05000333/LER-2019-001, Secondary Containment Differential Pressure Exceeded the Technical Specification Surveillance Requirement
| ML19074A293 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/15/2019 |
| From: | Timothy Peter Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-19-0037 LER 2019-001-00 | |
| Download: ML19074A293 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3332019001R00 - NRC Website | |
text
Exelon Generation T JAFP-19-0037 March 15, 2019 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001 James A. FitzPatrick NPP P.O. Box 110 Lycoming. NY 13093 Tel 315*349-6024 Fax 315-349-6480 Timothy C. Peter Plant Manager - JAF James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-059 NRG Docket No. 50-333
Subject:
Dear Sir or Madam:
LER: 2019-001, Transitory Secondary Containment Differential Pressure Excursion This report is being submitted pursuant to 1 O CFR 50.73(a)(2)(v)(C).
There are no new regulatory commitments contained in this report.
Questions concerning this report may be addressed to Mr. William Drews, Regulatory Assurance Manager, at (315) 349-6562.
Sincerely,~
dpt Plant Manager TCP/WD/mh
Enclosure:
LER: 2019-001, Transitory Secondary Containment Differential Pressure Excursion cc:
USNRC, Region I Administrator USNRC, Project Manager USNRC, Resident Inspector INPO Records Center (ICES)
NRC FORM 366 (04-2018)
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (04-2018)
LICENSEE EVENT REPORT (LER)
APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 1. Facility Name James A. FitzPatrick Nuclear Power Plant
- 2. Docket Number 05000333
- 3. Page 1 OF 3
- 4. Title Secondary Containment Differential Pressure Exceeded the Technical Specification Surveillance Requirement
- 5. Event Date
- 6. LER Number
- 7. Report Date
- 8. Other Facilities Involved
Month
Day
Year
Year Sequential Number Rev No.
Month Day Year Facility Name N/A Docket Number N/A 1
16 2019 2019 - 001 - 00 03 15 2019 Facility Name N/A Docket Number N/A
- 9. Operating Mode Month Day Year Yes (If yes, complete 15. Expected Submission date) No Abstract (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On January 16, 2019, at 0137, with James A. FitzPatrick Nuclear Power Plant at 100% power, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement 3.6.3.1.1 of greater than or equal to 0.25 inches of vacuum water gauge while isolating Reactor Building Ventilation. This condition existed for approximately ten (10) seconds and returned to within limits when the isolation sequence was completed.
The differential pressure did not exceed 0 so there was no unmonitored exfiltration and there were no actual radiological release events. When Secondary Containment did not meet the Technical Specification requirement for differential pressure, Secondary Containment was Inoperable. Therefore, this event is reportable under 10 CFR 50.73(a)(2)(v)(C).
A differential pressure excursion during transition from normal to isolation mode of the Reactor Building Ventilation System is an expected condition, and attributable to the design of the system. The cause of exceeding the differential pressure requirement has been determined not to be associated with a component failure or equipment malfunction.
Corrective actions include declaring Secondary Containment Inoperable when manually isolating Reactor Building Ventilation and increase the normal differential pressure in Secondary Containment.
(See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
(See Page 2 for required number of digits/characters for each block)
=
Background===
The Secondary Containment (SC) [EIIS identifier: NG] is a structure comprised of the Reactor Building that surrounds the primary containment and refuel equipment. Its safety function is designed to provide containment for postulated accident scenarios: loss-of-coolant accident and refueling accident. This structure forms a control volume that serves to hold up and dilute fission products in the event of an accident. The system was designed to include a differential pressure vacuum (DP) such that external atmosphere would leak into containment rather than fission products leak out.
The systems which maintain the DP is the normal Reactor Building Ventilation (RBV) system [VA] and the safety-related Standby Gas Treatment (SBGT) system [BH]. During a postulated accident scenario, the normal RBV isolates and the SBGT initiates to filter gas from SC to the environment. SBGT has the capacity to maintain DP. (Positive DP refers to a lower pressure inside SC in comparison to environmental air pressure)
Event Description
On January 16, 2019, at 0137, with James A. FitzPatrick Nuclear Power Plant (JAF) at 100% power, SC was manually placed into isolation mode using the operating procedure OP-51A to prepare for equipment maintenance. During this evolution, the DP exceeded the Technical Specification (TS) Surveillance Requirement SR 3.6.3.1.1 of greater than or equal to 0.25 inches of vacuum water gauge for approximately ten (10) seconds and returned to within limits by the SBGT system when the isolation sequence was completed. Operators identified this condition using the plant computer and recorded the lowest value at 0.14 inches of vacuum water gauge. An initial notification was submitted to the NRC by ENS 53828 per 10 CFR 50.72(b)(3)(v)(C).
Event Analysis
The SC DP tends to move towards zero, when the RBV is switched from normal to an isolation mode. The cause of the DP change during the transition phase is the difference in closure time for the RBV supply and exhaust isolation valves. The exhaust valves are designed to close within the first 5 seconds and the supply is designed to close within 15 seconds. Therefore, the supply fans keep bringing outside air in for up to 10 seconds after the exhaust valves have isolated, causing DP to lower. This may be observed from the readings obtained in the control room for the SC pressure during the transition phase.
When RBV is operating, DP is controlled much higher than the 0.25 inches of vacuum water gauge; such that if an isolation is initiated the change due to the transition may not exceed the SR limit. On January 11, 2019, work to clean RBV cooling coils was performed due to an increasing trend in SC DP. This lowered the average normal DP from about 1.5 to 0.8. Therefore, during this event, the isolation transition change had a lower starting point and was able to exceed the TS requirement.
The procedure used to isolate RBV during this event was OP-51A and it instructs starting SBGT to raise DP prior to isolating RBV. Operators did not expect that isolation transition to lower DP below the TS requirement because the normal and SBGT margin would normally be enough. Therefore, SC was not declared Inoperable in preparation for the planned maintenance. When SR 3.6.3.1.1 requirement of greater than or equal to 0.25 inches of vacuum water gauge was exceeded, TS 3.6.3.1 was not met, Secondary Containment was Inoperable, and this event is reportable per 10 CFR 50.73(a)(2)(v)(C). U.S. NUCLEAR REGULATORY COMMISSION (04-2018)
LICENSEE EVENT REPORT (LER)
CONTINUATION SHEET APPROVED BY OMB: NO. 3150-0104 EXPIRES: 03/31/2020
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. LER NUMBER James A. FitzPatrick Nuclear Power Plant 05000 - 333 YEAR SEQUENTIAL NUMBER REV N0.
2019
- - 001
- - 00 Page 3 of 3 (See NUREG-1022, R.3 for instruction and guidance for completing this form http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1022/r3/)
Cause
This event was caused by cleaning of the coolers which lowered normal running differential pressure such that when the isolation mode was initiated the transitory differential pressure change went lower than expected.
Secondary Containment differential pressure may exceed 0.25 inches water vacuum when the RBV is switched from normal to an isolation mode because the system is designed with a difference in closure time for the supply and exhaust isolation valves. However, the cause of this event is not associated with any component failure or malfunction.
Similar Events
FitzPatrick, LER: 2015-006-01, Transitory Secondary Containment Differential Pressure Excursions, JAFP-16-0002, dated February 4, 2016.
Corrective Actions
Completed Corrective Actions Revised OP-51A to declare SC Inoperable when manually isolating RBV.
Change DP controller 66DPC-112 setting to increase margin (WO 04759888).
Future Corrective Actions Submit License Amendment Request for TSTF-551, Revise Secondary Containment Surveillance Requirements.
Safety Significance
Nuclear safety - This event did not have any actual or potential impact on nuclear safety.
Industrial safety - This event did not have any actual or potential impact on industrial safety.
Radiological safety - In this event, the DP never exceeded 0 so there was no SC unmonitored exfiltration and there were no actual radiological release events during the period when the TS requirement was not met.
The potential for a radiological consequence is only applicable during the time that SC was below 0.25 inches water vacuum DP. During the Design Basis Loss of Coolant Accident event, Drywell High Pressure or Low Reactor Water Level signals would isolate SC. Any potential Fuel damage or release of radiological materials caused by this event scenario is not expected until after RBV isolation. During the Design Basis Refueling Accident event, a release of radioactive material by a dropped fuel assembly during refuel operations would be detected by Radiation Monitors and initiate SC isolation. The type of pressure changes reported in this LER could result in some exfiltration before the isolation was complete; however, the amount of exfiltration, consequentially the offsite and control room doses, would remain below regulatory limits as analyzed.
The difference in closing time between the supply and exhaust valves of the reactor building during transition from normal to isolate mode represents a potential exfiltration pathway for released activity. This potential exfiltration pathway has been conservatively quantified and is included in the JAF design basis accident analyses. Dose consequence results remain well below the 10 CFR 100 and 10 CFR 50.67 guidelines for all postulated accident conditions. This condition does not adversely impact that ability of RBV to isolate or SBGT to activate. Therefore, the capability of SC to mitigate the consequence of an accident is unaffected.
References JAF Issue Report - IR 04211279, January 16, 2019 Nuclear Safety Evaluation: JAF-SE-96-071, Revision 2