05000461/LER-2016-011
| Clinton Power Station, Unit 1 | |
| Event date: | 11-17-2016 |
|---|---|
| Report date: | 01-13-2017 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition |
| 4612016001R00 - NRC Website | |
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PLANT AND SYSTEM IDENTIFICATION
General Electric—Boiling Water Reactor, 3473 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX]
EVENT IDENTIFICATION
Incorrect Calculation Method Used to Demonstrate Control Room Habitability A. Plant Operating Conditions before the Event Unit: 1 Mode: 1 Event Date: 11/17/16 Mode Name: Power Operation Event Time: None Reactor Power: 99 percent
B. DESCRIPTION OF EVENT
During the NRC Component Design Basis Inspection (CDBI) on November 17, 2016, CPS determined that Calculation, C-020 "Reanalysis of Loss of Coolant Accident (LOCA) Using Alternate Source Terms," incorrectly took credit for the dual Main Control Room ventilation (VC) supply inlets being single failure proof (SFP). The calculation is used to demonstrate control room habitability in the CPS accident analysis. The calculation assumed dual air inlets for the emergency zones as the type of system used for the Main Control Room ventilation system (VC). The dual inlet type system allows for one of the calculated dose concentrations to be reduced by a factorof 4 in accordance with the guidance of Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants." The VC system is SFP, but the individual subsystems at the inlet as designed are not. The calculational error, when corrected, resulted in a calculated dose of greater than the 5 rem Total Effective Dose Equivalent (TEDE) allowable dose to occupants of the control room.
The cause investigation team established that the USAR sections which described the VC system were not clear regarding whether the intake structures were SFP. This condition was concluded as the cause of the event. A review of the design calculation established that C-020 was prepared by Sargent & Lundy. The individual responsible for the design calculation preparation considered that both VC intake structures were independently SFP based on his understanding of the USAR sections that described the VC system. This was an unverified assumption by the preparer which resulted in the use of an incorrect methodology when the calculation assumed the outside intakes for VC were SFP. The lack of technical rigor in the preparation of the design calculation was established a condition which contributed to the event described in this report.
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3. LER NUMBER
2016 - 00 011
C. CAUSE OF EVENT
The USAR sections that described the VC system were not clear regarding whether the intake structures were SFP. The design calculation preparers consideration that both outside VC intake structures were SFP is attributable to the lack of clarity of the USAR sections.
D. SAFETY ANALYSIS
There were no safety consequences associated with the event described in this report. The event is reportable under 10CFR50.73(a)(2)(ii)(B) as "any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.
The event described in this report did not involve degraded performance associated with any portion of the VC system or the Control Room Envelope boundary. It was limited to assumptions used in the performance of a design calculation that demonstrates control room habitability following a design basis accident. This design calculation used an incorrect method and represents a nonconforming condition as described in plant procedures since it did not align with the current licensing basis.
The redundant subsystems of the Control Room Ventilation System during this event remained OPERABLE to ensure that at least one was available, if a single active failure disabled the other subsystem as described in Technical Specification Basis 3.7.3. Plant systems required for the safe shutdown of the plant were also not affected by the event described in this report.
This event report does not identify any safety system functional failures.
E. CORRECTIVE ACTIONS
A standing order was issued for compensatory measures in the event of an emergency until calculations were prepared to demonstrate MCR operator dose limits can be met without compensatory measures. Actions have also been initiated to finalize the MCR Operator dose analysis assuming single failure of outside air intake dampers and revise the USAR to clarify that intake structures are not independently SFP.
comments regarding burden estimate to the FOIA, Privacy and Information Collections Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
F. PREVIOUS SIMILAR OCCURENCES
No previous Event Reports were identified associated with the event described in this report.
G. COMPONENT FAILURE DATA
This report does not involve the failure of components.