Read Online or Download Ageing of Nucl. Powerplant Compnts. - BWR Pressure Vessels (IAEA TECDOC-1470) PDF
Similar nonfiction_6 books
Observe this robust message and positioned the satan and all his assaults the place they belong, lower than your ft! faucet into the supernatural strength of religion and find out how to preserve Your Foot at the Devil's Neck!
- Jazz Piano for the Young Pianist: Exercises, Minuets, & Pieces #1
- Minorities in Greece: Aspects of a Plural Society
- Functional equations
- Pilots Handbook for Navy Model AF-2S Aircraft
Additional resources for Ageing of Nucl. Powerplant Compnts. - BWR Pressure Vessels (IAEA TECDOC-1470)
Due to its secondary and self-relieving nature, no safety factor is given for KIT. 5 during system hydrostatic testing. Because the pressure-temperature (°F) relationship of a BWR is controlled by the steam properties, brittle fracture concerns are limited to determining the test temperature. The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan. Appendix G to the ASME Codes specifies that for calculating the allowable limit curves for various heat-up and cool-down rates, the total stress intensity factor, KI, for the combined thermal and pressure stresses at any time during heat-up or cool-down cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time.
Intergranular and irradiation assisted stress corrosion cracking 4. General corrosion 5. Erosion corrosion. g. erosion corrosion), any such synergistic effect has been explicitly evaluated. The technical evaluation of a particular age related degradation mechanism and its effects on the continued safety performance of a particular BWR RPV component leads to one of two conclusions: (1) the degradation mechanism effects are potentially significant to that component and further evaluation is required relative to the capability of programs to effectively manage these effects; or (2) the age related degradation effects are not significant to the ability of that component to perform its intended safety function throughout the plant life.
Likewise, a thermal stress can often be allowed to reach a higher value than one which is produced by dead weight or pressure. Therefore, a new set of design criteria were developed which shifted the emphasis away from the use of standard configurations and toward the detailed analyses of stresses. The setting of allowable stress values required dividing stresses into categories and assigning different allowable values to different groups of categories. The failure theory used here is the maximum shear stress theory, which has been found appropriate to reactor vessel applications and has the advantage of simplicity.
Ageing of Nucl. Powerplant Compnts. - BWR Pressure Vessels (IAEA TECDOC-1470)