10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded

From kanterella
Jump to navigation Jump to search

Seriously Degraded

A degraded condition. “Any event or condition that results in:(A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or"

An LER is required for a seriously degraded principal safety barrier or an unanalyzed condition that significantly degrades plant safety. If not reported under 10 CFR 50.72(a), (b)(1), or (b)(2), an ENS notification is required under 10 CFR 50.72(b)(3) (an 8-hour report). On occasion, a “Degraded or Unanalyzed Condition” is discovered to have occurred in the past, but is not ongoing at the time of discovery. ENS notifications and LERs are required if a Degraded or Unanalyzed Condition occurred within 3 years of the date of discovery, even if the event is not on-going at the time of discovery.

Combined Criteria: 10 CFR 50.72(b)(3)(ii), Degraded or Unanalyzed Condition


(A) Nuclear Power Plant, Including Its Principal Safety Barriers, Being Seriously Degraded

This criterion applies to material (e.g., metallurgical or chemical) problems that cause abnormal degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, reactor coolant system (RCS) pressure boundary, or the containment). Abnormal degradation of a barrier may be indicated by the necessity of taking corrective action to restore the barrier’s capability, as is the case in some of the examples discussed below. Abnormal stress upon a barrier may result from an unplanned transient, as is the case in one of the examples discussed below. The following are examples of reportable events and conditions:

(1) Fuel cladding failures in the reactor, or in the storage pool, that exceed expected values, or that are unique or widespread, or that are caused by unexpected factors.
(2) Welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600, “Analytical Evaluation of Flaws,” or ASME Section XI, Table IWB-3410-1, “Acceptance Standards.”
(3) Serious steam generator tube degradation. A licensee’s plant-specific TS contain performance criteria for steam generator tube integrity, which includes structural integrity, accident induced leakage, and operational leakage. Steam generator tube degradation is considered serious only if either the steam generator structural integrity or accident-induced leakage performance criteria are not met.
In addition, one or more steam generator tubes satisfying the tube repair criteria and not plugged or repaired in accordance with the steam generator program is not considered to be serious steam generator tube degradation and therefore is not reportable as a “Degraded or Unanalyzed Condition,” as long as the structural integrity and accident-induced leakage performance criteria are both met.
(4) Low temperature over pressure transients in which the pressure-temperature relationship violates pressure-temperature limits derived from Appendix G, “Fracture Toughness Requirements,” to 10 CFR Part 50 (e.g., TS pressure-temperature curves).
(5) Loss of containment function or integrity, including containment leak rate tests in which the total containment as-found, minimum-pathway leak rate exceeds limits in the facility’s TS.[1]

(B) Unanalyzed Condition

(B) The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.


(1) Failures of Reactor Fuel Rod Cladding Identified during Testing of Fuel Assemblies

Radiochemistry data for a particular PWR indicated that a number of fuel rods had failed during the first few months of operation. Projections ranged from 6 to 12 failed rods. The end-of-cycle RCS iodine-131 activity averaged 0.025 microcuries per milliliter.

Following the end-of-cycle shutdown, iodine-131 spiked to 11.45 microcuries per milliliter. The cause was a significant number of failed fuel rods. Inspections revealed that 136 of the total 157 fuel assemblies contained failed fuel (approximately 300 fuel rods had through-wall penetrations), far exceeding the anticipated number of failures.

The defects were generally pinhole sized. The fuel cladding failures were caused by long-term fretting from debris that became lodged between the lower fuel assembly nozzle and the first spacer grid, resulting in penetration of the stainless-steel fuel cladding. The source of the debris was apparently a machining byproduct from the thermal shield support system repairs during the previous refueling outage.

The event is reportable because the cladding failures exceed expected values and are unique or widespread.

(2) Reactor Coolant System Pressure Boundary Degradation Due to Corrosion of a Control Rod Drive Mechanism Flange

While the plant was in hot shutdown, a total of six control rod drive mechanism (CRDM) reactor vessel nozzle flanges were identified as leaking. Subsequently, one of the flanges was found to be eroded and pitted. While removing the nut ring from beneath the flange, it was discovered that approximately 50 percent of one of the nut ring halves had corroded away and that two of the four bolt holes in the corroded nut ring half were degraded to the point that there was no bolt–thread engagement.

An inspection of the flanges and spiral wound gaskets, which were removed from between the flanges, revealed that the cause of the leaks was the gradual deterioration of the gaskets from age. A replacement CRDM was installed and the gaskets on all six CRDMs were replaced with new-design graphite-type gaskets.

The event is reportable because there is a material defect in the primary coolant system that cannot be found acceptable under ASME Section XI.

(3) Degradation of Reactor Fuel Rod Cladding

Identified during Fuel Sipping Operations With the plant in cold shutdown, fuel sipping operations appeared to indicate that a significant portion of cycle 2 fuel, type “LYP,” had failed; i.e., 4 confirmed and 12 potential fuel leakers. The potential fuel leakers had only been sipped once before the ENS notification was made. The licensee contacted the fuel vendor for assistance onsite in evaluating this problem.

An ENS notification was made because the fuel cladding degradation was thought to be widespread. However, additional sipping operations and a subsequent evaluation by the licensee’s reactor engineering department with vendor assistance concluded that no additional fuel failures had occurred; i.e., the abnormal readings associated with the potential fuel leakers was attributed to fission products trapped in the crud layer. Based on the results of the evaluation, the licensee concluded that the fuel cladding was not seriously degraded and that the event was not reportable. Consequently, after discussion with the regional office, the licensee appropriately retracted this event.


  1. The LCO typically employs La, which is defined in Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” to 10 CFR Part 50 as the maximum allowable containment leak rate at pressure Pa, the calculated peak containment internal pressure related to the design-basis accident. “Minimum pathway leak rate” means the minimum leak rate that can be attributed to a penetration leakage path; for example, the smaller of either the inboard or outboard valve’s individual leak rates.