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Solbrig, Charles W.

Publications and source records attributed to Solbrig, Charles W..

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Survivability of TREAT Zircaloy 3 Cladding

This chapter analyzes the effects of oxidation, growth, thermal stress, and fatigue on the Zircaloy-3 alloy used in TREAT fuel assembly cladding. The cladding surrounding the fuel is a protective shell which is used to prevent damage to the enclosed fuel. Therefore, its integrity is importance to guarantee this protection. All of the major factors which can affect the Zircaloy-3 cladding are considered in this chapter and investigated. Limits to operation are determined. The oxidation of Zircaloy-3 in air is of interest to air cooled reactors. The temperature range of interest is from room temperature where the fuel is stored for long periods of time, through the temperature range encountered in normal operation, to the highest temperatures which are possible in extreme accident situations. This temperature range is considered in this chapter to be from room temperature to 1100 C. Growth of the cladding can occur due to the buildup of the oxide layer which is less dense than the original metal and causes the metal to stretch. Thermal stress of the cladding can cause rupture of the cladding and therefore loss of protection to the fuel. Fatigue of the metal cladding occurs due to fuel handling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Thermal Stress in HFEF Hot Cell Windows Due to an In-Cell Metal Fire

This work investigates an accident during the pyrochemical extraction of Uranium and Plutonium from PWR spent fuel in an argon atmosphere hot cell. In the accident, the heavy metals (U and Pu) being extracted are accidentally exposed to air from a leaky instrument penetration which goes through the cell walls. The extracted pin size pieces of U and Pu metal readily burn when exposed to air. Technicians perform the electrochemical extraction using manipulators through a 4 foot thick hot cell concrete wall which protects them from the radioactivity of the spent fuel. Four foot thick windows placed in the wall allow the technicians to visually control the manipulators. These windows would be exposed to the heat of the metal fire. As a result, this analysis determines if the thermal stress caused by the fire would crack the windows and if the heat would degrade the window seals allowing radioactivity to escape from the cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Thermal Analysis of ZPPR High Pu Content Stored Fuel

The Zero Power Physics Reactor (ZPPR) operated from April 18, 1969, until 1990. ZPPR operated at low power for testing nuclear reactor designs. This paper examines the temperature of Pu content ZPPR fuel while it is in storage. Heat is generated in the fuel due to Pu and Am decay and is a concern for possible cladding damage. Damage to the cladding could lead to fuel hydriding and oxidizing. A series of computer simulations were made to determine the range of temperatures potentially occuring in the ZPPR fuel. The maximum calculated fuel temperature is 292°C (558°F). Conservative assumptions in the model intentionally overestimate temperatures. The stored fuel temperatures are dependent on the distribution of fuel in the surrounding storage compartments, the heat generation rate of the fuel, and the orientation of fuel. Direct fuel temperatures could not be measured but storage bin doors, storage sleeve doors, and storage canister temperatures were measured. Comparison of these three temperatures to the calculations indicates that the temperatures calculated with conservative assumptions are, as expected, higher than the actual temperatures. The maximum calculated fuel temperature with the most conservative assumptions is significantly below the fuel failure criterion of 600°C (1,112°F).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

EARTHQUAKE CAUSED RELEASES FROM A NUCLEAR FUEL CYCLE FACILITY

The fuel cycle facility (FCF) at the Idaho National Laboratory is a nuclear facility which must be licensed in order to operate. A safety analysis is required for a license. This paper describes the analysis of the Design Basis Accident for this facility. This analysis involves a model of the transient behavior of the FCF inert atmosphere hot cell following an earthquake initiated breach of pipes passing through the cell boundary. The hot cell is used to process spent metallic nuclear fuel. Such breaches allow the introduction of air and subsequent burning of pyrophoric metals. The model predicts the pressure, temperature, volumetric releases, cell heat transfer, metal fuel combustion, heat generation rates, radiological releases and other quantities. The results show that releases from the cell are minimal and satisfactory for safety. This analysis method should be useful in other facilities that have potential for damage from an earthquake and could eliminate the need to back fit facilities with earthquake proof boundaries or lessen the cost of new facilities.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Stress and Diffusion in Stored Pu ZPPR Fuel from Alpha Generation

ZPPR (Zero Power Physics Reactor) is a research reactor that has been used to investigate breeder reactor fuel designs. The reactor has been dismantled but its fuel is still stored there. Of concern are its plutonium containing metal fuel elements which are enclosed in stainless steel cladding with gas space filled with helium–argon gas and welded air tight. The fuel elements which are 5.08 cm by 0.508 cm up to 20.32 cm long (2 in × 0.2 in × 8 in) were manufactured in 1968. A few of these fuel elements have failed releasing contamination raising concern about the general state of the large number of other fuel elements. Inspection of the large number of fuel elements could lead to contamination release so analytical studies have been conducted to estimate the probability of failed fuel elements. This paper investigates the possible fuel failures due to generation of helium in the metal fuel from the decay of Pu and its possible damage to the fuel cladding from metal fuel expansion or from diffusion of helium into the fuel gas space. This paper (1) calculates the initial gas loading in a fuel element and its internal free volume after it has been brought into the atmosphere at ZPPR, (2) shows that the amount of helium generated by decay of Pu over 46 years since manufacture is significantly greater than this initial loading, (3) determines the amount of fuel swelling if the helium stays fixed in the fuel plate and estimates the amount of helium which diffuses out of the fuel plate into the fuel plenum assuming the helium does not remain fixed in the fuel plate but can diffuse to the plenum and possibly through the cladding. Since the literature is not clear as to which possibility occurs, as with Schroedinger’s cat, both possibilities are analyzed. The paper concludes that (1) if the gas generated is fixed in the fuel, then the fuel swelling it can cause would not cause any fuel failure and (2) if the helium does diffuse out of the fuel (in accordance diffusivities estimated from the literature), then it is unlikely that fuel element bulging will occur.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

ZPPR FUEL ELEMENT THERMAL STRESS-STRAIN ANALYSIS

The design temperature of high plutonium concentration ZPPR fuel assemblies is 600 degrees C. Cladding integrity of the 304L stainless steel cladding is a significant concern with this fuel since even small holes can lead to substantial fuel degradation. Since the fuel has a higher coefficient of thermal expansion than the cladding, an investigation of the stress induced in the cladding due to the differential thermal expansion of fuel and cladding up to the design temperature was conducted. Small holes in the cladding envelope would be expected to lead to the fuel hydriding and oxidizing into a powder over a long period of time. This is the same type of chemical reaction chain that exists in the degradion of the high uranium concentration ZPPR fuel. Unfortunately, the uranium fuel was designed with vents which allowed this degradation to occur. The Pu cladding is sealed so only fuel with damaged cladding would be subject to this damage. The thermal stresses that can be developed in the fuel cladding have been calculated in in this paper and compared to the ultimate tensile stress of the cladding. The conclusion is drawn that thermal stresses cannot induce holes in the cladding even for the highest storage temperatures predicted in calculations (292°C). In fact, thermal stress can not cause cladding failure as long as the fuel temperatures are below the design limit of 600 degrees C (1,112 degrees F).

99 GENERAL AND MISCELLANEOUS↗