less

Heat-transfer characteristics of flowing and stationary particle-bed-type fusion-reactor blankets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nietert, R.E.

1983-02-01

The following five appendices are included: (1) physical properties of materials, (2) thermal entrance length Nusselt number variations, (3) stationary particle bed temperature variations, (4) falling bed experimental data and calculations, and (5) stationary bed experimental data and calculations. (MOW)

Prospects of a baryon instability search in neutron-antineutron oscillations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Efremenko, Yu.; Kamyshkov, Yu.

1996-12-31

The purpose of this article is to review the current status and the future prospects for an experimental neutron-antineutron transition search. Traditional and new experimental techniques are discussed here. In the n {r_arrow} {anti n} search in experiments at existing reactors, it would be possible to increase the discovery potential up to four orders of magnitude for vacuum n {r_arrow} {anti n} transitions relative to the existing experimental level or to achieve the limit of {tau}{sub n-{anti n}{sup {approximately}}} 10{sup 10}s.. With dedicated future reactors and an ultimate experimental layout, it might be possible to reach the limit of 10{supmore » 11}s. Significant progress in an intranuclear n {r_arrow} {anti n} transition search expected to be made during the next decade by the SuperKamiokande and Icarus detectors. It can be matched, or even exceeded, in a new alternative approach, where unstable long-lived isotopes of technetium are searched in non radioactive deep-mined ores.« less

Dismantling the nuclear research reactor Thetis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michiels, P.

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storagemore » rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos. This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written). (authors)« less

Broadly Applicable Nanowafer Drug Delivery System for Treating Eye Injuries

DTIC Science & Technology

2016-09-01

Invited seminar: “Ocular Drug Delivery Nanowafer: Design and Applications,” Houston Methodist Research Institute, Houston, TX. Date: 08-08-2016. Books... Designed the experiments, reviewed and analyzed the experimental results. Funding Support Cystinosis Research Foundation, DOD Name Daniela...Annual PREPARED FOR: U.S. Army Medical Research and Materiel Command Fort Detrick, Maryland 21702-5012 DISTRIBUTION STATEMENT: Approved for

An Experimental Brain Missile Wound: Ascertaining Pathophysiology and evaluating Treatments to Lower Mortality and Morbidity

DTIC Science & Technology

1989-04-27

Narayan R, et al: Early insults to the injured brain. JAMA 240:439-442, 1978. 91 Neubauer JA, and Edelman N: Nonuniform brain blood flow response to...Research ( LAIR ), Bldg. 1110 ATI7I: SGRD-ULZ-RC Presidio of San Francisco, CA 94129-6815 1 copy Comander US Army Medical Research and Develop mnt Coand

Using Experimental Design and Data Analysis to Study the Enlisted Specialty Model for the U.S. Army G1

DTIC Science & Technology

2010-06-01

RECOMMENDATIONS FOR FUTURE STUDY ..................43 APPENDIX. OBJECTIVE FUNTION COEFFICIENTS ...................47 LIST OF REFERENCES...experiments are designed so an analyst can conduct simultaneous examination of multiple factors and explore these factors and their relationship to output...or more components. 46 THIS PAGE INTENTIONALLY LEFT BLANK 47 APPENDIX. OBJECTIVE FUNTION COEFFICIENTS

Multi-body Dynamic Contact Analysis Tool for Transmission Design

DTIC Science & Technology

2003-04-01

frequencies were computed in COSMIC NASTRAN, and were validated against the published experimental modal analysis [17]. • Using assumed time domain... modal superposition. • Results from the structural analysis (mode shapes or forced response) were converted into IDEAS universal format (dataset 55...ARMY RESEARCH LABORATORY Multi-body Dynamic Contact Analysis Tool for Transmission Design SBIR Phase II Final Report by

Flow past an axially aligned spinning cylinder: Experimental Study

NASA Astrophysics Data System (ADS)

Carlucci, Pasquale; Buckley, Liam; Mehmedagic, Igbal; Carlucci, Donald; Thangam, Siva

2017-11-01

Experimental investigation of flow past a spinning cylinder is presented in the context of its application and relevance to flow past projectiles. A subsonic wind tunnel is used to perform experiments on the flow past a spinning cylinder that is mounted on a forward sting and oriented such that its axis of rotation is aligned with the mean flow. The experiments cover a Reynolds number of range of up to 45000 and rotation numbers of up to 2 (based on cylinder diameter). Time-averaged mean flow and turbulence profiles in the wake flow are presented with and without spin along with comparison to published experimental data. Funded in part by the U. S. Army ARDEC, Picatinny Arsenal, NJ.

Comparison between coupled KZK-BHTE numerical simulations and scanned HIFU exposures in excised bovine liver

NASA Astrophysics Data System (ADS)

Andrew, Marilee A.; Brayman, Andrew A.; Kaczkowski, Peter J.; Kargl, Steven G.

2004-05-01

The use of moving high intensity focused ultrasound (HIFU) treatment protocols is of interest in achieving efficient formation of large-volume lesions in tissue. However, potentially unwanted thermal effects, such as prefocal heating, should be considered. A KZK acoustic model coupled with the BioHeat Transfer Equation has been extended to simulate multiple, moving scans in tissue. Simulation results are compared with experimental data collected over a range of exposure regimes for linear and concentric circular scans with a 3.5-MHz single-element transducer in ex vivo bovine liver. Of particular interest are investigating prefocal thermal buildup and ablating the central core of a circular pattern through conductive heating, that is without direct HIFU exposure. Qualitative agreement is observed between experimental and simulated data; limits of the predictive capability of the model in cavitation regimes will be discussed. [Support provided by the U.S. Army Medical Research Acquisition Activity through The University of Mississippi under terms of Agreement No. DAMD17-02-2-0014. The opinions expressed herein are those of the author(s) and do not necessarily reflect the views of U.S. Army Medical Research Acquisition Activity or The University of Mississippi.

Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

NASA Astrophysics Data System (ADS)

Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

2018-04-01

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Experimental investigations of a uranium plasma pertinent to a self-sustaining plasma source

NASA Technical Reports Server (NTRS)

Schneider, R. T.

1971-01-01

The research is pertinent to the realization of a self-sustained fissioning plasma for applications such as nuclear propulsion, closed cycle MHD power generation using a plasma core reactor, and heat engines such as the nuclear piston engine, as well as the direct conversion of fission energy into optical radiation (nuclear pumped lasers). Diagnostic measurement methods and experimental devices simulating plasma core reactor conditions are discussed. Studies on the following topics are considered: (1) ballistic piston compressor (U-235); (2) high pressure uranium plasma (natural uranium); (3) sliding spark discharge (natural uranium); (4) fission fragment interaction (He-3 and U-235); and (5) nuclear pumped lasers (He-3 and U-235).

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE Office of Scientific and Technical Information (OSTI.GOV)

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

DOE PAGES

Forsberg, Charles W.; Lam, Stephen; Carpenter, David M.; ...

2017-02-26

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt.more » The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. In this paper, we describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Finally, limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.« less

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stoller, R.E.

1987-12-01

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and alteredmore » mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.« less

Using Additive Manufacturing to Optimize FLiBe Coolant Blanket in Fusion Reactors

NASA Astrophysics Data System (ADS)

Fry, Vincent Michael

Fusion reactors have often been hailed as the holy grail of clean energy generation, though a power-generating reactor has never been built due to a multitude of limiting factors. One such factor is the immense 12-15 MW/m2 heat fluxes experienced by the inner wall of the reactor. Multiple groups have proposed the use of tungsten swirl tubes to withstand the heat generated within the reactor core. The primary focus of this investigation is to parameterize this 'first wall' interior structure to determine the highest achievable heat transfer coefficient given the many tungsten configurations enabled via additive manufacturing. Two general tube structures were considered: an orthogonal three-dimensional mesh of various diameters and spacings, as well as a swirl tube geometry with varying 'tape' thicknesses. The coolant liquid proposed is FLiBe (2LiF-BeF2) due to its high specific heat capacity as well as its ability to breed tritium, the fuel for the reactor. This was accomplished using theoretical calculations; computational fluid dynamics and conjugate heat transfer simulations in ANSYS Workbench; as well as an experimental setup to confirm tube pressure drop along the pipe. It was determined that heat transfer coefficients between upwards of 60,000 W/m 2K were readily achievable, keeping the first wall temperature around 1300 K. A multitude of designs proved to be feasible given the pumping power restrictions, though the suggested design going forward is a swirl tube with 2 mm 'tape' thickness and 3 m/s inlet velocity. Simulated pressure drop with water was accurate to within 30% of experimentally measured values, giving confidence in the credibility of the results.

How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

DOE PAGES

Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; ...

2015-06-18

A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. Themore » particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.« less

Department of the Army - The Fiscal Year 2008 Military Personnel, Army Appropriation and the Antideficiency Act

DTIC Science & Technology

2010-06-22

of the Army, U.S. Army Audit Agency, Budgeting for the Military Personnel, Army Appropriation, Report No. A-2010-0028- FFM (Jan. 6, 2010); Department...of the Army, U.S. Army Audit Agency, Military Personnel, Army FY 05 Subsistence Charges, Report No. A-2008-0037- FFM (Feb. 12, 2008); Department of

Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

PubMed

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

Flow reversal power limit for the HFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Lap Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Uncertainties about the afterheat removal capability during the flow reversal has limited the reactor operating power to 30 MW. An experimental and analytical program to address these uncertainties is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safemore » operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW.« less

Design and fabrication of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolytic oil production in Bangladesh

NASA Astrophysics Data System (ADS)

Aziz, Mohammad Abdul; Al-khulaidi, Rami Ali; Rashid, MM; Islam, M. R.; Rashid, MAN

2017-03-01

In this research, a development and performance test of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolysis oil production was successfully completed. The characteristics of the pyrolysis oil were compared to other experimental results. A solid horizontal condenser, a burner for furnace heating and a reactor shield were designed. Due to the pilot scale pyrolytic oil production encountered numerous problems during the plant’s operation. This fixed-bed batch type pyrolysis reactor method will demonstrate the energy saving concept of solid waste tire by creating energy stability. From this experiment, product yields (wt. %) for liquid or pyrolytic oil were 49%, char 38.3 % and pyrolytic gas 12.7% with an operation running time of 185 minutes.

The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

NASA Astrophysics Data System (ADS)

Scarlat, Raluca O.; Peterson, Per F.

2014-01-01

The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miao, Yinbin; Ye, Bei; Mei, Zhigang

Uranium silicide (U 3Si 2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U 3Si 2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U 3Si 2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuelmore » material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U 3Si 2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U 3Si 2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U 3Si 2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U 3Si 2 fuel as an accident-tolerant alternative for uranium dioxide.« less