(Proposed
New Regulatory Guide)
CONDUCTING
PERIODIC TESTING FOR BREAKAWAY
OXIDATION
BEHAVIOR
Background
In 1996, the NRC initiated a fuel-cladding research program intended to investigate the behavior of high-exposure fuel cladding under accident conditions. This research program included an extensive LOCA research and testing program at Argonne National Laboratory (ANL), as well as jointly funded programs at the Kurchatov Institute (Ref. 2) and the Halden Reactor Project (Ref. 3), to develop the body of technical information needed to evaluate LOCA regulations for high-exposure fuel.
Note: ACRS reviewed the NRC fuel-cladding research program during 2002, following are Blogger Leyse's notes from then.
Member Graham B. Wallis was especially enraged by the limited approaches to fuel integrity under LOCA conditions. In response to detailed descriptions of fracture of corroded specimens of cladding from irradiated power reactor fuel he asserted: “It seems to be that both these coursing tests and hitting tests, impact tests and the squeezing tests are not really typical of the loads imposed on the real cladding.. I keep wondering what the relevance of all these tests are to the real truth.” He also reacted to the discussions of runaway, “I think when you come back and talk about run-away to this committee you better have a criterion for run-away and not this sort of vagueness about heat transfer.”
Returning to the excerpts from DG-1261
Establishing the Onset of Breakaway Oxidation
The experimental procedure provided in Appendix A to this regulatory guide defines a procedure acceptable to the NRC staff to measure the onset of breakaway oxidation (Appendices B through E are provided to expand on critical aspects of the testing procedure in Appendix A). This experimental procedure may be used to characterize the onset of breakaway oxidation as a function of temperature for a zirconium cladding alloy.
A-7.2 Steamflow Rate
The average steamflow rate used in breakaway oxidation studies should be determined (and reported) from the mass of condensed water collected during these long-time tests or by the mass of water
that is input to the steam chamber divided by the test time and normalized to the net cross-sectional area of the steam chamber. The average steamflow rate should be in the range of 0.5 to 30 mg/square
centimeter per second (cm2 · s).
A-7.3 Steam Pressure
Breakaway oxidation tests should be conducted at a steam pressure at or slightly above atmospheric pressure.
At this point Blogger Leyse again interrupts the excerpts to convert 0.5 to 30 mg/square centimeter per second to meaningful units, like feet per second of steam flow.
1. First, convert 1 mg of steam at 100 degrees Centigrade and one atmosphere to cubic centimeters of steam:
So (0.001/18) x (373/273) x 22,400 = 1.7 cubic centimeters
2. Then 1 mg/square centimeter per second becomes 1.7 centimeters per second or 1.7/2.54 = 0.67 inches/second
3. So, an average steamflow rate that is in the range of 0.5 to 30 mg/square centimeter per second (cm2 · s) is 0.33 to 20.1 inches per second, or 0.03 to 1.67 feet per second.
The steaming conditions of this regulatory guide are thus far removed of the realities of the mixed thermal hydraulic conditions of severe LOCAs. The NRC uses the data in "improving" its series of codes such as TRACE. In this way, licensing of nuclear power plants proceeds including life extensions and power level increases.
The nuclear power industry and its lobbyists do not mind arguing about various aspects of proposed 50.46c as well as the assortment of proposed regulatory guides. In the meantime, aspects including Baker-Just, Cathcart-Powell, and 2200 degrees Fahrenheit are blessed in the proposed 50.46c, and the assertions of PRM-50-93 (ML093290250) are not addressed.
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Background
In 1996, the NRC initiated a fuel-cladding research program intended to investigate the behavior of high-exposure fuel cladding under accident conditions. This research program included an extensive LOCA research and testing program at Argonne National Laboratory (ANL), as well as jointly funded programs at the Kurchatov Institute (Ref. 2) and the Halden Reactor Project (Ref. 3), to develop the body of technical information needed to evaluate LOCA regulations for high-exposure fuel.
Note: ACRS reviewed the NRC fuel-cladding research program during 2002, following are Blogger Leyse's notes from then.
Runaway
Discussions at the ACRS (2002).
The USNRC is currently working on revisions
to rule 10 CFR 50.46 concerning emergency core cooling systems for
reactors. The process is called
risk-informing the regulation.
The ACRS discussions of Friday,
May 31, 2002, are revealing in that several aspects
of the revisions were discussed, however, the ubiquitous fouling of today’s
LWRs was not considered. This was a
combined meeting of three of the most influential subcommittees of the ACRS: Materials
and Metallurgy; Thermal Hydraulic Phenomena & Reliability and Probabilistic
Risk Assessment
Member Graham B. Wallis was especially enraged by the limited approaches to fuel integrity under LOCA conditions. In response to detailed descriptions of fracture of corroded specimens of cladding from irradiated power reactor fuel he asserted: “It seems to be that both these coursing tests and hitting tests, impact tests and the squeezing tests are not really typical of the loads imposed on the real cladding.. I keep wondering what the relevance of all these tests are to the real truth.” He also reacted to the discussions of runaway, “I think when you come back and talk about run-away to this committee you better have a criterion for run-away and not this sort of vagueness about heat transfer.”
Returning to the excerpts from DG-1261
Establishing the Onset of Breakaway Oxidation
The experimental procedure provided in Appendix A to this regulatory guide defines a procedure acceptable to the NRC staff to measure the onset of breakaway oxidation (Appendices B through E are provided to expand on critical aspects of the testing procedure in Appendix A). This experimental procedure may be used to characterize the onset of breakaway oxidation as a function of temperature for a zirconium cladding alloy.
A-7.2 Steamflow Rate
The average steamflow rate used in breakaway oxidation studies should be determined (and reported) from the mass of condensed water collected during these long-time tests or by the mass of water
that is input to the steam chamber divided by the test time and normalized to the net cross-sectional area of the steam chamber. The average steamflow rate should be in the range of 0.5 to 30 mg/square
centimeter per second (cm2 · s).
A-7.3 Steam Pressure
Breakaway oxidation tests should be conducted at a steam pressure at or slightly above atmospheric pressure.
At this point Blogger Leyse again interrupts the excerpts to convert 0.5 to 30 mg/square centimeter per second to meaningful units, like feet per second of steam flow.
1. First, convert 1 mg of steam at 100 degrees Centigrade and one atmosphere to cubic centimeters of steam:
So (0.001/18) x (373/273) x 22,400 = 1.7 cubic centimeters
2. Then 1 mg/square centimeter per second becomes 1.7 centimeters per second or 1.7/2.54 = 0.67 inches/second
3. So, an average steamflow rate that is in the range of 0.5 to 30 mg/square centimeter per second (cm2 · s) is 0.33 to 20.1 inches per second, or 0.03 to 1.67 feet per second.
The steaming conditions of this regulatory guide are thus far removed of the realities of the mixed thermal hydraulic conditions of severe LOCAs. The NRC uses the data in "improving" its series of codes such as TRACE. In this way, licensing of nuclear power plants proceeds including life extensions and power level increases.
The nuclear power industry and its lobbyists do not mind arguing about various aspects of proposed 50.46c as well as the assortment of proposed regulatory guides. In the meantime, aspects including Baker-Just, Cathcart-Powell, and 2200 degrees Fahrenheit are blessed in the proposed 50.46c, and the assertions of PRM-50-93 (ML093290250) are not addressed.
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Figure 1. E110 cladding test specimen
Source: NUREG/CR-6967
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