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The following information was provided by the licensee via email: Notification per 10 CFR 50.72 (b)(3)(v)(A) and (v)(D) At time 1630 CDT on 4/18/23, Comanche Peak Unit 1 entered TS (Technical Specification) 3.0.3 for 11 minutes due to declaring Train A component cooling water (CCW) inoperable in conjunction with a Train B centrifugal charging pump (CCP) inoperable for scheduled maintenance. This resulted in an event or condition that could have prevented fulfillment of a safety function, high head injection of the emergency core cooling system. CCP 1-02 and fan cooler were tagged out of service at 0400 CDT on 4/18/23 due to scheduled maintenance activities. Containment spray (CT) pump 1-03 seal oil cooler CCW leak was found by a watchstander at 0930 CDT on 4/18/23. Engineering determined that leakage was CCW from a pipe flange weld after insulation removal and could not (determine) operability and notified control room at 1630 CDT on 4/18/23. This placed unit 1 in a TS 3.0.3 condition from 1630 to 1641 CDT for approximately 11 minutes until CCP 1-02 was restored back to operable status. CCW was declared operable at 1912 after CT pump 1-03 seal oil cooler was isolated. CT pump 1-03 remained inoperable until weld repair completed. Train A CT pump 1-03 declared operable at 1211 CDT 4/19/23. ENS notification should have been made by 0030 CDT on 4/19/23. This report restores compliance. The NRC Resident Inspector has been notified.

• * * RETRACTION ON 6/7/2023 AT 1000 EDT FROM CASEY DAVIES TO BILL GOTT * * *

The following information was provided by the licensee via phone and email: Although Comanche Peak Unit 1 conservatively entered a limiting condition for operation action statement and performed repairs immediately, further engineering inspection and evaluation concluded that the CCW system was fully able to provide the needed flow to the 1-03 CT pump seal coolers from the time of discovery (0930 CDT) until which time the piping was isolated for repairs. During this period, structural integrity of the joint was maintained, CCW inventory loss remained within acceptable limits, and CCW could perform its intended design and safety functions. Based on this revised operability determination, train A CCW was always operable, and TS 3.0.3 did not apply. Therefore, reportability requirements per 10 CFR 50.72 (b)(3)(v)(A) and (v)(D) did not apply, and a 60 day LER will not be submitted. The licensee notified the NRC Resident Inspector. Notified R4DO (Young)

On March 28, 2017 at approximately 1957 CDT, a condition was discovered whereby a postulated moderate-energy line break (MELB) involving three fire protection (FP) pipe segments in the Safeguards Building did not contain MELB shielding. It was subsequently determined a postulated crack in one of the affected FP piping sections could adversely affect circuitry associated with the cooling support system for the train A RHR (Residual Heat Removal) pump room, potentially causing the ventilation system to be unavailable to support operation of the train A RHR pump. This condition is not consistent with the CPNPP licensing basis for the protection of essential safe shutdown RHR equipment. At approximately 1957 CDT train A RHR was declared inoperable but available and the unit entered a seventy-two hour LCO (Limiting Condition for Operation) Action Statement per Technical Specification 3.5.2 B pending completion of mitigative actions. Since Unit 1 train B RHR system components and related supporting equipment have been periodically declared inoperable at various times in the last three years for surveillance testing or maintenance, given the MELB condition, both trains of RHR and or support equipment could have been inoperable and this represents an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). At the time of discovery, train B RHR and support equipment were operable. Therefore, the identified condition is not reportable as a loss of safety function per 10 CFR 50.72(b)(3)(v). The Senior NRC Resident Inspector has been notified. Compensatory actions will include installing a spray shield on the affected cable trays.

• * * RETRACTION FROM JOHN ALEXANDER TO VINCE KLCO ON 5/23/17 AT 1720 EDT * * *

On 03/29/2017 Comanche Peak reported an ENS Report (no. 52646) related to the identification of potential moderate-energy line break (MELB) considerations in the Safeguards Building and the potential for adverse interaction with specified Unit 1 electrical equipment. The specific interactions of concern were related to ventilation equipment which would support operation of the Unit 1 A RHR train and several segments of fire protection piping. Subsequent investigations by Engineering have determined: (1) all but one of the suspected potential interactions were determined to not be credible, i.e., the potential MELB would not result in an adverse interaction with the 'target' equipment, and (2) for the remaining potential interaction, an assessment of piping stresses determined there was not a credible MELB source in the affected piping segment and therefore there was not a potential for adverse interaction with the ventilation support equipment. Based on the above, the condition described in ENS report no. 52646 is not considered to be an un-analyzed condition as described in 10 CFR 50.72(b)(3)(ii)(B). The licensee informed the NRC Resident Inspector. Notified the R4DO (Groom).

During a review of ongoing analyses related to postulated tornado missiles, a question was raised about the sentinel valve on the turbine driven auxiliary feedwater pump (TDAFW). The sentinel valve is designed as a warning system on steam equipment to warn personnel of increased back pressure. The valve is not an ASME component and its operation is not required to support TDAFW operation. The draft analysis predicts the TDAFW exhaust stack could be partially crimped by a tornado missile and the resultant back pressure on the turbine would increase to approximately 40 psi. This is higher than the set point for the sentinel valve (nominally 27 and 29 psi for Units 1 and 2, respectively). Therefore, in a design basis tornado with a design basis tornado missile striking the TDAFW exhaust stack, and in a condition where the TDAFW is demanded to run, the sentinel valve is expected to lift and allow steam to flow into the room. Vendor correspondence indicates that at approximately 40 psi the sentinel valve will conservatively pass 600 lbm/hr. Thus, it is conservatively considered operation of the TDAFW under such conditions would create an adverse steam environment which would be beyond that which the TDAFW pump has been analyzed to operate. Actions planned to alleviate the above condition would eliminate the potential for adverse environmental conditions. The steam supplies to the TDAFW have been isolated to affect repairs, which are expected to be limited to removal of the sentinel valve from each Unit and installation of a plug. Said activities are expected to be completed within the Allowed Out-of-Service Time (AOT) of the TDAFW of seventy-two hours per Technical Specification 3.7.5. NRC Resident Inspector has been informed.

• * * RETRACTION ON 10/27/2016 AT 1353 EDT FROM DANNY BRADFORD TO BETHANY CECERE * * *

On 09/01/2016 Comanche Peak reported an ENS Report (no. 52217) related to unanalyzed conditions related to the sentinel valve on the turbine driven auxiliary feedwater pump (TDAFW) during postulated tornado-based scenarios and non-tornado based scenarios. The technical concern was the potential for the sentinel valve to release steam into the TDAFW room and result in adverse environmental conditions within the room and potentially external to the TDAFW room for both tornado-based and non-tornado based scenarios. Subsequent investigations by Engineering have determined the sentinel valve would not be demanded to open during tornado based scenarios and would not result in adverse environmental conditions internal or external to the TDAFW room in any design bases scenario. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Farnholtz).

A potential scenario has been identified that has not been analyzed in the Comanche Peak Nuclear Power Plant Fire Safe Shutdown Analysis (FSSA). This situation is described below for Unit 1, but also applies to Unit 2. Listed below is the configuration for 1EA1. The basic configuration is typical for 1EA2, 2EA1 and 2EA2 as well (ref. E1-0001) - Safeguard Bus 1EA1 is a 6.9 kV switchgear with a Main-Tie-Main configuration. - The normal lineup has one feeder breaker closed, the other feeder breaker open and the tie breaker closed. - 1EA1 receives normal power from the secondary side of Startup Transformer XST2 through breaker 1EA1-1. - 1EA1 receives alternate power from the secondary side of Startup Transformer XST1 through breaker 1EA1-2. - 1EA1 receives emergency power from diesel generator 1EG1 through breaker 1EG1. - An alternate source of power for 1EA1 is also available from Train C through breaker 1EA1-3. The control wiring for the 1EA1-1 circuit breaker contains the following attributes that are important to understand the issue:

- Switch 43/1EA1-1, located in the Shutdown Transfer Panel (STP) is used to transfer control of 1EA1-1 from the Control Room (1-CB-11 switch CS-1 EA1-1) to the Hot Shutdown Panel (HSP) switch CS-1 EA-1 L. (Ref. E1-0031-01&02) - There is a trip circuit fuse located in the 6.9 kV switchgear compartment for 1EA1-1 for control of the trip circuit when the breaker is controlled at 1CB-11 and a separate fuse for the trip circuit when the breaker is controlled by the HSP. - The trip circuit for 1EA1-1 has a contact routed through the Control Room to the Solid State Protection System (SSPS) Cabinet. This contact is in the trip circuit when control is from the Control Room or when control is from the HSP. (Ref. E1-0031-01) The scenario is based on a fire in the Control Room. If the fire in the Control Room causes a ground in the wiring routed in the Control Room to the SSPS cabinet, the fuse for the 1E1-1 trip circuit would open. In the event of a fire in the Control Room, control of the plant is transferred to the HSP. When the control of breaker 1EA1-1 is transferred to the HSP and the ground condition in the SSPS wiring still exists, the second 1EA1-1 trip circuit would open. At this time there would be no way to remotely trip open 1EA1-1. The breaker could still be tripped mechanically at the breaker. Therefore, if 1EA1-1 is closed, 1EA1 could remain energized if off-site power is available. If off-site power is not available, but 1EA1-1 remains closed, bus 1EA1 would remain electrically connected to XST2. As part of the transfer of control from the Control Room to the HSP, operators start up the diesel generator and close 1EG1 to place 1EA1 loads on the generator. If 1EG1 is closed and 1EA1-1 breaker is still closed with off-site power available, the generator will be immediately connected to grid power through XST2 without synchronizing. If 1EG1 is closed and 1EA1-1 breaker is still closed with off-site power not available, the generator will be immediately connected to XST2 and attempt to energize XST2. This large current draw associated with energizing XST2 would likely stall and damage diesel generator 1EG1. Compensatory Action is being implemented through procedure revisions to preclude damage to the diesel generator during this scenario. The licensee will notify the NRC Resident Inspector.

* * * RETRACTION FROM TOM RUCKER TO PETE SNYDER AT 1713 ON 9/8/11 * * * 

At 1939 central daylight time on July 21, 2011, Luminant Power notified the NRC (Event No. 47086) of a Unanalyzed Condition per 50.72(b)(3)(ii)(B) regarding the vulnerability from a potential control room fire on 'A' Safeguards bus. The event report described a portion of a cable running from the Hot Shutdown Panel (HSP) to the Solid State Protection System (SSPS) cabinet in the Control Room (CR) that was not protected from a CR fire scenario for which a worst case Control Room/Cable Spreading Room fire induced short could result in the 1EA1-1 circuit breaker, connecting the 345 kV Startup Transformer to the grid, not tripping and damaging the EDG. Upon further review it has been determined that the Fire Safe Shutdown Analysis (FSSA) modeled that cable in the analysis since CPNPP began commercial operation. Additionally, the FSSA accounted for the fire-induced circuit ground on this cable and one of the specified manual actions is to trip the 1EA1-1 circuit breaker to assure the EDG would load the 'A' Safeguard bus to support the required fire scenario. The EDG output breaker is designed to auto close once the 1EA1-1 circuit breaker is tripped open. The previous version (prior to compensatory actions) of ABN-803A/B, the procedure used should a fire occur in the Control Room, directed the Reactor Operator to trip 1EA1-1 circuit breaker once the diesel was verified running, then ensure the EDG breaker closed. There is no procedural step directing personnel to manually close the EDG breaker. Since, in this scenario, indication for the 1EA1-1 circuit breaker at the Hot Shutdown Panel would be lost, it would be expected that the RO would direct the RRO to verify 1EA1-1 circuit breaker position. Based on the above, Luminant Power has concluded that the FSSA adequately modeled the plant and procedures were written which would not direct any action that would have caused the condition stated in the CR description. Based on the above, the conclusion is that CPNPP did not have an unanalyzed condition that significantly degraded plant safety per 50.72 (b)(3)(ii)(B) regarding the vulnerability from a potential control room fire on 'A' Safeguards bus. Therefore, this event is retracted. The licensee will notify the NRC Resident Inspector. Notified R4DO (Lantz).

The following event occurred with the unit in a "No Mode" defueled condition. On the 12th of April at 2215 CDT Comanche Peak Nuclear Power Plant Unit 2 had a Train B ESF Actuation (Black Out) reportable per 10CFR50.72(b)(3)(iv) for (the 6.9KV Train B) Emergency AC Electrical power systems including Emergency Diesel Generator (EDG). At the time of the event, Unit 2 was in 'No Mode' - the core was off loaded and post maintenance testing was being performed on the 2-02 Emergency Diesel Generator. The scope of the EDG testing was to verify proper operation of the EDG when it was the sole supply to the Unit 2 Train B 6.9KV safeguards buses. Unit 2 Train B 6.9 KV buses have 3 AC sources - XST1 (138KV) preferred source, XST2 (345KV) alternate source and the 2-02 Emergency Diesel Generator. The 2-02 EDG was started normally and was operating in parallel with XST1 (preferred AC source) just before the event. The event was triggered when the preferred AC supply breaker to the Train B 6.9KV Safeguard bus was opened per plan. The opening of the preferred feeder breaker resulted in the 2-02 EDG being the sole AC source to the Train B 6.9KV bus. Immediately following this breaker operation, the EDG tripped on an Auto Voltage Regulator failure. This caused the Train B 6.9KV safeguard bus to de-energize which resulted in the automatic closure of the alternate AC (XST2) feeder breaker. This in turn triggered the Train B Solid State Sequencer to sequence on Train B Black Out loads - as designed. The Train B Black Out Solid State Sequencer operated as designed. No abnormalities were noted. All available equipment operated as designed. The 2-02 EDG is currently tagged out and the problem associated with the (Auto Voltage Regulator) failure is being investigated. (The) NRC Resident Inspector has been informed and is currently located on site. No fuel movement was in progress during this event. Train A safeguard power was unaffected and available throughout. Non-safety related power was also not affected. The only operational impact was the loss of a Train B spent fuel pool cooling pump for approximately 10 minutes. The spent fuel pool temperature increased approximately 1/2 degree during before cooling was restored.

• * * RETRACTION FROM DAVID BUTLER TO HOWIE CROUCH AT 1830 EDT ON 4/14/11 * * *

Further investigation of this event has determined that the 2-02 Emergency Diesel Generator (EDG) auto voltage regulator functioned as designed. The EDG did not receive an emergency start signal, which was consistent with the design for the plant conditions present. In addition, Unit 2 was defueled and in a 'No Mode' condition, and there was no actuation of any of the systems listed in paragraph 50.72 (b)(3)(iv)(B) by the sequencer operation. Therefore, this event is not reportable because it did not meet the criteria for a system actuation per 50.72(b)(3)(iv)(A) and the guidance in NUREG-1022. The licensee will be notifying the NRC Resident Inspector. Notified R4DO (Miller).

At 0900 on July 30 2007, an Engineer noted during the review of a revision to the Comanche Peak Fire Safe Shutdown Analysis that a cable associated with the control circuitry for Train B of the Safety Chilled Water System may not be adequately protected from a potential fire. By design, electrical control cables for Trains A and B of the Safety Chilled Water System are located in the same fire zone. The original design specified that the Train B electrical control cables in this zone were to be protected with fire barrier material (thermolag). However, in this case the fire barrier material was found to be missing from the Train B electrical control cables. Upon discovery of this condition, a fire impairment was implemented for the affected fire zone. Engineering performed an evaluation of this condition and at 0900 on August 13, 2007 concluded that if a fire occurred in the affected fire zone, the required degree of separation for redundant safe shutdown trains was inadequate (i.e. both A and B trains were affected) and this would adversely affect the control circuitry and potentially prevent the Unit 1 Safety Chilled Water System from performing its intended safety function. The Unit 1 Safety Chilled Water Systems safety function at Comanche Peak is to remove heat dissipated from engineering safety features equipment and to maintain ambient temperatures in rooms containing safety related equipment below maximum design temperatures. This condition is similar to an example given in NUREG 1022, Rev. 2, Section 3.2.4 for an unanalyzed condition that significantly affects plant safety (fire barrier missing such that the required degree of separation for redundant safe shutdown trains is lacking). Therefore, this condition is reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee notified the NRC Resident Inspector.

• * * RETRACTION ON 08/30/07 AT 1249 EDT FROM RAUL MATINEZ TO MACKINNON * * *

CPNPP is retracting Event Notification 43562 based on the following: Further review of this issue by Engineering has determined that the required degree of separation for redundant safe shutdown trains was adequate and the Unit 1 Safety Chilled Water System was capable of performing its intended safety function. Therefore, this condition is not reportable per 10CFR50.72(b)(3)(ii)(B), 'The nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' CPNPP has informed the NRC Resident Inspector. R4DO (R. Nease) notified.