05000265/LER-2016-003
| 07-25-2016 | result, the Station determined that a drywell entry was required to investigate the condition, make any needed repairs, and refill the oil reservoir. NRC provided that Station personnel did not comply with Technical Specifications (TS) 3.6.2.5 (DW to Suppression.
Chamber DP) and 3.6.3.1 (Primary Containment 02 concentration) since while in MODE 1, at the end of the 32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> Completion Time (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action A, plus the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Action B) during the actual plant evolutions for power ascension, these Required Actions were not met because the associated Applicability for each TS were not met since the Unit remained in MODE 1. The cause of the issue was Station personnel understanding and application of the subject TS as used in context under this infrequent plant condition, differed from the NRC's understanding and application of the subject TS. The specific difference is with the application of the term, "start-up," as used in the LCO Applicability. Corrective actions included issuance of an Operations Standing Order, and revision of pertinent Operating procedures to ensure these Tech Specs are properly implemented. The safety significance of this event was minimal. Given the impact on the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications. ContentsPLANT AND SYSTEM IDENTIFICATIONGeneral Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX]. EVENT IDENTIFICATIONCompliance Issue with the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications A. CONDITION PRIOR TO EVENTUnit: 2 Reactor Mode: 1 Event Date: May 25, 2016 Event Time: 11:10 hours Mode Name: Power Operation Power Level: 100% B. DESCRIPTION OF EVENTOn 05/23/16 Low Level Alarm [LA] 2A Recirc [AD] Motor [MO] occurred due to a low oil level condition for the 2A recirculation pump [P] motor. As a result, the Station determined that a drywell [NH] entry was required to investigate the condition, make any needed repairs, and refill the oil reservoir [TK]. NRC provided that Station personnel did not comply with TS 3.6.2.5 (DW to Suppression Chamber DP) and 3.6.3.1 (Primary Containment 02 concentration [BB]) since while in MODE 1, at the end of the 32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> Completion Time (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action A, plus the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Action B) during the actual plant evolutions for power ascension, the Required Actions B for TS 3.6.2.5 and TS 3.6.3.1 were not met (at 1110 on 5/25/16, and 1123 on 5/25/16, respectively), because the Unit 2 LCO Applicability for establishing DW/Torus differential pressure and being fully inerted were not met since the Unit remained in MODE 1. NRC provided that these TS were not met when the Station improperly used the LCO Applicability (a), 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> "clock reset" allowance to proceed above 15% power "following startup" without setting the DW/Torus differential pressure (Dp) > 1 psid, and oxygen concentration contrary to the NRC's "plain language" interpretation of this associated TS Applicability, in that "following startup" was intended to mean "following MODE 2." Furthermore, the NRC provided that the Unit did not exit the Mode of Applicability just by dropping below 15% Rated Thermal Power (RTP), since the Unit was still in MODE 1, and a total of only 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> was available to re-achieve DW/Torus Dp and reinert while remaining in MODE 1. While under this interpretation, the resulting available options during this drywell entry were to either: 1) re-establish DW/Torus Dp and inerting prior to reaching 15% RTP during the power ascension, or 2) to exit MODE 1 to reset the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> clock (meaning to start power ascension from MODE 2). In this situation, the TS LCO Applicability is not clear, does not coincide with the Bases intent, and may be overly restrictive in that it uses the terms, "startup" and "shutdown. The cause of the issue was Station personnel understanding and application of the subject TS as used in context under this this infrequent plant condition, differed from the NRC's understanding and application of the subject TS. The specific difference is with the application of the term, "start-up," as used in the LCO Applicability. The safety significance of this event was minimal. Given the impact on compliance with the Drywell/Suppression Chamber Differential Pressure, and Primary Containment Oxygen Concentration Technical Specifications, this report is submitted for Unit 2 in accordance with the requirements of 10 CFR 50.73(a)(2)(i)(B), which requires the reporting of a past operation or condition which was prohibited by the plant Technical Specifications. C. CAUSE OF EVENTDuring the drywell entry and subsequent return to full power, the Station performed the power ascension under procedures, QCGP 3-1 (Reactor Power Operations) and QCOP 1600-20 (Nitrogen Inerting of Primary Containment Using the Vaporizer(s) and Reactor Building Ventilation System), for which under this infrequent plant condition, the context of "startup" was understood to refer to the "act of increasing reactor power," or "power ascension." Therefore, the apparent TS interpretation conflict occurred in the meaning and use of "startup," in the LCO Applicability, when during power ascension the Station proceeded above 15% RTP without resetting DW/Torus Dp and re-inerting within the 32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> maximum allowed Completion Time. This issue pertained to a reading of the language of the subject TS which in itself was not readily able to be consistently interpreted since the "plain language" did not match the TS Bases nor the NRC approved text of the Safety Evaluation (SE). This TS compliance interpretation issue occurred for only a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration (in excess of the 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> total Completion Time allowed while in MODE 1), pertaining to the DW/Torus Dp and oxygen concentration, respectively. D. SAFETY ANALYSISSystem Design TS Bases 3.6.2.5, Drywell-to-Suppression Chamber Differential Pressure Applicable Safety Analyses provides: "The purpose of maintaining the drywell at a slightly higher pressure with respect to the suppression chamber is to minimize the drywell pressure increase necessary to clear the downcomer pipes to commence condensation of steam in the suppression pool and to minimize the mass of the accelerated water leg. This reduces the hydrodynamic loads on the torus during the LOCA blowdown. The required differential pressure results in a downcomer waterleg of approximately 1 ft. Initial drywell-to-suppression chamber differential pressure affects both the dynamic pool loads on the suppression chamber and the peak drywell pressure during downcomer pipe clearing during a Design Basis Accident LOCA. Drywell-to suppression chamber differential pressure must be maintained within the specified limits so that the safety analysis remains valid. TS Bases 3.6.3.1, Primary Containment Oxygen Concentration Applicable Safety Analyses provides: "The UFSAR, Section 6.2.5 calculations assume that the primary containment is inerted when a Design Basis Accident loss of coolant accident occurs. Thus, the hydrogen assumed to be released to the primary containment as a result of metal water reaction in the reactor core will not produce combustible gas mixtures in the primary containment. Oxygen, which is subsequently generated by radiolytic decomposition of water, will not result in the primary containment becoming de-inerted within the first 30 days following an accident. Safety Impact TS Bases 3.6.2.5 Drywell-to-Suppression Chamber Differential Pressure LCO Applicability provides: "As long as reactor power is containment occurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a startup or within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a shutdown is low enough that these "windows," with the primary containment not inerted, are also justified. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting." For this event, since the period of time during which reactor power was > 15% RTP while the DW to Torus Dp was 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, the probability of an event that generates hydrogen or excessive loads on primary containment was low since this duration was less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, therefore, the safety impact of this condition was minimal. TS Bases 3.6.3.1, Primary Containment Oxygen Concentration LCO Applicability provides: "As long as reactor power is containment need not be inert. Furthermore, the probability of an event that generates hydrogen occurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a startup, or within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before a shutdown, is low enough that these "windows," when the primary containment is not inerted, are also justified. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting." For this event, since the period of time during which reactor power was > 15% RTP while the DW was not inerted (i.e., 02 concentration > 4%), was approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, the probability of an event that generates hydrogen was low since this duration was less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, therefore, the safety impact of this condition was minimal. Due to the language in the associated TS Bases and SE documentation for the actions that the Station took during the drywell entry and subsequent power ascension, this TS compliance interpretation issue is not a significant event/issue, since the interpreted non-compliance occurred for only a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration, pertaining to the DW/Torus Dp and oxygen concentration, respectively (i.e., 4 hour/2 hour in excess of the 32 'hours total Completion Time allowed while in MODE 1). Furthermore, this event was the first known recorded occurrence of non-compliance with these TS under this interpretation. Since the condition created no consequences, the safety impact of this condition was minimal. Risk Insights The plant Probabilistic Risk Assessment (PRA) model was reviewed with respect to this event. Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) were evaluated for impacts of oxygen concentration and DW/Torus Dp. Since the period of time during which reactor power was > 15% RTP while the DW was not inerted (i.e., oxygen concentration > 4%), was approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />, and since the period of time during which reactor power was > 15% RTP while the DW to Torus Dp was change in risk was minimal. In conclusion, the overall safety significance and impact on risk of this event were minimal. E. CORRECTIVE ACTIONSImmediate: 1. Issued an Operations Standing Order that provided clarifying information when using the subject Tech Spec for drywell entries. Follow-up: 2. The pertinent Operating procedures will be revised to ensure the subject Tech Specs are properly implemented for drywell entries. 3. This issue will be addressed under a proposed BWROG TSTF item for a potential future Tech Spec and Bases revision. 4. Operator Training will review this issue as an OPEX item, and for incorporation into appropriate lesson plans. F. PREVIOUS OCCURRENCESThe Station events database, LERs, and INPO Consolidated Event System (ICES) were reviewed for similar events at the Quad Cities Nuclear Power Station. This event was caused by Station personnel understanding and application of the subject Tech Specs as used in context under this infrequent plant condition, differed from the NRC understanding and application of the subject Tech Specs. The specific difference is with the application of the term, "start-up," as used in the LCO Applicability.
G. COMPONENT FAILURE DATAFailed Equipment: N/A Component Manufacturer: N/A Component Model Number: N/A Component Part Number: N/A This event has not been reported to ICES since there was no equipment failure. |
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| Quad Cities Nuclear Power Station Unit 2 | |
| Event date: | 05-25-2016 |
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| Report date: | 07-25-2016 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 2652016003R00 - NRC Website | |
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