Primary pipe rupture accident analysis for the Clinch River Breeder Reactor [electronic resource]
In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function...
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Online Access |
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| Corporate Author: | |
| Format: | Government Document Electronic eBook |
| Language: | English |
| Published: |
Upton, N.Y. : Oak Ridge, Tenn. :
Brookhaven National Laboratory ; distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy,
1976.
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| Subjects: |
MARC
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| 245 | 0 | 0 | |a Primary pipe rupture accident analysis for the Clinch River Breeder Reactor |h [electronic resource] |
| 260 | |a Upton, N.Y. : |b Brookhaven National Laboratory ; |a Oak Ridge, Tenn. : |b distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy, |c 1976. | ||
| 300 | |a Pages: 174 : |b digital, PDF file. | ||
| 336 | |a text |b txt |2 rdacontent. | ||
| 337 | |a computer |b c |2 rdamedia. | ||
| 338 | |a online resource |b cr |2 rdacarrier. | ||
| 500 | |a Published through SciTech Connect. | ||
| 500 | |a 04/01/1976. | ||
| 500 | |a "bnl-nureg-21656" | ||
| 500 | |a Bari, R.A.; Albright, D.C. | ||
| 520 | 3 | |a In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident. | |
| 536 | |b E(30-1)-16. | ||
| 650 | 7 | |a Breeder Reactors. |2 local. | |
| 650 | 7 | |a Epithermal Reactors. |2 local. | |
| 650 | 7 | |a Primary Coolant Circuits. |2 local. | |
| 650 | 7 | |a Reactor Shutdown. |2 local. | |
| 650 | 7 | |a Reactor Protection Systems. |2 local. | |
| 650 | 7 | |a Cooling Systems. |2 local. | |
| 650 | 7 | |a Reactors. |2 local. | |
| 650 | 7 | |a Liquid Metal Cooled Reactors. |2 local. | |
| 650 | 7 | |a Sodium Cooled Reactors. |2 local. | |
| 650 | 7 | |a Power Reactors. |2 local. | |
| 650 | 7 | |a Fbr Type Reactors. |2 local. | |
| 650 | 7 | |a Fast Reactors. |2 local. | |
| 650 | 7 | |a Clinch River Breeder Reactor. |2 local. | |
| 650 | 7 | |a Pipes. |2 local. | |
| 650 | 7 | |a Failures. |2 local. | |
| 650 | 7 | |a Shutdowns. |2 local. | |
| 650 | 7 | |a Loss Of Coolant. |2 local. | |
| 650 | 7 | |a Scram. |2 local. | |
| 650 | 7 | |a Reactor Components. |2 local. | |
| 650 | 7 | |a Ruptures. |2 local. | |
| 650 | 7 | |a Reactor Cooling Systems. |2 local. | |
| 650 | 7 | |a Accidents. |2 local. | |
| 650 | 7 | |a Reactor Accidents. |2 local. | |
| 650 | 7 | |a General Studies Of Nuclear Reactors. |2 edbsc. | |
| 650 | 7 | |a Specific Nuclear Reactors And Associated Plants. |2 edbsc. | |
| 710 | 2 | |a Brookhaven National Laboratory. |4 res. | |
| 710 | 1 | |a United States. |b Department of Energy. |b Office of Scientific and Technical Information. |4 dst. | |
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