Semantic search

On 8/20/2015 at 1710 CDT, a design flaw was discovered with the pressurizer power operated relief valve (PZR PORV) block valve control circuitry. Specifically, the circuit deficiency for which a design basis fire in the Main Control Room (MCR) or cable spreading room could prevent the PZR PORV block valves from being closed from the local control switch at their associated motor control center (MCC). Engineering has reviewed this issue and determined that a potential fire induced ground in the MCR or cable spreading room could clear the associated control power fuses which would prevent the block valves from operating at the local control switch. These valves are considered to form a High/Low pressure interface which requires postulating a proper polarity DC cable to cable fault. Engineering has reviewed the circuit design and cable routing associated with PORVs 1(2)RY455A and 1(2)RY456 and determined that their associated cables are routed with other DC circuit cables in the MCR control board and cable spreading room raceways, such that this postulated fault could potentially cause spurious opening of one of the PORVs even after the control power fuses have been removed as directed by the station abnormal operating procedures for control room inaccessibility. This identified block valve circuit deficiency prevents the credited safe shutdown action of locally closing the block valves to mitigate the spurious operation of a PORV. Hourly fire watches of the affected MCR and cable spreading room fire zones have been implemented. In addition, the MCR is continuously staffed and the affected cable spreading room fire zones are equipped with detection and automatic suppression. This event is being reported under 10CFR50.72(b)(3)(ii)(B) for 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee has notified the NRC Resident Inspector.

• * * UPDATE PROVIDED BY ROB SHERMAN TO JEFF ROTTON AT 1845 EDT ON 09/02/2015 * * *

During the extent of condition review, an additional design deficiency was identified with respect to the PZR PORV and PZR PORV Block valves. Specifically, the current mitigating strategy for removing PZR PORV control power fuses does not adequately prevent a PZR PORV from spuriously opening due to fire induced hot short. Furthermore, local actions to close the associated PZR PORV block valve at the motor control center (MCC) may not be effective because the MCC may not have electrical power during the design basis fire. Therefore, the credited safe shutdown action to remove the PZR PORV control power fuses does not prevent the PZR PORV from spuriously opening during design basis fires in some of the upper and lower cable spreading room fire zones. The affected Fire Zones are the same upper and lower spreading rooms previously identified and fire watches of the affected areas remain in place. The NRC Resident Inspector has been notified. Notified the R3DO (Skokowski)

On September 9, 2013, during the A1R17 Braidwood Station Unit 1 refueling outage, the as-found leakage of the controlled leakage devices (1RH01SA and 1RH01SB) for the safety injection (SI) system ECCS sump containment isolation valves (1SI8811A and 1SI8811B) were determined to not be 'leak tight' as described in the UFSAR. Since there was only minor leakage from the isolation valves or the associated residual heat (RH) system piping (1-2 drops/month from 1SI8811A and no leakage from 1SI8811B), there was no actual impact on offsite dose or long-term ECCS operation. However, further evaluation has concluded that there was a potential to exceed the assumed leakage limits of the Alternate Source Term (AST) calculation. The RH system is classified as a closed system outside containment meaning the system is designed to accommodate a single active failure (i.e., the failure of the 1SI8811B valve to close) and still maintain an adequate isolation barrier to release recirculation water outside containment. The encapsulation device is intended to capture and limit leakage from a potential leak in the 1SI8811A/B or piping. The controlled leakage device is built to the same standards as the remainder of the RH system recirculation water outside containment. The design function is to limit potential offsite dose due to leakage of recirculation water outside containment. This is not a specified safety function and there are no Technical Specification requirements for these devices. The encapsulation devices do not perform a containment function and are not a principle safety barrier. As there was only minor ECCS system leakage at the time of discovery, there was no impact on past offsite dose or long-term ECCS operation. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B) since the as-found leakage of the controlled leakage devices could have allowed RH leakage to exceed the calculated limits for ECCS systems outside containment. The NRC Resident Inspector has been informed.

• * * RETRACTION FROM RANDY RAHRIG TO HOWIE CROUCH AT 1556 EDT ON 10/31/13 * * *

Retraction of ENS 49371 dated 9/09/2013: The purpose of this report is to retract ENS report #49371 (September 9, 2013). This report was made during Braidwood's refueling outage (A1R17) for the as-found leakage on the controlled leakage devices (1RH01SA and 1RH01SB) for the safety injection (SI) system ECCS sump containment isolation valves (1SI8811A and 1SI8811B) that were determined not to be 'leak tight' as described in the UFSAR. When the ENS notification was made on 9/9/2013, the station determined that there was a potential to exceed the assumed leakage limits of the alternate source term (AST) calculation. The ENS notification was made under 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. At 1500 CDT on Thursday, October 17, 2013, the Braidwood Generating Station concluded that the prior ENS notification could be retracted based on the completion of Revision 2 of calculation BRW-13-0135-M, '1/2RH01SA/B Leak Rate Conversion and Test Pressure Determination'. The as-found pressure test results for the 1RH01SA and 1RH01SB valve containment assemblies would not have resulted in a total ECCS leakage outside containment in excess of that assumed in the AST dose calculation BRW-04-0038-M, 'Re-Analysis of Loss of Coolant Accident (LOCA) Using Alternate Source Terms (AST)'. The as-found valve containment assembly (VCA) pressure test results did not result in an unanalyzed condition that significantly degrades plant safety. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley).

At 1140 (CDT) on August 24th, Unit 2 received Essential Service Water (SX) discharge header pressure low and SX strainer delta pressure high alarms indicative of high flow. At the time, a 2B SX ASME surveillance was in progress which involved field operations by Equipment Operators (EO's). At the time of the event, SX discharge header pressure dropped to 65 psig, less than the 89 psig necessary for operability. The Control Room responded by directing the EO's to restore SX discharge header pressure, which was promptly restored. The 2B SX ASME surveillance sets initial conditions prior to data collection. The surveillance has the total SX flow be adjusted to 24000 gpm via the U2 Component Cooling Water (CC) heat exchanger outlet throttle valve, 2SX007. The subject flow was intended to be measured via an installed ultrasonic flow gauge 2FE-SX147. The EO's, instead used the U2 CC heat exchanger flow gauge 2FE-SX031. As a result, in an attempt to achieve 24000 gpm through the U2 CC heat exchanger, total SX flow exceeded the 24000 gpm since the U2 CC heat exchanger is but one of many loads the 2B SX pump is serving. For the 5 minutes described above, the SX system was in a lineup that may have prevented it to fulfill its safety function and placed Unit 2 in a potentially unanalyzed condition. This condition is still being evaluated. Site Engineering has determined no runout conditions existed. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM JOE KLEVORN TO VINCE KLCO ON 10/19/2011 AT 1626 EDT* * *

The evaluation of the condition has been completed. Based on Essential Service Water (SX) system flow model runs performed, the conditions that existed at the time of the low SX header pressure would have resulted in low flow supplied to multiple safety related components. However, the safety function to provide necessary cooling to required safety-related safe shutdown equipment would have been met under design basis conditions with the auto-start of the 2A SX pump. Therefore, this did not result in a condition that could have prevented fulfillment of a safety function or in an unanalyzed condition that significantly degraded plant safety. Therefore, ENS notification 46203 is being retracted. The licensee notified the NRC Resident Inspector. Notified the R3DO (Daley).