PARAMETRIC REACTIVITY TRANSIENT ANALYSES FOR THE FFTF NUCLEAR PROOF TEST REACTOR

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Fault tree techniques have been used to identify possible failure paths within the NPTR which could lead to core disassembly. The analysis o f the various faults has led to formulation of design requirements, protective system requirements, and administrative restraints required to prevent accidents from these faults. Transient analyses were performed using the heat transfer-nuclear kinetics codes, Nutiger-II, FORE-II, and MELT-II . To verify results, intercomparison studies were made between the codes. The codes were i n good general agreement. Each code was found to exhibit different advantages and disadvantage. Inherent reactivity feedback effects were assessed in the analysis. With ... continued below

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Schade, A. R. January 1, 1970.

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Fault tree techniques have been used to identify possible failure paths within the NPTR which could lead to core disassembly. The analysis o f the various faults has led to formulation of design requirements, protective system requirements, and administrative restraints required to prevent accidents from these faults. Transient analyses were performed using the heat transfer-nuclear kinetics codes, Nutiger-II, FORE-II, and MELT-II . To verify results, intercomparison studies were made between the codes. The codes were i n good general agreement. Each code was found to exhibit different advantages and disadvantage. Inherent reactivity feedback effects were assessed in the analysis. With the assumed core parameters, there appears to be sufficient Doppler to prolong a nuclear transient to allow protective action to prevent fuel from melting. The use of average values of the feedback coefficients smeared over the entire core does not appear to be an acceptable method with spacially dependent temperatures. In the thermal analysis, the fuel pin gap coefficient and sodium film coefficient do not appear to be highly sensitive parameters for transient analysis. Power transients resulting from reactivity insertions of from 2$/sec to 20$/sec have been examined in detail. Sodium will be molten before fuel melting occurs for accidents within this range. For the smaller ramp rates (< 4$/sec), sodium nay even reach vaporization temperatures before any fuel melts. Power transients terminated by effective protective action were investigated. It i s believed p o s s i b l e t o design a scram system, with the . present state of the art, to prevent sodium from melting for a reactivity ramp up to at least 6$/sec. This same system would prevent fuel melting for a reactivity ramp up to 15$/sec. Sodium thermal expansion will play a very important role in a core disassembly. When the average sodium temperature exceeds 250 {degrees}F, physical core distortion must result to relieve expansion pressures. Rupturing of the fuel assembly cans during a transient increases the probability of a sodium fire. Pressures and temperatures from a sodium fire could easily exceed 20 psig and 1000 {degrees}F. The design basis accident has not been identified. However, the lower limit is a sodium fire involving hot liquid sodium and possible sodium vapor. A fuel vapor explosion would require a large initiating reactivity ramp rate (> 20$/sec) with at least 3$ total reactivity worth. No mechanism for introduction of a reactivity insertion of this characteristic has been identified other than a dropped fuel assembly into a vacant core position. This mechanism is discounted as it is believed that sub-criticality of the reactor can be guaranteed during refueling. It is conceivable that a minor accident could be aggravated into an explosive accident by failure of protective system and positive feedback mechanisms. The possibility of this occurring is dependent upon what effects the confined sodium has on the core. It is desirable that the sodium would take the core to a disassembly condition or termination mode. Additional analysis will be necessary before this can be guaranteed.

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  • Related Information: FFTF DIVISION- DRAFT FORM DOCUMENT

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  • Report No.: BNWL-1111-FF1
  • Grant Number: AT(45-1)1830
  • DOI: 10.2172/1089652 | External Link
  • Office of Scientific & Technical Information Report Number: 1089652
  • Archival Resource Key: ark:/67531/metadc841295

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  • January 1, 1970

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  • May 19, 2016, 9:45 a.m.

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  • June 22, 2016, 1:18 p.m.

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Schade, A. R. PARAMETRIC REACTIVITY TRANSIENT ANALYSES FOR THE FFTF NUCLEAR PROOF TEST REACTOR, report, January 1, 1970; Richland, Washington. (https://digital.library.unt.edu/ark:/67531/metadc841295/: accessed May 23, 2019), University of North Texas Libraries, Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.