Text

Emergency License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating
	ML16287A609
Person / Time
Site:	Cook
Issue date:	10/11/2016
From:	Lies Q; AEP Indiana Michigan Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16287A615	List: AEP-NRC-2016-83, Emergency License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating; ML16287A610;
References
	AEP-NRC-2016-83
	Download: ML16287A609 (60)
	v • d • e

=

INDIANA MICHIGAN POWER.

A unit ofAmerican Electric Power Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 lndianaMichiganPower.com October 11, 2016 AEP-NRC-2016-83 10 CFR 50.91 (a)(5)

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 Emergency License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant Unit 1, proposes to amend Facility Operating License DPR-58. l&M proposes a one-time extension of the Completion Time for Technical Specification (TS) 3.8.1, "AC Sources - Operating." The proposed amendment is being requested on an emergency basis pursuant to 10 CFR 50.91 (a)(5).

TS 3.8.1 Required Action A.3 requires that, with the unit in Modes 1 through 4, one inoperable required offsite circuit be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the inoperable offsite circuit is not restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , TS 3.8.1 Condition G requires that the unit be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Mode 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . As detailed in Enclosure 2 to this letter, on October?, 2016, at approximately 2315 hours0.0268 days 0.643 hours 0.00383 weeks 8.808575e-4 months , a Taylor Machine Works Forklift was traveling along the plant service road outside of the protected area, the left rear wheel broke through a section of a below grade Plastibeton cable raceway. This raceway contains the 34.5Kv cables that feed the Train 'B' reserve feed to both Unit 1 and Unit 2. There was no load on the forks at the time and there was no obvious damage to the cables and no injuries. However, there is the likelihood that as a result of the tire penetrating the Plastibeton and coming to rest on the cables, that once the forklift and failed Plastibeton are removed, there will be some cable damage that needs to be repaired.

The best estimate schedule for deenergizing the run of 34.5Kv cable, removing the forklift, repairing damaged cables, and performing post-maintenance work is 71 hours8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months . Based on the uncertainty associated with these time estimates it is likely that this will put actions to complete the repairs over the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months TS 3.8.1 Required Action A.3 Completion Time. Based on the l&M time estimates for replacement of the cable, and the fact that this will be a first time evolution, l&M is requesting an additional 28 hours3.240741e-4 days 0.00778 hours 4.62963e-5 weeks 1.0654e-5 months , making the completion time 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . to this letter provides an affirmation affidavit pertaining to the proposed amendment. provides a detailed description and safety analysis to support the proposed amendment, including justification for approving the amendment on an emergency basis, an evaluation of significant hazards considerations pursuant to 10 CFR 50.92(c), and an environmental assessment. Enclosure 3 provides Probabilistic Risk Assessment (PRA) Technical Adequacy.

~DD\

. tJ~-~

U. S. Nuclear Regulatory Commission AEP-NRC-2016-83 Page 2 provides PRA calculation, Calculation PRA-QNT-007, Calculation of Regulatory Guide 1.177 Risk Parameters for Potential One-Time Emergency Technical Specification Completion Time Change for Unit 1 and Unit 2 Train B Reserve Feed. Enclosure 5 provides the applicable license page marked to the proposed change. Enclosure 6 provides the license page with the proposed changes incorporated.

The activities to address the identified condition are currently scheduled to start at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months on October 12, 2016. The start of these activities requires entry into TS 3.8.1, the A. 3 Required Action which the associated completion time is being requested to be changed by this license amendment request. l&M requests approval of the proposed amendment by 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on October 14, 2016, to allow for the . timely correction of the condition and preclude an unnecessary shutdown of Unit 1.

Copies of this letter and its attachments are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.

There are no new regulatory commitments in this letter. Should you have any questions, please contact Mr. Michael Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely,

?~J. Ea, Q(Shane Lies

¢ite Vice President DB/rdw

Enclosures:

1. Affirmation.

2. License Amendment Request for One-Time Extension of Completion Time for an Inoperable AC Source - Operating

3. Probabilistic Risk Assessment Technical Adequacy

4. Calculation PRA-QNT-007, Calculation of Regulatory Guide 1.177 Risk Parameters for Potential One-Time Emergency Technical Specification Completion Time Change for Unit 1 and Unit 2 Train B Reserve Feed

5. License Page - Marked to Show Proposed Change

6. License Page - Changes Incorporated c: R. J. Ancona - MPSC A. Dietrich, NRC Washington, D.C.

MDEQ- RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region Iii A. J. Williamson - AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2016-83 AFFIRMATION I, Q. Shane Lies, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (l&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of l&M, and that the statements made and the matters set forth herein pertaining to l&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company i::::1~

Site Vice President SW.ORN TO AND SUBSCRIBED BEFORE ME DANIELLE BURGOVN~ .

DAYOF ~~ ,2016 Notary Public, State o! M1ch1gan THIS \\ County of Be men My commission Expires otg4~:<:1 .,.~

8

~ .

,,,:,S)\h~\~,~~

. Notary Pu c Acting In the county of~ ~

My Commissio_n Expires D~ - C'.>'.\ ~ ~~ \ t

Enclosure 2 to AEP-NRC-2016-83 LICENSEAMENDMENT REQUEST FOR ONE-TIME EXTENSION OF COMPLETION TIME FOR AN INOPERABLE AC SOURCE - OPERATING

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1, proposes to amend Facility Operating License DPR-58. l&M proposes a one-time extension of the Technical Specification (TS) Completion Time for an inoperable AC Electrical Source. The proposed amendment is being requested on an emergency basis pursuant to 10 CFR 50.91 (a)(5).

2.0 PROPOSED CHANGE

The proposed change would add the following Footnote to the 72-hour Completion Time for TS 3.8.1 Condition A.3: *

"For Train 8 only, the Completion Time that Train 8 can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ," to support repair and restoration of the Train 8 Reserve Feed. Upon completion of the repair and restoration associated with this event which occurred on October 7, 2016, this footnote is no longer applicable."

3.0 BACKGROUND

AC Electrical Power Distribution System Description As shown on the attached sketches, the. onsite alternating current (AC) electric power*

distribution system for each unit contains four, 4160V (4.16 kV) non-safety-related buses designated 1A, 18, 1C, and 1D for Unit 1 and 2A, 28, 2C, and 20 for Unit 2. These buses are referred to as the RCP" buses because they power the reactor coolant pumps. Each of the non-safety-related RCP buses feed a downstream safety-related 4.16 kV bus. These safety-related buses are designated T11A, T118, T11 C, and T11 D for Unit 1 and T21A, T218, T21 C, and T21 D for Unit 2. These buses are referred to as the "T" buses. With the main generator on-line, the RCP buses are normally fed from the Unit Auxiliary Transformers (UATs), which receive power from the main generator.

Upon a trip of the main generator, the station auxiliaries are automatically fast transferred to the preferred offsite power source (i.e., to reserve auxiliary transformers {RATs) TR101AB and TR101CD for Unit 1 and TR201A8 and TR201CD for Unit 2) to assure continued power to equipmentwhen the main generator is off-line. The ESF Loads are sequenced onto the RATs, under accident conditions, using the same timing relays and sequence as used for the EOG sequencing. The RATs supply the reserve auxiliary power for both units.

The preferred offsite power source for both units can be arranged so that transformer No. 4 or transformer No. 9 can supply reserve auxiliary transformers (RATs) TR101CD and TR201CD to AEP-NRC-2016-83 Page 2.

and transformer No. 5, or transformer No. 9 supplies TR101AB and TR201AB. Under certain plant conditions, it is possible for transformer No. 4, transformer No. 5, or transformer No. 9 to feed the entire plant auxiliary load.

The other qualified circuit required to be operable by TS LCO 3.8.1.a is the alternate offsite circuit. The alternate qualified offsite circuit consists of a 69/4.16 kV transformer (TR12EP-1),

the cabling and switches to a 4.16 kV bus, designated as the EP Bus, which supplies breakers 1EP and 2EP, and the cabling, switches, and breakers to the T buses. Connection of the T buses to transformer 12EP1 requires manual switch operations in the control room. The alternate offsite power source has the necessary capacity to operate one train of the engineered safeguard equipment in one unit while supplying one train of the safe shutdown power in the other.. The T buses can also be powered from the emergency diesel generators (EDGs). TS LCO 3.8.1.a requires two qualified *circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System.

An additional independent onsite standby AC power source consisting of two supplemental diesel generators (SDGs) is provided to automatically supply power to the EP bus, which is normally supplied by the alternate qualified offsite circuit and can be manually aligned to directly supply the T buses.

to AEP-NRC-2016-83 Page 3 SKETCH 1 LEGEND 78SKV-345KV-69KV-34.5KV-28KV -

12KV-4KV -

800V -

~J h!ilR 201CD 2C> W& '1210 4 KV Vlt8i But Villil~

.,.. '111111!1.C '1111~ .,..Tlt21CMO to AEP-NRC-2016-83 Page 4 SKETCH 2 12EP1_Lr:--------

_ From Supplementa 4 kV EP Bus Diesel Generaors (See Sketch 1)

~EPI (:EPI ,,,--------------------. (See Sketch 1)

F'~ F'~ I I F'~.m

crA9 I1ABD...,IC*r*1 *r(:1I1CDO<<el. , ""i'° RCP RCP Frnm RCP Frnm RCP 1 ~12 RCP Li2A . ""i'" TckV Frnm RCP Frnm RCP c.. I 2ABD~ IC*1 ccrn ~12 Frnm RCP Bur

~12 Q,c., (:2Q,~1 ~12 Q,Cs2 (c2 Q,~1 1 84

.................~ \11~

r11A

~

r11c ~3 G a

_........ r 110

~"""'

T2 1c ~ 3 Ga T21D to AEP-NRC-2016-83 Page 5 Description of Events On October 7, 2016, at approximately 2315 hours0.0268 days 0.643 hours 0.00383 weeks 8.808575e-4 months , as a Taylor Machine Works Forklift was being driven along the plant service road outside of the protected area, the left rear wheel broke through a below grade section of a Plastibeton cable raceway. This raceway is situated along the shoulder of the service road and contains the 34.5Kv cables that are immediately downstream of the Circuit Breaker "12 AB", which is the common high side feed to the Unit 1 and Unit 2 RATs (TR101AB and TR201AB) (notated by a large star on Sketch 1). There was no load on the forks at the time. There is no indication of cable damage based on visual inspection and plant parameters and there were no injuries.

In the intervening time since the incident, l&M has been investigating to determine the cause of the event, evaluating the damage, preparing repair planning estimates, obtaining replacement parts, ensuring vendor resources, and conducting a risk analysis. Detailed scheduling of the replacement activities indicated that the cable could be replaced and the new cable declared operable within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

Cable repair and replacement activities are scheduled for 71 hours8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months . However, because the actual condition of the cable is unknown and the fact that this is a first-time evolution, the repair time could take longer.

TS 3.8.1 Required Action A.3 requires that one inoperable required offsite circuit be restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the inoperable offsite circuit is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , TS 3.8.1, Condition G requires that the unit be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Reason the Amendment is Requested on an Emergency Basis Regulation 10 CFR 50.91 (a)(5) states that where the U. S. Nuclear Regulatory Commission (NRC) finds that an emergency situation exists, in that failure to act in a timely way would result in derating or shutdown of a nuclear power plant, or in prevention of either resumption of operation, or of increase in power output up to the plant's licensed power level, it may issue a license amendment involving no significant hazards consideration without prior notice and opportunity for a hearing or for public comment. The regulation requires that a licensee requesting an emergency amendment explain why the emergency situation occurred and why the licensee could not avoid the situation. As explained below, an emergency amendment is needed to effect timely resolution of the issue and to preclude an unnecessary plant shutdown, and l&M could not have reasonably avoided the situation or made timely application for an amendment.

Reason Emergency Situation Has Occurred The emergency situation was initiated by the Taylor Machine Works Forklift unexpectedly dropping through the Plastibeton cable raceway onto the 34.5Kv cables. Detailed scheduling, and -accounting for uncertainties, indicates that the cable replacement could be accomplished within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , which is greater than the TS 3.8.1 Condition A.3 Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

l&M estimates that completion of cable replacement would likely exceed the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time of TS 3.8.1 Required Action A.3 if repair of all 3 phases of 34.5kV cable is to AEP-NRC-2016-83 Page 6 determined necessary, which would require that the unit be shutdown. Neither a routine nor an exigent amendment would allow for a timely resolution of the issue. Therefore, an emergency amendment is requested to effect repairs in a timely manner and to preclude a shutdown.

Reason the Situation Could Not Have Been Avoided The failure of the Plastibeton cable raceway could not have been foreseen. The Plastibeton cable raceway is designed to be installed below grade and the forklift had been successful in routinely driving this path. Any complications in replacing damaged cable also cannot be reasonably foreseen, as this will be an infrequent evolution, if it becomes necessary, for l&M to splice these 34.5Kv cables. Due to the uncertainty of the repairs that will be required, there are three restoration schedules. The worst case* schedule with a margin for uncertainties may require 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for completion. Personnel are aware of the activities and durations needed to accomplish a routine replacement.

The condition of the cable won't be known until the forklift tire and damaged Plastibeton pieces are removed to reveal the cable. Once the .cable in the affected section of the cable raceway is revealed, an assessment for any damage can be conducted. Visual inspection and tan-delta testing will be conducted, which could lead to three possible scenarios; no damage, minor damage to the cable outer jacket, or significant damage requiring the cable(s) to be replaced ..

The three possible outcomes and their time frames are described below, and the time estimates ,

are for known activities. The time estimates do not reflect contingency activities:

Visual inspection and Tan-delta cable diagnostic testing shows no repair is needed for the cable and the raceway will be repaired - this evolution is scheduled for 27 hours3.125e-4 days 0.0075 hours 4.464286e-5 weeks 1.02735e-5 months .

Visual inspection and Tan-delta testing indicates damage to outer jacket and the need to perform a repair of the outer jacket. The time estimate for this scenario is 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months .

On<;;e the damage* on the outer jacket has been repaired, Tan-delta testing will be performed to ensure cable integrity. Tan-delta testing could indicate the need to replace the cable.

Either the visual inspection or tan-delta testing leads to removal of the damaged cable sections for all three phases, replacing the damaged section(s), and performing up to six splices. After the splicing is complete a second Tan-delta test will be performed to verify the integrity of the "repaired" cable. This scenario is estimated to require 71 hours8.217593e-4 days 0.0197 hours 1.173942e-4 weeks 2.70155e-5 months for completion.

The worst case scenario, occurring if the visual inspection or the Tan-delta testing leads us to cable replacement and splicing, requires three steps in the project schedule: 1) cut out and remove the damaged cable for all three phases, 2) replace all three cables and splice in the new cables, 3) then retest with Tan-delta testing to verify cable integrity with the new "spliced" cables.

The reason for requesting 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for completion is due to the unknown nature of any delays that may occur in the planned activities. A few examples of the delays that could be experienced are: adverse weather, tooling issues, repair equipment failures, stuck material that to AEP-NRC-2016-83 Page 7 reveals itself to be excessively difficult to remove from the raceway. A delay greater than one hour of any type would therefore cause the exceedance of the 72-hour action statement.

Additionally, the language of the footnote expires the action statement at completion of any repairs and declaration of operability such that the entire 100-hour Completion Time would not be utilized if not required.

Considering that the cable was energized at the time of the event and power was never lost, interrupted, or showed any indication of degradation based on observable plant parameters, there is reasonable assurance that the cable is currently intact. However, because the level of damage, if any, is unknown, the extent of repair, if any, required cannot be ascertained until the forklift and damaged material are removed from the cable raceway.

l&M therefore considers that there is sufficient justification for requesting the proposed license amendment on an emergency basis.

4.0 TECHNICAL ANALYSIS

The proposed amendment to allow a one-time extension of the TS 3.8.1 Condition A.3 Completion Time for the current inoperability of the Unit 1 AC Power Source is supported by a quantitative risk evaluation.

4.1 Tier 1: Probabilistic Risk Assessment (PRA) Capability and Insights Technical Adequacy of the PRA model A full discussion and justification of the PRA technical adequacy is provided in Enclosure 3.

The CNP PRA model is generally robust and suitable to support this amendment. Specific issues identified by the recent peer review that may have a significant impact on the model have been addressed to reduce or eliminate their impact on the results. The ongoing PRA maintenance and update activities associated with the CNP PRA program ensure that the PRA models represent the as-built, as-operated plant moving forward. Therefore, the CNP PRA model has the technical adequacy required to support the amendment.

Quantitative Analysis The PRA risk impact of operation with Unit 1 and Unit 2 Train B Reserve Feed unavailable can be quantitatively estimated using the CNP Full Power Internal Events (FPIE) model of record and a Fire PRA model which has been updated for this application as describe in Enclosure 3, Section 4.1. The full analysis is provided in Enclosure 3 and is summarized in this section.

The risk assessment is quantified by analyzing the risk of the current plant configuration and then subtracting the risk from a base case (nominal equipment unavailability) core damage frequency (CDF) and large early release frequency (LERF) for both the FPIE and Fire PRA models. Since Unit 2 is currently offline for a scheduled refueling outage, only the risk from to AEP-NRC-2016-83 Page 8 Unit 1 is analyzed in this assessment. The following assumptions are used for the current plant configuration:

Unit 2 is currently in a scheduled refueling outage. The majority of the mitigating systems on Unit 2 Train B are currently unavailable and will remain unavailable during the duration of the Unit 1 and Unit 2 Train B Reserve Feed outage. Although it is possible that some Unit 2 Train B equipment will be available during the outage, it will not be credited for this analysis.

- The Unit 1 East - Unit 2 West ESW crosstie is scheduled to be closed for outage work on the Unit 2 West ESW pump. This crosstie valve will be assumed to be closed for this risk analysis.

No additional Unit 1 equipment will be unavailable for the duraUon of the Unit 1 and Unit 2 Train B Reserve Feed outage.

No surveillance testing will occur on Unit 1 PRA credited equipment.

The following compensatory measures are explicitly accounted for in the risk quantification assumed to be in effect for the duration of the Unit 1 and Unit 2 Train B Reserve Feed outage.

Additional compensatory actions, which are not quantitatively considered, are listed in Section 4.3:

- To the extent practicable and controllable, no other work is assumed to be undertaken that could jeopardize operation of Unit 1.

For example, main turbine valve testing or similar activities, or maintenance work on BOP components that have potential to initiate a unit trip, are assumed to be avoided while repair of Unit 1 and Unit 2 Train B Reserve Feed is in progress.

The following Unit 2 Train A equipment is available o Unit 2 East Motor-Drive Auxiliary Feedwater (AFW) Pump o Unit 2 East Component Cooling Water (CCW) Pump o Unit 2 East Charging Pump o Unit 2 East Essential Service Water (ESW) Pump

All PRA-related equipment for Unit 1 is available.

The compensatory measures are accounted for in the quantitative risk analysis by not assuming any unavailability of the protected or guarded equipment listed above. Test and maintenance is only considered for those systems identified above that were determined to be unavailable for the LCO period. Fire watch tours -and maintenance restrictions are not assumed to modify the likelihood of any fire or internal initiating events.

Since the plant electrical system normally receives power from the main generator, the units do not automatically trip off if offsite power is lost. For this reason, the likelihood of initiating events is not adjusted for the risk analysis. The loss of the Train A reserve auxiliary transformers is to AEP-NRC-2016-83 Page 9 already accounted for in the FPIE model as a random failure. Fire events are not considered to be any more likely due to the Train B Reserve Feed Outage.

FPIE Results The FPIE PRA model was re-quantified with the configuration information as discussed above.

Incremental Conditional Core Damage Probability {ICCDP) and Incremental Conditional Large Early Probability {ICLERP) are calculated for the full (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) duration of the LCO. The full calculation, which follows the approach of RG 1.177, is provided in Enclosure 4.

Internal Internal FPIE Cas.e Events .Events ICCDP IC LE RP CDF (/yr) LERF (/yr)

FPIE Base Case 2.0E-05 2.7E-06 Unit 1 and *unit 2 Train B Reserve 1.1 E-07 1.6E-08 Feed 3.0E-05 4.1E-06 Current Outage Configuration External Events - Fire PRA Results The Fire PRA model was re-quantified with the configuration information as discussed above.

ICCDP and ICLERP are calculated for the full (100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ) duration of the LCO. The full calculation, which follows the approach of RG 1.177, is provided in Enclosure 4.

Fire CDF Fire LERF Fire PRA Case ICCDP IC LE RP

(/yr) (/yr)

Fire PRA Base Case 5.4E-05 4.0E-06 Unit 1 and Unit 2 Train B Reserve

. 1.1 E-06 1.1 E-07 Feed 1.5E-04 1.4E-05 Current Outage Configuration External Events - Seismic Offsite power is generally considered to have a low seismic fragility due to the connection to the larger electrical grid. The unavailability of a single train of offsite power does not significantly affect seismic risk, because seismic events large enough to cause an automatic reactor trip would generally also result in a loss of offsite power. For this reason, the seismic risk due to the reserve feed outage time extension is considered to be negligible.

External Events - Other Other external events, such as high winds and external flooding, would be expected to result in a loss of offsite power during the event. Similar to the discussion for seismic events, significant external events that would cause a reactor trip would be expected to also cause a loss of offsite to AEP-NRC-2016-83 Page 10 power. For this reason, the external event risk due to the reserve feed outage time extension is considered to be negligible.

Conclusion Total CDF and LERF Results

.Ccise ICCOI? JC LE.RP FPIE 1.1 E-07 1.6E-08 Fire PRA 1.1E-06 1.1 E-07 Total 1.2E-06 1.3E-07 The thresholds for low risk in R.G. 1.177, Revision 1, are ICCDP < 1E-06 and ICLERP < 1E-07; however, the thresholds of ICCDP < 1E-05 and ICLERP < 1E-06 are acceptable should appropriate compensatory measure be implemented to reduce the sources of risk.

l&M has evaluated the risk implications of the proposed amendment. The risk assessment was performed assuming a 100-hour Completion Time. Therefore, the requested 28-hour extension in the allowed outage time is bounded by the risk assessment and associated compensatory actions.

4.2 Tier 2: Avoidance of Risk Significant Plant Configurations CNP plant risk associated with the proposed extended Train B Reserve Feed Completion Time is calculated using the CNP FPIE and Fire PRA models (including internal flooding). Associated actions to avoid or respond to these events on one or both units through function of onsite emergency backup power supplies, and inclusion of additional onsite emergency power, are discussed in Tier 3 information, below.

Ultimately for this extended Completion Time request, CNP provides assurance that any other risk significant plant equipment outage configurations will not occur during the extended Completion Time period by flatly ruling out elective maintenance on other PRA risk significant plant equipment and avoiding other activities that could challenge unit operation or cause fires in risk significant areas. Refer to actions discussed in Tier 3, below. The Tier 3 actions mitigate additional plant risk due to events beyond those associated with Train B Reserve Feed unavailability represented in the ICCDP and ICLERP values furnished in the Tier 1 discussion above.

A full analysis of the FPIE and Fire PRA model results is provided in Enclosure 3. A summary of the relevant risk contributors is provided in this section.

IMPACT ON INTERNAL EVENTS (IE)

The internal events risk impact is included in the ICCDP and ICLERP metrics provided in Tier 1.

The updated PRA model used in this assessment includes a typical Internal Events (IE) model and an internal flooding model. Based on a review of the results, risk management actions should focus on protecting the Unit 2 Train A equipment, and the Unit 1 ESW and CCW systems. Operators should focus on the most risk-significant actions related to a loss of CCW I ESW and the potential for loss of power due to only one train of reserve feed remaining.

to AEP-NRC-2016-83 Page 11 Equipment protection for electrical systems and CCW and ESW are of particular importance.

Although steam generator tube rupture was identified as a contributor, the additional risk from this event is primarily due to the potential for loss of power, so an emphasis on protecting equipment is appropriate.

IMPACT ON INTERNAL FLOODING Internal flooding as noted above is part of the updated PRA model used to determine the PRA metrics provided in Tier 1. Based on review of the model, internal flooding events were not determined to be risk-significant for this specific configuration. Compensatory measures discussed in Tier 3 are therefore focused on the more risk-significant initiating events.

IMPACT ON FIRE RISK As discussed above, the fire risk impact is included in the ICCDP and ICLERP metrics provided in Tier 1*. Based on the review of the results, risk management actions should focus on maintaining the Unit 2 Train A equipment as available, arid establishing hourly fire watches in areas identified as risk significant. Operator action review should focus on crosstie eves and tripping the RCPs to mitigate an RCP seal LOCA, both of which are contained within the fire emergency procedures. The Tier 3 information below includes actions to assure that fire detection and suppression systems for these areas are functional, that likelihood of fire initiation from work or operating equipment in the area is reduced/eliminated, and that flammable transient material is not in these. high risk areas.

SUMMARY

For the Risk Management Actions (RMA) presented in the Tier 3 discussion below, CNP will avoid risk significant plant configurations such as performing elective maintenance or intrusive surveillances on the listed plant equipment, and minimizing activities that could initiate plant transients or challenge continued operation of Unit 1. Unit 2 is currently shutdown for a scheduled refueling outage.

RG 1.177 indicates that actions modifying plant design or oper~ting procedures, or to obtqin additional backup equipment, should be considered in the Tier 1 evaluation. However, no plant modifications have been made to reduce the risks associated with these Tier 2 considerations.

An improvement estimate for some operational restrictions listed as Tier 3 actions are quantified as discussed in the quantitative analysis. Additional Tier 3 actions, which are developed to reduce the risk from risk-significant configurations identified in the quantitative analysis, are also included. *

4.3 Tier 3: Risk Informed Configuration Management Compensatory Measures .

Given the difficulty of identifying all possible risk-significant configurations, for this one time Unit 1 TS. 3.8.1, Condition A, Required Action A.3, Completion Time change, CNP will reduce plant risk exposure through a combination of RMAs that prevent planned high risk to AEP-NRC-2016-83 Page 12 configurations and other non-quantifiable risk-reducing actions to reduce risk through availability of additional power supplies requiring manual actions.

RMAs to prevent high risk configurations (due either to fire initiation or other significant plant events), and establish non-quantifiable actions to monitor for high risk (fire or other internal or external) events and provide readily usable alternate power sources are listed below:

Note: These actions include a provision that if emergent plant conditions require actions to stabilize the unit(s), and if any of those actions conflict with any of the RMAs below, then those actions should be taken without delay, and the RMA restored after the emergent condition has passed and the plant is stabilized.

Operation and Maintenance Restrictions Maintenance and testing during the Completion Time extension will be rescheduled for both units as warranted to minimize risk of unit transients. The development of these maintenance restrictions is based on an identification of risk-significant initiating events, operator actions, and equipment identified both before and after completion of the quantitative analysis. As discussed previously, some restrictions (specifically identified above) were credited in the quantitative analysis and others were developed by a thorough review of quantitative analysis results.

These actions will specifically include:

To the extent practicable and controllable, no other work is assumed to be undertaken that could jeopardize operation of Unit 1. For example, main turbine valve testing or similar activities, or maintenance work on BOP components that have the potential to initiate a unit trip, are assumed to be avoided while repair of Unit 1 and Unit 2 Train B Reserve Feed is in progress. *

Unit 1 Train A will be protected in accordance with plant procedures.

No additional Unit 1 PRA equipment will be voluntarily removed from service.

The Unit 2 RWST will maintain level at or above 35% to support the Unit 2 East -

Charging Pump capability to provide crosstie flow to Unit 1. This is the minimum level required by plant procedures for NFPA 805 crosstie support.

The following Unit 1 equipment will be guarded in accordance with plant procedures:

o Unit 1 CD EOG o Unit 1 AB & CD Batteries and Distribution Panels o Unit 1 East CCW Pump & Heat Exchanger

The following Unit 2 Train A equipment shall be available and guarded in accordance with plant procedures:

o Unit 2 East Motor-Driven Auxiliary Feedwater (AFW) Pump o Unit 2 East Component Cooling Water (CCW) Pump o Unit 2 East Charging Pump to AEP-NRC-2016-83 Page 13 o Unit 2 East Essential Service Water (ESW) Pump

The following Fire Zones are identified as high risk areas and will be protected (see discussion of activities in the next bullet):

o Unit 1 CD Emergency Diesel Generator Room - Fire Zone 15 o Unit 1 Train A Battery Room & Switchgear room cable vault- Fire Zones 55 and

56 o Unit 1 Train A Switchgear Room - Fire Zone 408 o Unit 1 Turbine building El' 609 Air Compressor Area - Fire Zone 91 o Screen house MCC ESW Equipment Area - Fire Zones 29G & 29E

For each listed fire zone, the following activities will serve to protect the fire zone:

a. No elective maintenance on fire detection or fire suppression equipment that will cause the fire detection or fire suppression equipment in the impacted fire zones to be inoperable.

b. Verify installed Fire Detection and Suppression systems will be available, as applicable AND-Establish an hourly fire watch tour of the area OR-Establish a continuous fire watch in the area

c. Transient combustible permits will be reviewed for the area and any unnecessary transients will be removed.

d. No hot work will be allowed in the area.

Operations crews will brief on the following procedures before entering the extended CT period o 1-0HP-4023-ECA-O-O, Loss of all AC Power o 1-0HP-4022-016-004, Loss of Component Cooling Water o 1-0HP-4025-001-001, Emergency Remote Shutdown o 12-0HP-4025-001-002, Fire Response Guidelines l&M will ensure the recovery of the Unit 1 AC electrical source is of the highest priority and will exit the proposed action following satisfactory completion of the final operability runs.

5.0 REGULATORY EVALUATION

Applicable Regulatory Requirements/Criteria 10 CFR 50.36 (c)(2)(ii), stipulates that a TS LCO must be established for each item meeting one or more of the following criteria:

1. Installed instrumentation that is used to detect, and indicate in the CR, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

to AEP-NRC-2016-83 Page 14

3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure, system, or component which operating experience or PRA has shown to be significant to public health and safety.

The proposed changes do not modify any plant equipment that provides emergency power to the safety-related 4160v buses in the event of a LOOP. This one-time amendment request to extend the CT for TS. 3.8.1, Condition A, has been prepared to comply with risk considerations from RG 1.177, Revision 1. Evaluation of the proposed changes has determined that the reliability of AC electrical sources is not significantly affected by the proposed changes and that applicable regulations and requirements continue to be met.

In conclusion, CNP has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements/criteria.

No Significant Hazards Consideration Determination Indiana Michigan Power Company (l&M) has evaluated whether or not a significant hazards consideration is involved with the proposed change by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No The proposed change is a one-time extension of a 72-hour Technical Specification (TS) required Completion Time for TS 3.8.1, Condition A'.3. The proposed change does not alter any plant equipment or operating practices in such a manner that the probability of an accident is increased. The proposed change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, the proposed completion time does not involve a significant increase in the probability of occurrence of an accident previously evaluated. *

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed amendment makes a one-time allowance of a 100-hour completion time for TS 3.8.1, Condition A.3. ** The proposed amendment does not introduce any new equipment, create any new failure modes for existing equipment, or create any new* limiting single failures. The plant equipment considered when evaluating the existing completion time remains unchanged. The extended completion time will permit completion of repair activities without incurring transient risks associated with performing a shutdown with one train of to AEP-NRC-2016-83 Page 15 reserve feed unavailable. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed lic1nse amendment makes a one-time allowance of a 100-hour completion time for TS 3.8.1, Condition A.3. The proposed completion time has been evaluated on a risk-informed basis. The proposed configuration controls and compensatory measures provide reasonable assurance that no significant reduction to the margin of safety will occur.

Therefore, the proposed change does not involve a significant reduction in margin of safety.

In summary, based upon the above evaluation, l&M has concluded that the proposed change involves no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified.

6.0 ENVIRONMENTAL CONSIDERATION

S l&M has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. l&M has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared concerning the proposed amendment.

7 .0 REFERENCES None.

Enclosure 3 Probabilistic Risk Assessment (PRA) Technical Adequacy 1 Overview PRA technical adequacy has been addressed through Nuclear Regulatory Commission Regulatory Guide (RG) 1.200, Revision 2 (Reference 2), which references the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard, RA-Sa-2009 (Reference 4), for internal events at power, internal flooding, and fire PRA models. External events (including Seismic Events) and shutdown risk impact will be considered quantitatively or qualitatively as described herein.

This enclosure demonstrates the technical adequacy of the Donald C. Cook Nudear Plant (CNP) PRA model to be used as the basis for the emergency license amendment request (LAR), consistent with the requirements of Section 3.3 and Section 4.2 of RG 1.200, Revision 2 (Reference 2).

2 Relationship of the PRA Model to the As-Built, As-Operated Plant The PRA model supporting the amendment for CNP Units 1 and 2 is the 2016 PRA model of record as documented in CNP calculation PRA-NB-QU Rev 2 (Reference 7), with an approved date of June 30, 2016, and the related PRA supporting documents. This is the most recent evaluation of the CNP internal events at-power risk profile, including internal flooding. To support the transition of its Fire Protection Program to National Fire Protection Association (NFPA) 805 (Reference 6), CNP developed a stand-alone Fire PRA model using the methodology defined in NUREG/CR-6850 (Reference 15). In October 2009, the initial Fire PRA was peer reviewed (Reference 9) against Section 4 of ASME RA-Sa-2009 (Reference 4), which includes the supporting requirements for fire PRAs. An additional focused-scope peer review of the Fire PRA with respect to modeling of Large *Early Release Frequency was performed in November 2015 (Reference 17). The full-power, internal events (including internal flooding)

PRA model and the Fire PRA model is maintained and updated under a PRA configuration control program in accordance with CNP procedures.

Plant changes, including physical and procedural modifications and changes in performance data, are reviewed and the PRA model is updated to reflect such changes periodically by qualified personnel, with independent reviews and approvals. An electronic tracking system documents such plant changes and performs Pending Change Evaluations for each, prioritizing them for inClusion into the PRA model of record as appropriate. The internal events PRA model update resulting in the June 2016 model of record evaluated all open Pending Change Evaluations and incorporated all changes identified to have a potential effect on the model results. The Fire PRA model was last updated in a working model in September of 2016. Refer to Section 4 for a detailed discussion on the Fire PRA.

The internal events PRA Model of Record is currently a CAFTA event tree I fault tree integrated model.

Enclosure 3 to AEP-NRC-2016-83 Page 2 3 Conformance with the ASME/ANS PRA Standard for Internal Events and Internal Flooding Indiana Michigan Power Company (l&M) considers the CNP Internal Events and Internal Flooding PRA as adequate to support the requested amendment. A Peer Review was conducted in July 2015 against ASME RA-Sa-2009 and RG 1.200, Rev. 2. The final Peer Review has been delivered to l&M and can be used to reliably evaluate the technical adequacy of the PRA against the Standard and RG (Reference 10). Previous peer reviews conducted prior to the 2015 peer review are considered to be superseded by the results of this latest peer review.

Each applicable supporting requirement (SR) in ASME RA-Sa-2009 was evaluated against a goal of Capability Category (CC) II. For each SR not meeting at least CC II, an evaluation is provided in the Technical Adequacy Justification Table, Section 8 of this enclosure, with respect to its impact on the proposed amendment.

The results of the peer review identify only 15 percent of the SRs as less than CC II. Many of these requirements are related to documentation, but a few technical issues are also identified.

The key technical issues identified in the table relate to the areas of modeling of actuation signals (SY-810), completeness of the pre-:-initiator human reliability analysis (HR-A1, A2, A3, 81, 82, C2), review of post-initiator human reliability analysis timing (HR-G4, GS), modeling of repair of essential service water (ESW) and component cooling water (CCW) systems (DA-C15), and ensuring the inclusion of all necessary internal flood scenarios (IFSN-A14, A16, A17, IFEV-A8, IFQU-A3). Resolution of the documentation issues is not expected to result in any impact to the PRA model or its associated metrics.

The resolution of the pre-initiator human reliability analysis (SRs HR-A1, A2, A3, 81, 82, C2)

F&Os was subsequently determined to be a PRA upgrade during a recent RAI response to the NRC (Reference 20). The resolution of the F&Os for the listed SRs was subject to a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that all SRs (SRs HR-A1, A2, A3, 81, 82, C2) were met at CC II or higher.

This discussion is provided for each SR in theTechnical Adequacy Justification Table in Section 8.

In addition to the documentation issues and technical issues already identified, several CC I SRs are related to modeling or documentation associated with large early release frequency (LERF). The approach to many of the PRA Large Early Release SRs was determined to be conservative or have minimal numerical impact on LERF. Therefore, CC I is considered to be sufficient for these LE SRs for the purposes of this amendment.

4 Fire Events and External Events

. *4.1 Fire PRA To support the transition of its Fire Protection Program to NFPA 805 (Reference 6), CNP developed a stand-alone Fire PRA model using the methodology defined in NUREG/CR-6850 (Reference 15). In October 2009, the initial Fire PRA was peer reviewed (Reference 9) against to AEP-NRC-2016-83 Page 3 Section 4 of ASME RA-Sa-2009 (Reference 4), which includes the. supporting requirements for fire PRAs.

The transition to NFPA 805 received NRC approval in October 2013 (Reference 5). The stand-alone Fire PRA model, updated in response to the peer review and NRC requests for additional information during the NFPA 805 review process, is documented in calculation PRA-FIRE-17663-005-LAR (Reference 8). This model has been further refined in a documented working model to use for applications (Reference 18). This current working Fire PRA model supports the NFPA-805 program.

The Fire PRA model was initially developed from a previous version of the Internal Events PRA model of record, and continued to be developed as a standalone model to support the transition as documented in calculation PRA-FIRE-17663-005-LAR (Reference 8). The elements of the Fire PRA model taken from the Internal Events model were not reviewed in the 2015 peer review against Part 2 of ASME RA-Sa-2009 (Reference 4) and RG 1.200, Revision 2 (Reference 2). However, the SRs from Part 2 of ASME RA-Sa-2009 and RG 1.200, Revision 2 not meeting at least CC II as shown in the Technical Adequacy Justification Table, Section 8, are considered to be applicable to the Fire PRA as well.

An additional focused-scope peer review of the Fire PRA with respect to modeling of LERF was performed in November 2015 (Reference 17). The focused peer review assessed only the LERF (LE) and related Plant Response Model (PRM) requirements as related to the LERF portion of the Fire PRA model. As with the Internal Events PRA, each applicable supporting requirement (SR) in ASME RA-Sa-2009 was evaluated against a goal of Capability Category (CC) II. For each SR not meeting at least CC II, an evaluation is provided in the Technical Adequacy Justification Table, Section 8 of this enclosure, with respect to its impact on the proposed amendment.

4.2 External Events The full-power, internal events PRA model does not include explicit consideration of external events such as seismic events, severe winds, and external flooding. The CNP Individual Plant Examination of External Events (IPEEE) (References 13 and 14) included an analysis of the following external events:

Seismic Events

External Flooding

Aircraft Accidents

Severe Winds (strong winds and tornadoes)

Ship Impact Accidents

Off-Site Hazardous Material Accidents

On-Site Hazardous Material Accidents

Turbine Missiles

External Fires The risk from external events is assessed qualitatively for the hazards listed.

to AEP-NRC-2016-83 Page4 5 Key Assumptions Based on evaluations supporting the 2016 Internal Events PRA model of record, two key assumptions were identified as key model uncertainties:

5.1 Westinghouse Generation Ill Reactor Coolant Pump (RCP) Shutdown Seals The modeling of the Westinghouse Generation Ill RCP shutdown seals is the first key model uncertainty for the CNP PRA. If the new RCP seals do not actuate or fail to remain actuated, severe accident sequences become much more likely. Risk metrics such as CDF and LERF increase significantly if failure of the shutdown seals is assured. The current PRA model utilizes the Pressurized Water Reactor Owners Group guidance (Reference 11) for PRA modeling of the shutdown seals, supported by the Westinghouse Owners Group 2000 RCP seal failure model (Reference 12), both of which are industry consensus models. The 2015 peer review also found the modeling of the shutdown seals acceptable. The RCP shutdown seals are only credited in the full-power, internal events model and are not credited in the CNP Fire PRA.

5.2 Component Repair Probabilities The repair probabilities for appropriate ESW and CCW component failures are the second key model uncertainty for the CNP PRA. If repair is not credited, some key accident sequences become more likely. The 2015 peer review also identified the importance of this assumption, noting that the recovery credit is based on analysis of information from NSAC-161 (published in 1992) and therefore, this data is not representative of current plant operations (DA-C15). The Technical Adequacy Justification Table, Section 8 of this enclosure, of supporting requirements also identifies this issue and provides the resolution. This credit is no longer included in the model of record.

6 Conclusions The CNP PRA models are generally robust and suitable to support this amendment. Specific issues identified by the recent peer review that may have a significant impact on the model have been addressed to reduce or eliminate their impact on the results. The ongoing PRA maintenance and update activities associated with the CNP PRA program ensure that the PRA models represent the as-built, as-operated plant moving forward. Therefore, the CNP PRA model has the technical adequacy required to support the amendment.

7 References

1. Not Used.

2. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk AssessmenfResults for Risk-Informed Activities," Revision 2, March 2009.

3. RG 1.174 "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision

on Plant-Specific Changes to the Licensing Basis," Revision 2! March 2011.

to AEP-NRC-2016-83 Page 5

4. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009. *

5. Letter from NRC to L. Weber, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c) (TAC NOS. ME6629 and ME6630)," dated October 24, 2013, ADAMS Accession Number ML13140A398.

6. NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition.

7. PRA-NB-QU, "Internal Events Quantification Notebook," Revision 2, 6/30/2016.

8. PRA-FIRE-17663-005-LAR, "DC Cook Fire PRA Fire-Induced Risk Model," Revision 1,

10/28/2014.

9. LTR-RAM-11-10-041, "Fire PRA Peer Review Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the DC Cook Fire Probabilistic Risk Assessment," July 20, 2010.

10. PWROG-15076-P, Peer Review of the D. C. Cook Nuclear Plant Internal Events Probabilistic Risk Assessment, September 2015.

11. PWROG-14001-P, "PRA Model for the Generation Ill Westinghouse Shutdown Seal, Revision 1, July 2014.

12. WCAP-15603, "WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs," Revision 1-A, June 2003.

13. "Individual Plant Examination, Other External Events Notebook," Revision 1, April 5, 1995.

14. "Individual Plant Examination, Addendum to Seismic Probability Risk Assessment Notebook," Revision 0, February 1995.

15. NUREG/CR-6850, "EPRl/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"

Volume 2, September, 2005.

16. NUMARC 91-06, "Guidelines for Industry Actions to Address Shutdown Management,"

December 1991.

17. ERIN Engineering and Research, Inc., "D. C. Cook Focused Scope Peer Review for Fire PRA," Document #00403140002-1515, November 2015.

18. PRA-NB-FIRE-W, "Fire PRA VVorking Model Notebook," Revision 1, September 2016.

19. Jensen Hughes, Inc. "D. C. Cook focused Scope Peer Review - Pre-Initiator HRA," Report No. 1BTl1V001-RPT-01, dated October 10, 2016.

to AEP-NRC-2016-83 Page 6

20. Letter from Q. Shane Lies, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission, "Donald C. Cook nuclear Plant Unit 1 and Unit 2, Follow-Up Response to Request for Additional Information Regarding License Amendment Request to Adopt TSTF-425, Relocate Surveillance Frequencies Program to Licensee Control-Risk Informed Technical Specification Task Force (RITSTF) Initiative 58," dated September 9, 2016.

8 Technical Adequacy Justification Table NOTE: Items listed below in the Supporting Requirements column specifically refer to supporting requirements in ASME RA-Sa-2009 (Reference 4).

to AEP-NRC-2016-83 Page 7 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IE-A5 I The basis for screening control air and Loss of 4kV bus initiating loss of 4160 volt (v) safety bus are not events have been added strong enough to assure that they can to the model.

be screened or combined with another Loss of Control Air initiator. The generic partitioning of modeling for AFW valves loss of offsite power (LOSP) into single has been improved. The and dual unit LOSP does not fit the initiating event frequency plant and should have more analysis is calculated individually done to show the chosen set of as one of several types of initiating events is conservative. Transients Without Steam Conversion Systems Available.

Investigation into plant-specific SLOOP/DLOOP split fractions determined that all or most LOOPs would be dual-LOOPs.

However, investigation into the actual industry events determined that the Plant-Centered and Switchyard-Centered events would not be expected to cause a LOOP for DC Cook.

Therefore, the current frequencies and split fraction for PC and SC are conservative. To address Grid-Related and Weather-Related LOOPs being likely to be DLOOP, PDLOOP-WR/GR fractions are set to 1.0 and PSLOOP-WR/GR fractions are set to zero in the current model.

IE F&Os have no impact on the Fire PRA since it has its own initiating events.

to AEP-NRC-2016-83 Page 8 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IE-83 Not Met Basis: See previous discussion with See the discussion in Loss of Instrument air, LOSP, and IE-A5. Excessive Loss 4160 v bus. Additionally Feedwater is expected to transients with steam generator overfill have minimal impact due potential (which would fail the turbine to the availability of two driven auxiliary feedwater pump motor driven auxiliary (TDAFP) have not been addressed. feed pumps from the accident unit and two from the opposite unit via the crosstie that can supply AFW in the event that the TDAFP is failed.

Therefore, the impact on this amendment is expected to be minor.

IE F&Os have no impact on the Fire PRA since it has its own initiating events.

to AEP-NRC-2016-83 Page 9 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IE-C12 Not Met Basis: PRA-NB-INIT Section 5.2 The explanation of documents the comparison of the differences is not posterior results and support system expected to result in any initiating event (SSIE) results to changes to the SSIEs, as generic values. However, the the modeling has been explanation of differences is not very reviewed in detail and thorough. For example, the Loss of determined to match the CCW initiating event frequency for plant systems. This CNP is approximately an order of issue is therefore a magnitude higher than the generic documentation issue value from NUREGICR-6928. The only.

explanation is that the difference is IE F&Os have no impact "reasonable based on the plant- on the Fire PRA since it specific cooling requirements of has its own initiating CCW" However, a more reasonable events.

explanation may be differences in the degree of redundancy in the system or differences in requirements for operator action to start standby pumps. It is recommended that information from other sources like WCAP-16464-NP, "Westinghouse Owner's Group Mitigating Systems Performance Index Cross Comparison" or EPIX be used to identify plant-specific differences that may affect the initiating event frequencies.

to AEP-NRC-2016-83 Page 1O Supporting Capability Peer Review Assessment Basis Assessment Requirement Category AS-A 1O I Basis: The CNP operator actions use ECCS injection m1ss1on a bounding timing to evaluate the times only impact passive operator actions associated with key failures of the refueling safety functions. In addition, the water storage tank motor timing for injection and recirculation operated valves (MOVs) modes of emergency core cooling for residual heat removal, system (EGGS) was assumed the safety injection, and same for all loss-of-coolant accidents charging. Since more (LOCAs) and Feed and Bleed valves are opened during scenarios. recirculation mode, a longer mission time for recirculation mode is conservative.

The modeled operator action time for ECCS recirculation time is different for small LOCA than medium or large LOCA. Primary Bleed and Feed operator action times are also modeled per initiator. This approach is conservative, and only minor benefits are expected if operator action timing was evaluated on different accident sequences within the event trees.

These issues have the same -impact on the .Fire PRA.

to AEP-NRC-2016-83 Page 11 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category SC-85 Not Met Basis: PRA-NB-SC Revision O does Since the MAAP results not document a reasonableness and have been reviewed in acceptability check with results from detail, and MAAP has not other sources for CNP specific been used outside of its analysis. There may be some applicability, the reasonableness check in earlier reasonableness check is Electric Power Research Institute not expected to result in (EPRI) reports (TR-100741, changes to success TR-1016750), however, those do not criteria. Therefore, the address modular accident analysis impact on this program (MAAP) 5 or plant-specific amendment is expected calculations. Recommend comparing to be minor.

the MAAP run results with Updated These issues have the Final Safety Analysis Report and other same impact on the Fire plant-specific Safetv Analysis. PRA.

SY-A4 Basis: Most of the notebooks indicate Plant walkdowns were that interviews with knowledgeable performed for all systems plant personnel were conducted to during initial model confirm that the systems analysis development and adequately reflected the as-built, as- subsequent plant operated plant and that plant-specific modifications have been data was appropriately collected tracked via the PRA where required. However, a record of configuration such interviews was neither management process.

referenced nor provided. Therefore, no model changes are expected Walkdowns are discussed in a generic during resolution of this walkdown document created in issue for either the FPIE June 1991. There is no record of or Fire PRA.

recent system walkdowns conducted with knowledgeable plant personnel.

SY-AS Not Met Basis: Definitions of component The assessment basis boundaries are found in an excel recognizes that the spreadsheet used in development of boundaries are included PRA-NB-DATA, but the spreadsheet in files used for was not included with the balance of development of the Data the documentation. There is no Analysis, so this

is a discussion related to ensuring that the documentation system model uses component improvement that would boundary definitions that match the not affect model results.

boundaries used in the data collection.

to AEP-NRC-2016-83 Page 12 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category SY-810 Not Met Basis: Actuation signals affecting Modeling of actuation multiple components are modeled in signals has been the 1CPA, 1CPB, 1RCA, 1RCB, expanded in the FPIE to 1SGA, 1SGB, 1SIA, and 1SIB fault meet CC-II.

trees and are documented in the These issues have the PRA-NB-SY-ESFAS notebook. same impact on the Fire However, the actuation logic PRA, but . are not components other than the final expected to have a master and slave relay for each signal significant impact on the are modeled as a single basic event risk calculation due to representing failure of the logic their very low random resulting in generation of an SI, failure rates.

Reactor Trip, etc. The calculation of the probability of signal failure which is applied across all actuation models is bflsed on common cause failure of two relays. On a train basis, the limiting component is typically the Safeguards Output Card or Undervoltage Output Card with a failure probability in the 10-4 range compared to the 5E-7 value being used. WCAP-15376, Table 8.25 shows SI signal failure probability with no operator action for 214 logic as 8. 96E-04 and Diverse RT probability as 2.2E-05 from NUREGICR-5500. These values are also significantly higher than the 5E-7 value being used for CNP.

In addition, logic associated with the undervoltage signal to execute load shed and start the diesels on LOOP is not modeled.

to AEP-NRC-2016-83 Page 13 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-A 1 Not Met Basis: CNP screened all component Pre-Initiator HFE mispositioning. Some of these were screening has been re-screened using post-maintenance test, evaluated to not pre-independent verification, or daily emptively screen checks. Individual failure of these mispositioning in the pre-initiator recovery actions or even current model.

combinations of these recoveries The Fire PRA does not results in higher probabilities than explicitly model pre-*

passive valve failures and should not initiator HFEs, but any be screened. such events (including some identified in the FPIE related to the diesel generators) are not expected to have a significant impact on the risk calculation due to their very low failure rates for risk-significant actions.

The resolution of the F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

to AEP-NRC-2016-83 Page 14 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-A2 Not Met Basis: Miscalibrations of control Pre-Initiator HFE instruments and/or miscalibration of screening has been re-instruments that are used to direct evaluated in the current operator actions (e.g., Refueling Water model to include Tank (RWT) Level alarms) are needed miscalibrations that can to support mitigating system models have an adverse impact and support applications like Fire PRA. on the automatic initiation There was no systematic review for of standby safety identification of misca/ibration events equipment as stated by documented. the SR.

For the impact on Fire PRA, see HR-A 1.

The resolution of the F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

to AEP-NRC-2016-83 Page 15 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-A3 Not Met Basis: Work practices that involve a Pre-Initiator HFE mechanism that simultaneously affects screening has been re-equipment in either different trains of a evaluated in the current redundant system or diverse systems model to include cross-are not specifically addressed. train and redundant Misca/ibrations of pressurizer pressure equipment, and and high containment pressure are miscalibrations.

included (although not directly linked Pressurizer pressure and to the ESFAS and containment containment pressure are isolation signals). However, other now linked to ESFAS via potentially important and common the new ESFAS miscalibrations, such as modeling.

miscalibrations of RWT level For the impact on Fire transmitters preventing operators from PRA, see HR-A 1.

diagnosing the need to swap EGGS suction to the sump, are not The resolution of the considered. F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

to AEP-NRC-2016-83 Page 16 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-81 I Basis: Rules were established for Pre-Initiator HFE screening of individual activities, but screening has been re-application of the rules to specific evaluated to not pre-procedures and activities using the emptively screen rules is not documented. The mispositioning in the screening process for component current model.

mispositioning and instrument Pre-Initiator HFE calibration goes beyond the criteria screening has been re-provided in CC II of this SR. This evaluated in the current process essentially screens classes of model to include activities (e.g., mispositioning events miscalibrations that can and most instrument calibrations). have an adverse impact on the automatic initiation of standby safety equipment as stated by the SR.

For the impact on Fire PRA, see HR-A1.

The resolution of the F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

to AEP-NRC-2016-83 Page 17 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-82 Not Met Basis: As noted in SR HR-81, the Pre-Initiator HFE screening of pre-initiator HRAs is screening has been re-aggressively beyond what is included evaluated to not pre-in this standard. This includes emptively screen misca/ibrations and mispositioning that mispositioning in the could . impact support systems that current model.

impact multiple trains or systems Pre-Initiator HFE screening has been re-evaluated in the current model to include miscalibrations that can have an adverse impact on the automatic initiation of standby safety equipment as stated by the SR.

Pre-Initiator HFE screening has been re-evaluated in the current model to include cross-train and redundant equipment, and miscalibrations.

Pressurizer pressure and containment pressure are now linked to ESFAS via the new ES FAS modeling.

For the impact on Fire PRA, see HR-A1.

The resolution of the F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

to AEP-NRC-2016-83 Page 18 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-C2 I Basis: The screening performed Pre-Initiator HFE eliminates pre-initiator failure modes screening has been re-that have been experienced in generic evaluated to not pre-operating experience and possibly emptively screen would have occurred during plant- mispositioning in the specific experience. current model.

Pre-Initiator HFE screening has been re-evaluated in the current model to include miscalibrations that can have an adverse impact on the automatic initiation of standby safety equipment as stated by the SR.

Pre-Initiator HFE screening has been re-evaluated in the current model to include cross-train and redundant equipment, and miscalibrations.

Pressurizer pressure and containment pressure are now linked to ESFAS via the new ES FAS modeling.

For the impact on Fire PRA, see HR-A 1.

The resolution of the F&Os this SR was the subject of a follow-on focused scope peer review (Reference 19) and the peer review determined that the resolutions were acceptable and that this SR was met at CC II or higher.

Enclosure 3 to AEP-NRC-2016-83 Page 19 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-03 I Basis: The performance shaping CNP eme~ency factors in the HRA Calculator address operating procedures procedure quality and administrative follow emergency controls, but do not justify the basis for response guidelines and determining that these PSFs are good. are generally of high Even the procedure references are quality. Control Room hidden in notes within the HRA event. layouts are routinely trained on and familiar to operators. This issue was addressed during the updates to the pre-initiator HRA discussed in HR-A and HR-B above.

For the impact on Fire PRA, see HR-A 1.

HR-E3 Basis: The operator interviews Post-initiator HRA provided in Appendix C of the HRA updates were performed report provide a walk-through of to include more detailed Emergency Operating Plan steps in operator interviews with general but does not provide a specific review of the detailed review of the operator actions necessary details. Post-modeled (including time to perform the initiator HRA updates actions, whether the action has to be were performed to performed outside the Main Control include talk-throughs with Room, or any special equipment operators to confirm needed to perform the action *(keys, response models. These jumpers, etc.) updates are already included in the current model.

For the Fire PRA, post-initiator HFEs were already updated and documented in response to the peer review and 805 RAls (Reference 8).

HR-E4 Basis: No simulator observations of See discussion for talk-throughs with operators to confirm HR-E3 the response models for the scenarios modeled were noted in PRA-NB-HRA.

iii._

to AEP-NRC-2016-83 Page 20 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-G4 Not Met Basis: It is not clear what the basis for Post-initiator HRA the system time windows for many updates were performed operator actions is. Some of the bases with detailed review of seem to be older Thrust analysis or stated timings for all generic studies when current MAAP events, ensuring that all analyses and other recent plant- timings have a basis or a specific evaluations are currently new basis is developed available. from operator interview and/or thermal-hydraulic calculations. Details are documented in the revision of the HRA notebook (and Success Criteria notebook for supporting MAAP runs).

For the Fire PRA, see discussion for HR-E3.

HR-G5 Basis: The completion times and Post-initiator HRA median response times appear to be updates were performed estimated for most Human Error with detailed review of Probabilities (HEPs). stated timings for all events, ensuring that all timings have a basis or a new basis is developed from operator interview and/or thermal-hydraulic calculations. Details are documented in the revision of the HRA notebook (and Success Criteria notebook for supporting MAAP runs).

For the Fire PRA, see discussion for HR-E3.

HR-G6 Not Met Basis: Table 0-1 in the HRA report An updated consistency provides a consistency check. check was performed However, there is no clear criteria along with the post-established to determine consistency initiator HRA update.

and HEPs seem to be more consistent Documentation . of with the method used to perform than consistency check is in the difficulty of the event and the time the revision of the HRA available to perform the action.

notebook.

For the Fire PRA, see discussion for HR-E3.

to AEP-NRC-2016-83 Page 21 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-11 Not Met Basis: .The current HRA analysis This SR is a references previous analyses as the documentation SR, so basis for event timing and in two cases there is no impact on this assigns current HEPs screening amendment. Refer to the values based on previous analysis other HR SRs listed (e.g., 1----FRH1-COGHE and 1K---- above for specific INJECTNHE2). This makes it difficult technical concerns with to review the current analysis without the pre and post-initiator recourse to the analysts. Since some human failure. events of the authors of the previous analyses (HFEs) (HR-A 1,-A2,-A3, -

are no longer employed by the utility, 81, -82, -C2 for pre-the basis of inputs that vary from the initiators and HR-03 -E3 norm cannot be verified. -E4, -G4,-G5, -G6 for post-initiators). The A lot of the HRA information was assumptions were apparently copied from earlier versions updated as part of other when entering in the HRA Calculator HR SR resolutions.

code. This makes -it difficult to For the Fire PRA, see determine if all of the information discussion for HR-E3.

needed for the HRA evaluation exists.

to AEP-NRC-2016-83 Page 22 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category HR-12 Not Met Basis: PRA-NB-HRA and the HRA This SR is a Calculator database taken together documentation SR, so document the HRA analysis. there is no impact on this HQwever,. a number of steps in the amendment. Refer to the process are not adequately other HR SRs listed documented. These include: above for specific technical concerns with

- Lack of information concerning the pre and post-initiator the process used to screen HFEs (HR-A 1,-A2,-A3, -

plant procedures and identify 81, -82, -C2 for pre-pre-initiator HFEs. initiators and HR-03, -E3,

- Source documents used to -E4, -G4,-G5, -G6 for establish HEP timing for a post-initiators).

number of actions. For the Fire PRA, see discussion for HR-E3.

- Incomplete identification of procedures applicable to HFEs in the HRA Calculator

- Dependence on previous analyses without verification that the timing information from the previous analysis is consistent with the current analysis.

Lack of criteria for establishing consistency of the final HRA results DA-81 Basis: Does not meet CC II. With Updates to the generic respect to the valves, the mission type data used for the (standby vs. operating) and service component data that condition (e.g., treated vs. raw water) does not meet CC-II is should factor into the grouping (to the not expected to extent supported by the data). significantly impact the model, since the most Examples: important component AV type code - Air Operated Valves events use plant-specific are used for the instrument air, aux. data which would meet feedwater and essential service water CC-II.

systems. For the Fire PRA random component failures are MV type code - MO Vs are used in both gen~rally not important,

clean and raw water applicatipns. and so will not have a significant impact on the risk results.

to AEP-NRC-2016-83 Page 23 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category DA-C8 I Basis: The CNP model estimates the While collecting pump-fraction of time a_ component is in specific run time data standby. For example, basic event, would improve overall ESW-1WEST2WEST, represents the model accuracy, it will not fraction of time Unit 1 AND Unit 2 west change the results ESW pumps operating at 0. 5. significantly. CC I is considered sufficient for this amendment for this SR.

DA-C10 Not Met Basis: The PRA-NB-DATA Revision 4 The model may be notebook is silent on the use of slightly conservative due surveillance test data. Based on a to this requirement, but discussion with ERIN Engineering any changes in resolving (Consulting company responsible for this issue are expected to the latest data update), the system be minor or negligible.

manager provided the estimated or For the Fire PRA random actual number of demands. It is component failures* are unclear if the system manager generally not important, reviewed actual surveillance test data and so will not have a with respect to the PRA modeling of significant impact on the possible failure modes. risk results.

DA-C15 Not Met Basis:. INIT-PRA-SY-ESW notes that Credit for repair has been "Recovery of the ESW system in the removed from the model loss of ESW event trees includes until a new basis is recovery of failures caused by valves developed.

or pumps." The same recovery factor Credit is not included in is used for CCW based on the Fire PRA.

INIT-PRA-SY-CCW The recovery credit is based on analysis of information from NSAC-161 to identify recoverable failures applicable to CNP. However, NSAC-161 was published in 1992 before implementation of the maintenance rule and mitigating systems performance index. In addition, the data is based on repair_ experience during normal plant operations which may not be applicable to conditions existing after an initiating event when the emphas_is is in putting the plant in a safe condition. Therefore, this data is not representative of current plant operations and does not meet the requirements of DA-C15 to AEP-NRC-2016-83 Page 24 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category QU-04 I Basis: Plant comparison performed in Resolution of this SR is Section 5.3. 7 of the quantification expected to involve notebook. documentation improvements only since The CNP results were compared to a the comparison was peer plant, but the comparison is at a already performed at a high level and does not present CNP high level. CC I is and peer plant results side-by-side in considered sufficient to the same format to allow identification support amendments for of similarities and differences. The this SR.

documentation is also unclear in some No unique impact on the cases. For example, the first table in Fire PRA.

PRA-NB-QU, Attachment 6 has three results columns, but no headings. As it is, it is difficult to determine if all significant differences are explained in Section 5. 3. 7 since we do not have the results presented using the same breakdown as used in the peer plant pie charts.

QU-F3 I Basis: The dominant contributors are Dominant contributors listed and can also be found from the are discussed and model results. Discussion of the reviewed in the events is less comprehensive than the quantification notebook first PRA quantification, but the (Reference 7), and no information is available if desired model changes are expected when the documentation is improved.

No unique impact on the Fire PRA.

LE-C1 Basis: The CNP LERF analysis follows CC I is considered to be methods in WCAP-16341 and sufficient to support NUREGICR-6595, Revision 1. amendments for this SR.

LE-C2 Basis: The CNP LERF analysis follows CC I is considered to be methods in WCAP-16341 and sufficient to support NUREGICR-6595, Revision 1 which is amendments for this SR.

considered conservative rather than realistic.

LE-C3 Basis: No repair of equipment after CC I is considered to be core damage was considered. sufficient to support amendments for this SR.

to AEP-NRC-2016-83 Page 25 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-C4 I Basis: The CNP LERF analysis follows CC I is considered to be methods in WCAP-16341 and sufficient to support NUREGICR-6595, Revision 1 and the amendments for this SR.

event trees developed in those reports.

LE-C5 I Basis: The CNP LERF analysis follows CC I is considered to be methods in WCAP-16341 and sufficient to support NUREGICR-6595, Revision 1 which is amendments for this SR.

considered conservative rather than realistic.

LE-C9 I Basis: No credit is taken for continued CC I is considered to be operation of equipment or operator sufficient to support actions in adverse environments. amendments for this SR.

LE-C10 I Basis: No credit is taken for CC I is considered to be survivability of equipment or operator sufficient to support actions in adverse environments. amendments for this SR.

LE-C11 I Basis: Containment failure equals CC I is considered to be LERF and ends the analysis. No sufficient to support events beyond containment failure are amendments for this SR.

postulated.

LE-C12 I Basis: Containment failure equals CC I is considered to be LERF and ends the analysis. No sufficient to support continued operation of equipment amendments for this SR.

beyond containment failure is postulated LE-C13 I Basis: Bypass was a deterministic CC I is considered to be event (YES or NO). No source terms sufficient to support or scrubbing or decontamination was amendments for this SR.

evaluated. All steam generator tube rupture (SGTR) sequences go to LERF.

LE-05 I Basis: Models from WCAP-16341 are CC I is considered to be used for Tl-SGTR and Pl-SGTR. sufficient to support SGTR initiator taken directly to amendments for this SR.

containment bypass Secondary Side Isolation is not considered to result in a direct containment bypass.

LE-E2 I Basis: Data is taken from CC I is considered to be NUREGICR-6595 or WCAP-16341. sufficient to support amendments for this SR.

to AEP-NRC-2016-83 Page 26 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-F2 Not Met Basis: The CNP results were Resolution of this SR is compared to a peer plant, but the expected to involve comparison is at a high level and does documentation not present CNP and peer plant improvements only since results side-by-side in the same format the comparison was to allow easy identification of already performed at a similarities and differences. Although high level. Containment differences are discussed, the cause Failure Probabilities of significant differences in the given hydrogen igniter contribution due to igniter failure was failure are taken from not fully explained NUREG/CR-6595, and no issues were noted with the hydrogei,- igniter system modeL to AEP-NRC-2016-83 Page 27 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IFSN-A16 I Basis: CNP IE HRA is performed in The quantitative PRA-FLOOD-013 Revision 1, which is screening has

not yet developed using the EPRI HRA been fully rechecked, but Calculator. The HEPs account for any new scenarios would flood indication in the **control room, not likely be affected by potential to isolate flood sources, and the specific equipment accounting for the likelihood of outage in the mitigative actions. Initial screening of amendment.

internal flooding events* is documented in PRA-FLOOD-012 Revision 0 Significant equipment Attachment A, Internal Flood Source related to recovery from Qualitative, Screening. This initial loss of offsite power was qualitative screening is based, in part, reviewed to\_ ensure the on time to consequential IE or time to flooding model failure of PRA equipment. For deficiencies are not example, FA-01-1a represents a flood expected to significantly from the Unit 1 ESW, which is impact this application.

screened based on "... only 1 train Battery and switchgear ESW Jost at a time, no IE or other PRA rooms are above grade damage for 170 min." This could meet and not a propagation the intent of IFSN-A14 CC I item (c) pathway for floods.

" ... time to the damage of safe Sprays and floods shutdown equipment is significantly initiated in these rooms greater than the expected time for are present in the PRA human mitigative actions to be . model.

performed ... " However based on the screening note, this does not meet the Only a large flood from a intent of IFSN-A14 CC II item (c) CW pipe is considered

" ... mitigative action can be performed credible to fail the EDGs.

with high reliability for the worst These large CW breaks flooding initiator. High reliability is are also currently established by demonstrating, for modeled.

example, that the actions are procedurally directed, that adequate time is available for response, that the area is accessible, and that there is sufficient manpower available to perform the actions. "

to AEP-NRC-2016-83 Page 28 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IFSN-A17 Not Met Basis: CNP performed internal Evaluation of the flooding walkdowns to verify the assumption related to accuracy of information. The doors could impact the walkdown notes are documented in model and results if new Attachment C of PRA-FLOOD-004 flooding pathways are Revision 1. For each flood area the identified. However, following is identified: 1) Flood Targets these would not likely be (PRA Equipment which is defined as affected by the specific "equipment that may be adversely equipment outage in the affected due to the accumulation of amendment.

water, spray, dripping, and steam damage. 'J, 2) Flood/High Energy Line Significant equipment Break (HELB) sources, 3) Barrier related to. recovery from Openings, and 4) Flood Mitigation loss of offsite power was Features. reviewed to ensure the flooding model Some of the assumptions made during deficiencies are not evaluation of flood propagation require expected to significantly additional analysis to substantiate. impact this application.

PRA-FLOOD-004 Revision 1 makes Battery and switchgear the following assumptions: rooms are above grade a) Doors that have a gap and not a propagation beneath them of less than 118 inch pathway for floods.

are considered adequately sealed Sprays and floods against possible propagation. initiated in these rooms are present in the PRA model.

b) Single doors that open toward Only a large flood from a a projected flood are assumed to CW pipe is considered remain intact when subjected to credible to fail the EDGs.

flood forces.

These large CW breaks c) The fixed doors of double are also currently doors at CNP are latched at the modeled.

top and bottom. It is assumed the fixed door being supported by the hinges and at the top and bottom on the free standing side would remain in place given a flood from either direction.

Specifically, an engineering analysis needs to be performed to verify the closed doors can handle the applied force when the water level has reached its maximum potential. CNP PRA meets (a) and (b), however, (c) requires additional analysis/documentation to verify propagation paths are appropriately analyzed.

to AEP-NRC-2016-83 Page 29 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IFEV-A8 Not Met Basis: Quantitative screening is The quantitative documented in PRA-FL000-014 screening has not yet Table 1, Unit 1 Internal Flood PRA been fully rechecked, but Bounding Screening Results, and any new scenarios would further analyzed in Table 2, Unit 1 not likely be affected by Flooding PRA Best Estimate. This the specific equipment quantitative screening analysis was outage in the performed in 2006. These results amendment.

were based on the core damage frequencies using pipe rupture Significant equipment initiating event frequencies from EPRI related to recovery from TR-1013141, "Pipe Rupture loss of offsite power was Frequencies for Internal Flooding reviewed to ensure the PRAs - Revision 1" March 2006 and in flooding model some cases EPRI TR-1 02266 "Pipe deficiencies are not Failure Study Update", April 1993. expected to significantly The generic pipe rupture frequencies impact this application.

have been updated a couple of times Battery and switchgear since these reports with more current rooms are above grade information and understanding of the and not a propagation pipe rupture frequencies. The latest pathway for floods.

CNP internal flooding analysis, PRA- Sprays and floods FLOOD-016, is based on the latest initiated in these rooms pipe rupture frequencies from EPRI are present in the PRA 3002000079, Pipe Rupture model.

Frequencies for Internal Flooding PRAs. Revision 3, 2013. Only a large flood from a CW pipe is considered The current quantitative screening credible to fail the EDGs.

analysis is based on outdated pipe These large CW breaks rupture failure rates. The latest pipe are also currently rupture frequencies from 2013 report modeled.

are significantly higher than the earlier values. It is recommended that the flood scenarios be updated with the latest values and reanalyze the screening.

to AEP-NRC-2016-83 Page 30 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category IFQU-A3 Not Met Basis: Quantitative screening of flood. The quantitative groups is documented in screening has not yet PRA-FLOOD-012, Attachments B and been fully rechecked, but.

G. No flood groups were screened if any new scenarios would the GDF using a bounding GGDP was not likely be affected by greater than 1E-9. the specific equipment outage in the However, the screening in amendment.

PRA-FLOOD-012 was based on flood source initiating frequencies calculated Significant equipment using EPRI pipe break data older than related to recovery from that used for the update of the loss of offsite power was

'dominant' flood groups and GGDP reviewed to ensure the values calculated using an earlier PRA flooding model model. A number of the flood groups deficiencies are not were screened with bounding GDF expected to significantly values just below 1E-9. These groups impact this application.

may be above the screening th_reshold Battery and switchgear using updated flood frequencies and rooms are above grade GGDP values. All of the information and not a propagation for the flood groups should be pathway for floods.

re-calculated using the updated data Sprays and floods and current model to ensure that the initiated in these rooms flood group screening represents the are present in the PRA current state of knowledge. model.

Only a large flood from a

CW pipe is considered credible to fail the EDGs.

These large CW breaks are also currently modeled.

New scenarios would be expected to be at or near the 1E-9 screening criteria arid would not be expected to significantly contribute to the proposed equipment outage risk.

to AEP-NRC-2016-83 Page 31 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-C1 The D. C. Cook LERF analysis follows CC I is considered (Fire PRA) methods in WCAP-16341 and acceptable for Fire PRA NUREGICR-6595, Revision 1. for this amendment for this SR. The CNP LERF model includes several conservatisms since a full generic containment model for ice condenser containments was not developed for WCAP-16341. These conservatisms overestimate LERF and thus have a conservative impact on decisions made using the Fire PRA LERF results. Severe accident containment failure modes are mostly phenomenological and thus do not depend on the type of initiating event that might cause a severe accident. Fire-induced containment bypass events, such as loss of containment isolation, are explicitly .

modeled. Fire-induced loss of the hydrogen igniters is also explicitly modeled.

LE-C2 The D. C. Cook LERF analysis follows CC I is considered to be (Fire PRA) methods in WCAP-16341 and sufficient to support this NUREGICR-6595, Revision 1 which is amendment for this SR.

considered conservative rather than See LE-C1.

realistic.

LE-C3 No repair of equipment after core CC I is considered to be (Fire PRA) damage was considered. sufficient to support this amendment for this SR.

See LE-C1.

to AEP-NRC-2016-83 Page 32 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-C4 I Basis: The 0. C. Cook LERF analysis CC I is considered to be (Fire PRA) follows methods in WCAP-16341 and sufficient to support this NUREG/CR-6595, Revision 1 and the amendment for this SR.

event trees developed in those See LE-C1.

reports.

LE-CS I The 0. C. Cook LERF analysis follows CC I is considered to be (Fire PRA) methods in WCAP-16341 and sufficient to support this NUREGICR-6595, Revision 1 which is amendment for this SR.

considered conservative rather than See LE-C1.

realistic.

to AEP-NRC-2016-83 Page 33 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-C7 Not Met Sections 6.5.3 and 6.5.6 of PRA-NB- Several specific issues (Fire PRA) FIRE-LE describes the operator are identified in the actions credited in the LERF Model. associated F&Os.

The level of detail of the analysis does The main impact of fire not reflect the use of the applicable on the identified HFEs is requirements in Section 2-2. 5 of the the level of stress during PRA Standard. a fire event. Since these are Level 2 HFEs that occur after severe core damage has already occurred or is imminent, stress is already expected to be high.

Therefore, fire events that result in core damage are considered to not be significantly more stressful than other events (such as large LOCAs, small LOCAs, SGTRs) once core damage has occurred.

Thus, ltisjudgedthatltis not necessary to adjust the value of the Level 2 HFEs generally for fire events.

Another impact of fire events is the possibility that operators would have to evacuate the control room due to a fire.

Where it is unlikely that the operators would be able to perform a function, no credit is taken for this action in main control room evacuation scenarios.

LE-C9 No credit is taken for continued CC I is considered to be (Fire PRA) operation of equipment or operator sufficient to support this actions in adverse environments. amendment for this SR.

See LE-C1.

to AEP-NRC-2016-83 Page 34 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-C10 No credit is taken for survivability of CC I is considered to be (Fire PRA) equipment or operator actions in sufficient to support this adverse environments. amendment for this SR.

See LE-C1.

LE-C11 Containment failure equals LERF and CC I is considered to be (Fire PRA) ends the analysis. No events beyond sufficient to support this containment failure are postulated. amendment for this SR.

See LE-C1.

LE-C12 Containment failure equals LERF and CC I is considered to be (Fire PRA) ends the analysis. No continued sufficient to support this operation of equipment beyond amendment for this SR.

containment failure is postulated. See LE-C1.

LE-C13 Bypass was a deterministic event CC I is considered to be (Fire PRA) (YES or NO). No source terms or sufficient to support this scrubbing or decontamination was amendment for this SR.

evaluated. All SGTR sequences go to See LE-C1.

LERF.

LE-01 CNP has a plant-specific containment CC I is considered to be (Fire PRA) fragility analysis (Attachment 1 of sufficient to support this PRA-L2 MODEL, Revision 0) that amendment for this SR.

predicts the ultimate containment The plant-specific capacity and the location of containment analysis is containment failure on pressure. now credited for However, it is not clear if and how this improving the total time calculation was factored into the available to energize the simplified Level 2 model documented igniters, allowing the in PRA-NB-FIRE-LE. Attachment 1 of recovery action to PRA-L2 MODEL is not cited in Section repower the igniters to be 8 of PRA-NB-FIRE-LE and not credited in more fire discussed in Section 6. 5. 1 of the areas. This was the most report. The simplified Level 2 model significant conservatism appears to be using NUREGICR-6595 present in the LERF if the igniters fail. model for the Fire PRA.

to AEP-NRC-2016-83 Page 35 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-02 CNP has a plant-specific containment CC I is considered to be (Fire PRA) fragility analysis (Attachment 1 of sufficient to support this PRA-L2 MODEL, Revision 0) that amendment for this SR.

predicts the ultimate containment See LE-01.

capacity and the location of containment failure on pressure.

However, it is not clear if and how this calculation was factored into the simplified Level 2 model documented in PRA-NB-FIRE-LE. Attachment 1 of PRA-L2 MODEL is not cited in Section 8 of PRA-NB-FIRE-LE and not discussed in Section 6. 5. 1 of the report. The simplified Level 2 model appears to be using NUREGICR-6595 if the igniters fail.

LE-03 CNP has a plant-specific containment CC I is considered to be (Fire PRA) fragility analysis (Attachment 1 of sufficient to support this PRA-L2 MODEL, Revision 0) that amendment for this SR.

predicts the ultimate containment See LE-01.

capacity and the location of containment failure on pressure.

However, it is not clear if and how this calculation was factored into the simplified Level 2 model documented in PRA-NB-FIRE-LE. Attachment 1 of PRA-L2 MODEL is not cited in Section 8 of PRA-NB-FIRE-LE and not discussed in Section 6. 5. 1 of the report. The simplified Level 2 model appears to be using NUREGICR-6595 if the igniters fail.

LE-05 Models from WCAP-16341 are used CC I is considered to be (Fire PRA) for Tl-SGTR and Pl-SGTR. SGTR sufficient to support this initiator taken directly to containment amendment for this SR.

bypass. See LE-01.

to AEP-NRC-2016-83 Page 36 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-E1 Not Met Sources for parameter values are See LE-C7.

(Fire PRA) shown in Table 2 of PRA-NB-FIRE-LE.

Appropriate parameter values were selected consistent with the requirements of technical element DA.

Operator actions identified in Sections

6. 5. 3 and 6. 5. 6 were not selected in accordance with Section 2-2. 5 of the PRA Standard.

LE-E2 Basis: Data is taken from NUREGICR- CC I is considered to be (Fire PRA) 6595 or WCAP-16341. sufficient to support this amendment for this SR.

Containment Isolation modeling is taken from the Internal Events model, with some specific Fire PRA components added as necessary.

This is related to LE-C8, which was Met.

LE-F1 Not Met The results of the LERF quantification The updated Fire PRA (Fire PRA) and cutset reviews are provided in model includes the PRA-FIRE-17663-014-LAR-R1-final- quantification results by 1017, Tables 5-1, 5-7, 5-9, 5-17 and 5- failure mechanism.

19. The results do not provide contributions by LERF PDS designation and LERF failure mechanism.

to AEP-NRC-2016-83 Page 37 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-F2 Not Met The CNP results were not compared The ASME standard (Fire PRA) to a peer plant. specifically states in SR FQ-E1, that SR QU-03 (plant comparison of results) is not applicable to a Fire PRA (LE-F2 is not discussed, but the derived requirement is related).

Additionally, LE-F2 states that a reasonableness check is required (i.e.

review the results to ensure excessive conservatism has not skewed the results).

While a comparison to a peer plant could function as a reasonableness check for an internal events model, Fire PRAs.

are inherently different since the spatial location of cables varies widely at other plants. Thus, a general discussion of the results is provided in the working model notebook, including some review of results from another plant. This discussion is focused at a high level to provide insights into possible conservatisms, as opposed to a direct corrioarison of results.

LE-G3 Not Met The results of the LERF quantification The updated Fire PRA (Fire PRA) and cutset reviews are provided in model includes the PRA-FIRE-17663-014-LAR-R 1-final- quantification results by 1017, Tables 5-1, 5-7, 5-9, 5-17 and 5- failure mechanism.

19. The results do not provide contributions by LERF PDS .:

designation and LERF failure mechanism.

to AEP-NRC-2016-83 Page 38 Supporting Capability Peer Review Assessment Basis Assessment Requirement Category LE-G6 Not Met Sections 5.2 and 5.3 of PRA-FIRE- The internal events LERF (Fire PRA) 17663-014-LAR-R1-final-1017 provide notebook now provides a a quantitative definition used for definition of LERF significant core damage accident sufficient to meet the progression sequence that is standard.

consistent with Part 1-2 of the standard. However, there is no equivalent definition for LERF.

PRM-82 Not Met An assessment of Internal Event PRA As this is linked to the (Fire PRA) peer review deficiencies is required to internal events model, evaluate the impact on the Fire PRA. CC I is considered to be sufficient to support this amendment for this SR.

PRM-814 Not Met Provide documentation demonstrating The potential for multiple (Fire PRA) an evaluation for this SR. Evaluate containment isolation the potential for screened LERF valve failures to result in scenarios impacting the Fire PRA, a large early release is e.g., LERF bypass pathway screened now included in the based on size, where a fire may analysis and working impact multiple pathways where the model notebook.

sum of the pathway sizes may exceed the LERF bypass pathway screening criteria.

PRM-815 Not Met Provide documentation demonstrating The potential for multiple (Fire PRA) an evaluation for this SR. containment isolation valve failures to result in a large e?rly release is now included in the analysis and working model notebook.

Enclosure 5 to AEP-NRC-2016-83 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGE MARKED TO SHOW PROPOSED CHANGES

AC Sources - Operating 3.8.1 ACTIONS

N 0 TE-----------------------------------------------------------

L CO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A One required offsite A.1 ---------------NOTE--------------

circuit inoperable. Not applicable if a required Unit 2 offsite circuit is inoperable.

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite

circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months rcseeJ circuit to OPERABLE IFootnote)I status.

AND 17 days from discovery of failure to meet LCO 3.8.1.a orb Footnote: For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ," to support repair and restoration of the Train B Reserve Feed. Upon completion of the repair and restoration associated with this event which occurred on October 7, 2016, this footnote is no longer applicable.

Cook Nuclear Plant Unit 1 3.8.1-2 Amendment No. 2&7, 291

Enclosure 6 to AEP-NRC-2016-83 DONALD C. COOK NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATION PAGE CHANGESINFORPORATED

AC Sources - Operating 3.8.1 ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 ---------------NOTE--------------

circuit inoperable. Not applicable if a required Unit 2 offsite circuit is inoperable.

Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Declare required feature(s) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from with no offsite power discovery of no offsite available inoperable when power to one train its redundant required concurrent with feature(s) is inoperable. inoperability of redundant required feature(s)

AND A.3 Restore required offsite 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (See circuit to OPERABLE Footnote) status.

AND 17 days from discovery of failure to meet LCO 3.8.1.a orb Footnote: For Train B only, the Completion Time that Train B can be inoperable as specified by

. Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " up to "100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months ," to support repair and restoration of the Train B Reserve Feed. Upon completion of the repair and restoration associated with this event which occurred on October 7, 2016, this footnote is no longer applicable.

Cook Nuclear Plant Unit 1 3.8.1-2 Amendment No. ~. 2-9+,

- - - - - ----

AEP-NRC-2016-83, Emergency License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

Contents

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

SUMMARY

5.0 REGULATORY EVALUATION

6.0 ENVIRONMENTAL CONSIDERATION

Navigation menu

AEP-NRC-2016-83, Emergency License Amendment Request for One-Time Extension of Completion Time for Inoperable AC Source - Operating

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

SUMMARY

5.0 REGULATORY EVALUATION

6.0 ENVIRONMENTAL CONSIDERATION

Navigation menu

Search