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Federal Register / Vol. 86, No. 4 / Thursday, January 7, 2021 / Proposed Rules Issue 1: Calculated Maximum Fuel Element Cladding Temperature Limit Background for Issue 1
Under 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, of 10 CFR, light-water nuclear power reactors fueled with uranium oxide pellets within cylindrical Zircaloy cladding must be provided with an ECCS that must be designed so that its calculated cooling performance following postulated loss of coolant accidents LOCAs 1 conforms to the criteria specified in 50.46b.2 Under 50.46b1, the calculated maximum fuel element cladding temperature shall not exceed 2,200 F. In addition, 50.46b2 through 5, respectively, contain requirements for calculations involving: Maximum cladding oxidation, maximum hydrogen generation, changes in core geometry, and long-term cooling.

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Petitioners Arguments and Requests Related to Issue 1
The petitioner asserted that data from multirod assembly severe fuel damage experiments indicate that the calculated maximum fuel element cladding temperature limit of 2,200 F specified in 50.46b1 is not conservative.
Although not its intended purpose, the NRC previously determined that this limit provides a conservative safety margin from an area of Zircaloy cladding oxidation behavior known as the autocatalytic regime. An autocatalytic condition occurs when the heat released by the metal-water reaction of zirconium with steam is greater than the heat that can be transferred away from the Zircaloy cladding. This causes the Zircaloy cladding temperature to rise, thereby increasing the diffusion of oxygen into the metal, which in turn raises the rate at which the zirconium-steam oxidation reaction occurs. As the metal-water reaction rate continues to increase, the temperature of the Zircaloy cladding continues to rise, eventually resulting in an uncontrolled reaction and 1 Under 50.46c, LOCAs are hypothetical accidents that would result from the loss of reactor coolant, at a rate that exceeds the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary.
2 Criterion 35 of appendix A to 10 CFR part 50, General Design Criteria for Nuclear Power Plants, further requires that a system to provide abundant emergency core cooling shall be provided and that the system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that: 1 Fuel and cladding damage that could interfere with continued effective core cooling is prevented and 2 the cladding metal-water reaction is limited to negligible amounts.

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temperature excursion. The petitioner asserted that data from cited experiments indicate that such autocatalytic metal-water oxidation reactions and uncontrolled temperature excursions involving Zircaloy cladding have occurred at temperatures below 2,200 F. The petitioner provided this assertion as evidence that the 2,200 F
limit is not conservative, and requested that the NRC amend 50.46 to require that the calculated maximum fuel element cladding temperature not exceed a limit based on data from cited experiments, instead of the 2,200 F
limit specified in 50.46b1.
Issue 2: Metal-Water Reaction Rate Equations for ECCS Evaluation Models Background for Issue 2
To evaluate conformance with the criteria specified in 50.46b, ECCS
cooling performance must be calculated using an acceptable evaluation model 3
for a range of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are evaluated. On September 16, 1988, the NRC amended the requirements of 50.46 and appendix K, ECCS Evaluation Models, to 10 CFR part 50 to reflect an improved understanding of ECCS
performance during reactor transients that was obtained through extensive research performed after promulgation of the original requirements 53 FR
35996. Under 50.46a1, licensees or applicants may use one of two acceptable ECCS evaluation model options: 1 A best-estimate or realistic evaluation model 4 or 2 a conservative evaluation model. Each ECCS
evaluation model option is summarized below.
Option 1: Best-Estimate or Realistic ECCS Evaluation Model Section 50.46a1i of 10 CFR
specifies that a best-estimate evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a LOCA. Comparisons to 3 Regulatory Guide RG 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, issued May 1989, states that the term evaluation model refers to a nuclear plant system computer code or any other analysis tool designed to predict the aggregate behavior of a reactor during a loss of coolant accident. It can be either best-estimate or conservative and may contain many correlations or models.
4 RG 1.157 states that the terms best-estimate and realistic have the same meaning. Both terms are used to indicate that the techniques attempt to predict realistic reactor system thermal-hydraulic response.

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applicable experimental data must be made and uncertainties must be identified and assessed so that the uncertainty in the calculated results can be estimated to 1 account for the uncertainty in comparing the calculated ECCS cooling performance to the criteria specified in 50.46b; and 2
assure that there is a high probability of not exceeding these criteria.
RG 1.157 describes models,5
correlations,6 data, model evaluation procedures, and methods that are acceptable to the NRC staff for meeting the requirements for: 1 A realistic or best-estimate calculation of ECCS
cooling performance during a LOCA; 2
estimating the uncertainty in that calculation; and 3 including uncertainty in the comparisons of the calculated results to the criteria of 50.46b to assure a high probability that the criteria would not be exceeded.
Other models, data, model evaluation procedures, and methods can be considered if they are supported by appropriate experimental data and technical justification.
To be considered acceptable under RG
1.157, evaluation models should account for identified sources of heat including the metal-water reaction ratein performing best-estimate calculations. In particular, the rates of energy release, hydrogen generation, and Zircaloy cladding oxidation from the metal-water reaction of zirconium with steam should be calculated in a best-estimate manner using one of two procedures, depending on the cladding temperature:
1 If the cladding temperature is less than or equal to 1,900 F, correlations to be used to calculate metal-water reaction rates should: a Be checked against a set of relevant data and b recognize the effects of steam pressure, pre-oxidation of the cladding, deformation during oxidation, and internal oxidation from both steam and uranium oxide fuel.
2 If the cladding temperature is greater than 1,900 F, the Cathcart-Pawel equation and the underlying empirical data used to derive it are considered acceptable for calculating the rates of energy release, hydrogen generation, and cladding oxidation.
5 RG 1.157 states that the term model refers to a set of equations derived from fundamental physical laws that is designed to predict the details of a specific phenomenon.
6 RG 1.157 states that the term correlation refers to an equation having empirically determined constants such that it can predict some details of a specific phenomenon for a limited range of conditions.

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