Information Notice 2012-12, HVAC Design Control Issues Challenge Safety System Function
| ML12115A012 | |
| Person / Time | |
|---|---|
| Issue date: | 07/24/2012 |
| Revision: | 0 |
| From: | Dudes L A, McGinty T J Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking |
| To: | |
| Garmon-Candelaria D | |
| References | |
| IN-12-012 | |
| Download: ML12115A012 (5) | |
ML12115A012 July 24, 2012 NRC INFORMATION NOTICE 2012-12: HVAC DESIGN CONTROL ISSUES CHALLENGE SAFETY SYSTEM FUNCTION
ADDRESSEES
All holders of an operating license or construction permit for a nuclear power reactor or a non-power (research or test) reactor issued under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities," except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vesse All holders of and applicants for a power reactor early site permit, combined license, standard design certification, standard design approval, or manufacturing license under 10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants."
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform addressees about certain events involving heating, ventilation, and air conditioning (HVAC) system design control issues that challenged, or potentially challenged, safety system function The NRC expects recipients to review the information contained within for applicability to their facilities and consider actions, as appropriate, to avoid similar occurrence Suggestions contained within this IN are not NRC requirements; therefore, no specific action or written response is require DESCRIPTION OF CIRCUMSTANCES Susquehanna Steam Electric Station (Susquehanna) HVAC Controller On January 3, 2011, PPL, the licensee for Susquehanna, identified a single-point vulnerability in the reactor building HVAC syste The vulnerability was that a failure of a nonsafety-related temperature controller coincident with outside ambient air temperatures below 10 degrees Fahrenheit (oF) could result in a spurious steam leak detection (SLD) system isolation on high differential temperature (T), causing simultaneous isolation of main steam isolation valves (MSlV), the high pressure coolant injection system, and the reactor core isolation cooling syste This vulnerability was common to both Susquehanna Units 1 and 2 and had been in existence since the plants began licensed operation PPL initially reported the issue through an event notification (EN) (EN 46519) under 10 CFR 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors," as an unanalyzed condition (10 CFR 50.72 (b)(3)(ii)(B)) and an accident mitigation concern (10 CFR 50.72 (b)(3)(v)(D)). However, on February 28, 2011, PPL submitted an updated EN that removed the accident mitigation consideration based on the low likelihood of a reactor building temperature controller failure during a period when outside temperature was below 10 oF (both conditions are required for the deficient SLD system isolation on high T to occur). PPL provided additional information pertaining to this issue in the form of a 10 CFR 50.73, "License Event Report [LER] System," for an unanalyzed condition (LER 3872011001). The LER stated that the single-point vulnerability was discovered during the preparation of a 10 CFR 50.59, "Changes, Tests and Experiments," determination for an engineering change to remove the SLD high T isolation function to address obsolescence of the function's component The licensee attributed the issue to a "less than adequate single-failure analysis performed during the original plant design." The original single-failure analysis was performed consistent with accepted practices during the period of the initial plant desig In 2007, Susquehanna engineers received training on failure modes and effects analysis (FMEA) technique This training updated the expectations for FMEAs performed on nonsafety system Consequently, Susquehanna engineers used the new techniques when evaluating the impact of removing the SLD isolation function and, in the process, identified the single-point vulnerability deficienc The corrective actions for this issue included removing the isolation function of the SLD system T instrumentation and performing a FMEA on all nonsafety systems that could cause an isolation of the emergency core cooling system or MSIVs as an extent of condition assessmen The report, "Susquehanna Steam Electric Station - NRC Integrated Inspection Report 05000387/2011003 and 05000388/2011003 and Exercise of Enforcement Discretion," dated August 10, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112220409), provides the results of the NRC inspection related to this issu Diablo Canyon Power Plant Auxiliary Building Ventilation System Actuation Logic Diablo Canyon Nuclear Power Plant (DCNPP) completed modifications to its auxiliary building ventilation systems (ABVS) in November 201 These modifications included replacement of relay-based actuation logic with a programmable logic controller (PLC). The licensee implemented the modification to address problems with reliability and availability (i.e. obsolescence). The licensee reviewed the modification design to ensure applicable single-failure criteria were me Notwithstanding the licensee's review, on January 10, 2011, during containment spray pump quarterly testing, a deficiency in the actuation logic of the recently installed PLC resulted in a complete loss of the Unit 2 ABVS when a damper failed to open as required because of leakage past a piston sea This led one of the two ABVS exhaust fans to trip and prevented the other exhaust fan from starting; thus ABVS became inoperabl The loss of the ABVS required the licensee to take action in accordance with Technical Specification Limiting Condition for Operation 3.0.3 (i.e., action statement to reduce mode of plant operation) for approximately 20 minutes until operators restored the ABVS system through manual action The failure of the piston seal was attributed to using the seal beyond its defined service life, contrary to the requirements of the licensee's preventive maintenance program for the sea DCNPP initially reported this event through a 10 CFR 50.72 EN (EN 46531) as an unanalyzed condition (10 CFR 50.72(b)(3)(ii)(B)) and an accident mitigation concern (10 CFR 50.72(b)(3)(v)(D)). The licensee provided additional information in the form of a 10 CFR 50.73 LER for an unanalyzed condition and safety system functional failure (LER 2752011002). In the LER, the licensee incorrectly attributed the cause of the loss of the ABVS to a nonconforming single-failure vulnerability in the ABVS system design that existed as part of the original design for both DCNPP Unit It was later determined that the 2010 modifications to the ABVS control logic introduced a single-failure vulnerability, where ABVS exhaust fans tripped when a system damper was not fully opene The corrective actions for this issue consisted of modifying the design of both DCNPP units to satisfy the single-failure design criteria, revising the design change process to include a design evaluation of new and old failure modes based on the current licensing and design bases, and revising the licensing basi The report, "Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011002 and 05000323/2011002," dated May 11, 2011 (ADAMS Accession No. ML111310608), provides the results of the NRC inspection related to this issu Point Beach Nuclear Plant (Point Beach) Control Room Emergency Filtration Fan Thermal Overload On February 3, 2007, Point Beach lost operability of the control room emergency filtration system (CREFS) because of an inadequately designed modification (LER 2662007001). In October 2006, the licensee installed a modification (high efficiency CREFS fan motors) for the purpose of increasing the low flow margi During the design of this modification, an incorrect assumption was made that outside temperature had a negligible effect on motor current draw, so no compensation for low temperature was included in the motor thermal overload desig On February 3, 2007, with outside temperature at 6 oF, a CREFS fan tripped during a Technical Specification surveillance test because of a thermal overload relay tri After evaluating the cause of the trip, the licensee declared both CREFS fans inoperable because the fan motors had inadequately sized thermal overload heater element The corrective actions for this issue included replacing the overload heater elements with elements having trip current setpoints adjusted to values that considered design requirement The report, "Point Beach Nuclear Power Plant, Units 1 and 2, NRC Integrated Inspection Report 05000266/2007002 and 05000301/2007002," dated April 12, 2007 (ADAMS Accession No. ML071020081), provides the results of the NRC inspection related to this even
BACKGROUND
Criterion III of Appendix B to 10 CFR Part 50 requires, in part, that licensees ensure that applicable regulatory requirements and design basis are "correctly translated into specifications, drawings, procedures, and instructions." Furthermore, "design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design..."