Nothing Special   »   [go: up one dir, main page]

Henager, 1966 - Google Patents

Design criteria for phase II dummy hydraulic core for containment systems experiment reactor simulator

Henager, 1966

View PDF
Document ID
3963701127032350591
Author
Henager C
Publication year

External Links

Snippet

The design of a Phase II dummy hydraulic core for the CSE reactor simulator is described. The core will be a scale model having features typical of PWR or BWR reactor cores, and will include control rods and associated support mechanisms, fuel elements, a thermal …
Continue reading at www.osti.gov (PDF) (other versions)

Classifications

    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GASES [GHG] EMISSION, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors
    • Y02E30/40—Other aspects relating to nuclear fission
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/10—Structural combination of fuel element, control rod, reactor core, or moderator structure with sensitive instruments, e.g. for measuring radioactivity, strain
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/02—Devices or arrangements for monitoring coolant or moderator
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/003—Remote inspection of vessels, e.g. pressure vessels
    • G21C17/013—Inspection vehicles
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
    • G21C3/02—Fuel elements
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
    • G21C15/24—Promoting flow of the coolant
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C11/00—Shielding structurally associated with the reactor
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C1/00—Reactors
    • G21C1/04—Thermal reactors; Epithermal reactors
    • G21C1/06—Heterogeneous reactors, i.e. in which fuel and moderator are separated
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C13/00—Pressure vessels; Containment vessels; Containment in general
    • G21C13/02—Details
    • G21C13/032—Joints between tubes and vessel walls, e.g. taking into account thermal stresses
    • G21C13/036—Joints between tubes and vessel walls, e.g. taking into account thermal stresses the tube passing through the vessel wall, i.e. continuing on both sides of the wall

Similar Documents

Publication Publication Date Title
CN101939795B (en) Method of determining the value of a parameter representative of the operability of a nuclear reactor, determining system, computer program and corresponding medium
Hetsroni Use of hydraulic models in nuclear reactor design
Henager Design criteria for phase II dummy hydraulic core for containment systems experiment reactor simulator
Ouma et al. Flow measurements in rod bundle subchannels with varying rod-wall proximity
Feldman et al. Semiscale Mod-1 program and system description for the blowdown heat transfer tests (Test series 2)
Olson et al. The Fort St. Vrain high temperature gas-cooled reactor: X. Core temperature fluctuations
Allemann et al. EXPERIMENTAL HIGH ENTHALPY WATER BLOWDOWN FROM A SIMPLE VESSEL THROUGH A BOTTOM OUTLET.
Dubourg et al. Model experimentation and analysis of flow-induced vibrations of PWR internals
Russcher et al. LOCA simulation in the NRU reactor: materials test-1
Wittenbrock et al. Plutonium Recycle Test Reactor Final Safeguards Analysis
Collins et al. Experiment data report for Semiscale Mod-1 tests S-05-2A and S-05-2B (alternate ECC injection tests)
Burnette et al. Blowdown Heat Transfer Program. Task C-1 informal report, preliminary system design description of two-loop test apparatus
Hambric et al. Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures
Müller et al. Liquid metal cooled fast breeder reactor development and its mechanical-structural requirements and problems
Waddell Jr Measurement of light-water reactor coolant channel reduction arising from cladding deformation during a loss-of-coolant accident
Crapo et al. Experiment data report for Semiscale Mod-1 Test S-29-1 (integral test with asymmetrical break).[PWR]
DOE Project Description ORNL PWR Blowdown Heat Transfer Separate-Effects Program—Thermal-Hydraulic Test Facility (THTF)
Smith et al. PM-1 NUCLEAR POWER PLANT PROGRAM 3RD QUARTERLY PROGRESS REPORT FOR SEPTEMBER 1 TO NOVEMBER 30, 1959
Haire et al. Semiscale Blowdown and Emergency Core Cooling (ECC) Project
de Leija Measurement and qualification of the local heat transfer and pressure losses in high turbulent gas flows in channels with different artificial surface structures
Curry SEMISCALE PROJECT TEST DATA REPORT: TEST 851.
Crapo et al. Experiment data report for Semiscale Mod-1 Test S-02-4 (blowdown heat transfer test)
MHMH BLANK PAGE
Allemann et al. COOLANT BLOWDOWN STUDIES OF A REACTOR SIMULATOR VESSEL CONTAINING A PERFORATED SIEVE PLATE SEPARATOR.
Kiss et al. 4: 4 US Døe-Ge-Anl Flow Induced Vibration Technology Program.