AEP-NRC-2012-61, Enforcement Discretion Regarding Engineered Safety Feature Actuation System Steam Line Isolation Automatic Actuation Logic and Actuation Relays for Steam Generator Stop Valve Dump Valve

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Enforcement Discretion Regarding Engineered Safety Feature Actuation System Steam Line Isolation Automatic Actuation Logic and Actuation Relays for Steam Generator Stop Valve Dump Valve
ML12215A013
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 07/23/2012
From: Gebbie J
Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2012-61
Download: ML12215A013 (13)
v • d • e


Text

z INDIANA MICHIGAN Indiana Michigan Power R One Cook Place PIEM Bridgman, MI 49106 A unit ofAmerican Electric Power Indiana Mic higa nPower.com July 23, 2012 AEP-NRC-2012-61 Docket Nos.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Unit 1 Enforcement Discretion Regarding Engineered Safety Feature Actuation System Steam Line Isolation Automatic Actuation Logic and Actuation Relays for Steam Generator Stop Valve Dump Valve On July 19, 2012, Indiana Michigan Power the licensee for Donald C. Cook Nuclear Plant verbally requested the U. S. Nuclear Regulatory Commission (NRC) staff to grant enforcement discretion regarding the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Times required by Technical Specification (TS) 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation" Condition I. TS 3.3.2 Condition I applies when one train of TS Table 3.3.2-1 Function 4.b (Steam Line Isolation Automatic Actuation Logic and Actuation Relays) is inoperable and has not been restored within the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> specified by TS 3.3.2, Condition C. TS 3.3.2 Condition I, Required Action 1.1, requires that the unit be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and Required Action 1.2 requires that the unit be in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The request for enforcement discretion was made to preclude an unnecessary unit shutdown due to the failure of electrical control power to one of two dump valve trains for one of four steam generator stop valves (SGSVs). The dump valve control power failure occurred at 0734, July 19, 2012, as evidenced by control room indication. Investigation determined that fuses were blown for Train B 250 volt direct current control power to two dump valves, each associated with one of four SGSVs. With no control power, the associated dump valve cannot open and thereby close the associated SGSV. However, closure of the two SGSVs would still be achieved by operation of the second train of dump valves for each SGSV.

Control power was subsequently restored to one of the dump valves. However, more extensive work was needed to restore control power to the second dump valve. As a result, the dump valve control power failure existed more than the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowed by TS 3.3.2 Condition C, and the unit began a power reduction pursuant to TS 3.3.2 Condition I. Discussions with the NRC staff regarding enforcement discretion for the Required Actions of TS 3.3.2 Condition I were commenced at approximately 1530 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.82165e-4 months <br /> EDT.

U. S. Nuclear Regulatory Commission AEP-NRC-2012-61 Page 2 The enforcement discretion was requested to extend the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time specified by TS 3.3.2 Required Action 1.1, and the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time specified by TS 3.3.2 Required Action 1.2, by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> each, thereby providing additional time to complete repairs to the second dump valve while the unit remained in Mode 1. At approximately 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br /> the same day, the NRC staff verbally granted the requested enforcement discretion. Repairs to the control power to the second dump valve were completed and TS 3.3.2 Condition I was exited at approximately 2030 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.72415e-4 months <br />, July 19, 2012. The enclosure to this letter describes information discussed with representatives of the NRC in a teleconference on July 19, 2012. This letter fulfills the requirement to submit a written request for enforcement discretion within two working days of verbal approval by the NRC. In the July 19, 2012, teleconference, it was also determined that a follow-up license amendment was not required because the enforcement discretion will be in effect for a short duration and the action represents a one-time deferral with no permanent change to TS needed.

This letter contains no new or modified regulatory commitments. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President J RW/dmb

Enclosures:

Request for Enforcement Discretion Regarding Engineered Safety Feature Actuation System Steam Line Isolation Automatic Actuation Logic and Actuation Relays for Steam Generator Stop Valve Dump Valve.

c: C. A. Casto, NRC Region III J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosure MDEQ - WHMD/RPS NRC Resident Inspector P. S. Tam, NRC Washington, DC

Enclosure to AEP-NRC-2012-61 Request for Enforcement Discretion Regarding Engineered Safety Feature Actuation System Steam Line Isolation Automatic Actuation Logic and Actuation Relays for Steam Generator Stop Valve Dump Valve This enclosure describes information discussed in a teleconference between Indiana Michigan Power (I&M) representatives and U. S. Nuclear Regulatory Commission (NRC) staff members on July 19, 2012, regarding the subject request for enforcement discretion.

I&M reviewed NRC Regulatory Issue Summary 2005-01, "Changes to Notice of Enforcement Discretion (NOED) Process and Staff Guidance", and the accompanying NRC Inspection Manual Part 9900 Technical Guidance, "Operations - Notices of Enforcement Discretion." I&M concluded that Part 9900, Section B.2.1, "Situations Affecting Radiological Safety - Regular NOEDs," Criterion 2 was satisfied. This criterion applies to a plant in power operation for which an NOED is requested to avoid unnecessary transients as a result of compliance with the license and, thus, minimize potential safety consequences and operational risks. The basis for this conclusion and other information required to support a request for NOED is provided below.

1. The Technical Specification or other license condition that will be violated.

The Technical Specifications (TS) that will be violated are the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Times for TS 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation" Required Actions 1.1 and 1.2 which apply when one train of TS Table 3.3.2-1 Function 4.b (Steam Line Isolation Automatic Actuation Logic and Actuation Relays) is inoperable and has not been restored within the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> specified by TS 3.3.2, Condition C, "One train inoperable."

I&M, the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1, requests NRC regional enforcement discretion such that the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> TS 3.3.2, Required Action 1.1 Completion Time for placing the unit in Mode 3 be extended to 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and the Required Action 1.2 Completion Time for placing the unit in Mode 4 be extended to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

I&M is requesting the enforcement discretion to provide additional time to correct the condition that caused blown fuses on the Unit 1 250 volt direct current (VDC) control power to a Train B dump valve for one Steam Generator Stop Valve (SGSV) and restore the dump valve to operability.

2. Circumstances surroundinqthe situation:including likely causes; the need for prompt action:

action taken in an attempt to avoid the need for an NOED: and identification of any relevant historicalevents.

Enclosure to AEP-NRC-2012-61 Page 2

System Description

Each of the four Steam Generators has an SGSV located downstream of the safety valves.

The SGSVs are parallel slide gate valves and are capable of closing rapidly in the event of a main steam line rupture occurring anywhere in the piping between the steam generator and turbine. The SGSVs are designed to close against flow in either the normal or reverse direction to limit the effect of a steam line rupture to the blowdown of the one affected steam generator; assuming, conservatively, the failure of one of the four valves to close.

The SGSV design incorporates a piston, which is attached to the valve stem. The steam above and below the piston is normally at main steam line pressure. The cylinder volume above the piston is piped through a three-way valve into a pair of redundant, air-operated dump valves. Upon receipt of a signal for the SGSVs to close, the dump valves open and vent the steam from the cylinder. The steam pressure in the valve body below the piston forces the piston to move rapidly and close the valve. The dump valves, which open to vent the steam from the SGSV are air operated and depend on a solenoid being energized to open the dump valve which in turn causes the stop valve to go closed.

Protective logic is supplied to isolate all steam generators by rapid closure of the four stop valves for any of the following conditions:

a) Containment spray actuation signal initiation Hi-Hi pressure.

b) High steam flow coincident with Lo/Lo Taverage.

c) Steam line pressure low.

d) In addition, emergency closure can be initiated by operator actuation of the dump valves in the steam generator stop valve control system.

Circumstances Surrounding the Situation At 0734, July 19, 2012, a control room annunciator for SGSV dump valve failure actuated, and indicating lights for 1-MRV-212 and 1-MRV-222 extinguished simultaneously. These are the dump valves for SGSVs 1-MRV-210 and 1-MRV-220, respectively. Investigation determined that fuses were blown for Train B 250VDC control power to the solenoid for 1-MRV-212 and 1-MRV-222. With no 25OVDC control power, the dump valve solenoid valves could not have opened to allow the air operated dump valves to open and vent the steam from the cylinder, and thereby close the associated SGSV. However, closure of SGSVs 1-MRV-210 and 1-MRV-220 would still have been achieved by operation of dump valves 1-MRV-211 and 1-MRV-221, since their 250VDC control power is provided by Train A.

The fuse for 1-MRV-212 was replaced following visual inspection of the associated solenoid, and the valve was tested with acceptable results. The probable cause of the fuse blowing was considered to be a voltage spike resulting from a ground on circuitry related to 1-MRV-222.

Enclosure to AEP-NRC-2012-61 Page 3 When the fuse for 1-MRV-222 was replaced, the replacement fuse also blew. A ground was detected, but it was not known if the ground was in the position indication or valve actuation circuitry. Activities to identify the specific grounded component were in progress at the time of the I&M - NRC teleconference.

Need For Prompt Action The inoperability of the Train B dump valves 1-MRV-212 and 1-MRV-222 for SGSVs 1-MRV-210 and 1-MRV-220 resulted in entry into TS 3.3.2 Condition C. The Required Action for Condition C is that the inoperable train must be restored to operable status within a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time. Although 1-MRV-212 was restored to operable status, 1-MRV-222 remained inoperable when the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> required Completion Time limit was reached at 1334, July 19, 2012. This resulted in entry into TS 3.3.2 Condition I, which requires that the unit be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The investigation into the cause of the fuses blowing in 1-MRV-222 indicated that there was a ground in the cable or the solenoid. It was determined that the ground was not in the solenoid. Based on previous cable replacement activities, replacement of all or part of the 180 foot length of cable was estimated to take as long as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> specified by TS 3.3.2 Condition 1.1 for reaching Mode 3. Therefore, prompt action was needed to allow the unit to remain in Mode 1 while repairs were made to eliminate the ground, including the potential replacement of all or part of the cable.

Subsequent to the teleconference, it was determined that the ground was likely in the splice between the cable and the solenoid, since remaking this splice eliminated the ground and 1-MRV-222 was declared operable at 2030 July 19, 2012. However, enforcement discretion was requested based on the potential need to replace the entire cable.

Action Taken In an Attempt to Avoid the Need For An NOED Upon receipt of the alarms and loss of indication at 0734, CNP personnel immediately began efforts to determine the cause and correct the condition. This included prompt formation of a multidiscipline failure investigation team, with management oversight.

Identification of Any Relevant Historical Events.

A search was performed of condition report data extending back to 1995 for relevant information relating to steam dump valves or ESF actuations. No condition reports were identified that documented conditions similar to those known to be involved with the inoperability of 1-MRV-212 and 1-MRV-222.

3. Information to show that the cause and proposed path to resolve the situation are understood by the licensee, such that there is high likelihood that planned actions to resolve the situation can be completed within the proposed NOED time frame.

As described in the response to Item 2 above, the investigation into the cause of the fuses blowing in 1-MRV-222 indicated that there was a ground in the cable or the solenoid. It was

Enclosure to AEP-NRC-2012-61 Page 4 subsequently determined that the ground was not in the solenoid. Therefore, I&M had to plan for the worst case scenario, which included replacing all or part of the 180 foot cable.

I&M also planned to simultaneously install a temporary cable in case unexpected difficulties were encountered replacing all or part of the existing cable. Based on previous experience, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension assured that there was high likelihood that planned actions to resolve the situation could be completed within the proposed NOED time frame.

4. Safety basis for the request, including an evaluation of the safety siqnificance and potential consequences of the proposedcourse of action and the qualitative and quantitative aspects.

NRC Regulatory Information Summary 2005-01 section 5 provides guidance for demonstrating a condition represents an acceptable level of risk by asking seven specific questions (section 5, a - g.) The probabilistic risk assessment (PRA) response to these questions is provided below:

a. Use the zero maintenance PRA model to establish the plant's baseline risk and the estimated risk increase associated with the period of enforcement discretion. For the plant-specific configuration the plant intends to operate in during the period of enforcement discretion, the incremental conditional core damage probability (ICCDP) and incrementalconditional large early release probability(ICLERP) should be quantified and compared with guidance thresholds of less than or equal to an ICCDP of 5E-7 and an ICLERP of 5E-8. These numerical guidance values are not pass-fail criteria.

Response

The zero maintenance PRA model for CNP Unit 1 was used to establish the plant's baseline risk. No other maintenance or equipment unavailability was planned for the same time period as the 1-MRV-222 repair. The zero-maintenance Mode 1, full power, core damage frequency (CDF) and large early release frequency (LERF) values for Unit 1 are 1.14E-5/yr and 2.39E-6/yr. The ICCDP and ICLERP due to a total of 1 day of 1-MRV-212 and 1-MRV-222 unavailability was calculated to be:

Unit 1: ICCDP is 3.86E-8, and ICLERP is 7.37E-9.

Note the assumption that both 1-MRV-212 and 1-MRV-222 were unavailable was conservative in that 1-MRV-212 had been restored to operable status at the time of the teleconference.

During the time frame for this work, no risk-significant equipment was allowed to be electively taken out-of-service during the enforcement discretion period. Since the plant configuration risk during the 1-MRV-222 unavailability period was effectively represented by the calculated ICCDP and ICLERP values shown above, which are less than the Maintenance Rule (a)(4) ICCDP and ICLERP threshold values for Green (i.e., 1E-6 and 1E-7, respectively), the plant MR(a)(4) risk profile was projected to be Green.

b. Discuss the dominant risk contributors (cut sets/sequences) and summarize the risk insights for the plant-specific configuration the plant intends to operate in during the

Enclosure to AEP-NRC-2012-61 Page 5 period of enforcement discretion. This discussion should focus primarily on risk contributors that have changed (increasedor decreased)from the baseline model as a result of the degraded condition and resultant compensatory measures.

Response

Unit 1 Dominant Risk Contributors The dominant risk contributors were determined from the top 20 CDF cutsets associated with the plant-specific configuration (1-MRV-212 and 1-MRV-222 out of service, other dump valves' failure probabilities increased to 0.08, and alignments indicated in the assumptions).

Initiating No. Frequency Event Event Event Event 1 5.44E-6 IE-SLB-5 1E-AV--MRV211FO 1E-AV--MRV221FO 2 1.34E-6 IE-SLB-1 1E-AV--MRV211FO 1E-AV--MRV221FO 3 1.34E-6 IE-SLB-2 1E-AV--MRV211FO 1E-AV--MRV221FO 4 7.26E-7 IE-SLO 1K---INJECTNHE2 5 5.51E-7 IE-ESW4 BPM-RUNFR-CCF1-4 ESW-CC-----I E ESW-PUMP-RECOV 6 5.17E-7 IE-SLO IRH108FO--CCF12 7 5.OOE-7 IE-SLO Q-CC-SI-SIGNAL 8 4.35E-7 IE-SLB-5 1E-AV--MRV211FO 1E-AV--MRV231FO 1E-AV--MRV232FO 9 4.35E-7 IE-SLB-5 1E-AV--MRV221FO 1E-AV--MRV241FO 1E-AV--MRV242FO 10 4.35E-7 IE-SLB-5 1E-AV--MRV211FO 1E-AV--MRV241FO 1E-AV--MRV242FO 11 4.35E-7 IE-SLB-5 1E-AV--MRV221FO 1E-AV--MRV231FO 1E-AV--MRV232FO 12 4.24E-7 IE-ISL1 1------ DYNAMHE 13 3.75E-7 IE-SLO 1Y---RECIRCCGHE2 14 3.43E-7 IE-SLO IPP35PR---CCF12 15 3.10E-7 IE-TRA 1---SHUTDOWNHE 1P-CD-SCRAM-ECM 16 3.OOE-7 IE-VEF 17 2.58E-7 IESGTR-1 1------.-.- E3HE 1 -------

E3-CDHE SSV-SUCCESS 18 2.58E-7 IE-SGTR-2 1--------- E3HE 1-------

E3-CDHE SSV-SUCCESS 19 2.58E-7 IE-SGTR-3 1--------- E3HE 1------E3-CDHE SSV-SUCCESS 20 2.58E-7 IE-SGTR-4 1---------- E3HE 1-------

E3-CDHE SSV-SUCCESS As evident by comparing the top 20 cutset listing for the MRV-unavailable case with a similar listing for the Unit 1 base case, cutsets 1 through 3, and 8 through 11 are sequences (for a total of 7 of the top 20 cutsets, that account for about 70% of the ACDF associated with this MRV unavailability) with increased importance for the plant configuration. Sequences 1 - 3 and 8 - 11 involve failures that disable the remaining steam dump valves on SGs 1 and 2 and these scenarios are assumed to proceed directly to core damage.

Enclosure to AEP-NRC-2012-61 Page 6 Initiating Event Risk Insight Review of the impact of the 1-MRV-222 repair based on Initiating Events showed there is no direct impact on initiating event frequencies. Common cause failure of the remaining SGSVs or loss of power supplies to the remaining SGSVs have made Main Steam Line Break sequences more important.

c. Explain compensatory measures that will be taken to reduce the risk associated with the specified configuration. Compensatory measures to reduce plant vulnerabilities should focus on both event mitigation and initiating event likelihood. The objectives are to:
i. reduce the likelihood of initiating events Risk mitigating actions (RMAs) taken to reduce the likelihood of initiating events are as follows:
  • No other work that jeopardizes plant operation, such as alignment changes (except in response to emergent plant equipment failures) or balance-of-plant function testing, or significant switchyard work will be allowed. Thus, the potential for a unit trip and the related electrical bus transfers will be reduced for items under plant control.
  • To the extent practicable and controllable, no other work will be undertaken that could jeopardize unit operation. For example, main turbine valve testing or similar activities, or maintenance work on balance-of-plant components that have potential to initiate a unit trip, is avoided while the repair is in progress. No RPS testing or maintenance will be performed.
  • The other SGSVs, dump valves, AB and CD Battery rooms and chargers, CCV-CD 250VDC distribution cabinet (i.e., Train A), EDGs, SDGs and switchyard will be treated as "guarded" in accordance with plant procedures. No maintenance will be performed on DC systems except that needed to accomplish repairs to the Train B 250 VDC control power supply.

Enclosure to AEP-NRC-2012-61 Page 7 ii. reduce the likelihood of unavailabilityof trainsredundant to the equipment that is out-of-service during the period of enforcement discretion RMAs taken to reduce the likelihood of unavailability of trains redundant to the equipment that is out-of-service are as follows:

  • No other PRA related equipment maintenance is planned.
  • The unit's Switchyards shall have access prevented, except as needed in response to emergent failures or conditions that develop (i.e., no elective switchyard work).

" No maintenance will be performed on the operable dump valves or associated power supplies.

  • No maintenance (other than data gathering for required surveillances) will be performed on DC systems except that needed to accomplish repairs to the Train B 250 VDC control power supply.

iii. increase the likelihood of successful operator recovery actions in response to initiatingevents CNP has emergency operating procedures that require verification of SGSV closure when needed for initiating events such as a steam generator tube rupture, secondary system pipe break, or reactor trip.

d. Discuss how the proposed compensatory measures are accounted for in the PRA.

These modeled compensatory measures should be correlated, as applicable, to the dominant PRA sequences identified in item b. above. In addition, other measures not directly related to the equipment out-of-service may also be implemented to reduce overall ,plant risk and, as such, should be explained. Compensatory measures that cannot be modeled in the PRA should be assessedqualitatively.

Response

None of the compensatory measures (RMAs) discussed above, were credited in the PRA analysis. The compensatory measures were additional risk management actions to

Enclosure to AEP-NRC-2012-61 Page 8 keep plant focus on the 1-MRV-222 repair, and assure that any initiating events (to the extent practicable) are avoided.

e. Discuss the extent of condition of the failed or unavailable component(s) to other trains/divisionsof equipment and what adjustments, if any, to the related PRA common cause factors have been made to account for. potential increases in their failure probabilities. The method used to determine the extent of condition should be discussed. It is recognized that a formal root cause or apparent cause is not required given the limited time available in determining acceptability of a proposed NOED.

However, a discussion of the likely cause should be provided with an associated discussion of the potential for common cause failure.

Response

The I&M failure investigation team identified the fault as being in the dump valve solenoid, associated cabling, or spliced connection. This extent of condition is supported by the following:

  • No other control room alarms or indications were received at the time of the failure.
  • Investigation results determined the 1-MRV-212 circuit and 1-MRV-222 circuit to be electrically isolated from each other.
  • Following confirmation of electrical separation, the 1-MRV-212 circuit was re-energized by replacing fuses. Subsequent testing provided reasonable assurance of the circuit being operable and lead to successful completion of the applicable surveillance to return 1-MRV-212 to service.
  • Analyzed field data from the 1-MRV-222 circuit provided strong indication of a fault within the circuitry.

" Isolation of the 1-MRV-222 solenoid, its cabling, and splice from the 1-MRV-222 circuit eliminated indication of a fault.

  • Replacement of the solenoid and rebuilding of the splice eliminated the fault indications.

Based on the above, the team concluded with reasonable assurance that the extent of condition had been properly bounded and the appropriate actions were in place to be successful in resolution. This was supported by the successful return of 1-MRV-222 to operation.

The PRA modeling documented in this enclosure was performed prior to the above determinations and were performed in a conservative manner to encompass possible situations. Specifically, both 1-MRV-212 and 1-MRV-222 were considered unavailable, even though 1-MRV-212 had successfully been returned to service. In addition, a common cause failure of the other similar SGSV steam dump valves (i.e., 1-MRV-211, 1-MRV-221, 1-MRV-231, 1-MRV-232, 1-MRV-241, and 1-MRV-242) from a similar cause was considered possible, and the failure probabilities for these valves were increased from 2.015E-3 to 0.08.

Enclosure to AEP-NRC-2012-61 Page 9 f Discuss external event risks for the specified plant configuration. An example of external event risk is a situation where a reactor core isolation cooling (RCIC) pump has failed and a review of the licensee's Individual Plant Examination of External Events or full-scope PRA model identifies that the RCIC pump is used to mitigate certain fire scenarios. Action may be taken to reduce fire ignition frequency in the affected areas or reduce human error associated with time-critical operator actions in response to such scenarios.

Response

There are no specific external event risks (seismic, fire, or flood) that are related to this failure of 1-MRV-222 that has led to this request.

The existing weather in the area was not expected to contribute to the potential for fire in the protected areas surrounding the plant, its switchyards, and the associated local high voltage (345 & 765 kiloVolt) transmission lines.

Seismic events are of extremely low probability and the 1-MRV-222 repair activities had no impact on seismic capability of any other related plant equipment.

g. Discuss forecasted weather conditions for the NOED period and any plant vulnerabilities related to weather conditions.

Response

The forecasted weather was for scattered showers and thunderstorms, partly sunny, with a high temperature near 81 degrees Fahrenheit. The wind was forecast as northwest, about 10 mph. The chance of precipitation was forecast as 30%.

If a forecast for severe weather occurred, the resulting ICLERP values would accommodate 31.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> of severe weather duration before plant risk exceeded the RIS 2005-01 risk status threshold.

5. The iustification for the duration of the noncomnpliance.

As described in the response to Item 2 above, the investigation into the cause of the fuses blowing in 1-MRV-222 indicated that it may have been necessary to replace all or part of the 180 foot length of cable. Based on previous cable replacement activities, replacement of all or part of the 180 foot length of cable was estimated to take as long as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> beyond the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> specified by TS 3.3.2 Condition 1.1 for reaching Mode 3. The enforcement discretion was requested based on the potential need to replace the entire cable.

Enclosure to AEP-NRC-2012-61 Page 10

6. The condition and operationalstatus of the plant (including safety related equipment out of service or otherwise inoperable).

At the time of the request for enforcement discretion, Unit 1 was decreasing power from 100% due to entry into TS 3.3.2 Condition I. There was no other Unit 1 safety related equipment inoperable. Unit 2 was operating at nominal power of 100%.

7. The status and potential challenges to off-site and on-site power sources.

In accordance with normal operating configuration, the Unit 1 electrical loads were being powered from the main generator via the Unit Auxiliary Transformers, and offsite power was available via a fast transfer to the Reserve Auxiliary Transformers. The grid was verified to be stable with no alerts or warnings. The EDGs, SDGs, and safety related batteries were operable.

8. The basis for the licensee's conclusion that the noncompliance will not be of potential detriment to the public health and safety.

A risk assessment has determined there is no net increase in radiological risk to the public.

The requested enforcement discretion does not increase the probability of a plant transient.

The compensatory actions described above will be implemented to further minimize risk including the protection of key systems and deferring production activities that would increase risk.

9. The basis for the licensee's conclusion that the noncompliance will not involve adverse consequences to the environment.

The proposed enforcement discretion does not involve any adverse consequences to the environment. As noted above, the proposed action does not represent a potential detriment to the public health and safety. A bounding risk assessment determined that the calculated risk is consistent with the site's normal work control levels, and therefore there is no net increase in radiological risk to the public. In addition, there is no significant change in the types, or a significant increase in the amounts, of any effluent that may be released offsite, since the proposed actions do not affect the generation of any radioactive effluent nor do they affect any of the permitted release paths. Finally, there is no significant increase in individual or cumulative occupational radiation exposure. The actions proposed in this request for enforcement discretion will not significantly affect plant radiation levels, and therefore do not significantly affect dose rates and occupational exposure.

Enclosure to AEP-NRC-2012-61 Page 11

10. A statement that the request has been approved by the facility orqanization that normally reviews safety issues (Plant On-site Review Committee or its equivalent).

This request for enforcement discretion was reviewed and approved by the Plant Operations Review Committee on July 19, 2012.

11. The specific NOED criteria for plant conditions specified in Section B (NRC Inspection Manual Part 9900: Technical Guidance) that is satisfied and how it is satisfied.

This request is for a "regular" NOED for plant conditions that satisfy the criterion set forth in NRC Inspection Manual, Part 9900, paragraph B.2.1.1.a. This criterion states that a NOED is intended to avoid unnecessary transients as a result of compliance with the license condition and, thus, minimize potential safety consequences and operational risks. This criterion is satisfied in that the requested enforcement discretion will preclude the unnecessary shutdown of Unit 1 as a result of compliance with TS 3.3.2, thereby minimizing the potential safety consequences and operational risks resulting from a shutdown.

12. Commitment for submittal of the written NOED within 2 workinq days of verbal approvaland follow-up amendment within 4 workinq days of verbal approval I&M stated that it would submit a written request for enforcement discretion within two working days of NRC approval. This letter satisfies that commitment.

In the July 19, 2012, teleconference between NRC staff and CNP staff, it was determined that a follow-up license amendment was not required based on the following:

  • This action represents a one-time deferral with no permanent change to TS needed.
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