ML11290A170

Root Cause Evaluation Re Dual Unit Trip Following August 23, 2011 Earthquake

ML11290A170
Person / Time
Site:	North Anna
Issue date:	10/12/2011
From:	Grecheck E Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
11-578
Download: ML11290A170 (133)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 10 CFR 100, Appendix A October 12, 2011 U.S. Nuclear Regulatory Commission Serial No.: 11-578 Attention: Document Control Desk NL&OS/GDM R2 Washington, DC 20555 Docket Nos.: 50-338/339 License Nos.: NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 ROOT CAUSE EVALUATION DUAL UNIT TRIP FOLLOWING THE AUGUST 23, 2011 EARTHQUAKE During a September 30, 2011 telephone conversation between Dominion personnel and NRC staff, the NRC requested information regarding the root cause evaluation (RCE) for North Anna Power Station that addressed the dual unit trip following the August 23, 2011 Mineral, Virginia earthquake. Pursuant to the NRC request, a summary of excerpted portions of the RCE is provided as an attachment to this letter along with selected attachments to the RCE.

Separately, in regard to the "Sequence of Events Validation" discussion, unexpected instrumentation responses were evaluated, and an investigation determined that the alarms received were valid for the existing plant conditions.

If you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Sincerely, E. S. Grecheck Vice President - Nuclear Development

Attachment:

1. Summary of Excerpted Portions of the Root Cause Evaluation - Dual Unit Trip Following Magnitude 5.8 Earthquake LAAAAAAAAA Commitments made in this correspondence: None 2 .

VICKI L. HULL

.0mm"nuU MOf Virgini COMMONWEALTH OF VIRGINIA ) 140642

) MYC "uinM Expire May 31, 2014 COUNTY OF HENRICO ............ I The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by E. S. Grecheck who is Vice President - Nuclear Development, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this L2Eday of 2011.

My Commission Expires: /)I.

A,4=

W1i Notary Public

Serial Number 11-578 Docket Nos. 50-338/339 Page 2 of 2 cc: Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave. NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station M. Khanna Branch Chief - Mechanical and Civil Engineering Branch U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 R. E. Martin NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 P. G. Boyle NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, MD 20852-2738 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.

Glen Allen, Virginia 23060

Serial Number 11-578 Docket Nos. 50-338/339 Attachment Summary of Excerpted Portions of the Root Cause Evaluation - Dual Unit Trip Following Magnitude 5.8 Earthquake Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

• pV Domi niion Summary of Excerpted Portions Root Cause EvaluationRCEO01061 Rev I Dual Unit Trip Following Magnitude 5.8 Earthquake North Anna Power Station RCE 001061, Rev. 1 Page 1 of 130

Table of Contents 1.0 Executive Summary 4 1.1 Problem Statement 4 1.2 Root Causes 4 1.3 Contributing Cause(s) 4 1.4 Corrective Action 4 1.4.1 Corrective Action(s) to Prevent Recurrence (Not Included in Summary of Excerpted Portions) 1.4.2 Recommendation for Contributing Causes (Not Included in Summary of Excerpted Portions) 1.4.3 Compensatory or Short Term Corrective Action (Not Included in Summary of Excerpted Portions) 1.4.4 Other Insights That Warrant Corrective Action (Not Included in Summary of Excerpted Portions) 2.0 Detailed Report 4 2.1 Team Members (Not Included in Summary of Excerpted Portions) 2.2 Event Investigation and Analysis 4 2.3 Organizational and Programmatic Review(Not Included in Summary of Excerpted Portions) 2.4 Extent of Condition (Not Included in Summary of Excerpted Portions) 2.5 Assessment of Safety Consequences 16 2.6 Assessment of Safety Culture (Not Included in Summary of Excerpted Portions) 2.7 Repeat Event Review (Not Included in Summary of Excerpted Portions) 2.8 Operating Experience 17 2.9 Extent of Cause 23 2.10 Equipment Reliability/PM Adequacy 24 2.11 Personnel Interviewed (Not Included in Summary of Excerpted Portions) 2.12 Documents Reviewed (Not Included in Summary of Excerpted Portions) 2.13 Causal Factors 24 RCE 001061, Rev. 1 Page 2 of 130

Attachment(s): : Nuclear Instrumentation Core Orientation : Why Chart : Dual-Unit Seismically-Induced Reactor Trip - Indicated Nuclear Power Transient Cause Matrix (Not Included in Summary of Excerpted Portions) : Timeline of Events Unit 1 : Timeline of Events Unit 2 : Electrical Timeline : Failure Mode Effects Analysis (FMEA) (Not Included in Summary of Excerpted Portions) : Organizational and Programmatic Deficiencies Fishbone (Not Included in Summary of Excerpted Portions) : Cause To Corrective Action Matrix (Not Included in Summary of Excerpted Portions) 0: Analysis Of Dranetz Event Data vs Plant Computer System 1: Purdue Assessment 2: Response Time Assessment (Not Included in Summary of Excerpted Portions) 3: 1H Bus & Seismic Activity RCE 001061, Rev. 1 Page 3 of 130

1.0 Executive Summary 1.1 Problem Statement "On August 23, 2011 at 1351 North Anna Power Station experienced a magnitude 5.8 earthquake followed by a dual unit trip and a loss of offsite power. The purpose of this Root Cause Evaluation is to:

0 Identify the cause for the reactor trips on NAPS Unit 1 and Unit 2."

1.2 Root Causes 1.2.1 "The Direct Cause for the both the Unit 1 and Unit 2 reactor trip was the initiation of the rate Nuclear Instrument (NIs) Power (PWR) Range Hi Flux Rate Reactor Trip. Both Unit 1 and Unit 2 met the required coincidence of 2 out of 4 Power Range Nuclear Instruments (PRNI) with greater than a 5 % change in 2.25 seconds.

(RC1) The Root Cause of this event was a synergistic combination of seismically induced conditions which include core barrel movement, detector movement, and small reactivity effects from core movement and thickening of the thermal-boundary layer along the fuel rods. The additive effects of the combined conditions resulted in momentary under moderated core conditions as evidenced by the oscillatory but overall decreasing flux profiles from both Unit 1 and Unit 2."

1.3 Contributing Cause(s):

1.3.1 "No Contributing Causes are applicable to this event. All potential candidates for Contributing Causes were investigated in detail and determined to not be credible."

1.4 Corrective Action "As part of this investigation, a detailed review was performed to determine the validity of the Dranetz Sequence of Event Recorder data and PCS alarms for Unit 1 and Unit 2. The results of this review determined that all data points were valid, equipment operated as expected, with the exception of 5 points that are being tracked by NAPS Corrective Action Program. Notable Equipment related Narrative Logs and Condition Reports submitted immediately following the Seismic Event are referenced in this report and resolution of concerns is tracked through the Corrective Action Process."

2.0 Detailed Report 2.2 Event Investigation & Analysis "On August 23, 2011 at 13:51 hours, a magnitude 5.8 earthquake occurred approximately 11 miles WSW of North Anna Power Station. The earthquake caused a series of reactor trip signals to both Unit 1 and Unit 2 reactors, as well as a total loss of offsite power to the station. The "First Out" reactor trip signals for both Units were "High Flux Rate Reactor RCE 001061, Rev. 1 Page 4 of 130 I

Trip". The "First Out" turbine trip signal for Unit 1 was "Main Transformer Lock Out". The "First Out" for Unit 2 was "Reactor Trip/Turbine Trip".

To facilitate the investigation the following investigative techniques were used;

1. Fault Diagram - "Why" chart (Attachment 2) ...

2. Event Timeline (Attachments 4, 5 and 6) ...

3. Analysis Of Dranetz Event Data vs Plant Computer System (Attachment 10)"

Detailed Discussion Introduction "The RCE Team began by gathering all available information already obtained by the Event Review Team as well as data that was not previously available. A critical source of information made available to the RCE Team was the Transient Response Analysis (TRA) data which provides thirty-three sample points per second from one hour before the Reactor Trip Breakers opened until ten minutes after the start of data acquisition. The data captures Plant Computer System (PCS) monitored points on a much higher than normal sampling rate.

It is important to note that normal post trip analysis at NAPS would not routinely involve scrutinizing post trip data on a thirty-three times per second sampling interval. The normal sampling rate did not show a negative rate trip when reviewed. The period in question is extremely short in duration, approximately one second long. With the assistance of Nuclear Analysis and Fuels (NAF) the Unit 1 and 2 average PRNI traces were then plotted. The graph below (Figure 1) shows the results.

Unit 1 and 2 Power vs Time During Earthquake (Average of 4 NI signals)

I ___

75 o 2 ...... ----- -I-95 -

Time(mm:ss.s)

Figure 1 Unit 1 and Unit 2 PRNI indication had very similar trends. Given the fact that both Units PRNI traces exhibited the similar plots shown in the above graph the first task of the RCE Team was to either accept the PRNI as valid indicators of reactor power or some type of instrument error either caused by or induced from the earthquake. The Team also reviewed RCE 001061, Rev. 1 Page 5 of 130

Intermediate Range Nuclear Instrumentation (IRNI) traces and they also produced a similar trend when plotted (See Figures 2 and 3).

105 -3 100 U IR & PR Comparisons -3.05 p 95 - -3.1 o 90* -V' 0 -3.15 I 90 R e 8 85 -3.2 r V ,A 80 -. 5M R 75,, -, -*U1 N-41 -33

-3. P a S U1 N-42 n n 70---,U1 70 - N-43 V .* -3.35 9 65 - -U1 u11 R N-44 TI P3 .

e - U1 N-36 60 -U1N-35 -3.45 51:09.9 51:10.3 51:10.8 51:11.2 51:11.6 51:12.1 51:12.5 Time Figure 2 105 2.9

% 100 -2.q55 95 -3 P

0 -3.05 1 90 W R

-3.1 e 85 r -3.15 A 80 M

-3.2 R P 75 a -3.25 S n 70 3

g e 65 3.35 60 -3.4 51:07.9 51:08.8 51:09.6 51:10.5 51:11.4 51:12.2

'Time Figure 3 It was also observed that PRNI output for detectors that are located in the same position with regard to the core on both Units (for example 1-NI-CHA-41 and 2-NI-CHA-42 see ) exhibited similar flux traces. This would also be in agreement with the analyzed direction of seismic waves experienced at NAPS. Electrical busses supplying the PRNIs were reviewed by the RCE Team to determine if any anomalies were present and none were identified. PCS data on the PRNI upper and lower detector voltage signals as well RCE 001061, Rev. 1 Page 6 of 130

as the Gamma-metrics excore detector signals (-one second sampling rate) did not provide any useful data. The RCE Team concluded that the likelihood of an instrument failure/fault impacting Unit 1 and 2 PRNIs as well as IRNIs at the exact same time and in the exact same manner was deemed non-credible ... Based on this, the focus of the RCE Team was to identify which seismic related events could impact either indicated or true reactor power.

Several Short Term Corrective Actions (STCA) and Enhancements were created to verify that the Nuclear Instruments were not damaged during this event."

RCE Team Methodology "The RCE Team postulated various theories to explain the shape of the PRNI traces... If the theory was deemed plausible, every effort was made to quantify the impact on either indicated power or real power.

Each theory was reviewed by the RCE Team, discussed with Dominion Nuclear Analysis and Fuels (NAF) personnel and Westinghouse personnel brought on site to assist the RCE Team.

Westinghouse personnel also requested support from their corporate support staff to assist with the vetting process.

Specific Indicated Nuclear Power Transient Cause Discussion "Operating experience and testing associated with Japanese nuclear reactors indicates core reactivity changes can occur as a result of small changes in overall fuel geometry caused by earthquakes. This phenomenon would cause a down power created by seismic vibrations disrupting the laminar sub-layer along the cladding wall resulting in rapid and transitory bubble bursts that would add negative reactivity due to the void defect. The RCE Team noted that the down power at the start of the event coincides with the start of seismic activity. Once the Reactor Trip Breakers opened there were several more oscillations which coincided with the highest seismic peaks noted during the earthquake. Figure 4 and 5 show the correlation between the oscillations and seismic activity on both Units. Also plotted on the graphs is the time that reactor trip breakers opened.

RCE 001061, Rev. 1 Page 7 of 130

Unit 1 PRNI vs Time During Earthquake (Average NI signals) i3o

.2 a

75 Time (mmms.s)

Figure 4 Unit 2 PRNI vs Time During Earthquake (Average NI signals) 85 a 58 75

-3 Time (MMýss.s)

Figure 5 Note that the peaking of the average PRNI indication for both Units appears to follow the measured seismic motion. Upon further observation it was also noted that the two PRNIs located on opposite sides of the core, 1(2)-NI-CHA-41 and 1(2)-NI-CHA-42, exhibited RCE 001061, Rev. 1 Page 8 of 130

power oscillations 180 degrees out of phase with each other during some parts of the event.

These PRNIs are located in a plane along the same axis in both Units (See Figure 6).

Figure 6 By overlaying the NI plots with the seismic motion recorded at NAPS a strong correlation is noted.

(See Figures 7 and 8).

RCE 001061, Rev. 1 Page 9 of 130

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5 Figure 8 This was observed on both Unit 1 and Unit.2. This supports the theory that the observed indications were directly related to the seismic event in that the PRNIs were impacted based on location with regard to the core. 1(2)-NI-CHA-41 and 1(2)-NI-CHA-42 were the channels that tripped the reactor on both Unit 1 and 2.

RCE 001061, Rev. 1 Page 10 of 130

Another graph used to compare the difference between N-41/42 and N-43/44 is shown in Figures 9 and 10. It is a plot of the magnitude of the difference between these two sets of NIs at any given time. As an example Unit 1 shows that N-41 and N-42 were out of synch for the majority of the event. It also shows the predominate effect was in the east -west plane.

Figure 9 I,

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!,Oulh ur* 2 T3/4,w Figure 10 Impacts Due To Voiding The original proposed voiding mechanism was that seismic motion would result in rapid bubble formation that would result in a negative reactivity being added to the core.

Subsequent analysis ... deemed that this would not produce significant amounts of negative reactivity on the time scale and magnitude necessary to explain the total negative reactivity observed. The assessment did however state that some amount of fuel assembly gap reduction could produce thickening of the thermal-boundary layer along the fuel rods. A RCE 001061, Rev. 1 Page 11 of 130

CASMO calculation to determine the effect of reducing the assembly pitch shows that moving all assemblies in the core closer together by 31 mils (the inter-assembly gap) reduces reactivity by 50 pcm, which could produce a power reduction of 10%. A much smaller effect could occur locally with a smaller set of assemblies. Secondary flow in the core may be possible resulting in a possible explanation of the localization of thermal-boundary layers.

Existing bubbles would grow or new bubble generation could occur. According to the research results in Japan, these bubbles can collapse significantly when the frequency is higher than 10 Hz. The Purdue Assessment is included in this report (Attachment 11)."

Core Compression "Movement of internal core components due to the earthquake could close the gaps between assemblies in the core which would add negative reactivity. The subsequent decompression would then add positive reactivity. Westinghouse Core Design experts estimated that this could add as much as 100 to 200 pcm. Subsequent evaluation determined that the core compression during a seismic event would occur after the time of the reactor trip and could not explain the entire power behavior; however, it could be a minor contributor. Additionally the Purdue Assessment also discusses the impact of fuel assembly motion on localized changes in the fuel rod thermal-boundary layer, which can produce bubble growth and bubble generation that could be a minor potential contributor to the power transient."

Core Barrel Movement "Based on the fact that there was a correlation between PRNI and seismically induced motion the RCE Team discussed what could cause indicated PRNI to change due to motion of the core. One possible theory was that the gap between the core barrel and the vessel wall increased by seismic motion, thereby decreasing the flux seen by the PRNI detectors due to more moderation. The outer most fuel assemblies produce the majority of flux seen at the detectors and if the moderator thickness was increased this would cause a down power indication. This is due to the fact that the freedom of movement for core components is limited and Westinghouse experts estimated only 0.1" of movement could be expected during an event of the magnitude experienced at NAPS. This could result in 3 to 6 % change in indicated power. Based on the analysis this was deemed possible but would not fully explain the total decreasing trend. Additionally this impact is directionally dependent and would not explain why all four PRNIs showed the same decreasing trend."

PRNI Movement "The PRNI detectors are located in dry wells against the inner wall of the Primary Shield Tank as shown in Figure 11.

RCE 001061, Rev. 1 Page 12 of 130

Figure 11 Westinghouse reports that an analysis performed for Beaver Valley, a sister plant to NAPS, indicates that it is possible for detector movement within the wells to cause a 5% variation in indicated reactor power. Based on the analysis this was deemed possible but would not cause the trip by itself."

Rod Control Motor Generator Set Output Breakers "Initially it was thought that the event was caused by opening of the Rod Control Motor Generator set output breakers. This was based on the fact that all four Rod Control Motor Generator set output breakers had an instantaneous over current trip locked in and were verified open as noted in post event walk downs. NAF reviewed the PRNI traces and noted that the rate of power decrease at the start of the event was not high enough to support this theory based on computer modeling. At this point the RCE Team concluded that rod movement started after the Reactor Trip Breakers opened and that in all likelihood the output breaker drops came after the breakers opened and therefore could not be the cause of the initial down power excursion. Although the Rod Control Motor Generator output breakers were deemed as not causing the event the RCE Team did perform additional checks and testing to determine the cause of the actuation of the relay. The relays on Unit 2 were tested satisfactorily IAW 2-IPM-1821-01 and 02 to ensure that the as found setpoints were correct for the relay. One of the relays that tested satisfactorily was sent for seismic fragility testing.

This relay tripped during low frequency testing and was deemed seismically fragile by the results of the test. Based on the test results the RCE Team believes that the relay actuated due to ground motion vice fault actuation... (Figure 12) shows the comparison between the 1995 Rod Control Motor Generator set output breaker trip and this event. Note that the 1995 event reduced power much faster than the event in question.

RCE 001061, Rev. 1 Page 13 of 130

NAPS Reactor Trips - Indicated Nuclear Power Profiles 120 NOTE:

All data synchronized to RTBopening 100 time. 2011 seismic trip data for Unit 1 1

! * ** and Unit 2 artificially appear to be out-i--

80 S

• , *The overlay of the 1995 MG Set event demonstrates

__6o the earthquake events were not caused by 4the MG Output Breaker eoThe 1995 Ul Dropped Rod events matches the earthquake events however, Rod Drop has been refuted as a cause based on the FMEA 40.and investigation of rod drop times.

20 - N ,

0 0

-1 -0.5 0 0.5 1 1.5 2 Time Relative to RTBOpening (sec) 1995 U2 M-GSet 1996 Ul Dropped Rod - 2011 U1 Seismic NFRT - 2011 U2 Seismic NFRT

- - 1995 U2 M-G Set Shifted 2007 U2 SI - - - ReactorTrip Breaker Figure 12 The RCE Team also noted that had the output breaker indication been available on PCS or the "Rod Control MG Sets Trouble" annunciator been recorded the exact time could have been fixed."

Dropped Rod or Group of Rods "NAF recommended that the RCE Team look at the possibility of a rod or group of rods dropping before the Reactor Trip Breakers opened on "High Flux Rate Reactor Trip." When the RCE Team compared rod drop times from this event with previous events it was discovered that the times versus position data were very close (within 2.5 steps) when measured from the point where Reactor Trip Breakers opened. Additionally, the faster eight rod drop times for this event were compared with normal trip data from a Unit 2 trip in 2007.

No inconsistencies were noted with regard to the drop times."

Sequence of Events Validation "As part of the Dual Unit Trip Root Cause investigation, a detailed review was performed to determine the validity of the Dranetz Sequence of Event Recorder data for Unit 1 and Unit 2.

The results of this review determined that all data points were valid, equipment operated as expected, with the exception of the following which were determined to warrant further investigation by I&C."

RCE 001061, Rev. 1 Page 14 of 130

"Unit 1 Dranetz point 0031 - RWST CHEM ADD TK LO TEMP The Trip for LO temperature is 25 degF, Reset is 26.5 degF. This may have been caused by a connection issue with the field or cabinet wiring providing an intermittent partial short across the RTD causing a low temperature input to the NRA card TM-QS201 (CF-121). I&C checks of wiring and calibration recommended.

Dranetz point 0036/PCS point T0444A, T0497A - Loop 1C HI DELTA T DEVIATION Actual Delta T Deviation never exceeded the setpoint, so the reason for this actuation is UNKNOWN. Recalibration did not show any anomalous data. Same event on Unit 2.

Additional I&C checks recommended Dranetz point 0041 - FIRE WATER SYSTEM INITIATED Fire system pressure changes during seismic event may have caused this indication. This is indicated by initial fast Init/Normal toggling time of < 400ms due to earthquake. No plant log entries found for Fire System Initiated on 08/23/2011. Although not indicated in plant logs, RCE team post trip discussion with Operations personnel indicated that the fire water system actuated. Additional I&C checks recommended.

Unit 2 Dranetz point 0031 - RWST CHEM ADD TK LO TEMP The Trip for LO temperature is 25 degF, Reset is 26.5 degF. This may have been caused by a connection issue with the field or cabinet wiring providing an intermittent partial short across the RTD causing a low temperature input to the NRA card TM-QS101 (CF-121).

Additional I&C checks recommended Dranetz point 0036/PCS point T0444A, T0497A - Loop 1C HI DELTA T DEVIATION Actual Delta T Deviation never exceeded the setpoint, so the reason for this actuation is UNKNOWN. Recalibration did not show any anomalous data. Same event on Unit 1.

Additional I&C checks recommended Conclusion Based on the detailed review conducted by the RCE Team several plausible explanations were deemed to be capable of producing the PRNI Traces observed during the event with regard to the magnitude of the PRNI changes as well as the speed at which it could occur.

Although a singular effect could not be supported as producing the indicated PRNI traces, a cumulative affect is plausible. The RCE Team determined that the High Flux Rate Reactor Trip was a valid trip ... Even if the High Flux Rate Reactor Trip (decreasing) had not tripped the Units other trip signals would have resulted in a trip for both Units. Attachment 10 contains the review performed to evaluate the validity of the sequence of events alarms. The tripping of both Units was conservative in that they occurred prior to other trip set points were reached. Based on this report all systems actuated as expected and desired."

RCE 001061, Rev. 1 Page 15 of 130

2.5 Assessment of Safety Consequences "The observed negative and positive reactivity excursion of the event was postulated to have occurred due to synergistic effects of the seismic event. Core motion induced reactivity changes resulted in a net negative reactivity addition and observed power decrease followed by a net positive reactivity addition and observed power increase. The power increase on Unit 1 was observed to peak and turn downward without rod motion, on Unit 2 the positive peak was arrested by the control motion. At no time did reactor power increase above 100%

following the initial decrease in power. There were no safety consequences as a result of this event and the reactor was shut down as a result of a negative flux rate trip. All protection equipment responded as designed as documented in the OP-AA-105, Post Trip Review Report."

Safety Analysis Summary of Reactor Trip Data "The Plant Computer System data was reviewed relative to the requirements of the North Anna UFSAR safety analyses. Based on a review of this data, the global RCS response is consistent with a normal reactor trip from full power followed by an RCP trip coastdown from the loss of power to the supply buses. Although there are some core power variations prior to full control rod insertion, power decreases from the initial value and at no time exceeds 100% power. RCS temperatures trend smoothly toward hot zero power values as expected with no perturbations. There were some variations in pressurizer pressure and level early in the event whose validity could not be confirmed; however, the overall trends were reasonable and the magnitudes, even considering the variations, were well within safety analysis response values.

Safety analysis events most applicable to this event include UFSAR 15.2.7 "Loss of External Electrical Load and/or Turbine Trip" (LOEL), UFSAR 15.2.9 "Loss of Offsite Power to the Station Auxiliaries" (LOOP), and UFSAR 15.3.4 "Complete Loss of Reactor Coolant Flow" (CLOF). Since the reactor trip, turbine trip and reactor coolant pump trips occurred at essentially the same time during the plant event, the transient response was easily bounded by the safety analysis response for these events. In particular, all safety analysis requirements relative to core cooling/DNB criteria, RCS and main steam pressure, and pressurizer level were met. It is also noted that the LOOP analysis, which demonstrates a long-term secondary heat sink is available, only credits operation of the motor-operated AFW pumps and therefore bounds the planned unavailability of the turbine-driven AFW pump.

Corporate Engineering (I&C) reviewed response times for the negative rate trip and why the positive rate trip did not come in during this event. The circuit was verified as having operated within design limits. Corporate Engineering (I&C) has reviewed the sequence of events with regard to reactor trip breaker opening and has no concerns."

"In summary, the down power event and subsequent reactor trip is bounded by the North Anna UFSAR safety analyses. The integrity of the core is maintained by operation of the reactor protection system and natural circulation flow through the RCS loops and reactor core. In addition, pressure relief valves and/or sprays maintain primary and secondary pressures well below safety analysis allowable values."

RCE 001061, Rev. 1 Page 16 of 130

2.8 Operating Experience "Note: Attempts were made to obtain Fukushima OE, however no data could be obtained that applied to this report.

A search of internal and external OE which may be related to this event was conducted. The OE which has been reviewed for the RCE report is listed below. None of the Operating Experience reviewed could have predicted or prevented the reactor trips at North Anna. The major lesson learned was to improve our station procedure for responding to a seismic event to meet industry standards. This specific issue will be addressed in the "Organizational and Programmatic Review" section of the RCE. The specific response of the Nuclear Instrumentation during the earthquake at NAPS was not identified in other plant OE. One possible reason for this, which was brought up in calls with others, is that the resolution of the data which the ERT and RCE team looked at (33.3 times per second) is not obtainable at other sites with their installed computer systems. No internal and external events similar to NAPS in regards to initial plant response (RX trip on high flux rate) could be found either.

Other sites have had seismic events. Some have resulted in reactor trips and others have not.

For the OE looked at, those trips were a result of electrical issues, manual trips, or automatic trips as a result of trip signals keyed off of seismic activity."

LER 87-075-00 Vogtle - Unit 1 Missing Screws in the Nuclear Instrumentation Drawers "On December 22, 1987, at 1406, with Unit 1 at approximately 99 percent of rated thermal power (RTP), an Instrumentation and Control (I&C) technician, while taking test readings in a power range Nuclear Instrumentation (NI) drawer, identified that the screws were missing from a hold down (cover) plate on a printed circuit (PC) card rack. The plate functions as a hold down plate for the card assemblies to aid in the restriction of card movement.

The exact circumstances by which the screws came to be missing were not known. It is thought that this event occurred because a failure to initially install these screws was not discovered during the Construction Acceptance Test Program or during the subsequent calibration of the NI's."

"This OE was evaluated as being potentially applicable to this event because loose or missing screws could cause an unexpected response of the nuclear instrumentationduring a seismic event. As a result, both units NIS were inspected by a member of the ERT Photographs were taken and inspection results were documented. Although one screw was missing... there was no apparent deficiencies that would have contributed to the response that was seen on the nuclear instrumentation."

"This OperatingExperience could not have predicted orprevented the reactortrips at North Anna Power Station."

OE17568 - 6.5M San Simeon Earthquake Impact on Diablo Canyon "On December 22, 2003, at 1116 PST, with Unit 1 and Unit 2 operating at 100 percent power, a 6.5 magnitude earthquake occurred approximately 50 km NNW of Diablo Canyon RCE 001061, Rev. 1 Page 17 of 130

Power Plant (DCPP). Ground motion was felt and recognized as an earthquake by the control room operators. The earthquake force monitor recorded greater than 0.01g for the seismic event. Operations personnel declared an Unusual Event at 1122 PST. (Reference NRC Event Notification Number 40408.) On December 23,2003, at 1212 PST, the Unusual Event was terminated upon confirmation that no damage to the plant occurred. There was no adverse effect to public health and safety, or upon facility features important to safety.

The main shock was felt in the Units 1 and 2 Control Room. It triggered the basic seismic system analog recorder (Kinemetrics SMA) in the Control Room and the Kinemetrics digital recorders (SSA) at the Unit 1 containment base, top of containment, the Auxiliary Building, and the free field pit locations (near the Fitness Trailer). The supplemental system was out of service at the time of the earthquake, however, three temporary accelerometers located in the Auxiliary and Turbine Buildings and a permanent instrument in the basement of the 500 kV Switching Center triggered. The supplemental system was unavailable and replacement parts are obsolete; both the basic and supplemental systems are scheduled to be replaced in January 2004 with new instrumentation."

"This OE was evaluated as being applicable to this event. Two areas of concern in this OE are that at Diablo Canyon, the installed seismic supplemental system was unavailabledue to maintenance and the main turbine experienced rotormovement during the event, and various tank levels alarmed at Diablo Canyon. At NAPS, a portion of the installed seismic instrumentation did not have power available from when the emergency bus was de-energized until it was restored by its respective Emergency Diesel Generator (EDG)...

Temporary Modification #1845 will install an uninterruptible power supply (UPS) on the Earthquake Monitoring Panel 1-EI-CB-151 in the Control Room to address the powering issue.

"This OperatingExperience could not have predicted orprevented the reactortrips at North Anna Power Station."

ENR D-1998-00628, Emergency Power System Breakers May Trip During Seismic Event Due to Improperly Oualified Relays Installed "On May 15, 1998 at Darlington, with Units 1, 2 and 3 operating at 98% power, and Unit 4 in guaranteed shutdown state, a review of the emergency power system relays found that the sub-systems qualified for category B design basis earthquake contains relays which are qualified for category C. As a result, the category C relays will chatter during the category B seismic event. The situation could potentially cause trips of normally closed breakers in 4kV and 600V emergency power systems. The emergency power system operating manual was revised to instruct operators to close about twenty breakers that may have tripped during the seismic event. This re-closure can be accomplished from the common secondary control area at the same control panel used to trip and close the other breakers. A subsequent technical operability evaluation determined that the operability of the 4kV and 600V emergency power system was not affected, and the system will perform its design function following a design basis earthquake. The event is not significant as it did have an actual impact on the operability of the safety system. However, it is NOTEWORTHY because it identifies weakness in configuration control. Also, manual operation of the breakers is a work around.

Too many work arounds can negatively affect operator's performance during an emergency."

RCE 001061, Rev. 1 Page 18 of 130

"This OE was evaluated as being potentially applicable to this event. During the event, at NAPS, breakers were observed to have tripped and in addition there was a tripping of the A/B/C RSS due to sudden pressure relay actuations.Also, it was observed that both units rod drive motor generator sets output breakers have phase over current trip flags and each of these breakers were determined to have tripped at some time during the event. The impact that the actuation of the phase over current trips on the rod drive motor generatorset output breakers is being evaluated as part of the RCE."

"This OperatingExperience could not have predictedor prevented the reactor trips at North Anna Power Station."

OE6860 - Earthquake Near Humboldt Bay Power Plant "On September 1, 1994 at 8:16 a.m. a 7.2 magnitude Earthquake occurred off the north coast of California, approximately 90 miles from Humboldt Bay Power Plant. Unit 3 at this plant is a permanently shutdown BWR. All the fuel is off-loaded in a spent fuel pool. The seismic event resulted in peak on-site accelerations up to 0.05 g. The peak acceleration at the fuel pool was 0.04 g. The acceleration generated 6 inch waves, which were fully contained within the pool. Follow-up inspections verified there was no fuel or fuel pool liner damage caused by this event. Note: this event is relatively minor compared to an April 1992 earthquake which generated peak acceleration at the plant of 0.24 g. This April 1992 Quake was of similar magnitude, but was located much closer to the plant. The Design Basis Earthquake for the plant is 0.25 g."

"Following the event at NAPS, a walkdown of the fuel building, spent fuel pool and spent fuel pool cooling systems was completed. It was reported-by individuals in the NAPS fuel building at the time of the event that there was significant wave action in the spent fuel pool due to the seismic event, but no apparentdamage was observed and all water was contained within the spent fuel pool. No abnormalities were noted based on the inspections that were conducted. No further action is requiredas a result of this OE."

"This OperatingExperience could not have predicted orprevented the reactor trips at North Anna Power Station" EAR TYO 06-022; (CHASNUPP 1 - PWR in Pakistan) Earthquake during Refueling Outage "At 0852 hours0.00986 days <br />0.237 hours <br />0.00141 weeks <br />3.24186e-4 months <br /> on 8th October, 2005 when the plant was in "refueling outage" and preparations were underway for unloading of the irradiated core, an earthquake of 7.6 magnitude on the Richter's scale, with epicenter 325 km due north-east, struck the plant site.

Plant seismic instrumentation recorded local ground acceleration of 0.31 m/s2. An OBE indication however appeared on local seismic instrumentation from one of the sensors inside the Reactor building and a mini tsunami was observed in the Refueling Water Pool which was filled up to 12.5m with borated water. On feeling the tremors, all operations were immediately suspended and the Upper Internals' lifting rig, hanging over the Reactor Pressure Vessel (RPV) for removal of Upper Internals from the RPV, was immediately RCE 001061, Rev. 1 Page 19 of 130

moved away from the core and safely placed back on its storage stand. A plant standby emergency was declared that prompted activation of level-1 of the Emergency Response Organization who reached Emergency Control and Technical Support Centers as per procedure. A detailed inspection of all critical areas and safety equipment of the plant was carried out and, after confirming normalcy, the "standby emergency" was terminated at 1213 hrs the same day and maintenance work was allowed from 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> onward. Fuel unloading operations however remained suspended for subsequent 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in view of continued intermittent aftershocks."

CONSEQUENCES:

"Appearance of minor cracks in non-load-bearing walls, damage of an underwater light and dislocation of another in Refueling Water Pool, splash / loss of insignificant quantity of Refueling Pool water through ventilation ducts, delay in Fuel unloading, and other work execution.

The fuel conveyor car got stuck on its railing inside transfer tube between Reactor and Fuel buildings when it was tested functionally in the evening of 8th October, 2005. Some grinding work on the anti-seismic cramp plates corrected the problem. Thereafter the empty conveyor car got stuck once again inside the transfer tube during actual fuel unloading when it was being moved back after completing unloading of 15th fuel assembly at 2140 hours0.0248 days <br />0.594 hours <br />0.00354 weeks <br />8.1427e-4 months <br /> on l1th October, 2005. Interference of one anti-seismic cramp plate, installed on front of the car, with the railing was found to be the reason. Both anti seismic cramp plates from front of car (facing towards the Reactor building) were temporarily removed in consultation with designer to expedite completion of unloading as partial loaded core inside RPV was under risk from aftershocks."

"Following the event at NAPS, a walkdown of the fuel building, spent fuel pool and spent fuel pool cooling systems was completed. It was reported by individuals in the NAPS fuel building at the time of the event that there was significant wave action in the spent fuel pool due to the seismic event, but no apparent damage was observed and all water was contained within the spent fuel pool. No abnormalities were noted based on the inspections that were conducted. This OE wasforwarded to the Supervisor Fuel Handling to review the issues with the fuel conveyor car to determine if any actions will be required at NAPS. It should be noted that NAPS fuel handling group will conduct dry and wet checkouts of the fuel transfer system using a dummy fuel assembly prior to actualfuel movement. "

"This OperatingExperience could not have predicted orprevented the reactortrips at North Anna Power Station."

OE16691 - Seismic Event "On Monday, June 30, 2003 Perry Nuclear Power Plant, while operating at 100 percent power, declared an Unusual Event. No major plant problems were identified attributable to the seismic event and at 2233 the Shift Manager terminated from the Unusual Event, after consulting with personnel at the State and Counties. The State and Counties were notified at 2239 and the NRC was notified at 2241 that the event had been terminated."

RCE 001061, Rev. 1 Page 20 of 130

"This OperatingExperience could not have predicted or prevented the reactor trips at North Anna Power Station."

OE5406 - Seismic Events, Seismic Ground Motion at SONGS "At 0458 on June 28, 1992, strong ground motion, measured at .038g, was felt at SONGS, as the result of an earthquake in Landers, CA (approximately 100 miles from the plant).

Annunciator "Seismic Recording System Actuation", Seismic Alarm Annunciator Alarm "Strong Motion Acceleration System Activation" was received. Numerous plant alarms were received from various plant systems. At 0805 strong ground motion was once again felt, measured at .042g, and the same alarms as mentioned before were received. This was the result of second earthquake in Big Bear, CA (approximately 70 miles from the plant).

At 0502 an Unusual Event was declared after the first series of alarms were received and closed out at 1035 upon completion of the required checks after the second one. No significant abnormal plant conditions were noted. Unit 2's Spent Fuel Pool filter has been plugging up frequently due to debris stirred up in the water. There was no significant damage reported in the media within a 50 mile radius of the plant."

"At NAPS, water clarity issues were identified in the spent fuel pool which requiredplacing the spentfuel pool purificationsystem and skimmer in service. "

"This OperatingExperience could not have predicted or prevented the reactortrips at North Anna Power Station."

SEN 269 Earthquake at Kashiwazaki-Kariwa "At 10:13 on July 16, 2007, an earthquake of magnitude 6.8 on the Richter scale shook the northwest coast of Japan in the Niigata Prefecture near Kashiwazaki City. The epicenter of the earthquake was 16 kilometers from Tokyo Electric Power Company's (TEPCO)

Kashiwazaki-Kariwa Nuclear Power Station. The station includes seven boiling water reactors, with advanced boiling water reactors on units 6 and 7, and has an installed capacity of 8,212 MWe. Units 3, 4, and 7 automatically scrammed from 100 percent power on exceeding their seismic high-level scram setpoints. Unit 2 also automatically scrammed during startup operations. Units 1, 5, and 6 were already shut down for planned outages at the time of the earthquake.

Key lessons learned from this event include the following:

" An integrated emergency response strategy and alternate methods of communication can improve the response to sitewide events with multiple challenges.

• On-site fire protection systems and local fire department response may be challenged during natural disasters.

• Unexpected radiological liquid and gaseous releases can occur following natural disasters.

RCE 001061, Rev. 1 Page 21 of 130

" Seismic events can impact the integrity of radioactive waste storage drums or other items that are stacked without restraints.

• Alternate means of personnel contamination monitoring may need to be established following a natural disaster."

"EPRI Technical Report NP 6695, "Guidelines for Nuclear Plant Response to an Earthquake," which provides additional guidance for post-earthquake response actions is cited in the SEN. It is this report that is being used as the basis for our post earthquake inspections and recovery plans."

"This OperatingExperience could not have predicted or prevented the reactortrips at North Anna Power Station."

MER TYO 07-067 Reactor Manual Trip Due to Intensity Level V Earthquakes "On December 26, 2006, Maanshan Unit 1 & 2 were operating at 100% thermal power. At 20:26, an earthquake of magnitude (ML) 7.0 occurred at about 25 km southwest of plant site with a depth of 44.1 km. At 20:34 (8 minutes later), a second earthquake of magnitude 7.0 occurred at about 25 km west of plant site and 50.2 km in depth. The maximum intensity for both quakes measured at station was level V. The intensity reached 0.25 g for local area nearby.

During the earthquakes, many alarms including turbine and reactor coolant pump high vibration alarms were actuated. But the trip set points were not exceeded. Due to intense shaking in the control room and many alarms showing on the annunciator panel, unit 2 shift supervisor made a conservative decision and ordered a manual reactor trip. All safety systems were actuated as designed and the unit was stabilized at hot shutdown condition.

Similar scenario occurred in Unit 1 control room, all shift personnel were on highly alert but did not commence a reactor trip. Operation crew initiated a walk down inspection for all building structures and equipment in power block according to the procedure in response to earthquake. The inspection items included all instrument (panels, tubing and elements),

electrical equipment (transformers, generators, motors, breakers, chargers and batteries), and low level waste storage areas."

CONSEQUENCES:

"After extensive inspection, no significant damage on building structures and no major fault on system equipment were found. The followings were observed:

Spent fuel pool water was a little turbid. Less than one gallon of fuel oil spilled out from over-flow pipes of emergency diesel generator fuel oil storage tanks. Some piping thermal insulators were dropped down. Some alarms actuated during earthquakes and were reset without problem. Two cable conduits of RCP space heater and current transformer were disconnected. 55 drums of below regulatory concerned radwaste fell down with no damage."

ANALYSIS/COMMENTS:

RCE 001061, Rev. 1 Page 22 of 130

"According to plant procedure, manual reactor trip is required if the earthquake intensity exceeds the OBE settings. The OBE settings range from 0.25 g for containment base to 0.37 g for higher level floor. The recorded maximum intensity for these earthquakes was 0.17 g at power block. Determination of OBE by free-field response spectrum and accumulative absolute velocity (CAV) according to RG 1.166 need to be evaluated for applicability in response to the actual earthquake energy release."

CORRECTIVE ACTIONS:

"In addition to the extensive inspection performed, a further investigation for all building structures' response to the big earthquake was initiated.

Further survey of control room non-safety panels and ceilings were completed.

A task force was assigned to thoroughly review the procedures in response to the earthquake, post earthquake inspection program and determination criteria of an OBE."

"This OperatingExperience could not have predicted orprevented the reactor trips at North Anna PowerStation."

2.9 Extent of Cause "Object of Cause: Nuclear Instrumentation Cause: Earthquake induced changes Application: Reactor Critical Operations

• Tier 1 - Same Object - Same Application: Are there other nuclear instrumentation issues that have the same cause during reactor critical operations? No. All potential effects on nuclear instrumentation as a result of an earthquake's were evaluated as part of this RCE.

• Tier 2 - Same Object - Other Application(s): Are there other nuclear instrumentation issues that have the same cause during other modes of operation? No, when the reactor is subcritical, the source range instrumentation is used to monitor the reactor core. The theory of operation of the source range detectors is different from the power range or intermediate range, therefore the same type of response would not be expected from the source range detectors.

• Tier 3 - Similar Object - Same Application: Are there other instruments that could have the same cause during reactor critical operations? Yes. A review of instrumentation traces from the PCS computer indicates that fluctuations were observed in level and flow instrumentation. Examples include Pressurizer level, RCS loop flows, SG levels, RCP seal leakoff flows, and chemical addition tank levels. After the seismic activity ceased, the fluctuations in the instrumentation ceased. There is no evidence that any of these transmitters failed as a result of the fluctuations.

• Tier 4 - Similar Object - Other Application(s): Are there other instruments that could have the same cause during other modes of operation? Yes. As stated in the Tier 3 RCE 001061, Rev. 1 Page 23 of 130

discussion, there were instrumentation fluctuations, but these fluctuations ceased following the earthquake. These same fluctuations could occur during other modes of operation."

Extent of Cause Basis "The extent of cause was bounded only to the instruments that are used to monitor the reactor while at power or shutdown conditions since they may be affected by the earthquake. During the reactor trip of both North Anna units, other equipment issues were observed which were a result of the earthquake. This included affects on level instrumentation, spiking on radiation monitors, failures of the sudden pressure relays for the reserve station transformers, and operation of the 2H EDG (list not all inclusive). Each of these and other issues will be evaluated outside of this RCE using the stations corrective action program."

2.10 Equipment Reliability/PM Adequacy "Review of the Equipment Reliability and PM 's for the unit 1 & 2 Nuclear Instrumentation System Power Range channels and the reactor vessel internals has not shown any concerns.

The Power Range channels and the reactor vessels are properly classified and have established maintenance strategies. There is no actual failure mode to consider as the neutron flux indications seen on the Power Range channels for each unit were an actual seismic induced indication of flux. A review of the applicable components for equipment classification, performance monitoring, preventative maintenance, work practices, design/operation, parts/vendor quality, and aging/obsolescence has not revealed areas that were determined to be inadequate or would have had an impact in mitigating the event."

2.13 Causal Factors "RC1: GVB, Vibration - Failure due to vibration of the equipment or components, often as a result of unbalanced loading, mechanical looseness, excessive clearances or unexpected harmonics."

RCE 001061, Rev. 1 Page 24 of 130

Attachment 1 Nuclear Instrumentation Core Orientation RCS Summary. Rev 1 B"Loop Inlet UN "B"Loop OtiV T2** UNIT 1* "B"Loop Inlet A., B C\ E F G 1IJ KL Pr4R P R P N IA L K J H G F E D C 15 I V I I ý: C"C Loon Outlet -- MEMMLIII 1 4i1

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/ SA"Loop Outlet INCORE TUJNNEL

/ IQ "A"Loop Inlet INCOFE TUN,.NEL Note: Those cells gray incolor are confirmed with 11716-FE-47A-1O and 12050-FE-47A-7 Note: Akes (AR)(1-15) come fromthe onloadoffload core maps used in refueling, which are very good references, as the SR are clearly defined on these maps Note: Regardless of unit#. thirn-le N-13 is in quadrant N-4e., C-13 is in 1*42, etc. Therefore Ceo*is configuration holds truefor QPTR even with U1 and U2 being 18Y out.

N-411 N-.t N-41 N-42 RCE 001061, Rev. 1 Page 25 of 130

Attachment 2 HUnit 1/2 Trips F RN IFU AER IPWR RNG N42 HI FLUX BISTABLE (NEG) I IPWR RNG N41 HI FLUX BISTABLE (NEG) I I Indicated Negative Neutron Flux I or more Dropped Rods STEM MG Set Bkr Opens MG Set Output Over Current Loss of SS Gripper loss of equip seismically Trip signal to pwr to MG set disengages due voltage to affected MG Set to seismic stationary Output Bkr activity gripper coils T i i Flywheel coast down only Not a cause Not a Cause Logic Cabinet Loose LOSS OF Blown fuses Power Cabinet Failures Breaker Rela Electrical maintains Design is robust. Per Westinghouse, connections CONTACT failures Fault voltage and Hz Statistically power cycling off and on the CRDM ON FUSE I for -I second unlikely to have on induced logic errors, I cable to head HOLDER both Units SAT Results from 2- connection Not a cause Regulation Firing Card Phase Not.aqCaus-e experience the Same style as IPM-RCS-G-001B No blown Card Failure Control Card same mechanical Testing will rule out NOT a Rx Trip Bkr Not a cause fuses failure of the cause which is Flux proftie Not a cause definitively CRDM at the SAT test SQUG overlay w/ Timeline same time Not a cause NOTa results from Qualified previous shows loss of RPI traces cause. NOT a NOT a NOT a cause.

SS occurred show no Visual I-IPM-known trip cause, cause.

abnormalities checks RCS-G-not a match after initial trip 2-IPM- 2-IPM- G-OO1B will Jackshaft SAT. RCS-G- RCS-G- check for OOIA drionned CARE 01B will OOIB wll indicate no proper phase from spider Testing by [ check check for control card issues Not a Cause 9/20 will VEER to proper output firing Fragility Testing Results confirm determine if output voltages determined seismically

>6.2 VDC firing which feed the fragile. Flux profile overlay wt previous voltages firing card Not a Cause Weight known trip not a match checks SAT PaCe 0026 of, Rev.I Page 26 of 130

Attachment 2 Not a Cause NI traces indicate actual change in flux.

Equipment responded as expected. SAT Results from Channel Analyses performed for Operational Testing will Beaver Valley 1 sister rule out definitively plant in terms of NSSS and NST design) show that the maximum possible variation in indicated power would be 5%. This is for the case of the detector tube contacting the well wall and is believed to be a very bounding estimate.

RCE 001061, Rev. 1 Page 27 of 130

Attachment 2 Seismic motion resulting in Not a Cause closing assembly gaps would NAF analyses showed that it would require on result in a decrease in core the order of a 40% reactivity in an undermoderated Not a Cause Nota Cause lattice. Subsequent opening of Seismic motion increases reduction in total RCS gaps when motion changes .

in magnitude during and flow over 0.5 seconds to The required reduction direction would be expected to after the reactor trip. It cause the power in moderation in the add positive reactivity to core.

is difficult to explain why reduction. A nmechanism vessel downcomer gap voids are formed due to to cause this reduction, to result in the seismic activity, then followed by a near observed cease at the same time in instantaneous increase decrease/increase in both units even though back to full flow (all not power is estimated at the seismic motion seen on high resolution between 0.5" continues to increase. TRA (Transient Response (Westinghouse) and Analysis) flow data for 1.0" (NAF).

any loop) is not plausible.

Westinghouse internals experts have quantified the maximum reduction in the gap as being on the order of 0.1".

RCE 001061, Rev. 1 Page 28 of 130

Attachment 4 Timeline of Events Unit 1 F;781-1' I 1+

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RCE 001061, Rev. 1 Page 29 of 130

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RCE 001061, Rev. 1 Page 31 of 130

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RCE 001061, Rev. 1 Page 33 of 130

Attachment 5 Timeline of Events Unit 2 SW 08N B IlS1~M4F SThIGENC H~STh$Ml WIS5 LOOP' C Tft~PPEII CHrV (>ill LOFICMO

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Attachment 5 Timeline of Events Unit 2 S&*4CPBUS~A Ut~RVOLTAG ETRPPED

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RCE 001061, Rev. 1 Page 35 of 130 Electrical Timeline Unit 1 RCE 001061, Rev. 1 Page 36 of 130 Electrical Timeline Unit 2 RCE 001061, Rev. 1 Page 37 of 130

Attachment 10 North Anna Power Station Units 1 and 2 Analysis of Dranetz event data versus the Plant Computer System (PCS)

For August 23, 2011 RCE 001061, Rev. 1 Page 38 of 130

Attachment 10 Executive Summary Purpose The purpose of this analysis is to determine the validity of the Dranetz SOE Recorder events for North Anna Units 1 and 2 during the earthquake that occurred on August 23, 2011. The following tables summarize the analysis of the sequence of events for Units 1 and 2, respectively. The Unit 1 analysis begins at time = 13:51:10.224 and ends at time =

14:00:14.359. The Unit 2 analysis begins at time = 13:51:11.072 and ends at time = 14:16:18.714. To perform this analysis, the Unit 1 and Unit 2 Dranetz Sequence of Events (SOE) Recorder event times were normalized to the Plant Computer System (PCS). For Unit 1, 1.0 Minute and 20.703 Seconds was added to each Dranetz SOE event time. For Unit 2, 3.0 Minutes and 26.327 Seconds was added to each Dranetz SOE event time. This normalization process was necessary because only PCS data sources were used to perform the analysis of the Dranetz SOE events.

Summary of Results North Anna Unit 1 Valid Dranetz SOE Recorded events: 183 Outlier Dranetz SOE Recorded events: 25 (8 items are Electrical Power Events)

North Anna Unit 2 Valid Dranetz SOE Recorded events: 231 Outlier Dranetz SOE Recorded events: 38 (11 items are Electrical Power Events)

The Unit 1 and Unit 2 Outlier Dranetz SOE Recorded events are addressed in Attachments 3 and 4, respectively. The review of the Outliers by alternate means verification has eliminated many of them; however, some remain with recommended actions as indicated in Attachments 3 and 4.

Methodology The PCS data source of choice was to use the PCS TRA (i.e., the PCS Transient data base) Analog and Digital Points to validate the recorded Dranetz SOE time stamped events. The TRA Points were chosen because they are recorded and saved to the data base at a time interval of 0.030 Seconds per data point. When TRA Points were not available for use for a particular event analysis, then the normal PCS Analog and Digital Points were used. The normal PCS Analog and Digital Points are recorded at a time interval of > 1.0 Second per data point. Also, the PCS Analog Points have a dead band assigned to each specific point that will prevent the PCS from recording a process change at a specific time if the process signal does not move outside of the dead band. Finally, the Dranetz SOE Recorder, the PCS TRA Points, and the normal PCS Analog and Digital Points are not synced to the same clock. That means recorded data from each PCS source and the Dranetz SOE Recorder is not referenced to the same clock (i.e., time stamp) and the data points used to perform a given analysis may be recorded at different time intervals. In many cases where the process parameter signal actually caused a TRIP and/or NOT TRIP event recording on the Dranetz SOE, the normalized Dranetz SOE time data was skewed when referenced to the PCS data sources due to the items discussed above.

A majority of the Unit 1 and Unit 2 Dranetz SOE event data was processed by the Westinghouse 7300 Protection and Control System. Also, a majority of the 7300 process data was sent to the Westinghouse Solid State Protection System (SSPS) for further processing, i.e., to develop trip functions, control room alarms/annunciators, and PCS/Dranetz alarms/events. The actual measurement and development of the Trip and/or Reset function within the 7300 Protection and Control System is performed by Westinghouse 7300 Analog Comparator (NAL) Cards (i.e., Model Number 2837A13G01, 2837A13G02, or 2837A13G03). Upstream of the NAL Card'in a typical 7300 Instrument Loop is a Loop Power Supply (NLP) Card (i.e., Model Number 2837A12G01, 2837A12G02, or 2837A12G05). Both of these cards have RCE 001061, Rev. 1 Page 39 of 130

low pass filters embedded in their on-board circuitry. This means that trip setpoint accuracy for any given function will be affected when the instrument loop is subjected to relatively high frequency input signals (i.e., between 3 and 7 hz for this event) and fast ramp rates (i.e., > 300 VDC / Minute "nominal" and > 2000 VDC / Minute in some cases for this event). In addition, a small lag component (i.e., = 10 milliseconds) is attributed to the Solid State Protection System for the processing of the digital point signals sent to the PCS and the Dranetz SOE Recorder.

0 When the inaccuracy of the 7300 trip signals is combined with the skewed PCS data sources, judgment had to be used in some cases to determine if the Dranetz event was valid or not. The justification for the determination of the validity of the event function is contained in the Notes section for each line item in the tables.

The rest of the Dranetz SOE event data was developed by individual plant switches and sensors that send on-off electrical signals directly to SSPS, the Dranetz SOE Recorder, and/or to the PCS. In general, the Dranetz SOE event data developed by the Westinghouse 7300 Protection and Control System and the Solid State Protection System was able to be accurately quantified in the attached analysis as compared to the inputs developed from individual plant switches and sensors. The analysis of Dranetz SOE event data developed by individual plant switches and/or sensors was based on the best available TRA or PCS Analog and/or Digital Point data.

In the tables that follow, in the Valid Trip column, if a 'Yes" is entered then the Dranetz event function is considered valid, taking into account the limitations noted above. If a "No" is entered then the Dranetz event function is not considered valid due to PCS Data not supporting the event status or because no PCS or TRA Points were available that could be used to determine the status versus the time stamp. All of the events designated with a "No" are considered to be Outliers and they will be addressed in Attachment 3 (Unit 1) and Attachment 4 (Unit 2).

RCE 001061, Rev. 1 Page 40 of 130

Attachment 10 North Anna Power Station, Unit 1 Analysis of Dranetz event data versus the Plant Computer System (PCS)

For August 23, 2011 RCE 001061, Rev. 1 Page 41 of 130

Attachment 10 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz I PCS) (Status) (Yes / No)

PCS 13:51:10.224 0078/ PRZ HI LEVEL, BU HTRSON ON YES Between t = 13:51:10.224 and t = 13:51:11.784, LiRC001A/2A/3A Pressurizer Level was spiking between a low point of 13:51:10.231 NORMAL YES 58% and a high point of 75%. (PZR HI LVL, BU HTRS ON signal will be generated whenever Pressurizer 13:51:10.786 ON YES Level exceeds 5% of program level which was at 64%

during this time period). Heaters turning on then 13:51:10.800 NORMAL YES returning to normal during the time span are likely.

13:51:11.059 ON YES 13:51:11.091 NORMAL YES 13:51:11.585 ON YES 13:51:11.617 NORMAL YES 13:51:11.685 ON YES 13:51:11.750 NORMAL YES 13:51:11.784 ON YES RCE 001061, Rev. 1 Page 42 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:11.801 0044 / F1 RC003A RC LOOP 1A LO FLOW CH III LO YES At t = 13:51:11.800 RCS flow was spiking down and reached a peak of 92% on the PCS (not the low setpoint of 90%). At t = 13:51:11.829, RCS flow was on an upward spike at t = 13:51:11.829 with level 13:51:11.829 0044 / F1RC003A NORMAL YES indication of 96% for Ch Ill.

13:51:11.840 0078/ PRZ HI LEVEL, BU HTRSON NORMAL YES At t = 13:51:11.840, Pressurizer Level was spiking LiRC001A/2A/3A downward with a minimum value of 56%. Heaters turning off during the time span are likely.

13:51:11.873 0005/ NIS PWR RGE HI FLUX RATE TRIP YES X1RD035D indicates trip a t = 13:51:12.024, Ni's show M1NM007A,8A,9A, RX TRIP negative rate event at approx t= 13:51:11.870 to 10A, X1 RD035D 13:51:11.982 13:51:11.888 0082 / MAIN TRANS SUD PRESS TRIP NO

• There was not an electrical fault within the transformer.

RELAY (63X ABC) The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

13:51:11.888 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES At t = 13:51:11.888 RCS flow was on a decreasing trend to t = 13:51:11.950, CH III flow indication was at 90%.

RCE 001061, Rev. 1 Page 43 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:11.892 0177/ MAIN TRANS LO RELAY TURB TRIP YES 86 = Lockout Relay, Main Transformer 86T relay TRIP (86T) activates with 2 out of 3 SPR actuations (per transformer) causing SST Lockout.

13:51:11.916 0041 / FIRE WATER SYSTEM INIT NO

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:11.918 0195/ MAIN REACTOR TRIP TRIP YES t = 13:51:12.008 BKR Tripped X1RD036D BREAKER B 13:51:11.926 0196/ MAIN REACTOR TRIP TRIP YES t= 13:51:12.024 BKR Tripped XlRD035D BREAKER A 13:51:11.927 0044/ RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:11.930 F1RC003A = 110%

Fl RC003A 13:51:11.935 0120 / Y0390D/ REACTOR TRIPPED-TURB TRIP YES "B" / "A" Rx Trip Breakers opened at t = 1-3:51:12.510 Y0006D / TRIP with a Turbine Trip at t = 13:51:12.558.

Y0007D 13:51:11.944 0079/ SWYD PCB 1C TRIPPED YES Breaker G102-1 and G102-2 are the Unit 1 Generator X1SY001D / output breakers in the switchyard. These breakers XlSY002D tripped due to the GSU/SST SPR/FPR actuation, as I_ __ _desired.

RCE 001061, Rev. 1 Page 44 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:11.952 0072/ SWYD PCB 11 TRIPPED YES Breaker G102-1 and G102-2 are the Unit 1 Generator X1SV001D/ output breakers in the switchyard. These breakers XlSVO02D tripped due to the GSU/SST SPR/FPR actuation, as desired.

13:51:11.971 0082/ MAIN TRANS SUD PRESS NORMAL NO

• There was not an electrical fault within the transformer.

RELAY (63X ABC) The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

13:51:11.975 0041 / FIRE WATER SYSTEM NORMAL NO

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:11.991 0044/ RC LOOP 1A LO FLOW CH 111 LO YES t = 13:51:11.991 RCS flow was spiking downward with F1 RC003A a low flow indication at t = 13:51:12.063 of 93%. PCS indication was not to the low flow setpoint of 90% but the data was cut off at this point.

13:51:11.994 0096/ TRANS 1C SUD PRESS RELAY TRIP NO

• There was not an electrical fault within the transformer.

(63X-1) The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

RCE 001061, Rev. 1 Page 45 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.000 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:12.000, Pressurizer Level was on a L1RC001A/2A/3A downward spike with an immediate upward spike and Pressurizer Level indicated 69% at t = 13:51:12.070.

Heaters turning on during the time span are likely.

13:51:12.003 0176/ SS TRANS 1C LO RELAY TRIP YES 86 = Lockout Relay, Main Transformer 86T relay TURB TRIP activates with 2 out of 3 SPR actuations (per transformer) causing SST Lockout.

13:51:12.015 0096/ SS TRANS 1C SUD PRESS NORMAL NO

• There was not an electrical fault within the transformer.

RELAY (63X-1) The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

13:51:12.015 0044/ RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:12.014 F1RC003A = 99%

F1 RC003A 13:51:12.033 0119 / Y0390D SOLENOID TURB TRIP TRIP YES This is the 7Y/AST Solenoid Position Indicator switch (window 1E-F1). It is valid, given the turbine trip on PCS at t = 13:51:12.558.

RCE 001061, Rev. 1 Page 46 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.047 0107/ RSS TRANS B SUD PRESS RELAY TRIP NO* There was not an electrical fault within the (63A) transformer. The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

13:51:12.050 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.050, Pressurizer Level was Li RC001 A/2A/3A indicating 58% on the PCS. Heaters returning to normal during the time span are likely.

13:51:12.055 0189/X1TGO02D GEN BKR G12 TRIPPED YES t = 13:51:12.114 PCS indicates Breaker Tripped 13:51:12.056 0058 / Y0390D/ TURBINE TRIP RX TRIP TRIP YES t = 13:51:12.557 PCS indicates Tripped Xl RD035D /

XlRD036D 13:51:12.058 0106/ RSS TRANS A SUD PRESS RELAY TRIP NO* There was not an electrical fault within the (63A) transformer. The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

RCE 001061, Rev. 1 Page 47 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.079 0038/ STM GEN 1A HI-HI LEVEL TURB TRIP YES At t = 13:51:12.080 Steam Generator Levels were L1FW002N3A/1A TRIP on an increasing trend with max indications at 72%

on CH II. At t = 13:51:12.172, Steam Generator Levels spiked upward with a max indication at 78%.

(PCS data on CH II was cut off at 71% on increasing spike.) SG 1A HI-HI LEVEL TURB TRIP setpoint is 75% increasing levels.

13:51:12.082 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES At t = 13:51:12.080 RCS flow indication was on a decreasing trend to 87% at t = 13:51:12.143.

13:51:12.090 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:12.096, Pressurizer Level was on an L1RC001A/2A/3A upward trend with max indication of 87%. Heaters turning on during the time span are likely.

13:51:12.094. 0038/ STM GEN 1A HI-HI LEVEL TURB NORMAL YES At t = 13:51:12.094 Steam Generator Level L1FW002A/3A/1A TRIP indications were at 51% which is below the HI-HI Level trip setpoint of 75%.

13:51:12.118 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES At t = 13:51:12.119 F1RC003A = 97%

RCE 001061, Rev. 1 Page 48 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID (Dranetz Point Description Message Valid Trip Notes Normalized PCS) (Status) (Yes/ No) to PCS 13:51:12.120 0027/L1RC001A/2A/3A PRESSURIZER HI LEVEL RX TRIP TRIP YES Pressurizer Level spiked up with a peak reading of 99% at t = 13:51:12.101 with a downward trend at t

13:51:12.120. (PZR HI LVL RX TRIP setpoint

92%)

13:51:12.125 0045/X1RC001D RCP 1ACH 1 BKR OPEN YES t = 13:51:12.188 PCS indicates "A" RCP Breaker Open 13:51:12.133 0076 /L1QS001A RWST CHEM ADD TK LO LEVEL LO NO* t = 13:51:12.132 L1QS001A = 88% (Low Level Trip setpoint = 85%, RWST CAT Level did not change/decrease below trip setpoint) 13:51:12.134 0027/L1RC001A/2A/3A PRESSURIZER HI LEVEL RX TRIP NORMAL YES At t = 13:51:12:132, Pressurizer Level Ch I, II, and III were indicating 61%, 72%, and 68%. (PZR HI LVL RX TRIP reset = 91% decreasing level) 13:51:12.144 0129/L1RC001A/2A/3A LOSS OF COOL FLOW PWR >30% TRIP YES At t = 13:51:12.120, RCS Flow was spiking

/M1NMO07A/ downward, at t = 13:51:12.164 RCS flow reached M1NM008A/ low peaks of 87% on CH III and 94% on CH II with M1 NMO09A NI Pwr > 30% (Loss of Cool Flow Pwr > 30% trip MlNMO10A setpoint is 2 out of 3 RC Low Flow Trips on 1 out of 3 loops with Pwr greater than P8)

RCE 001061, Rev. 1 Page 49 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.158 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.158, Pressurizer Level Ch 1,11, and L1RC001A/2A/3A III were indicating 67%, 63%, and 66% which are all below the 5% above Program Level trip setpoint.

Heaters returning to normal during the time span are likely.

13:51:12.177 0044 /F1RC003A RC LOOP 1A LO FLOW CH III LO YES t = 13:51:12.180 F1RC003A = 87%

13:51:12.188 0076 / L1QS001A RWST CHEM ADD TK LO LEVEL NORMAL YES t = 13:51:12.188 L1QS001A = 88%, with a low level reset setpoint of 86%. (Level did not change) 13:51:12.194 0134/ FEED WTR PP BKRS OPEN TURB TRIP YES t = 13:51:13.550 PCS indicates "B" & "C" X1 FW011D,012D, TRIP Feedwater pump breakers open 013D, 14D,01 5D,01 6D 13:5112.205 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:12.204 F1RC003A = 105%

13:51:12.212 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:12.211, Pressurizer Level Ch II and III LI RC001 A/2A/3A were indicating 88% and 82%. Heaters turning on normal during the time span are likely.

RCE 001061, Rev. 1 Page 50 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.213 0041 / FIRE WATER SYSTEM INITIATED INIT NO

• No TRA or PCS Computer Points for this function.

13:51:12.242 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.240, Pressurizer Level minimum Li RC001 A/2A/3A value was indicating 58%. Heaters returning to normal during the time span are likely.

13:51:12.242 0013 / T0499A/ TAVG > < TREF DEV DEVIATION YES At t = 13:51:12.242, Tref = 580.6 and Tavg = 580.4 T0496A (Diff of 0.2);

At t = 13:52:13.470 Tavg = 580.4 and Tref = 573.4 (Diff of > 5 degrees). Tref starts trending down at t

= 13:51:12.242 13:51:12.303 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:12.301, Pressurizer Level signal was Li RC001 A/2A/3A spiking up with a peak reading of 78%. Heaters turning on during this time are likely.

13:51:12.352 0044/F1RC003A RCLOOP1ALOFLOWCHIII LO YES Loop A, Ch Ill, t = 13:51:12.421, Valley =

90.6007%.

RCE 001061, Rev. 1 Page 51 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.352 0038/ STM GEN 1A HI-HI LEVEL TURB TRIP YES At t = 13:51:12.352 Steam Generator Level CH II Li FW002A/3A/1A TRIP and III were on an increasing trend and indicating 71% and 61% (while the HI-HI Level Trip setpoint is 75%, the data for CH II and III were cut off at 719%/61%).

0038 /

13:51:12.369 L1FW002A/3A/1A STM GEN 1A HI-HI LEVEL TURB NORMAL YES SG Level indications were trending downward from t = 13:51:12.364 to t = 13:51:12.382 with CH II and TRIP III indicating 71% and 65%.

13:51:12.380 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:12.388 F1RC003A = 104%

13:51:12.426 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.330, PZR Level spiked downward Li RC001 A/2A/3A with a low peak of 30% and then spiked upwards where t = 13:51:12.426, max PZR Level was 88%.

Heaters returning to normal during this time are likely..

13:51:12.443 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES At t = 13:51:12.441 RCS flow was on a decreasing trend with F1 RC003A indicating 90%. (PCS data was cut off at 90%)

RCE 001061, Rev. 1 Page 52 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.482 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:12.482, Pressurizer Level max value Li RC001 A/2A/3A was indicating 86%. Heaters turning on during this time are likely.

13:51:12.482 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES At t = 13:51:12.480 RCS flow was on an increasing trend with F1 RC003A indicating 105%

13:51:12.509 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.510, Pressurizer Level minimum LiRC001A/2A/3A value indicated was 53%. Heaters returning to normal during this time are likely.

13:51:12.516 0116/V0321D/ STASERV2OF3BUSESUV-RX TRIP YES 13:51:12.556 SS/RCP BUS 1A / 1B indicated V0322D TRIP tripped 13:51:12.542 0044 / F1 RC003A RC LOOP 1A LO FLOW CH III LO YES t = 13:51:12.540 F1 RC003A = 84%

13:51:12.558 0108/ RSS TRANS C SUD PRESS RELAY TRIP NO* There was not an electrical fault within the (63A) transformer. The SPR (i.e., FPR) actuated due to the earthquake which it is not designed to intentionally detect.

13:51:12.578 0044/F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:12.578 F1RC003A = 105%

RCE 001061, Rev. 1 Page 53 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.618 0078/ PRZ HI LEVEL, BU HTRSON ON YES At t = 13:51:12.600 Pressurizer Level spiked LiRC001A/2A/3A upwards with a max level indication of 74% with an immediate downward spike to t = 13:51:12.618 with Pressurizer Level indication of 54%. Heaters turning on during this time are likely.

13:51:12.657 0002/ STM GEN 1A LEVEL ERROR ERROR YES At t = 13:51:12.657 Ch I was 47.8%, Ch II was Li FW002A/3A/1A 62.6% and Ch III was 56.2%. Level error is +/- 5%

from program Level = 44%.

13:51:12.695 0004/C0099D, RPI ROD BOTTOM ROD DROP DROP NO

• At RPI Rod Bottom/Rod Drop PCS point indicates C0002A,C001 5A, Dropped at t = 13:51:13.520 with RPI PCS C0026A, C0038A/ indications dropping to the Rod Bottom Setpoint XlRD036D / after the alarm occurs. Alarm occurred at the same V1EE019A time as "B" Rx Trip Breaker went open. MCC 1H1-1 voltage source for the RPI was at 478.99 volts at this time which shows a decrease in RPI source voltage.

13:51:12.700 0031 / RWST CHEM ADD TK LO TEMP LO NO

• No TRA or PCS Computer Points for this function.

RCE 001061, Rev. 1 Page 54 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.710 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Pressurizer Level was spiking between t =

LiRC001A/2A/3A 13:51:12.700 to t = 13:51:12.884 with max low and high indications were 63% and 78%. Heaters 13:51:12.747 0078 / PRZ HI LEVEL, BU HTRS ON ON YES turning on then returning to normal during the time L1 RC001 A/2N3A span are likely.

13:51:12.811 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Li RC001 A/2A/3A 13:51:12.827 0044 /F1 RC003A RC LOOP 1A LO FLOW CH III LO YES At t = 13:51:12.826 RCS flow was on a decreasing trend indicating 91% (PCS indication did not decrease below the low flow setpoint of 90%, but PCS data was cut off at 91%)

13:51:12.845 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES RCS flow was on an increasing trend from t =

13:51:12.845 to t = 13:51:12.858 with indication of 93%

1.3:51:12.883 0078/ PRZ HI LEVEL, BU HTRS ON ON YES Pressurizer Level was spiking, between t =

Li RC001A/2A/3A 13:51:12.700 to t = 13:51:12.884 with max low and high indications were 63% and 78%. Heaters turning on during the time are likely.

13:51:12.886 0076 / L1QS001A RWST CHEM ADD TK LO LEVEL LO NO

• t = 13:51:12.886 L1QS001A = 88% (Low Level Trip setpoint = 85%, RWST CAT Level did not change/decrease below trip setpoint)

RCE 001061, Rev. 1 Page 55 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:12.891 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:12.892, Pressurizer Level minimum LiRC001A/2A/3A value indication was 51%. Heaters returning to normal during this time are likely.

13:51:12.915 0044/F1RC003A RCLOOP1ALOFLOWCHIII LO YES Loop C, CH III, t = 13:51:12.972 Valley =

91.1806%, t = 13:51:12.991 F1RC003A =

13:51:12.943 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES 92.0246% (RC LO FLOW TRIP = 90% decreasing, Reset = 90.8% increasing level) 13:51:12.964 0076 / L1QS001A RWST CHEM ADD TK LO LEVEL NORMAL YES At t = 13:51:12.964 L1QS001A = 88% (Low Level reset point = 86%, Level did not change) 13:51:13.002 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES Loop C, CH III decreasing trend to 91.1806% at t =

13:51:12.980, slight increasing trend to 92.0246%

at t=13:51:13.002 13:51:13.010 0078/ PRZ HI LEVEL, BU HTRSON ON YES At t = 13:51:12.980, Pressurizer Level was on a Li RC001 A/2A/3A downward spike with max level indication of 78%.

At t = 13:51:13.010, Pressurizer Level was indicating 67%. Heaters turning on during this time L____ are likely.

RCE 001061, Rev. 1 Page 56 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.034 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:13.034 F1RC003A = 99%

13:51:13.094 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES RCS flow was on a decreasing trend to 91% at t =

13:51:13.082 13:51:13.104 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Between t = 13:51:13.001 and t = 13:51:13.104, LiRC001A/2A/3A Pressurizer Level spiking: Valley = 50% and Peak =

74%. Heaters returning to normal during this time are likely.

13:51:13.129 0044/ F1 RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t =13:51:13:128 FlRC003A=100%,

13:51:13.259 0021 /X2TGO01D EXCITATION LOSS OF POWER POWER YES Since SPR's activated and the 86T relay actuated (ALl -2, AL4-5) for the GSUs, the Exciter field breaker tripped causing the Excitation Loss of Power to go from Normal to Power (alarm condition).

RCE 001061, Rev. 1 Page 57 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.276 0044 / F1RC003A RC LOOP 1A LO FLOW CH III LO YES t = 13:51:13.264 RCS flow was spiking down then reached an upward spike at t = 13:51:13.318. Note 13:51:13.284 0043/ F1RC002A RC LOOP 1A LO FLOW CH II LO YES PCS indicates a low spike of 92% level (not to the low flow setpoint of 90 %) but the PCS data was 13:51:13.315 0129/ P0398A, LOSS OF COOL FLOW PWR > 30% NORMAL YES cutoff at this point. PCS indicates at t =

P0399A 13:51:13.296 1 st stage pressure was 582 psig at t =

13:51:13.529 1 st stage pressure was 342 psig.

13:51:13.316 0058 / Y0390D / TURBINE TRIP RX TRIP NOT TRIP YES t = 13:51:13.470 PCS indicates not tripped X1 RD035D /

X1 RD036D 13:51:13.322 0043 / F1RC002A RC LOOP 1A LO FLOW CH II NORMAL YES t = 13:15:13.322 RCS flow was on an upward spike with level indication 92.85% and 93.44% for Ch II 13:51:13.323 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES and II1.

13:51:13.329 0078 / PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:1.322 Pressurizer Level started spiking Li RC001 A/2A/3A up from 53% level to a max peak of 81% level at t=

13:51:13.371 PRZ HI LEVEL, BU HTRS ON NORMAL YES 13:51:13.424. Level then spiked down to 64% at t =

0078/ 13:51:13.458. Heaters turning on then returning to L RC001 A/2A/3A normal during the time span are likely.

RCE 001061, Rev. 1 Page 58 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.383 0043 / F1RC002A RC LOOP 1A LO FLOW CH II LO YES t = 13:51:13.380 CH II and III were below the trip setpoint of 90% flow with indication of a start of an 13:51:13.398 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES upward trend. At t = 13:51:13.416 Ch II flow was above reset point i.e. 94.23%.

13:51:13.416 0043 F1RC002A F RC LOOP 1A LO FLOW CH II NORMAL YES 13:51:13.419 0078/ PRZ HI LEVEL, BU HTRS ON ON YES t = 13:51:13.414 Pressurizer Level was spiking up LiRC001A/2A/3A and reached it max value of 81% at t =

13:51:13.424 which is 5% greater than program level.

13:51:13.425 0047 / F1RC005A RC LOOP 1B LO FLOW CH II LO YES t = 13:51:13.390 Ch I and II of "B" RCS Flow was on a downward spike with the trip setpoint being 0130/ LOSS OFCOOL FLOW PWR > 10 % TRIP YES reached at approximately t = 13:51:13.498. PCS 13:51:13.426 F1RC004A/5A/6A indicates at t = 13:51:13.296 1 st stage pressure was 582 psig at t = 13:51:13.529 1 st stage pressure was 13:51:13.428 0046/ F1RC004A RC LOOP 1B LO FLOW CH I LO YES 342 psig.

t = 13:51:13.428 top of upward spike with Ch III "A" RCS flow indicating 96%.

13:51:13.429 0044/F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES RCE 001061, Rev. 1 Page 59 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.446 0078/ PRZ HI LEVEL, BU HTRSON NORMAL YES t = 13:51:13.446 Pressurizer Level was spiking Li RC001A/2A/3A down and reached the programmed Level at t =

13:51:13.458.

13:51:13.470 0043/F1 RC002A RC LOOP 1A LO FLOW CH II LO YES t = 13:51:13.470 "A" RCS flow was on a downward spike below Low Flow setpoint of 90% for both Ch II 13:51:13.481 0044/F1RC003A RC LOOP 1A LO FLOW CH III LO YES and I11. RCS flow was cycling above and below setpoint until t = 13:51:13.600 at which time flow 13:51:13.508 0043/ F1RC002A RC LOOP 1A LO FLOW CH II NORMAL YES was at the peak of an upward spike with an equivalent flow of 92.5% for Ch II and 95% for Ch 13:51:13.521 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES Ill.

13:51:13.554 0043 / F1RC002A RC LOOP 1A LO FLOW CH II LO YES 13:51:13.561 0044/ F1RC003A RC LOOP 1A LO FLOW CH Ill LO YES 13:51:13.592 0048 / F1RC006A RC LOOP 1B LO FLOW CH Ill LO YES t = 13:51:13.568 Ch Ill "B" RCS flow was spiking downward with the trip setpoint of 90% being 13:51:13.599 0043 I F1RC002A RC LOOP 1A LO FLOW CH II NORMAL YES reached at t = 13:51:13.644 13:51:13.600 0078/ PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:13.600 Pressurizer Level was at the LiRC001A/2A/3A peak of a downward spike then started an upward spike reaching the max peak 69.5% at t =

_ __ _13:51:13.708 which could cause the ON status.

RCE 001061, Rev. 1 Page 60 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.608 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:13.606, F1RC003A = 92%

13:51:13.622 0078 / PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:13.708 Pressurizer Level was at a max Li RC001 A/2A/3A peak of 69.5% then spiked down reaching programmed level at approximately t =

13:51:13.714 and level of 63.89%.

13:51:13.637 0043/ F1RC002A RC LOOP 1A LO FLOW CH II LO YES t = 13:51:13.630 RCS flow was on a downward trend and reached the low flow setpoint of 90% at t 13:51:13.649 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES = 13:51:13.640 and then started an upward trend and reached the peak of 91.4% on Ch II at t =

13:51:13.690 0043/ F1RC002A RC LOOP 1A LO FLOW CH II NORMAL YES 13:51:13.694.

13:51:13.691 0078/ PRZ HI LEVEL, BU HTRS ON ON YES From t = 13:51:13'672 to t = 13:51:13.758 Li RC001 A/2A/3A Pressurizer Level was spiking from 65% to 73%

RCE 001061, Rev. 1 Page 61 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz I PCS) (Status) (Yes / No) to PCS 13:51:13.699 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:51:13.694 RCS Flow was at a max peak of 95.8% on Ch III and started a downward trend at 13:51:13.732 0043 / F1RC002A RC LOOP 1A LO FLOW CH II LO YES which the low peak was reached at t =

13:51:13.728, where Ch II reached a value of 13:51:13.743 0044 / F1RC003A RC LOOP 1A LO FLOW CH III LO YES 86.2% and Ch III was 84.83%.

13:51:13.773 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:13.772 Pressurizer Level was at a peak L1RC001A/2A/3A value of 72.19% going down and reached the reset value at approx. t = 13:51:13.794 13:51:13.793 0043 / F1RC002A RC LOOP 1A LO FLOW CH II NORMAL YES t = 13:51:13.790 RCS flow was on a upward spike with flow starting below the trip setpoint of 90% for 13:51:13.798 0044 / F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES both Ch II and Ill. RCS flow was cycling above and below setpoint with Ch I and Ch II staying below the 13:51:13.801 0042 / F1RC001A RC LOOP 1A LO FLOW CH I LO YES trip setpoint until t = 13:51:13.910 at which time flow for Ch III Was at the peak of an upward spike 13:51:13.811 0043/ F1RC002A RC LOOP 1A LO FLOW CH II LO YES with an equivalent flow of 91.7%.

13:51:13.827 0044 / F1RC003A RC LOOP 1A LO FLOW CH III LO YES 13:51:13.885 0044 / F1RC003A RC LOOP 1A LO FLOW CH Ill NORMAL YES 13:51:13.910 0044/F1RC003A RC LOOP 1A LO FLOW CH Ill LO YES RCE 001061, Rev. 1 Page 62 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:13.960 0078 / PRZ HI LEVEL, BU HTRS ON ON YES At t = 13:51:13.902 Pressurizer Level appears to be LiRC001A/2A/3A steady at 62%, then starts a downward spike at t =

13:51:13.984 to a low value of 48% then spikes upward to a max value of 80% at t = 13:51:14.054 13:51:13.992 0044/ F1RC003A RC LOOP 1A LO FLOW CH III NORMAL YES t = 13:15:13.984 a rapid downward spike was noted on Ch II and Ch III and the minimum peak was 13:51:13.996 0044/ F1RC003A RC LOOP 1A LO FLOW CH III LO YES reached at t = 13:51:14.010 with Ch III being at 84.6% which is below the trip setpoint of 90%.

13:51:13.998 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:13.998 Pressurizer Level is spiking L1RC001A/2A/3A downward then starts an upward spike at t =

13:51:14.010 reaching a peak value of 81% at t =

13:51:14.085 0078 / PRZ HI LEVEL, BU HTRS ON ON YES 13:51:14.054. Level then spikes downward and at t LiRC001A/2A/3A = 13:51:14.100 LVL is at 57%.

13:51:14.099 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Li RC001 A/2A/3A 13:51:14.279 0052/ F1RC009A RC LOOP 1C LO FLOW CH III LO YES t = 13:51:14.280 Ch III was on a downward trend with Low flow setpoint being reached at t =

13:51:14.342.

13:51:14.327 0051 / F1RC008A RC LOOP 1C LO FLOW CH II LO YES t = 13:51:14.320 Ch II was on a downward trend with Low flow setpoint being reached at t =

13:51:14.408.

RCE 001061, Rev. 1 Page 63 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:14.337 0078/ PRZ HI LEVEL, BU HTRSON ON YES Between t = 13:51:14.334 and t = 13:51:14:412, Li RC001 A/2A/3A Pressurizer Level was spiking with Valley = 58.4%

and Peak = 71%.

0078 /

13:51:14.376 L1RC001A/2A/3A PRZ HI LEVEL, BU HTRS ON NORMAL YES At t = 13:51:14.461, Pressurizer Level spiked down, Valley = 56%.

13:51:14.421 0050 / F1RC007A RC LOOP 1C LO FLOW CH I LO YES Ch I was trending downward with the trip setpoint being reached at t = 13:51:14.488 13:51:14.693 0078/ PRZ HI LEVEL, BU HTRSON ON YES At t = 13:51:14.680 Pressurizer Level reaches a LiRC001A/2A/3A max value of 66% Level from an upward spike and then the trace flattens out until t = 13:51:14.710 at 13:51:14.711 0078 / PRZ HI LEVEL, BU HTRS ON NORMAL YES which it starts trending downward reaching a peak Li RC001 A/2A/3A value of 57% at t = 13:51:14.874. Level then starts an upward spike to 70% at t = 13:51:15.058 then 13:51:14.947 0078/ PRZ HI LEVEL, BU HTRSON ON YES spikes down to 58% at t = 13:51:15.074 and L RC001 A/2A/3A continues spiking with the largest peak of 68.78%

at t= 13:51:15.330.

13:51:14.978 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Li RC001 A/2A/3A 13:51:15.231 0078/ PRZ HI LEVEL, BU HTRS ON ON YES Li RC001A/2A/3A 13:51:15.244 0078/ PRZ HI LEVEL, BU HTRS ON NORMAL YES Li RC001A/2A/3A RCE 001061, Rev. 1 Page 64 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:15.530 0041 / FIRE WATER SYSTEM INITIATED NORMAL NO

• No TRA or PCS Computer Points for this function.

13:51:15.703 0078/ PRZ HI LEVEL, BU HTRSON ON YES At t = 13:51:15.672 Pressurizer Level starts an Li RC001A/2A/3A upward spike from 57% to 65% at t = 13:51:15.710 and then spikes downward to 55% at t =

13:51:15.721 0078 / PRZ HI LEVEL, BU HTRS ON NORMAL YES 13:51:15.730.

Li RC001 A/2A/3A 13:51:15.820 0071/L1FW006A STM GEN 1B LO-LO LEVEL CH I-I1-111 LO-LO YES At t = 13:51:15.818 all three S/G Level Channels were 20% and spiking down with CH III spiking to 16.7% at t = 13:51:15.868 and Ch I & II at 18% and 19%.

13:51:15.827 0003 / L1FW006A STM GEN 1B LEVEL ERROR ERROR YES At t = 13:51:15.826 Ch I was 20%, Ch II was 20.6%

and Ch III was 19%. Level error is +/- 5% from Program Level.

13:51:15.840 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-Il-Ill LO-LO YES At t = 13:51:15.840 S/G Level Ch I was 18.81% on a downward spike and reached 18% at t =

13:51:15.870 0062/L1FW002A STM GEN 1A LO-LO LEVEL CH I-I1-111 NORMAL YES 13:51:15.844. The Level started an upward spike at t

= 13:51:15.874 and reached a peak value of 20.33%

13:51:15.930 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-Il-Ill LO-LO YES at t = 13:51:15.900 then spiked downward to 17.73%

at t = 13:51:15.944.

RCE 001061, Rev. 1 Page 65 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:15.939 0029/ LIFW006A STN GEN 1B LO-LO LEVEL RX TRIP TRIP YES At t = 13:51:15.940 "B" S/G Level was CH I =

17.8%, Ch II = 18.6% and Ch III = 17.7%

13:51:15.977 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-I1-111 NORMAL YES At t = 13:51:15.944 CH I S/G Level was 17.73%

and flat lining until t = 13:51:16.028 when. Level starts an upward spike to 18.87% at t =

13:51:16.013 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-I1-111 LO-LO YES 13:51:16.084 then spikes down to 16.95% at t =

13:51:16.108 13:51:16.088 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-I1-11 NORMAL YES 13:51:16.103 0062 / L1FW002A STM GEN 1A LO-LO LEVEL CH I-I1-111 LO-LO YES RCE 001061, Rev. 1 Page 66 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz I (Status) (Yes ! No) to PCS PCS) 13:51:16.437 0073/ STM GEN 1C LO-LO LEVEL CH I-I1-111 LO-LO YES At t = 13:51:16.436 Ch I = 20%, Ch II = 21% and Ch IIl =

L1FW01OA 22.7% with Ch III spiking down. At t = 13:51:16.470 Ch I

& II peaked high at 20.9% and 22% with Ch III starting 13:51:16.466 0073/ STM GEN 1C LO-LO LEVEL CH I-Il-Ill NORMAL YES an upward spike. At t = 16.504 Ch I, II, and Ill were L1FW01OA spiking down with Ch I reading a low peak of 17.95% at t

= 13:51:16.514. At t = 13:51:16.520 all 3 channels had 13:51:16.505 0073/ STM GEN 1C LO-LO LEVEL CH I-I1-11 LO-LO YES spiked down to a low value of 17.95% Ch 1, 20.2% Ch II L1FW01OA and 19.6% Ch III at t = 13:51:16.564 Ch II & III were 20.9% and 21.8% and at t = 13:51:16.610 Ch I was 13:51:16.540 0124/ STM GEN 1C LO-LO LEVEL RX TRIP TRIP YES 17.2% and Ch II & III were spiking down to approx. 19%

LIFW01OA Level at t = 13:51:16.654 Level then spiked up and was falling and were indicating approx. 19.8% on Ch II and 13:51:16.568 0124 / STM GEN 1C LO-LO LEVEL RX TRIP NORMAL YES 19.97% on Ch III.

L1FW01OA 13:51:16.608 0124/ STM GEN 1C LO-LO LEVEL RX TRIP TRIP YES L1 FW01OA 13:51:16.651 0124/ STM GEN 1C LO-LO LEVEL RX TRIP NORMAL YES L1FW01OA 13:51:16.665 0028/ STM GEN 1A LO-LO LEVEL RXTRIP TRIP YES At t = 13:51:16.664 "A" CH I = 14.85% and Ch III is Li FW002A spiking down to a value of 19.8%

RCE 001061, Rev. 1 Page 67 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz I (Status) (Yes / No)

PCS PCS) 13:51:16.669 0014/ STM GEN 1C LEVEL ERROR ERROR YES At t = 13:51:16.670 Ch I = 15% Level, Ch II = 18.5%

L1FW01OA Level and CH III = 19.79% Level and all three were 13:51:16.672 STM GEN 1C LO-LO LEVEL RX TRIP TRIP YES spiking down. The RX Trip setpoint was reached at t =

0124/ 13:51:16.686 L1EFW01OA 13:51:16.694 0028/ STM GEN 1A LO-LO LEVEL RXTRIP NORMAL YES At t = 13:51:16.694 CH I is 13.7% Level and CH III Li FW002A spikes up to 21% Level.

13:51:16.732 0124/ STM GEN 1C LO-LO LEVEL RXTRIP NORMAL YES At t = 13:51:16.732 Ch I = 16.9% and Ch II and III were L1FW01OA 19.7% and 19.3%.

13:51:16.736 0028/ STM GEN 1A LO-LO LEVEL RXTRIP TRIP YES At t = 13:51:16.736 "A" CH I = 12.88% and Ch III is Li FW002A spiking down to a value of 18.6%

13:51:16.756 0124/ STM GEN 10 LO-LO LEVEL RX TRIP TRIP YES At t = 13:51:16.756 Ch I= 15.9%, Ch 11 = 18.3%, and Ch L1FW010A III = 18%. At t = 13:51:16.812 Ch I = 15.6%, Ch II =

18.8% and Ch III = 19.3% and all three channels were 13:51:16.812 0124/ STM GEN 1C LO-LO LEVEL RXTRIP NORMAL YES on an upward spike. At t = 13:51:16.836 Ch I = 14.9%,

L1FW01OA Ch 11 = 18.2% and Ch III = 18.4% and all three were on a downward spike.

13:51:16.836 0124/ STM GEN 1C LO-LO LEVEL RX TRIP TRIP YES 1FW01OA RCE 001061, Rev. 1 Page 68 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:16.867 0002/ L1FW002A STM GEN 1A LEVEL ERROR ERROR YES At t = 13:51:16.867 CH I Level = 11.4%, CH II was 20.66%, and CH III was 16.4%. All are < 5% of the programmed value of 33% Level.

13:51:16.907 0124/LiFW010A STM GEN 1C LO-LO LEVEL RX NORMAL YES At t = 13:51:16.906 Ch II & III were on an upward spike TRIP reading 17.2% and 16.6% at t = 13:51:16.920 peak of upward spike Ch II and III were 19% and 18.6% then 13:51:16.917 0124/LFW01OA STM GEN 1C LO-LO LEVEL RX TRIP YES spiked down below the trip setpoint at t = 13:51:16.932 TRIP 13:51:17.013 0041 / FIRE WATER SYSTEM INIT NO

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:18.508 0001 / SMOKE DETECTION SYSTEM TROUBLE NO

• No TRA or PCS Computer Points for this function.

RCE 001061, Rev. 1 Page 69 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz! PCS) (Status) (Yes / No)

PCS 13:51:19.452 0036 / T0444A, LOOP 1C HI DELTAT HI NO

• At t = 13:51:19:466 Loop C Control = 85.13%, Loop C T0443A T0497A DEVIATION Protection = 97.73% and Median Delta T = 84.27%.

Protection to Median Delta T Difference was 13.46%,

however, Control to Median Delta T Difference was only 0.86%. The comparison of Protection to Median Delta T would have actuated this alarm, however the difference between Control and Median would not have due to the difference between Loop and Median Delta T Deviation Trip setpoint must be a +3.2% difference. There is a discrepancy between the protection and control readings during this event.

13:51:20.141 0031 / RWST CHEM ADD TK LO TEMP NORMAL NO

• Could not find a computer point to validate 13:51:20.623 0025 / Y0004D, MANUAL INITIATION RX TRIP TRIP YES t = 13:51:21:536 PCS indicates TRIP Y0005D 13:51:21.609 0025 / Y0004D, MANUAL INITIATION RX TRIP NORMAL YES t = 13:51:22.546 PCS indicates Not Tripped Y0005D 13:51:22.740 0001 / SMOKE DETECTION SYSTEM NORMAL NO

• No TRA or PCS Computer Points for this function.

13:51:25.092 0001 / SMOKE DETECTION SYSTEM TROUBLE NO

• RCE 001061, Rev. 1 Page 70 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:25.349 0132 / MANUAL INITIATION TURB TRIP TRIP NO

• Could not find a computer point to validate 13:51:25.884 0132/ MANUAL INITIATION TURB TRIP NORMAL . NO

• 13:51:27.657 0036 / T0443A, LOOP 1C HI DELTA T NORMAL YES At t = 13:51:27.660 Loop C Control = 35.53%, Loop C T0444A, DEVIATION Protection = 44.19% and Median Delta T = 34.06%. The T0497A Protection to Median Delta T Difference was 10.13%,

however, Control to Median Delta T Difference was only 1.47%. The comparison of Protection to Median Delta T would not have allowed the point to go to normal, however the difference between Control and Median would have gone back to normal condition. The difference between Loop and Median Delta T Deviation reset must be less than +1.7% difference. There is a discrepancy between the protection and control readings during this event.

13:51:32.476 0013/ T0499A / TAVG > <TREF DEV NORMAL YES At t = 13:51:32.430 Tref = 557.73 and Tavg = 563.07, at T0498A t = 13:51:34.529 Tref = 557.73 and Tavg = 562.15 (PCS processing time)

RCE 001061, Rev. 1 Page 71 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz I PCS) (Status) (Yes / No)

PCS 13:51:33.151 0041 / FIRE WATER SYSTEM NORMAL NO

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:38.218 0053 / RCP CH 3 BKR OPEN YES "C" RCP Breaker indicates OPEN at t = 13:51:38.284 XlRC003D 13:51:43.494 0188/ AMSAC INITIATED YES At t = 13:51:43.256 AMSAC was not initiated per the X2RX004D/ PCS Computer . But between t = 13:51:45.420 and X2RX005D 13:51:46.506 AMSAC did initiate.(PCS scan rate is different than Drantez) 13:51:44.531 0049 / RCP 1B CH 2 BKR OPEN YES "B" RCP breaker indicates open at t = 13:51:44.594 XlRC002D RCE 001061, Rev. 1 Page 72 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:52:09.867 0013 / T0499A/ TAVG > < TREF DEV DEVIATIO YES At t = 13:52:09.860, Tref = 556.16 and Tavg = 560.8 (Diff T0498A N of 4.64) at t = 13:52:10.552 Tref = 555.8 and tref = 560.8 (Diff of 5) 3:53:21.070 0033/ LOOP 1A LO DELTA T LO YES At t = 13:53:21.084 Loop A Control = 38.60%, Loop A T0403A,T0404A, DEVIATION Protection = 38.43% and Median Delta T = 41.73%. The T0497A Protection to Median difference = 3.3% and the Control to Median Difference = 3.13%. The Delta T Deviation Trip setpoint +/- 3.2% difference between the Loop and Median Delta T signals.

At t = 13:53:39.596 Loop A Control = 47.02%, Loop A 13:53:36.544 0033 / NORMAL YES Protection = 45.69% and Median Delta T = 47.94%. The T0403A,T0404A, LOOP 1A LO DELTA T Protection to Control difference = 2.25% and the Control T0497A DEVIATION to Median difference = 0.92% The Loop Delta T signal was trending down and would have cleared the trip signal.

13:56:10.211 0116 / V0321 D/ STA SERV 2 OF 3 BUSES UV- NORMAL NO

• PCS Points out of service / No data available from 23-V0322D RX TRIP AUG-2011 t=1 3:52:09.987 to 14-SEP-2011 t=10:25:24.246. Points currently indicating "NOT H_ TRIPPED" RCE 001061, Rev. 1 Page 73 of 130

Attachment 10 North Anna Power Station Unit 1 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:56:10.212 0130/ P0398A / LOSS OF COOL FLOW PWR > NORMAL YES At t = 13:56:09.260 RCS Flow is zero and Turbine 1 st P0399A / FIRC001A/ 10% stage pressure is 47 and 48 psig. Which is <10% turbine F1 RC002A / load.

F1 RC003A 13:56:50.170 0033 / T0403A / LOOP 1A LO DELTA T LO YES At t = 13:56:52.596 Control Loop A = 61.85%, Protection T0404A / T0497A DEVIATION Loop A = 62.64, Median Delta T = 65.08%, Protection to Median Difference = 2.44%, Control to Median Difference = 3.23%.

0033 /T0403A/ At t = 14:00:14.510 Control Loop A = 57.11, Protection 14:00:14.359 LOOP 1A LO DELTA T NORMAL YES Loop A = 57.07%, Median Delta T = 58.70%. Protection T0404A / T0497A DEVIATION to Median Difference = 1.63%, Control to Median I_ I I I Difference = 1.59%

• See Attachment 3 RCE 001061, Rev. 1 Page 74 of 130

Attachment 10 North Anna Power Station Unit 2 North Anna Power Station, Unit 2 Analysis of Dranetz event data versus Plant Computer System (PCS) data For August 23, 2011 RCE 001061, Rev. 1 Page 75 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:11.072 0076 /L2QS001A RWST CHEM ADD TK LO LEVEL LO No

• Level constant at 86.249% till time 13:51:13.170 13:51:11.224 0076 /L2QS001A RWST CHEM ADD TK LO LEVEL NORMAL No

• at which point level increases to 87.090%

13:51:11.554 0076 /L2QS001A RWST CHEM ADD TK LO LEVEL LO No

• constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:11.559 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:11.678 L2RC002A Tripped 75.36%,

13:51:11.598 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes peak. Time 13:51:11.710 L2RC002A Reset 59.398%, valley. Pressurizer level signal cyclic 13:51:11.685 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes within time period.

13:51:11.732 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes 13:51:11.768 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes 13:51:11.778 0046 /F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:11.854 F2RC004A Tripped 88.620%,

w/ valley of 86.229% at 13:51:11.857.

Time 13:51:11.866 F2RC004A Reset 91.813%,

13:51:11.795 0046 /F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes w/ peak of 109.432% at 13:51:11.930.

13:51:11.823 0078 /L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:11.901 L2RC002A 55.980%, Reset 13:51:11.839 0005/M2NMC008A NIS PWR RGE HI FLUX RATE TRIP Yes Time 13:51:11.844 M2NMC008A= 92.404%

M2NMC007A RX TRIP Time 13:51:11.700 M2NMC007A= 92.127% 2 out of 4 Rx Trip RCE 001061, Rev. 1 Page 76 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:11.858 0076/L2QSO01A RWST CHEM ADD TK LO LEVEL NORMAL No

• Level constant at 86.249% till time 13:51:13.170 at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:11.863 0046 /F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:11.948 F2RC004A Tripped 89.409%,

w/valley of 79.397% at 13:51:11.960.

13:51:11.873 0052 /F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:11.860 - 11:910 F2RC009A data 13:51:11.887 0052 /F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes lost/clipped descending trend 96.170% w/ previous peak of 105.841% in to next peak of 109.111% at time 13:51:11.924.

13:51:11.887 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:11.970 F2RC004A Reset 93.140%,

w/peak of 107.847% at time 13:51:12.007.

13:51:11.887 0196 /X2RD035D MAIN RX TRIP BREAKER A TRIP Yes Time 13:51:12.002 X2RD035D Tripped.

13:51:11.891 0195 /X2RD036D MAIN RX TRIP BREAKER B TRIP Yes Time 13:51:12.002 X2RDO36D Tripped.

13:51:11.900 0120/X2RD001B REACTOR TRIPPED- TURB TRIP TRIP No

• Time 13:51:14.000 Tripped Y0005D (Man. Rx trip)

Y0005D/YO004D Time. 13:51:23.000 Tripped X2RD001B.

RCE 001061, Rev. 1 Page 77 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:11.951 0046/ F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.024 F2RC004A Tripped 87.651%,

w/ valley of 81.881 % at time 13:51:12.028.

13:51:11.961 0052/ F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:12.038 F2RC009A some data lost/clipped, Tripped, 92.330% descending to time 13:51:12.058.

13:51:11.969 0120 / MAIN TRANS LO RELAY TURB TRIP TRIP Yes Main Trans 86T Lockout relay actuated do to 2/3 (86T) SPRs/FPRs on Main Transformers, Data from Control Ops.

13:51:11.975 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:12.068 F2RC004A Reset 91.092%, w/

peak of 108.075%.

13:51:11.976 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:12.034 L2RC002A Tripped 73.691%.

13:51:11.987 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes Time 13:51:12.058 F2RC009A Reset 92.330%.

13:51:12.006 0027 /L2RC003A PRESSURIZER HI LEVEL RX TRIP TRIP Yes Time 13:51:12.106 L2RC001A at 93.762% and L2RC001 A L2RC003A 96.637%.

RCE 001061, Rev. 1 Page 78 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.015 0072 / X2SY002D 500KV BREAKER 21 OPEN Yes Time 13:51:13.950 Bkrs closed and data clipped to 13:51:12.016 0079 / X2SY001 D 500KV BREAKER 2C OPEN Yes 13:51:20.950 Bkrs open this time. STATMON time from Control Ops 13:51:41 Bkrs open. Time 13:51:12.019 0072 / X2SY002D 500KV BREAKER 21 CLOSED No

• differential due to STATMON polling. Contact bounce No

• cause of closed alarms, known occurrence verified 500KV BREAKER 2C CLOSED Nof Cowntol vps.

13:51:12.021 0079 / X2SY001D from Control Ops.

13:51:12.023 0027 /L2RC003A PRESSURIZER HI LEVEL RX TRIP NORMAL Yes Time 13:51:12.116 L2RC001A at 86.242% and L2RC001A L2RC003A at 84.995%.

13:51:12.024 0072 / X2SYoo2D 500KV BREAKER 21 OPEN No

• Contact bounce cause of bkr open/closed alarms, 13:51:12.028 0072 / X2SYOO2D 500KV BREAKER 21 CLOSED No

• known occurrence verified from Control Ops.

13:51:12.029 0079 / X2SY001D 500KV BREAKER 20 OPEN No

• 13:51:12.032 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:12.056 L2RC002A Reset 67.724%.

13:51:12.039 0072 / X2SY002D 500KV BREAKER 21 OPEN No

• Contact bounce cause of bkr open alarms, known occurrence from Control Ops.

13:51:12.046 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.128 F2RC004A Tripped 88.0449%

w/valley of 88.0449% at 13:51:12.130.

RCE 001061, Rev. 1 Page 79 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCs 13:51:12.048 0107/ RSS TRANS B SUD PRESS (63A) TRIP No* No electrical fault in transformer. SPRs actuated due to earthquake. Verified by Control Ops.

13:51:12.051 0052/ F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:12.158 F2RC009A Tripped 93.615%.

13:51:12.056 0046/ F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:12.139 F2RC004A Reset 92.322%

w/peak of 103.450% at 13:51:12.166.

13:51:12.057 0119 Y0390D SOLENOID TURB TRIP TRIP Yes Trip is 74/AST solenoid position indicator switch valid due to Turb. Trip (Y0390D) time 13:51:13.000.

13:51:12.057 0106 / RSS TRANS A SUD PRESS (63A) TRIP No

• No electrical fault in transformer. SPRs actuated due to earthquake. Verified by Control Ops.

13:51:12.057 0058 / Y0390D TURBINE TRIP RX TRIP TRIP Yes Time 13:51:13.000 Tripped Y0390D (Turbine trip) 13:51:12.061 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:12.139 L2RC002A Tripped 71.142%,

w/peak of 83.388% at 13:51:12.154.

13:51:12.078 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes Time 13:51:12.164 Reset 94.428% w/peak of 107.259% at 13:51:12.230.

13:51:12.108 0039/ L2FW006A STM GEN 1B HI-HI LEVEL TURB TRIP Yes Time 13:51:12.134 L2FW006A at 78.384% and 13:51:12.118 L2FW005A TRIP L2FW005A 80.499%

0039 / L2FW006A STM GEN 1B HI-HI LEVEL TURB NORMAL Yes Time 13:51:12.146 L2FW006A 93.925% and L2FW005A TRIP L2FW005A 70.491%.

RCE 001061, Rev. 1 Page 80 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCs 13:51:12.130 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:12.172 L2RC002A Reset 65.351%

w/valley of 45.566% at 13:51:12.184.

13:51:12.138 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:12.324 F2RC009A 91.795% from peak of 104.269% at 13:51:12.308, data clipped till 13:51:12.358.

13:51:12.149 0076 / L2QSO01A RWST CHEM ADD TK LO LEVEL LO No

• Level constant at 86.249% till time 13:51:13.170 at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:12.161 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes Time 13:51:12.358 F2RC009A Reset at 91.795%, to peak of 104.100% at 13:51:12.380.

13:51:12.191 0004/C0099D RPI ROD BOTTOM ROD DROP DROP No

• Time 13:51:12.002 X2RD035D Main Rx Trip bkr X2RD035D Tripped. Time 13:51:12.950 Tripped C0099D Rod Dropped.

13:51:12.195 0134 /Y2322D FEED WTR PP BKRS OPEN TRIP Yes Time 13:51:12.250 Y2322D/Y2323D (2-FW-P-1 B)

Y2323D TURB TRIP breakers Tripped 13:51:12.225 0013 / T0496A TAVG > < TREF DEV DEVIATION Yes Time 13:51:12.220 Median TAVG 580.'F and TREF T0499A 580.'F. Time 13:51:13.186 Median TAVG 580.3 OF and TREF = . 570.9 OF.

RCE 001061, Rev. 1 Page 81 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.234 0076 / L2QS001A RWST CHEM ADD TK LO LEVEL NORMAL No

• Level constant at 86.249% till time 13:51:13.170 at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:12.297 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.334 F2RC004A at 92.215% from a peak of 107.161% at 13:51:12.290.

13:51:12.302 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:12.319 L2RC002A Tripped at 69.562% w/

peak of 88.663% at 13:51:12.384.

13:51:12.313 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:12.344 F2RC004A reset 95.968% data clipped at peak of 107.226% at 13:51:12.356.

13:51:12.356 0027 / L2RC001A PRESSURIZER HI LEVEL RX TRIP TRIP Yes Time 13:51:12.360 L2RC001A at 100.814% and L2RC002A L2RC002A at 88.663% data clipped but signal is 13:51:12.368 0027 / L2RC002A PRESSURIZER HI LEVEL RX TRIP NORMAL Yes increasing. Time 13:51:12.372 L2RC001A at L2RC001A 89.599% and L2RC002A at 77.983%.

13:51:12.389 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:12.406 L2RC002A at 67.426% Reset w/valley of 45.945% at 13:51:12.434.

13:51:12.495 0003 /L2FW005A STM GEN 1B LEVEL ERROR ERROR Yes Time 13:51:12:509 L2FW005A at 56.523%,

L2FW006A L2FW006A at 60.239%, and L2FW007A at 67.860%

L2FW007A 13:51:12.557 0108/ RSS TRANS C SUD PRESS (63A) TRIP No

• No electrical fault in transformer. SPRs actuated due to earthquake. Verified by Control Ops.

13:51:12.589 0116/ STA SERV 2 OF 3 BUSES UV-RX TRIP Yes 13:51:13.300 SS/RCP UV V0320D, V0321D, V0322D TRIP Tripped.

RCE 001061, Rev. 1 Page 82 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.595 0041 / FIRE WATER SYSTEM INITIATED INIT No

• No PCS point 13:51:12.600 0078/ L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:12.696 L2RC002A Tripped 69.535%

w/peak of 93.301% at 13:51:12.706.

13:51:12.625 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.731 F2RC004A Tripped at 90.562%,

13:51:12.646 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes this is valley. Time 13:51:12.744 F2ROOO4A Reset at 94.332% w/ peak of 105.642% at 13:51:12.756.

13:51:12.673 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:12.746 L2RC002A Reset at 67.457%

w/valley of 60.591% at 13:51:12.756.

13:51:12.713 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.780 F2RC004A Tripped at 90.717%,

13:51:12.734 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes valley. Time 13:51:12.816 F2RC004A Reset 91.261% w/ peak of 105.309% at 13:51:12.860.

13:51:12.777 0076 / L2QS001A RWST CHEM ADD TK LO LEVEL LO No

• Level constant at 86.249% till time 13:51:13.170, Lo alarm setpt85%, at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170.

13:51:12.797 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:12.872 F2RC004A Tripped at 89.892%

13:51:12.822 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes w/valley of 87.322% at 13:51:12.880. Time 13:51:12.896 F2RC004A Reset at 91.388% w/ peak 13:51:12.882 0046/ F2RC004A RC LOOP 1B LO FLOW CH I LO Yes of 105.342% at 13:51:12.958. Time 13:51:12.970 F2RC004A Tripped at 89.921% w/ valley of 82.352%

at 13:51:12.974.

RCE 001061, Rev. 1 Page 83 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.885 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:12.970 L2RC002A Tripped at 71.235% w/

peak of 79.414% at 13:51:12.988.

13:51:12.910 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:13.000 F2RC004A Reset at 91.461%

w/peak of 102.0219% at 13:51:13.024.

13:51:12.932 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:13.016 L2RC002A Reset at 67.694% w/

valley of 47.654% at 13:51:13.040.

13:51:12.941 0041 / FIREWATER SYSTEM INITIATED NORMAL No

• No PCS point 13:51:12.969 0046/ F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:13.050 F2RC004A Tripped at 89.743%

w/valley of 84.832% at 13:51:13.060.

13:51:12.971 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:13.052 L2RC002A Tripped at 70.256% w/

peak of 81.557% at 13:51:13.058.

13:51:12.972 0129 / F2RC001A LOSS OF COOL FLOW PWR > 30% TRIP Yes Time 13:51:13.034 Tripped F2RC001A at 87.724%

F2RC002A and F2RC002A at 89.665% with M1 NMO07A at 83.679% and M1NMO08A at 90.187% Pwr.

M1NM007A/

MlNMO08A 13:51:12.972 0043 / F2RC002A RC LOOP 1A LO FLOW CH II LO Yes Time 13:51:13.068 F2RC002A Tripped 89.665% this is valley, from a peak of 95.132% at 13:51:13.040.

13:51:12.973 0140/Y0322D STA SERV 2 OF 3 BUSES UF-RX TRIP Yes Time 13:51:13.970 2A and 21 SS/RCP UF Tripped Y0321 D/Y0320D TRIP Y0320D and Y0321D RCE 001061, Rev. 1 Page 84 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.981 0129/F2RC001A LOSS OF COOL FLOW PWR > 30% NORMAL Yes Time 13:51:13.074 Reset F2RC001A at 89.610%

F2RC002A and F2RC002A at 91.355% with M1NMO07A at 83.678% and M1NMO08A at 90.187% Pwr.

P0398A/P0399A 13:51:12.983 0043 / F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes Time 13:51:13.074 F2RC002A Reset at 91.597% w/

peak of 96.425% at 13:51:13.084.

13:51:12.998 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:13.078 F2RC004A at 91.081% w/ peak of 98.409% at 13:51:13.124.

13:51:13.009 0076 / L2QS001A RWST CHEM ADD TK LO LEVEL NORMAL No

• Level constant at 86.249% till time 13:51:13.170 at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:13.024 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON NORMAL Yes Time 13:51:13.024 L2RC002A Reset at 66.893% w/

valley of 60.890% at 13:51:13.131.

13:51:13.056 0129//F2RC001A LOSS OF COOL FLOW PWR > 30% TRIP Yes Time 13:51:13.136 Tripped F2RC001A at 84.915%

F2RC002A and F2RC002A at 89.469% with M1NMO07A at 83.678% and M1NMO08A at 90.187% Pwr.

P0398A/P0399A 13:51:13.056 0043/ F2RC002A RC LOOP 1A LO FLOW CH II LO Yes Time 13:51:13.134 F2RC002A Tripped at 89.469%

signal clipped till 13:51:13.170.

13:51:13.058 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:13.146 F2RC004A Tripped at 89.501% w/

valley of 82.820% at 13:51:13.158.

RCE 001061, Rev. 1 Page 85 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:13.071 0129//F2RC001A LOSSOFCOOLFLOW PWR>30% NORMAL Yes Time 13:51:13.174 Reset F2RC001A at 88.010% and F2RC002A F2RC002A at 91.389% with M1 NMO07A at 83.678%

and M1NMO08A at 90.187% Pwr.

M1NM007A/

Ml NMO08A 13:51:13.071 0043 / F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes Time 13:51:13.174 F2RC002A Reset at 91.664% w/

peak of 97.151% at 13:51:13.190.

13:51:13.072 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:13.174 F2RC009A Tripped at 91.488, valley from peak of 107.964% at 13:51:13.124.

13:51:13.086 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:13.170 F2RC004A Reset at 92.590% w/

peak of 100.947% at 13:51:13.210.

13:51:13.098 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes Time 13:51:13.188 F2RC009A Reset at 96.634% w/

peak at 109.496% at 13:51:13.224.

13:51:13.132 0043 / F2RC002A RC LOOP 1A LO FLOW CH II LO Yes Time 13:51:13.274 F2RC002A Tripped at 90.017% at valley from peak of 93.012% at 13:51:13.228.

13:51:13.140 0129/I F2RC001A LOSS OF COOL FLOW PWR > 30% TRIP Yes Time 13:51:13.264 Tripped F2RC001A at 85.943%

F2RC002A and F2RC002A at 90.017% with M1NM007A at 83.678% and M1NMO08A at 90.187% Pwr.

P0398A/P0399A 13:51:13.153 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:13.270 F2RC004A Tripped at 89.477% w/

valley of 85.861% at 13:51:13.280.

13:51:13.159 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:13.274 F2RC009A Tripped at 92.063%

valley from peak of 109.496% at 13:51:13.210.

RCE 001061, Rev. 1 Page 86 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:13.162 0129/ F2RC001A LOSS OF COOL FLOW PWR >30% NORMAL Yes Time 13:51:13.274 Reset F2RC001A at 88.372% and F2RC002A F2RC002A at 91.150% with M1NMO07A at 83.678%

and M1NMO08A at 90.187% Pwr.

M1NM007A/

MlNMO08A 13:51:13.163 0043 / F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes Time 13:51:13.282 F2RC002A Reset at 91.959% w/

peak of 94.548% at 13:51:13.300.

13:51:13.181 0046/ F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:13.310 F2RC004A Reset at 91.604% w/

peak of 95.473% at 13:51:13.324.

13:51:13.190 0076 / L2QS001A RWST CHEM ADD TK LO LEVEL LO No

• Level constant at 86.249% till time 13:51:13.170 at which point level increases to 87.090% constant from 13:51:13.495-14.170. Level starts decreasing to 84.180% at time 13:51:15.170. Lo alarm setpt 85%.

13:51:13.190 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes Time 13:51:13.296 F2RC009A reset at 92.735% w/

peak of 104.538% at 13:51:13.330.

13:51:13.232 0129/F2RC001A LOSS OF COOL FLOW PWR > 30% TRIP Yes Time 13:51:13.346 Tripped F2RC001A at 89.858%

F2RC002A and F2RC002A at 91.211% with M1 NMO07A at 83.678% and M1NM008A at 90.187% Pwr.

M1NM007A/

M1 NMO08A 13:51:13.232 0043 / F2RC002A RC LOOP 1A LO FLOW CH II LO Yes Time 13:51:13.314 F2RC002A Tripped at 90.562%

data clipped to 13:51:13.350.

13:51:13.241 0046/ F2RC004A RC LOOP 1B LO FLOW CH I LO Yes Time 13:51:13.350 F2RC004A Tripped at 89.909% w/

valley of 87.684% at 13:51:13.358.

RCE 001061, Rev. 1 Page 87 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:13.242 0021 / EXCITATION LOSS OF POWER POWER Yes Time 13:51:11.969 Main Trans Lo Relay Turb (AL1-2, AL4-5) Trip(86T) Tripped. Time 13:51:12.950 X2TGO01 B, Gen Exciter Field Breaker, Tripped.

13:51:13.246 0129/ F2RC001A LOSS OF COOL FLOW PWR >30% NORMAL Yes Time 13:51:13.355 Reset F2RC001A at 89.466% and F2RCO02A F2RC002A at 91.643% with M1NMO07A at 83.678%

and M1NMO08A at 90.187%

Pwr.

Ml1NM007A/

M1 NM008A 13:51:13.247 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes Time 13:51:13.388 F2RC009A at 93.652% data clipped till 13:51:13.410 at 13:51:13.414 F2RC009A at 95.717%.

13:51:13.247 0043/ F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes Time 13:51:13.356 F2RC002A Reset at 91.303% w/

peak of 94.251% at 13:51:13.400.

13:51:13.257 0078 / L2RC002A PRZ HI LEVEL, BU HTRS ON ON Yes Time 13:51:13.296 L2RC002A Tripped at 70.093% w/

peak of 78.628% at 13:51:13.308.

13:51:13.273 0046/ F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes Time 13:51:13.374 F2RC004A Reset at 91.488% w/

peak of 94.325% at 13:51:13.424.

RCE 001061, Rev. 1 Page 88 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.276 0052/F2RCOO9A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:13:260 92.063%

F2RC009A, MAX t=13:51:13.280 93.238%

13:51:13.293 0058/YO390D TURBINE TRIP RX TRIP NOT TRIP Yes PCS Indicates TRIP on Y0390D on 1 second X2RDO35D scan data (13:51:12.818 to 13:51:13.906)

X2RDO36D Y0390D not available in TRA data.

RTB A/B TRIPPED t=13:51:12.014 13:51:13.310 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13:300 94.548%

F2RC002A, MIN t=13:51:13:330 90.562%

13:51:13.318 0078/L2RCOO1A PRZ HI LEVEL, BU HTRS ON NORMAL Yes L2RC001A, VAL t=13:51:13:318 74.668%

L2RC002A L2RC002A, VAL t=13:51:13.318 67.670%

L2RC003A L2RC003A, VAL t=13:51:13.318 72.579%

Several Changes over a 100 ms time interval over a range of 55-75 % Level 13:51:13.326 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A, MAX t=13:51:13:300 94.548%

F2RC002A, MIN t=13:51:13:330 90.562%

13:51:13.331 0046/F2RCOO4A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:330 95.473%

F2RC004A, MIN t=13:51:13.360 87.684%

13:51:13.343 0076/1L2QSOO1A RWST CHEM ADD TK LO LEVEL NORMAL Yes L2QSOO1A, MAX t=13:51:12.030 87.090%

L2QS001A, MIN t=13:51:14.030 84.188%

13:51:13.356 0078/L2RCOO1A PRZ HI LEVEL, BU HTRS ON ON Yes L2RCO1A, VAL t=13:51:13:360 64.578%

L2RC002A L2RC002A, VAL t=13:51:13.360 60.727%

L2RCOO3A L2RC003A, VAL t=13:51:13.360 72.471%

Several Changes over a 100 ms time interval over a range of 55-75 % Level 13:51:13.367 0046/F2RCOO4A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MIN t=13:51:13:352 89.206%

F2RC004A, MAX t=13:51:13.370 91.488%

RCE 001061, Rev. 1 Page 89 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.378 0078/ L2RC001A PRZ HI LEVEL, BU HTRS ON NORMAL Yes L2RC001A, VAL t=13:51:13:386 64.578%

L2RC002A, VAL t=13:51:13.386 60.727%

L2RCOO2A L2RC003A, VAL t=13:51:13.386 72.471%

Several Changes over a 100 ms time L2RC0O3A interval over a range of 55-75 % Level 13:51:13.397 0031/ RWST CHEM ADD TK LO TEMP LO No

• No PCS Point - No change in temperature.

13:51:13.402 0043/F2RC0O2A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.380 94.251%

F2RC002A, MIN t=13:51:13.410 89.783%

13:51:13.423 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:420 94.325%

F2RC004A, MIN t=13:51:13.440 87.000%

13:51:13.426 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A MIN t=13:51:13.422 90.273%

F2RC002A MAX t=13:51:13.430 90.601%

13:51:13.438 0076/ RWST CHEM ADD TK LO LEVEL LO No

• L2QSOO1A, MAX t=13:51:12.030 87.090%

L2QSOO1A L2QS001A, MIN t=13:51:14.030 84.188%

13:51:13.459 0046/F2RCOO4A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MIN t=13:51:13.440 87.000%

F2RC004A, MAX t=13:51:13:500 96.023%

13:51:13:467 0043/F2RC002A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.448 90.554%

F2RC002A, MIN t=13:51:13.490 89.978%

13:51:13.514 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:500 96.023%

F2RCO04A, MIN t=13:51:13.532 86.133%

13:51:13.525 0052/F2RCOO9A RC LOOP 1C LO FLOW CH III LO Yes F2RC009A, MAX t=13:51:13:500 103.25%

F2RC009A, MIN t=13:51:13.550 88.243%

13:51:13.548 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A, MIN t=13:51:13.540 89.626%

F2RC002A, MAX t=13:51:13.570 90.445%

13:51:13.551 0047/F2RC005A RC LOOP 1B LO FLOW CH II LO Yes F2RC005A, MIN t=13:51:13.530 90.64%

F2RC005A, MAX t=13:51:13.610 88.561%

RCE 001061, Rev. 1 Page 90 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes I No)

PCS 13:51:13.556 0078/L2RCOO1A PRZ HI LEVEL, BU HTRS ON ON Yes L2RC001A, VAL t=13:51:13:555 49.417%

L2RC002A L2RC002A, VAL t=13:51:13.555 49.851%

L2RC003A L2RC003A, VAL t=13:51:13.555 78.980%

Several Changes over a 100 ms time interval over a range of 55-75 % Level 13:51:13.558 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MIN t=13:51:13.520 85.287%

F2RC004A, MAX t=13:51:13:562 86.084%

13:51:13:559 0052/F2RCO09A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:13:500 88.243%

F2RC009A, MAX t=13:51:13.580 88.561%

13:51:13.560 0076/1L2QSOO1A RWST CHEM ADD TK LO LEVEL NORMAL Yes L2QS001A, MAX t=13:51:12.030 87.090%

L2QS001A, MIN t=13:51:14.030 84.188%

13:51:13.572 0048/F2RC006A RC LOOP 1B LO FLOW CH III LO Yes F2RC006A, MIN t=13:51:13:530 90.794%

F2RC006A, MAX t=13:51:13.580 90.445%

13:51:13.576 0044/F2RCO03A RC LOOP 1A LO FLOW CH III LO Yes F2RC003A, MIN t=13:51:13:530 90.988%

F2RC003A, MAX t=13:51:13.580 90.562%

13:51:13.577 0078/L2RCOo1A PRZ HI LEVEL, BU HTRS ON NORMAL Yes L2RC001A, VAL t=13:51:13:574 64.999%

L2RC002A L2RC002A, VAL t=13:51:13.574 51.071%

L2RC003A L2RC003A, VAL t=13:51:13.574 66.382%

Several Changes over a 100 ms time interval over a range of 55-75 % Level 13:51:13.601 0043/F2RCO02A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.570 90.445%

F2RC002A, MIN t=13:51:13.600 87.684%

13:51:13.605 0130/ LOSS OF COOL FLOW PWR > 10% TRIP Yes RC LOOP A CH II, and III LO RCE 001061, Rev. 1 Page 91 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.606 0078/L2RC001A PRZ HI LEVEL, BU HTRS ON ON Yes L2RC001A, VAL t=13:51:13:604 66.477%

L2RC002A L2RC002A, VAL t=13:51:13.604 64.022%

L2RC003A L2RC003A, VAL t=13:51:13.604 60.727%

Several Changes over a 100 ms time interval over a range of 55-75 % Level NOT AVAILABLE 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MIN t=13:51:13.600 85.204%

F2RC004A, MAX t=13:51:13:630 92.823%

13:51:13.616 0052/F2RCO09A RC LOOP IC LO FLOW CH III LO Yes F2RC009A, MIN t=13:51:13:610 88.561%

F2RC009A, MAX t=13:51:13.630 100.07%

13:51:13.637 0078/L2RCOO1A PRZ HI LEVEL, BU HTRS ON NORMAL Yes L2RC001A, VAL t=13:51:13:635 60.184%

L2RC002A L2RC002A, VAL t=13:51:13.635 63.873%

L2RC003A L2RC003A, VAL t=13:51:13.635 59.154%

T2RCO38A Flat Level over a 120 ms time interval T2RCO37A (0.620 - 0.740 sec)

T2RC040A @13:51:13.634 Tave T2RC039A A Loop Tave 580.390 T2RCO42A B Loop Tave 580.325 T2RCO41A C Loop Tave 579.105 RCE 001061, Rev. 1 Page 92 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.647 0052/F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:13:630 100.07%

F2RC009A, MAX t=13:51:13.734 100.07%

NOT AVAILABLE 0043/F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A, MAX t=13:51:13.632 91.642%

F2RC002A, MIN t=13:51:13.730 91.642%

13:51:13.653 0130/ LOSS OF COOL FLOW PWR > 10% NORMAL Yes RC LOOP A CH II NORMAL 13:51:13.654 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MAX t=13:51:13:630 92.823%

F2RC004A, MAX t=13:51:13:734 92.823%

13:51:13.685 0130/ LOSS OF COOL FLOW PWR > 10% TRIP Yes RC LOOP B CH I, and III LO 13:51:13.689 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.680 91.642%

F2RC0O2A, MIN t=13:51:13.760 87.604%

13:51:13.694 0046/F2RCOO4A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:630 92.823%

F2RC004A, MAX t=13:51:13:734 92.823%

Flat Spot in PCS Data 13:51:13.708 0052/F2RC009A RC LOOP IC LO FLOW CH III LO Yes F2RC009A, MIN t=13:51:13:698 100.07%

F2RC009A, MAX t=13:51:13.734 100.07%

Flat Spot in PCS Data 13:51:13.733 0078/L2RCOO1A PRZ HI LEVEL, BU HTRS ON ON Yes L2RCOO1A, VAL t=13:51:13:733 60.184%

L2RC002A L2RC002A, VAL t=13:51:13.733 63.873%

L2RC003A L2RC003A, VAL t=13:51:13.733 59.154%

T2RCO38A Flat Level over a 120 ms time interval T2RCO37A (0.620 - 0.740 sec)

T2RCO40A @13:51:13.733 Tave T2RCO39A A Loop Tave 580.390 T2RCO42A B Loop Tave 580.325 T2RCO41A C Loop Tave 579.105 RCE 001061, Rev. 1 Page 93 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.735 0043/F2RCOO2A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A, MAX t=13:51:13.730 91.642%

F2RC002A, MIN t=13:51:13.742 90.027%

13:51:13.736 0130/ LOSS OF COOL FLOW PWR > 10% NORMAL Yes RC LOOP B CH I NORMAL 13:51:13.738 0052/F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:13:734 100.07%

F2RC009A, MAX t=13:51:13.764 94.816%

13:51:13.745 0046/ F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MAX t=13:51:13:734 92.823%

F2RC004A, MIN t=13:51:13.758 89.256%

13:51:13.752 0130/1P0398A LOSS OF COOL FLOW PWR > 10% TRIP Yes RC LOOP A, B, C NORMAL P0399A P0398A TIP III 260.6 PSIG (>10% TIP)

P0399A TIP IV 391.3 PSIG(>10% TIP) 13:51:13.753 0043/F2RC002A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.742 90.027%

F2RC002A, MIN t=13:51:13.760 87.604%

13:51:13.755 0078/L2RC001A PRZ HI LEVEL, BU HTRS ON NORMAL Yes L2RC0O1A, VAL t=13:51:13:750 60.184%

L2RCO02A L2RC002A, VAL t=13:51:13.750 63.873%

L2RCO03A L2RC003A, VAL t=13:51:13.750 59.154%

Several Changes over a 100 ms time interval over a range of 55-75 % Level RCE 001061, Rev. 1 Page 94 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes / No)

PCS 13:51:13.790 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:740 92.823%

F2RC004A, MIN t=13:51:13.790 84.208%

13:51:13.815 0043/F2RC002A RC LOOP 1A LO FLOW CH II NORMAL Yes F2RC002A, MIN t=13:51:13.798 88.680%

F2RC002A, MAX t=13:51:13.840 89.075%

13:51:13.817 0130/ LOSS OF COOL FLOW PWR > 10% NORMAL Yes RC LOOP A, B, C NORMAL 13:51:13.837 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MAX t=13:51:13:820 92.709%

F2RC004A, MIN t=13:51:13.870 83.832%

13:51:13.847 0043/F2RC002A RC LOOP 1A LO FLOW CH II LO Yes F2RC002A, MAX t=13:51:13.840 89.075%

F2RC002A, MIN t=13:51:13.870 84.583%

13:51:13.848 0130/ LOSS OF COOL FLOW PWR > 10% TRIP Yes RC LOOP A CH II, Ill LO 13:51:13.873 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13:870 83.832%

F2RC004A, MIN t=13:51:13.900 83.832%

13:51:13.880 0052/F2RC009A RC LOOP IC LO FLOW CH III LO Yes F2RC009A, MAX t=13:51:13:820 96.970%

F2RC009A, MIN t=13:51:13.896 91.986%

13:51:13.914 0052/F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:13:896 91.986%

F2RC009A, MAX t=13:51:13.930 97.404%

13:51:13.917 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MIN t=13:51:13:900 83.832%

F2RC004A, MAX t=13:51:13.930 92.253%

13:51:13.961 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MAX t=13:51:13.930 92.253%

F2RC004A, MIN t=13:51:13.962 85.170%

13:51:13.962 0025/YO004D MANUAL INITIATION RX TRIP TRIP Yes Manual Operator Action - Y005D (SW Y0005D 2-1)YO05D t=13:51:14.030 13:51:13.965 0052/F2RC009A RC LOOP 1C LO FLOW CH Ill LO Yes F2RC009A, MIN t=13:51:13:930 97.404%

1_ 1_ 1 1 1F2RC009A, MAX t=13:51:13.980 89.351%

RCE 001061, Rev. 1 Page 95 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes [ No)

PCS 13:51:13.976 0078/L2RC001A PRZ HI LEVEL, BU HTRS ON ON Yes L2RC001A, VAL t=13:51:13:604 66.477%

L2RC002A L2RC002A, VAL t=13:51:13.604 64.022%

L2RC003A L2RC003A, VAL t=13:51:13.604 60.727%

Several Changes over a 100 ms time interval over a range of 55-75 % Level 13:51:14.006 0052/F2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes F2RC009A, MAX t=13:51:13:988 89.351%

F2RC009A, MIN t=13:51:14.030 100.25%

13:51:14.007 0078/L2RC001A PRZ HI LEVEL, BU HEATERS ON NORMAL Yes L2RC001A, VAL t=13:51:14:004 62.734%

L2RC002A L2RC002A, VAL t=13:51:14.004 65.758%

L2RC003A L2RC003A, VAL t=13:51:14.004 61.486%

Several Changes over a 100 ms time interval over a range of 55-75 % Level RCE 001061, Rev. 1 Page 96 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz/PCS) (Status) (Yes/ No)

PCS 13:51:14.009 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MIN t=13:51:13.980 85.467%

F2RC004A, MAX t=13:51:14.010 93.765%

13:51:14.049 0046/F2RC005A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MIN t=13:51:14.040 82.138%

F2RC004A, MAX t=13:51:14.072 88.793%

13:51:14.052 0052/F2RC009A RC LOOP 1C LO FLOW CH III LO Yes F2RC009A, MAX t=13:51:14.010 100.25%

F2RC009A, MIN t=13:51:14.052 88.608%

13:51:14.086 0052/F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes F2RC009A, MIN t=13:51:14.052 88.608%

F2RC009A, MAX t=13:51:14.090 97.368%

13:51:14.101 0046/F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes F2RC004A, MAX t=13:51:14.090 94.734%

F2RC004A, MIN t=13:51:14.120 81.278%

13:51:14.137 0046/F2RC004A RC LOOP 1B LO FLOW CH I LO Yes F2RC004A, MIN t=13:51:14.120 81.278%

F2RC004A, MAX t=13:51:14.170 92.974%

13:51:14.164 0041/ FIRE WATER SYSTEM INITIATED INIT No

• No PCS Point RCE 001061, Rev. 1 Page 97 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:14.186 0052 / F2RC009A RC LOOP IC LO FLOW CH III LO Yes Loop C, Ch 3, t = 13:51:14.188, Peak = 92.582 %, t =

13:51:14.216 0052 / F2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes 13:51:14.276, Valley = 89.448 %. Loop 13:51:14.248 0051 / F2RC008A RC LOOP IC LO FLOW CH II LO Yes C, Ch 2, t = 13:51:14.282, Valley = 90.002%.

13:51:14.251 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes 13:51:14.252 078 / L2RC001A PRZ HI LEVEL, BU HTRS ON ON Yes Ch 3, t = 13:51:14.100, Peak = 72.36 %,t = 13:51:14.247, Valley

= 53.625 %.

13:51:14.274 051 / F2RC008A RC LOOP 1C LO FLOW CH II NORMAL Yes Loop C, Ch 2, t = 13:51:14.250, Peak = 93.69 %, t =

13:51:14.276, Valley = 92.839 %.

13:51:14.275 078 / L2RC001A, PRZ HI LEVEL, BU HTRS ON NORMAL Yes @ t = 13:51:14.276, all three channels between 59 % to 60 %

L2RC002A, Level.

L2RC003A 13:51:14.300 0052 / F2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes Loop C, Ch 3, t = 13:51:14.350, Peak = 91.604 %, t 13:51:14.332 0051 / F2RC008A RC LOOP 1C LO FLOW CH II LO Yes 13:51:14.368, Valley = 84.542 %. Loop C, Ch 2, t =

13:51:14.335 0052 / F2RC009A RC LOOP IC LO FLOW CH III LO Yes 13:51:14.318, Peak = 94.66 %, t = 13:51:14.364, Valley =

,13:51:14.362 0051 / F2RC008A RC LOOP 1C LO FLOW CH II NORMAL Yes 89.939%. LoopB ,Ch 1, t= 13:51:14.386, Peak= 90.134 %, t 13:51:14.380 0046 / F2RC004A RC LOOP 1B LO FLOW CH I NORMAL Yes = 13:51:14.342, Valley = 84.276 %.

13:51:14.381 041 / N/A FIRE WATER SYSTEM NORMAL No

• No TRA or PCS Computer Points for this function.

INITIATED I_ _

RCE 001061, Rev. 1 Page 98 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:14.384 0052 / F2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes Loop C, Ch 3, t = 13:51:14.418, Peak = 91.795 %, t =

13:51:14.389 0046 / F2RC004A RC LOOP 1B LO FLOW CH I LO Yes 13:51:14.388, Valley = 84.542 %. Loop C, Ch 2, t =

13:51:14.402 0050 / F2RC007A RC LOOP IC LO FLOW CH I LO Yes 13:51:14.418, Peak = 93.276 %, t = 13:51:14.470, Valley =

13:51:14.420 0051 /F2RC008A RC LOOP IC LO FLOW CH II LO Yes 89.351 %. Loop B, Ch 1, t = 13:51:14.418, Peak = 91.795 %,

13:51:14.427 0052 / F2RC009A RC LOOP IC LO FLOW CH III LO Yes t = 13:51:14.388, Valley = 84.542 %.

13:51:14.461 0051 / F2RC008A RC LOOP IC LO FLOW CH II NORMAL Yes 13:51:14.488 0052 / F2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes 13:51:14.512 0051 / F2RC008A RC LOOP IC LO FLOW CH II LO Yes 13:51:14.518 025 / X2RDOO1B MANUAL INITIATION RX TRIP NORMAL Yes @ t = 13:51:14.544, X2RDOO1B indicates NOT TRIPPED @ t Y0005D = 13:51:14.372 on Y0005D indicated Reset 13:51:14.522 0052 / F2RC009A RC LOOP IC LO FLOW CH III LO Yes Loop C, Ch 3, t = 13:51:14.538, Peak = 91.782 %, t =

13:51:14.557 0051 / F2RC008A RC LOOP IC LO FLOW CH II NORMAL Yes 13:51:14.654, Valley = 84.204 %. Loop C, Ch 2, t =

13:51:14.579 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes 13:51:14.594, Peak = 91.393 %, t = 13:51:14.638, Valley =

13:51:14.596 0051 / F2RC008A RC LOOP 1C LO FLOW CH II LO Yes 87.613%.

13:51:14.653 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes 13:51:14.663 0052 / F2RC009A RC LOOP 1C LO FLOW CH III NORMAL Yes 13:51:14.679 0051 / F2RC008A RC LOOP IC LO FLOW CH 1I LO Yes 13:51:14.690 0052 / F2RC009A RC LOOP 1C LO FLOW CH III LO Yes 13:51:14.711 0003 / L2FW005A, STM GEN 1B LEVEL ERROR NORMAL Yes @ t = 13:51:14.711, all three channels between 32 % and 34 L2FW006A,  % Level.

L2FW007A RCE 001061, Rev. 1 Page 99 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:14.738 052 / L2RC009A RC LOOP IC LO FLOW CH III NORMAL Yes Loop C, Ch 3, t = 13:51:14.738, Peak = 90.017 %, t =

13:51:14.775 052 / L2RC009A RC LOOP IC LO FLOW CH III LO 13:51:14.776, Valley= 88.219 %.

13:51:14.925 078 / L2RC003A PRZ HI LEVEL, BU HITRS ON ON Yes Ch 2, L2RC003A @ t = 13:51:14.970 = 71.549 % Level 13:51:14.946 078 / L2RC003A PRZ HI LEVEL, BU HTRS ON NORMAL Yes and @ t = 13:51:14.947 = 60.483 % Level.

13:51:15.309 041 / N/A FIRE WATER SYSTEM INITIATED INIT No

• No TRA or PCS Computer Points for this function.

13:52:15.530 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No

• RCE 001061, Rev. 1 Page 100 of 130

Attachment 10 North Anna Power Station Unit 2 13:51:15:775 0062/ L2FW001A STM GEN 1A LO-LO LEVEL CH I-II- LO-LO Yes SG A, Ch 1, t = 13:51:15.790, Peak = 19.540 % Level , t 13:51:15.805 0062 / L2FW002A III STM GEN 1A LO- NORMAL Yes = 13:51:16.140, Valley = 18.525 %Level. SG A, Ch 2, t 13:51:15.775 0062 / L2FW003A LO LEVEL CH I-II-Ill LO-LO Yes = 13:51:15.780, Peak = 21.249 % Level , t =

13:51:15.870 0062 / L2FW001A STM GEN IA LO-LO LEVEL CH 1-II- NORMAL Yes 13:51:16.140, Valley = -17.358 % Level. SG A, Ch 3, t =

13:51:15.905 0071 / L2FW005A III STM GEN IA LO-LO Yes 13:51:15.790, Peak = 21.701 % Level , t = 13:51:16.140, 13:51:15.928 0071 / L2FW006A LO-LO LEVEL CH I-II-III NORMAL Yes Valley = 17.819 % Level. SG B, Ch 1, t = 13:51:15.780, 13:51:15.955 0062 / L2FW001A STM GEN 1B LO-LO LEVEL CH 1 LO-LO Yes Peak = 22.701 % Level , t = 13:51:16.140, Valley =

13:51:16.004 0028 / III STM GEN lB TRIP Yes 20.152 %Level. SG B, Ch 2, t = 13:51:15.780, Peak=

13:51:16.012 0071 / L2FW007A LO-LO LEVEL CH I-II-III LO-LO Yes 22.955 % Level , t = 13:51:16.140, Valley = 17.772 %

13:51:16.019 0073/ L2FW009A STM GEN IA LO-LO LEVEL CH 1 LO-LO Yes Level. SG B, Ch 3, t = 13:51:15.780, Peak = 20.821 %

13:51:16.039 0073 / L2FW010A III STM GEN 1A LO-LO NORMAL Yes Level , t = 13:51:16.140, Valley = 17.413 % Level. SG 13:51:16.048 0071 / L2FW005A LEVEL RX TRIP NORMAL Yes C, Ch 1, t = 13:51:15.780, Peak =23.208 % Level, t =

13:51:16.100 0071 / L2FW006A STM GEN lB LO-LO LEVEL CH I LO-LO Yes 13:51:16.140, Valley = 19.447 % Level. SG C, Ch 2, t =

'13:51:16.107 0073 / L2FW009A III STM GEN IC LO-LO Yes 13:51:15.780, Peak = 21.336 % Level , t = 13:51:16.140, 13:51:16.121 0029 / L2FW007A LO-LO LEVEL CH I-II-II TRIP Yes Valley = 17.711 % Level. SG C, Ch 3, t = 13:51:15.780, 13:51:16.122 0124/ 0029 STM GEN IC LO-LO LEVEL CH I-II- TRIP Yes Peak = 24.763 % Level , t = 13:51:16.140, Valley =

13:51:16.142 III STM GEN lB LO-LO NORMAL Yes 17.684 % Level.

LEVEL CH I-II-III STM GEN 1B LO-LO LEVEL CH I-II-III STM GEN IC LO-LO LEVEL CH I-II-III STM GEN 1B LO-LO LEVEL CH 1 III STM GEN IC LO-LO LEVEL RX TRIP STM GEN 1B LO-LO LEVEL RX TRIP RCE 001061, Rev. 1 Page 101 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:16.150 033 / T0404A, LOOP 1A LO DELTA T DEVIATION LO Yes @ t = 13:51:16.150, Median Delta T = 99.45 % and Loop T0497A A Delta T = 99.0 %. @ t = 13:51:17.309, Median Delta T

= 95.38 % and Loop A Delta T = 91.7 %.

13:51:16.159 0071 / L2FW005A STM GEN lB LO-LO LEVEL CH I-II- NORMAL Yes SG B, Ch 1, t = 13:51:16.460, Peak = 20.206 % Level , t =

13:51:16.164 0124/ 0124 III STM GEN 1C LO- NORMAL Yes 13:51:16.922, Valley = 14.981% Level. SG B, Ch 2, t =

13:51:16.191 / 0071 / LO LEVEL RX TRIP TRIP Yes 13:51:16.460, Peak = 17.846 % Level , t = 13:51:16.922, 13:51:16.192 L2FW006A STM GEN 1C LO-LO LEVEL RX TRIP LO-LO Yes Valley = 13.615 % Level. SG B, Ch 3, t = 13:51:16.460, 13:51:16.250 0014/ 0029 STM GEN lB LO-LO LEVEL CH I-II- ERROR Yes Peak = 24.030 % Level , t = 13:51:16.922, Valley = 19.760 13:51:16.275 / 0029/ III STM GEN 1C TRIP Yes  % Level. SG C, Ch 1, t = 13:51:16.370, Peak = 24.247 %

13:51:16.319 0029 / LEVEL ERROR NORMAL Yes Level, t = 13:51:16.885, Valley = 19.308 % Level. SG C, 13:51:16.370 0029/ STM GEN lB LO-LO LEVEL RX TRIP TRIP Yes Ch 2, t = 13:51:16.370, Peak = 16.816 % Level , t =

13:51:16.399 0029/ STM GEN 1B LO-LO LEVEL RX TRIP NORMAL Yes 13:51:16.885, Valley = 12.618 % Level. SG C, Ch 3, t =

13:51:16.454 0029 / STM GEN 1B LO-LO LEVEL RX TRIP ERROR Yes 13:51:16.370, Peak = 18.335 % Level , t = 13:51:16.885, 13:51:16.479 0002 / STM GEN lB LO-LO LEVEL RX TRIP NORMAL Yes Valley = 12.298 % Level.

13:51:16.498 0029 / 0029 STM GEN 1B LO-LO LEVEL RX TRIP ERROR Yes 13:51:16.507 0029/ STM GEN IB LO-LO LEVEL RX TRIP TRIP Yes 13:51:16.548 0003 / STM GEN lA LEVEL ERROR ERROR Yes 13:51:16588 STM GEN lB LO-LO LEVEL RX TRIP TRIP Yes 13:51:16.936 STM GEN 1B LO-LO LEVEL RX TRIP ERROR Yes STM GEN 1B LO-LO LEVEL RX TRIP STM GEN 1B LEVEL ERROR 13:51:18.510 0001/ *ROG ALARM No

• RCE 001061, Rev. 1 Page 102 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:19.439 0036/T0444A LOOP 1C[IIDELTAT HI No* @ t= 13:51:19.410, Median Delta T =85.1% andLoop T0497A DEVIATION C Delta T =83.6%.

13:51:20.543 0031 / N/A RWST CHEM ADD TK LO TEMP NORMAL No

• No TRA or PCS Computer Points for this function.

13:51:22.739 0001 / *ROG NORMAL No

• 13:51:23.303 0041 / N/A FIRE WATER SYSTEM INIT No

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:24.492 0132/X2RD0O1B MANUAL INITIATION TURB TRIP Yes @ t = 13:51:23.120, X2RDOO1B (RX) and X2TGOO1B X2TG001B TRIP (Turb) changed state, i.e., Reset, Trip, Reset.

13:51:25.093 0001 /N/A *ROG ALARM No

• 13:51:25.780 0132 / X2TGO01B MANUAL INITIATION TURB NORMAL Yes @ t = 13:51:25.840, X2TG001B indicated NOT TRIPPED.

TRIP 13:51:27.436 0041 / N/A FIRE WATER SYSTEM NORMAL No

• No TRA or PCS Computer Points for this function.

INITIATED 13:51:30.886 0033 / T0404A LOOP IA LO DELTA T NORMAL Yes @ t = 13:51:30.889, Median Delta T = 27.04 % and Loop T0497A DEVIATION A Delta T = 24.28 %. @ t = 13:51:31.279, Median Delta T

= 25.5 %, Loop A Delta T = 23.23 %.

RCE 001061, Rev. 1 Page 103 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:51:30.887 0013 / T0496A T0499A TAVG > < TREF DEV NORMAL Yes @ t = 13:51:30.880, Med TAVG = 561.4 °F and TREF=

556.9 OF.

13:51:43.411 0188 / X2RX004D AMSAC INITIATED ALARM Yes @ t = 13:51:46.100, AMSAC, Trains A & B indicated X2RX005D TRIPPED (off by + 3.5 Seconds) 13:51:45.407 0049 / Y2430D RCP 1B CH 2 BKR OPEN Yes @ t = 13:51:46.315, Y2430D indicated OPEN 13:51:49.447 0053 / Y2431D RCP IC CH 3 BKR OPEN Yes @ t = 13:51:50.405, Y2431D indicated OPEN 13:51:49.521 0045 / Y2801D RCP 1A CH 3 BKR OPEN Yes @ t = 13:51:50.405, Y2801D indicated OPEN 13:52:16.290 0013 / T0499A, TAVG > < TREF DEV DEVIATIO Yes @ t = 13:51:16.276, TREF = 554.8 OF, Median TAVG =

T0496A N 559.48 OF.

13:56:00.488 0130 / F2RC001A LOSS OF COOL FLOW PWR > 10% NORMAL Yes @ t = 13:56:01.040, all three RCS Loop Flows @

F2RC004A, 0.0 %, and Turb First Stage Pressure Ch 3 = 51.46 F2RC007A P0398A / PSI and Ch 4 = 52.8 PSI.

P0399A 13:56:00.495 0116 / V0320D, STA SERV 2 OF 3 BUSES UV-RX NORMAL No * @ t = 13:51:09.537, all three buses indicated NOT V0321D, V0322D TRIP TRIPPED, @ t = 13:51:19.082, 3 of 3 buses indicated TRIPPED. @ t = 13:56:00.497, 3 of 3 buses indicated TRIPPED.

RCE 001061, Rev. 1 Page 104 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Notes Normalized (Dranetz / PCS) (Status) (Yes / No) to PCS 13:56:00.497 0116 / Y0320D, STA SERV 2 OF 3 BUSES UF-RX NORMAL No

• t = 13:51:55.964, all three buses indicated Y0321D, Y0322D TRIP TRIPPED and Y0324D indicated TRIPPED, @ t =

Y0324D 13:56:05.416, 3 of 3 buses indicated TRIPPED and Y0324D indicates NOT TRIPPED.

13:58:13.147 0139 / P2500A, LOW VACUUM TURB TRIP TRIP Yes @ t = 13:58:13.066, P2500A = 9.2325 INHGA, P2510A P2510A = 9.9262 INHGA, the Trip Setpoint is 4.0 INHGA 14:04:54.505 078 / L2RCOO1A PRZ HI LEVEL, BU HTRS ON ON Yes @ t = 14:04:55.890, TAVG between 555.89 'F and 556.26 14:10:27.545 078 / L2RC002A 078 PRZ HI LEVEL, BU HTRS ON NORMAL Yes 'F. @ t = 14:04:53.710, PZR Level Ch 1 = 43.748, %,

/ L2RC003A Ch 2 = 44.75 %, Ch 3 = 44.535 %. @ t = 14:10:29.970, TAVG between 560.26 °F and 560.66 'F. @ t =

14:10:29.300, PZR Level Ch 1 = 43.748, %, Ch 2 =

44.75 %, Ch 3 = 44.534 %.

14:15:40.042 028 / N/A STM GEN 1A LO-LO LEVEL RX NORMAL Yes SG A @ t = 14:15:40.118, Channels between 14:16:18.714 062 / L2FW005A, TRIP STM GEN 1B LO-LO LEVEL NORMAL No

• 17.819 % to 18.633 % (Reset occurs @ t =

L2FW006A, CH I-II-III 14:15:48.980). SG B @ t = 14:16:18.012, Channels L2FW007A were at = 15.5 % Level (did not reach 18 % Level until 14:19:26.909).

• For Unit 2 Outliers, See Attachment 4.

RCE 001061, Rev. 1 Page 105 of 130

Attachment 10 North Anna Power Station Unit 1 North Anna Power Station, Unit 1 Dranetz event Outliers For August 23, 2011 Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.695 004/ COOXXA RPI ROD BOTTOM ROD DROP DROP No Yes* Electrical Power lost on 24Vdc Lamp feeding input to Dranetz point. TRA Points COOXXA show that the Rods started to drop when the RTB opened and the MCC 1 H1-1 Voltage Source for RPI system dropped out causing voltage to 24 Vdc Dranetz point to drop out at t=13:51:12.695. The TRA points COOXXA (Rod Position Points) show a sharp increase at t=13:51:12.720 when power was restored from the backup supply. Valid DROP indication per alternate evaluation.

RCE 001061, Rev. 1 Page 106 of 130

Attachment 10 North Anna Power Station Unit 2 13:51:12.700 031 / N/A RWST CHEM ADD TK LO TEMP LO No The Trip for LO temperature is 25 degF, Rest is 13:51:20.141 031 / N/A RWST CHEM ADD TK LO TEMP NORMAL No 26.5 degF. This may have been caused by a connection issue with the field or cabinet wiring providing an intermittent partial short across the RTD causing a low temperature input to the NRA card TM-QS201 (CF-121). I&C checks of wiring and calibration recommended.

13:51:19.452 036/T0444A LOOP IC HI DELTA T DEVIATION HI No Actual Delta T Deviation never exceeded the T0497A setpoint, so the reason for this actuation is UNKNOWN. Recalibration did not show any anomalous data. SAME EVENT ON OTHER UNIT Additional I&C checks recommended 13:51:11.916 041 / N/A FIRE WATER SYSTEM INITIATED INIT No Yes* Fire system pressure changes during seismic 13:51:11.975 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No event may have caused this indication. This is 13:51:12.213 041 / N/A FIRE WATER SYSTEM INITIATED INIT NG Yes* indicated by initial fast Init/Normal toggling time of 13:51:15.530 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No < 400ms due to earthquake. No plant log entries 13:51:17.013 041 / N/A FIRE WATER SYSTEM INITIATED INIT No Yes* found for Fire System Initiated on 08/23/2011.

13:51:33.151 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No Although not indicated in plant logs, RCE team post trip discussion with Operations personnel indicated that the fire water system actuated.

Additional I&C checks recommended

• Yes per Alternate Evaluation RCE 001061, Rev. 1 Page 107 of 130

Attachment 10 North Anna Power Station Unit 2 Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.133 076 / N/A RWST CHEM ADD TK LO LEVEL LO Ne Yes* Unit 2 indicated that level changed therefore it is 13:51:12.886 076 / N/A RWST CHEM ADD TK LO LEVEL LO Ne Yes* reasonable to expect that Unit 1 had level changes although they were not indicated on the PCS 1 PCS checks second scan point. Trip=85.0%, Reset=86.0%

recommended Level = 87.958%. Need to evaluate why PCS did not show indication as Unit 2 did.

I&C checks on PCS recommended.

13:56:10.211 116 / V0320D STA SERV 2 of 3 BUSES UV -RX NORMAL Ne Yes* Unit 2 Data shows valid response.

V0321D TRIP Unit 1 UNKNOWN - No Data Available on PCS.

V0322D Possible PCS Chatter program action on points V0320D, V0321 D, and V0322D caused data to be eliminated. I&C checks on PCS recommended.

13:51:11.888 0082 / N/A MAIN TRANS SUD PRESS RELAY TRIP Ne Yes* No electrical fault in transformer. SPRs actuated 13:51:11.971 0082 / N/A (63X ABC) NORMAL Ne Yes* due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication.

13:51:11.994 0096 / N/A SS TRANS 1C SUD PRESS RELAY TRIP Ne Yes* No electrical fault in transformer. SPRs actuated 13:51:12.015 0096 / N/A (63X-1) NORMAL Ne Yes* due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

13:51:12.047 0107/ N/A RSS TRANS B SUD PRESS RELAY TRIP. Ne Yes* No electrical fault in transformer. SPRs actuated (63A) due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

13:51:12.058 0106 / N/A RSS TRANS A SUD PRESS RELAY TRIP Ne Yes* No electrical fault in transformer. SPRs actuated (63A) due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

13:51:12.558 0108 / N/A RSS TRANS C SUD PRESS RELAY TRIP Ne Yes* No electrical fault in transformer. SPRs actuated (63A) due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation RCE 001061, Rev. 1 Page 108 of 130

Attachment 10 North Anna Power Station Unit 2

• Yes per Alternate Evaluation Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:18.508 0001 / N/A SMOKE DETECTION SYSTEM TROUBLE No Yes* The Fire Water System was affected by the 13:51:22.740 0001 / N/A SMOKE DETECTION SYSTEM NORMAL No earthquake; it is not unreasonable to expect that 13:51:25.092 0001 / N/A SMOKE DETECTION SYSTEM TROUBLE No Yes* other portions of the Fire Protection System were affected also. I&C checks recommended.

13:51:25.349 0132 / N/A MANUAL INITIATION TURB TRIP TRIP NG Yes* Operations initiated manual turbine trip as per 13:51:25.884 0132 / N/A MANUAL INITIATION TURB TRIP NORMAL No Yes* OP-E.0 Reactor Trip Response Procedure. Valid TRIP/NORMAL indication per alternate evaluation.

• Yes per Alternate Evaluation RCE 001061, Rev. 1 Page 109 of 130

ATTACHMENT 10 North Anna Power Station Unit 2 North Anna Power Station, Unit 2 Dranetz event Outliers For August 23, 2011 Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCS 13:51:12.019 0072/ X2SY002D 500KV BREAKER 21 CLOSED No Yes* Time 13:51:13.950 Bkrs closed and data clipped to 13:51:12.021 0079 / X2SY001D 500KV BREAKER 2C CLOSED No Yes* 13:51:20.950 Bkrs open this time. STATMON time from Control Ops 13:51:41 Bkrs open. Time differential due to STATMON polling. Contact bounce cause of closed alarms, known occurrence from Control Ops. NOT AN ISSUE 13:51:12.024 0072 / X2SY002D 500KV BREAKER 21 OPEN Ne Yes* Contact bounce cause of bkr open/closed alarms, 13:51:12.028 0072 / X2SY002D 500KV BREAKER 21 CLOSED No Yes* known occurrence verified by Control Ops. NOT 13:51:12.029 0079 / X2SYOO1D 500KV BREAKER 2C OPEN NG Yes* AN ISSUE 13:51:12.039 0072 / X2SY002D 500KV BREAKER 21 OPEN No Yes* Contact bounce cause of bkr open/closed alarms, known occurrence verified from Control Ops. NOT AN ISSUE RCE 001061, Rev. 1 Page 110 of 130

Attachment 10 North Anna Power Station Unit 2 13:51:12.048 0107 / RSS TRANS B SUD PRESS (63A) TRIP No Yes* No electrical fault in transformer. SPRs actuated due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

13:51:12.057 0106/ RSS TRANS A SUD PRESS (63A) TRIP No Yes* No electrical fault in transformer. SPRs actuated due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

13:51:12.557 0108/ RSS TRANS C SUD PRESS (63A) TRIP No Yes* No electrical fault in transformer. SPRs actuated due to earthquake induced oil pressure variations.

Verified by Control Operations. Valid TRIP indication per alternate evaluation.

• Yes per Alternate Evaluation Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz / PCS) (Status) (Yes / No)

PCs 13:51:11.900 0120 / Y0390D REACTOR TRIPPED - TURB TRIP TRIP Yes PCS 1 second data showed that REACTOR TRIPPED - TURB TRIP happened within the previous second as shown on the Y0390D point.

Valid TRIP indication per alternate evaluation 13:51:18.510 001 / N/A *ROG ALARM No Undefined Point - May be loose wiring at input to 13:51:22.739 001 / N/A *ROG ALARM No Dranetz causing intermittent inputs during seismic 13:51:25.093 001/N/A *ROG ALARM No event - I&C check recommended.

RCE 001061, Rev. 1 Page 111 of 130

Attachment 10 North Anna Power Station Unit 2 13:51:12.191 004 / COOXXA RPI ROD BOTTOM ROD DROP DROP No Yes* Electrical Power lost on 24Vdc Lamp feeding input to Dranetz point. TRA Points COOXXA show that the Rods started to drop when the RTB opened and the MCC 2H1-1 Voltage Source for RPI system dropped out causing voltage to 24 Vdc Dranetz point to drop out at t=13:51:12.191. The TRA points COOXXA (Rod Position Points) show a sharp increase at t=13:51:12.220 when power was restored from the backup supply. Valid DROP indication per alternate evaluation.

13:51:13.397 031 / N/A RWST CHEM ADD TK LO TEMP LO No The Trip for LO temperature is 25 degF, Rest is 13:51:20.543 031 / N/A RWST CHEM ADD TK LO TEMP NORMAL No 26.5 degF. This may have been caused by a connection issue with the field or cabinet wiring providing an intermittent partial short across the RTD causing a low temperature input to the NRA card TM-QS101 (CF-121).

I&C check recommended 13:51:19:439 0036/T0444A LOOP 1C HI DELTA T DEVIATION HI No Actual Delta T Deviation never exceeded the T0497A setpoint, so the reason for this actuation is UNKNOWN. Recalibration did not show any anomalous data. SAME EVENT ON OTHER UNIT Additional I&C checks recommended

• Yes per Alternate Evaluation Time Point ID Point Description Message Valid Trip Alternate Evaluation Normalized to (Dranetz I PCS) (Status) (Yes / No)

PCS 13:51:12.595 041 / N/A FIRE WATER SYSTEM INITIATED INIT No Yes* Fire system pressure changes during seismic 13:51:12.941 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No event may have caused this indication. This is 13:51:14.164 041 / N/A FIRE WATER SYSTEM INITIATED INIT No Yes* indicated by initial fast Init/Normal toggling time of 13:51:14.381 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No < 400ms. No plant log entries found for Fire 13:51:15.309 041 / N/A FIRE WATER SYSTEM INITIATED INIT NG Yes* System Initiated on 08/23/2011. Although not 13:51:15.530 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No indicated in plant logs, RCE team post trip 13:51:23.303 041 / N/A FIRE WATER SYSTEM INITIATED INIT No Yes* discussion with Operations personnel indicated 13:51:27.436 041 / N/A FIRE WATER SYSTEM INITIATED NORMAL No that the fire water system actuated. Additional I&C checks recommended RCE 001061, Rev. 1 Page 112 of 130

Attachment 10 North Anna Power Station Unit 2 14:16:18.714 062 / N/A STM GEN 1B LOLO LEVEL CH-l- NORMAL No SG 1B Level did not reach 18 % until Il-Ill 14:19:26:.909. All other indications prior to and after this time were accurate.

UNKNOWN - I&C Check Calibration.

13:51:11.072 076 / N/A RWST CHEM ADD TK LO LEVEL LO Ne Yes* PCS L2QS001 A 1 second data showed changes 13:51:11.224 076 / N/A RWST CHEM ADD TK LO LEVEL NORMAL Ne Yes* starting at 13:51:10.7 ending at 13:51:17.3 with 13:51:11.554 076 / N/A RWST CHEM ADD TK LO LEVEL LO Ne Yes* level changing between MAX of 87.090% to MIN 13:51:11.858 076 / N/A RWST CHEM ADD TK LO LEVEL NORMAL No Yes* of 84.188%. TRIP=85.0% and RESET=86.0%.

13:51:12.149 076 / N/A RWST CHEM ADD TK LO LEVEL LO No Yes* Level changes due to earthquake causing 13:51:12.234 076 / N/A RWST CHEM ADD TK LO LEVEL NORMAL No Yes* LO/NORMAL events are expected.

13:51:12.777 076 / N/A RWST CHEM ADD TK LO LEVEL LO No Yes* Valid LO/NORMAL indications per alternate 13:51:13.009 076 / N/A RWST CHEM ADD TK LO LEVEL NORMAL No Yes* evaluation.

13:51:13.190 076 / N/A RWST CHEM ADD TK LO LEVEL LO Ne-Yes*

13:51:13.438 076 / N/A RWST CHEM ADD TK LO LEVEL LO No Yes*

13:56:00.495 116 / V0320D STA SERV 2 of 3 BUSES UV -RX NORMAL No Yes* After further electrical evaluation, indications are V0321 D TRIP considered valid.

V0322D 13:56:00.497 116/ V0320D STA SERV 2 of 3 BUSES UF -RX NORMAL Ne Yes* After further electrical evaluation, indications are V0321 D TRIP considered valid.

V0322D

• Yes per Alternate Evaluation RCE 001061, Rev. 1 Page 113 of 130

ATTACHMENT 11 Purdue Assessment October 6, 2011 Evaluation of the Dominion's Root Cause Analysis for the North Anna Power Plants Scram Due to a Seismic Event on August 23, 2011 Mamoru Ishii and Takashi Hibiki School of Nuclear Engineering Purdue University 400 Central Drive, West Lafayette, IN 47907-2017, USA

1. Introduction 1.1 Brief Introduction of North Anna Power Plants Scram Due to a Seismic Event 1.1.1 North Anna Power Station [1]

North Anna generates 1,806 megawatts from its two units - enough electricity to power 450,000 homes. Unit 1 began commercial operation in June, 1978 and Unit 2 followed in December 1980. North Anna is located in Louisa County in central Virginia, northwest of Richmond. The facility was named after the North Anna River, which was dammed to form the 9,600-acre Lake Anna reservoir and the 3,400-acre Waste Heat Treatment Facility, used to provide cooling water for the station. The Lake Anna reservoir and the Waste Heat Treatment Facility have become a popular outdoor recreational area, whose shoreline is dotted with homes and cabins.

There are a number of marinas, campgrounds and a large state park on the Lake Anna reservoir [1].

1.1.2 Virginia Earthquake [2]

At 1:51pm on August 23, 2011, a 5.8 magnitude earthquake occurred, centered south of Mineral, Virginia, and, 11 miles from the North Anna Power Station. The Associated Press reported the quake "was felt as far north as Rhode Island, New York City and Martha's Vineyard, Mass." The reactors automatically shut down and, because of a loss of offsite power, four diesel generators started up to supply electricity to safety systems. The plant reported an "Alert" status, the second lowest level of four NRC emergency classifications, until 11:16am on August, 24, 2011. One of the diesel generators experienced a coolant leak and was secured by operators.

The station blackout diesel generator was activated to replace the secured diesel generator, which was repaired.

Offsite power was restored later on August 23. Dominion also reported that the aftershocks did not affect the power plant. Also on August 24, Dominion announced that it had ended the "Notice of Unusual Event", the least serious of the NRC emergency classifications, at the North Anna Power Station following inspection of equipment susceptible to seismic activity.

RCE 001061, Rev. 1 Page 114 of 130

ATTACHMENT 11 Purdue Assessment 1.2 Objective of This Report The earthquake caused both North Anna units to trip off-line. The protection signal that caused the trip was the high negative flux rate trip. The North Anna Root Cause Evaluation Team (NARCET) has been investigating the root cause of this trip from a variety of different angles. One of possible root causes proposed by NARCET is bubble burst due to the earthquake. The purpose of this report is to review and evaluate if the bubble burst can be a significant contributor to the reactor response.

2. Key Information to Evaluate the Root Cause (Provided by Dominion) 2.1 Reactor Information

" Nominal core power: 2940 MWt

• Nominal pressure: 2250 psia (15.3 MPa)

" Nominal flow: 310,000 gpm

" Nominal inlet and average temperatures: Tm=549 OF (287 °C) and Tav=580.8 °F (304.9 °C)

• Saturation temperature: Tsat=652.66 OF (344.8 °C)

• No. of fuel assemblies: 157

• Rod array: 17 by 17

• Fuel rods per fuel assembly: 264

• Guide tubes per fuel assembly: 24

• Instrument tubes per fuel assembly: 1

• Grids: 11 grids, top and bottom grids without mixing vanes above and below fuel region, 6 equally spaced structural grids with mixing vanes in fuel region, 3 mid-span mixing grids between the upper four structural grids.

• Each of the fuel assemblies are held in place by two alignment pins on the lower core plate and two alignment pins of the upper core plate.

• Nominal component dimensions for fuel assemblies are given in Table 1.

" A fuel assembly cross section for the 17x17 fuel array is shown in Fig. 1.

Table 1 Nominal Component Dimensions for Fuel Assemblies.

Component Cold dimensions Fuel rod OD 0.374 in (9.50 mm)

Pin pitch 0.496 in (12.60 mm)

Grid tube OD 0.482 in (12.24 mm)

Instrument tube OD 0.482 in (12.24 mm)

Bundle pitch 8.466 in (215.0 mm)

Bundle flow area 38.1089 in2 (24586 mm2 )

RCE 001061, Rev. 1 Page 115 of 130

ATTACHMENT 11 Purdue Assessment CLSE C 0ONTROLRO AC

-. CONTRL CU.IER E~O FUEL RODS (264-)

O0 = 0.374 CLAD THICXNESS = 0.0225 CLAD MATERIAL - ZIRC-4 OR ZIRLO RECONSTITUTED FUEL ASSY'S MAY HAVE LESS THAN 264 z FUELED RODS Figure 1 Fuel Assembly Cross Section 17x17.

2.2 Core Power Information

• Hot channel census: 36 % of the core has a hot channel relative power greater than 1.20. The detailed information on the percentage of the core occupied by different hot channel factors is given in Table 2.

• The 0.5 % void fraction change for a 10 % power decrease is a core wide void fraction (volume average void fraction). The void generation would only be credible in the top part of the fuel assemblies, so the local void fraction in the hottest assemblies would have to be significantly larger than 0.5 %.

• Sub-cooling (°F) for 1.5, 1.4, 1.3, 1.2 and 1.1 radial peaking at the end of the heated length of the fuel assembly are 6.06, 11.86, 17.66, 23.66 and 29.76 °F, respectively.

• Sub-cooling (QC)for 1.5, 1.4, 1.3, 1.2 and 1.1 radial peaking at the end of the heated length of the fuel assembly are 3.36, 6.59, 9.81, 13.14 and 16.53 'C, respectively.

RCE 001061, Rev. 1 Page 116 of 130

ATTACHMENT 11 Purdue Assessment Table 2 Percentage of Core Occupied by the Hot Channel with Certain Relative Power.

Hot Channel Relative power Percent of Core 1.45 1%

1.43 2%

1.40 5%

1.35 5%

1.30 7%

1.25 7%

1.20 9%

2.3 Earthauake Information Figure 2 (a) and (b) show Engdahl data from the containment basemat recorder in horizontal and vertical directions, respectively. Please note that Engdahl is the data collection equipment that is installed at the North Anna Power Station for collecting seismic data.

Engdahl Data for North Anna Unit 1 & 2 Containment Basemat Elevation 0.6 0.5 0.4 0

4-0 0.3 0.2 0.1 0.0 0.1 1 10 100 Frequency [Hz]

Figure 2 (a) Engdahl Data from the Containment Basemat Recorder - Horizontal Direction.

RCE 001061, Rev. 1 Page 117 of 130

ATTACHMENT 11 Purdue Assessment Engdahl Data for North Anna Unit 1 & 2 Containment Basemat Elevation 0.40 0.35 0.30 0.25 0.20 o 0.15 0.10 0.05 0.00 0.1 1 10 100 Frequency [Hz]

Figure 2 (b) Engdahl Data from the Containment Basemat Recorder - Vertical Direction.

" In Figs.2 (a), blue and red lines indicate OBE (Operating Bases Earthquake) and DBE (Design Bases Earthquake), respectively. For comparison purposes 2% dampening has been applied both OBE and DBE.

The amount of dampening is placed on the seismic fluctuations in performing analytical analysis in structural codes (e. g., ANSYS). Green and purple lines with symbols indicate the data in the horizontal directions of East-West and North-South, respectively. In Figs.2 (b), blue and red lines indicate OBE (Operating Bases Earthquake) and DBE (Design Bases Earthquake) with 2 % of dampening, respectively.

Green line with symbols indicates the data in the vertical direction.

" Maximum accelerations in horizontal and vertical directions are about 500 Gal and 300 Gal, respectively.

" Frequencies in horizontal and vertical directions in horizontal and vertical directions exceed 10 Hz at maximum accelerations.

" Figure 3 shows the time history of North Anna Unit 1 during earthquake.

• In Fig.3, black broken and red solid lines indicate the power averaged by the measurements at N41, N42, N43 and N44, and at N43 and N44, respectively. Black chain line indicates the trip signal. Blue, purple and green lines indicate the N-S seismic, E-W seismic and vertical seismic accelerations.

" The instrumentation locations are shown in Fig.4. N41, N42, N43 and N44 are locations of power range and N31 and N32 are the locations of source range, respectively.

" Based on the data shown in Fig.3, the earthquake oscillation frequency is approximately 10-to-20 Hz.

RCE 001061, Rev. 1 Page 118 of 130

ATTACHMENT 11 Purdue Assessment N1 Power vs Time During Earthquake (Average NI signals) 105 100 0.5 95 0.4 0.3 85 0.2 £ CL 80 0.1 "

75 0 70 -0.1 65 -0.2 60 1i S4 4 4- 4 - 4-0.3 51:11.12 51:11.32 51:11.52 51:11.72 51:11.92 51:12.12 51:12.32 51:12.52 51:12.72 51:12.92 Time [min:seconds.s]

Figure 3 Time History of North Anna Unit 1 during Earthquake.

RCS SLurImary. Rey I "S' Loop Inlet

~A/

4 7270, 19" Lo~opkti Figure 4 Detector Orientations.

RCE 001061, Rev. 1 Page 119 of 130

ATTACHMENT 11 Purdue Assessment 2.4 Core Power Oscillation Information S Figure 5 shows the time history of North Anna Unit 1 during earthquake.

S In Fig.5, dark blue, red, green and purple lines indicate the power measured at N41, N42, N43 and N44, respectively. Orange line indicates the power averaged by the measurements at N41, N42, N43 and N44.

Light blue line indicates the reactor trip signal.

N1 Power vs Time During Earthquake 105 1 0.9 100 0.8 95 0.7 Lj) 4-J 0.6 (10 90 4g-J 0.5 L..

0 85 0.4 1n 0.3 80 0.2 75 0.1 70 f--- - 0 51:07.5 51:13.5 Time [min : seconds s]

Figure 5 Time History of North Anna Unit 1 during Earthquake.

• As shown in Fig.5, a 13 % core power decrease on average was observed before the scram on Unit 1.

• Movement of the core barrel and the nuclear instrumentation (NI) detectors may explain approximately 5 %

of the magnitude of the observed power oscillations, especially in the direction of stronger movements.

This may explain some of the non-symmetric behavior of the core.

2.5 Other Important Information The distance between the fuel assemblies is approximately 0.070 inches (1.78 mm) at hot conditions.

Westinghouse has performed analyses for the Design Basis Earthquake (DBE), also referred to as the Safe RCE 001061, Rev. 1 Page 120 of 130

ATTACHMENT 11 Purdue Assessment Shutdown Earthquake (SSE). The DBE consists of strong motion over ten seconds. The results of the evaluation indicate that the first impact of fuel assembly to an adjacent fuel assembly is relatively long (-3 seconds) after the time of NI oscillations. During the time of the NI oscillations, each of the fuel assemblies are bowed approximately the same amount so that there is no closure of the gap between the fuel assemblies.

Westinghouse has confirmed that there is no reduction in the gap between fuel assemblies in the time frame of interest based on their calculation.

3. Bubble Burst Theory Proposed by North Anna Root Cause Evaluation Team 3.1 Basic Concept of Bubble Burst Theory The reactor trips on negative flux rate due to power oscillations created by seismic vibrations disrupting the laminar sublayer along the cladding wall resulting in rapid and transitory bubble bursts that add negative reactivity due to the void defect and positive reactivity when they subsequently collapse.

3.2 Phenomenological Explanation of Reactor Trip Using Bubble Burst Theory

• On August 23, 2011 both units at the North Anna Power Station tripped on negative flux rate as they were being shaken by an earthquake. Power oscillations were observed prior to the reactor trips. The last power decreases prior to the trips were of sufficient magnitude and rate to generate the negative rate trip on both units. There is a high degree of correlation between the power oscillations and the seismic waves striking the station. The Root Cause Evaluation Team has postulated that the major cause of the power oscillations was void defect created by bursts of bubbles as the seismic waves disrupted the laminar sublayer adjacent to the fuel cladding wall. The improved convection heat transfer was responsible for the bubble bursts. The bubbles collapsed very quickly since the coolant bulk was substantially subcooled. The collapse of the bubbles in the bulk coolant added positive reactivity which caused the power increases which followed the power decreases. The reactor trips stopped the power oscillations on both units as the rapid insertion of the control rod reactivity overwhelmed the void reactivity effect.

• It is noteworthy that the power oscillations were more pronounced on Unit 2 which was near end of core life and therefore had a larger void coefficient than Unit 1. It is postulated that a substantial part of the fuel cladding surface is just below the threshold for nucleate boiling. As the seismic waves shake the laminar sublayer the boiling curve (see Fig.6) above shifts down while the heat flux hardly changes (note that the vertical scale is logarithmic). A large portion of the cladding surface is suddenly in the nucleate boiling region that was formerly subcooled single phase convection. This causes the bubble burst.

RCE 001061, Rev. 1 Page 121 of 130

ATTACHMENT 11 Purdue Assessment H PO ST.DRV OUT HEAT TAAN Fir- RRFGMU H

FORCED CCýNVECTI ON

'VAIPOflIZATION H-EA'r

-FAMPE1 FULLY DEVELOPED POSTr-CflYOU7iPcOh CON YE-:7IV' H E~AT "AN 3 E R FORCED

~nURMTION SOTCTU ISINGLE-PHASE - - - - - - - - - -

FORCQED CONVECTIPON MNfl L,

Figure 6 Comparison of Typical Convective and Pool-Boiling Curves [3].

4. Assessment of Bubble Burst Theory 4.1 Assessment Assuming No Effect of Earthquake on Water Gap

" As mentioned in the section 2.5, Westinghouse has confirmed that there is no reduction in the gap between fuel assemblies in the time frame of interest based on their calculation. In this section, it is assumed that there is no reduction in the gap between fuel assemblies in the time frame of interest.

• According to the North Anna Root Cause Evaluation Team, "It is postulated that a substantialpart of the fuel cladding surface is just below the thresholdfor nucleate boiling. As the seismic waves shake the RCE 001061, Rev. 1 Page 122 of 130

ATTACHMENT 11 Purdue Assessment laminarsublayer the curve above shifts down while the heatflux hardly changes (note that the vertical scale is logarithmic). A large portion of the cladding surface is suddenly in the nucleate boiling region that was formerly subcooled single phase convection. This causes the bubble burst." However, the basis is not clear why the boiling curve shifts down due to seismic waves shaking the laminar sublayer.

In this assessment, a key parameter is the "thermal-boundary layer" which is defined as the layer from the fuel pin wall to the location where the coolant temperature reaches to saturation temperature. Vapor bubbles created by wall heat flux can exist only within the thermal-boundary layer and when bubbles grow beyond the thermal-boundary layer, the part of the bubbles outside the thermal-boundary layer is collapsed due to subcooled coolant. Thus, the thermal-boundary layer is a key to the discussion of void fraction in subcooled boiling flow. Figure 7 depicts typical images of bubble collapse taken by using atmospheric pressure boiling loop with an internally heated annulus [4]. The red dotted line in Fig.7 indicates the thermal-boundary layer considered in the assessment of [4]. The bubble diameter in millimeters is shown below in Fig. 7.

Tie0m* 0.2 0.4 0.6 0.8 "1 1.2 Diameter 0.1812 0.2093 0.2097 0.2097 D.1974 0.1803 0,1658 (nMM):

Time (ms): 1.4 1.6 1.8 2 2.2 2.4 2.6 Diameter 0.1510 0.1362 0.1168 0.0984 0.0867 0.0718 Figure 7 Typical Images of Bubble Collapse after Bubble Detachment from Heater Wall (Inlet Temperature=98 "C, Heat Flux=28.6 kW/m 2, Water Velocity=0.595 m/s, Pressure=0.103 MPa).

The time frame for bubble formation and collapse in a subcooled liquid is dependent on physical properties and thermal and flow conditions [5-7]. However, the time frame is typically very short. For example, the time frame for bubble collapse indicated in Fig.7 is on the order of milliseconds. Thus typical time frame for bubble formation and collapse may be considered to be two orders of magnitude faster than the observed power oscillations. If the formation of bubbles dominates the power oscillation, the power oscillation pattern would be synchronized with the seismic oscillation pattern. However, the frequency of the power oscillation is much less than the frequency of the seismic vibration. It should be noted here that the effect of simulated seismic vibration on bubble growth and condensation process is negligible [8].

RCE 001061, Rev. 1 Page 123 of 130

ATTACHMENT 11 Purdue Assessment According to the North Anna Root Cause Evaluation Team, "Reactortrips on negativeflux rate due to power oscillationscreated by seismic vibrations disruptingthe laminarsublayer along the cladding wall resulting in rapid and transitorybubble bursts (suddenformation of bubble) that add negative reactivity due to the void defect and positive reactivity when they subsequently collapse." However, in order to make bubble bursts, thermal-boundary layer should be sufficiently thick. The order of magnitude in created bubble size is approximately estimated by Db : ýs/gDr =0.986 mm. Thus, the thermal-boundary layer should be thicker than 1 mm to enable bubble burst. Beyond the thermal boundary layer, the formation of bubbles is not possible due to the subcooled liquid and condensation. Based on the information such as fuel rod OD of 9.50 mm and pin pitch of 12.60 mm, hydraulic diameter is estimated to be 11.70 mm. The minimum gap between fuel pins is 3.10 mm. Even if the laminar sublayer along the cladding wall is disrupted, the thermal-boundary layer should be thickened to sustain bubble burst. The basis is not clear where the energy to thicken the boundary layer come from. It is noted that the fuel rod may not be able to provide the energy to the thermal boundary layer in the observed time frame.

• If the core is completely single phase flow, the flow in the core is basically incompressible flow. If the calculation result by Westinghouse (no gap change) is adopted, the thermal-boundary layer indicated by red layers in Fig.8 would follow the moving fuel rods indicated by dark blue circles in Fig.8.

iOe

• 0@

Figure 8 Schematic Diagram of Synchronized Behavior of Fuel Rods and Thermal-Boundary Layer against Seismic Vibration.

When 1 mm thermal boundary layer is created in sub-channel, the area ratio of the thermal boundary layer to the subchannel is 17.8 %. In order to create a core average void fraction 0.5 %, the local void fraction should be 2.8 %. If it is very conservatively assumed that void is created along 1/3 of fuel length in hot channel with hot channel relative power higher than 1.2 create void fraction (36 % of core), the core volume where void is generated is approximately 12 %. In order to achieve a core average void fraction of 0.5 %,

RCE 001061, Rev. 1 Page 124 of 130

ATTACHMENT 11 Purdue Assessment the average sub-channel void fraction where void is generated should be 4.17 % (=0.5 % x 100/12). Then, local void fraction should be 23.4% (=4.17% x 100/17.8)" 12. Hot channels with hot channel relative powers of 1.2 and lower may not be able to participate in the generation of local voids due to local channel subcooling margins. Note that sub-cooling (C) for 1.2 peaking at the end of the heated length is 13.14 'C.

As discussed above, it may not be plausible to generate sufficient void (core wide void fraction of 0.5 %) to decrease the core power up to 10 %. Thus, the bubble burst theory may not be plausible to create sufficient void fraction to change the power. It can be concluded that the bubble burst theory alone does not explain the magnitude of change or the asymmetric quadrant behavior of the nuclear instrumentation observed in North Anna Units 1 and 2.

4.2 Assessment Assuming Some Effect of Earthquake on Water Gap

• As mentioned in section 2.5, Westinghouse has confirmed that there is no reduction in the gap between fuel assemblies in the time frame of interest based on their calculation. In this section, it is assumed that initially small subcooled bubbles exist near the fuel rod. It is also assumed that the fuel assemblies are bowed and gap change occurs.

" When the fuel assemblies are non-structurally moved (i.e., no longer engaged by the alignment pins in the upper and lower core plates), the pushed coolant at the core baffle needs to flow upward or downward resulting in secondary flow as shown in Fig.9. The secondary flow rate is dependent on the change in gap.

Thus, secondary flow in the core can be possible and due to the seismic vibration, the thermal-boundary layer may be localized (see Fig. 10), where the bubble can grow in the thicken thermal-boundary layer, resulting in void fraction increase, namely reactivity decrease. If the thermal-boundary layer thickness becomes double, the void fraction is approximately increased 8 times (-c 23). Nariai and Tanaka [9]

performed an experiment to address the effect of seismic vibration on void fraction in subcooled boiling flow. Based on the experimental data, the void fraction is decreased with the frequency higher than 10 Hz, which means that the bubble collapse is enhanced with the frequency higher than 10 Hz. The detached bubbles are collapsed, resulting in increased reactivity. The details of the experiment can be found in Appendix.

• If movement of fuel assemblies were possible due to the seismic vibrations, the fuel assemblies may move in a way indicated in Fig. 9. If this happen, several reasons to explain the trip may be possible.

(1) Secondary flow in the core (see Fig.9) may be possible resulting in potential explanation of the localization of thermal-boundary layer as shown in Fig. 10. Then, initially existing tiny bubbles can grow or new bubble generation may be possible. According to the research results in Japan, the bubble can collapse significantly when the frequency is higher than 10 Hz [9].

(2) Due to the gap change, some neutron leakages may be possible.

1 If it is assumed that void is created along 1/3 of fuel length in hot channel with hot channel relative power higher than 1.35 create void fraction (13 % of core), the core volume where void is generated is approximately 4.33 %. In order to achieve a core average wide void fraction of 0.5 %, the average sub-channel void fraction where void is generated should be 11.5 % (=0.5 % x 100/4.33).

Then, local void fraction should be 64.6 % (=11.5% x 100/17.8).

2 If we assume that only the hot channel with hot channel relative power higher than 1.4 creates voids, local void fraction in thermal-boundary layer exceeds 100  %.

RCE 001061, Rev. 1 Page 125 of 130

ATTACHMENT 11 Purdue Assessment (3) Ratio of moderator to fuel rod may change locally.

/Core baffle Flow

__________ .1J, Fuel avem blv Time Figure 9 Schematic Diagram of Induced Secondary Flow Mechanism.

Seismic vibration 4 Thermal boundary layer Fuel rod Figure 10 Schematic Diagram of Localized Thermal Boundary Layer Creation Mechanism.

5. Conclusions From section 4.1, with no significant reduction in the gap between fuel assemblies, the bubble burst theory is not plausible to create sufficient void fraction to explain the magnitude of change in power or the asymmetric quadrant behavior.

RCE 001061, Rev. 1 Page 126 of 130

ATTACHMENT 11 Purdue Assessment

• From section 4.2, under the assumed conditions, the thermal-boundary layer localization may.produce some growth of bubbles, which could be a potential minor contributor to the power oscillation seen at North Anna Power Plants.

" According to some additional information about the effect of the seismic motions on the core barrel and nuclear instrumentation detectors provided by the North Anna Root Cause Evaluation Team, the movement of the core barrel and the nuclear instrumentation detectors could explain approximately 5 % of the magnitude of the observed power oscillations, especially in the direction of stronger movements. This can explain some of the non-symmetric behavior of the core in Fig. 5.

6. Appendix - Additional Information on Effect of Seismic Vibration on Bubble Behavior

" Fukushima BWR plant and Onagawa BWR plant in Japan scrammed due to high neutron flux signal in 1987 and 1993, respectively. Then, they tried to understand the mechanism why the reactors were scrammed.

First they considered that this scram would be due to the reactivity increase caused by the collapse of vapor bubbles in subcooled region. They considered that the collapse of vapor bubbles was due to the earthquakes.

• Nariai and Tanaka performed a simulation experiment to confirm this mechanism [9]. They utilized vertical upward boiling water flow in 6 mm stainless heater in transparent polycarbonate. This experiment was performed under atmospheric pressure and the tested amplitudes were 4 and 6 mm and the tested shaking frequency range is 0-to 12.5 Hz. It turned out that this seismic vibration effect on void fraction would be small (see Fig.A1). As shown in Fig.A2, Ch-1 and Ch-2 mean the void fraction measurement in oscillation direction and direction normal to oscillation, respectively. It was observed that the void fraction was almost not affected by the oscillation up to the frequency less than 10 Hz. However, the void collapse phenomenon was observed at the frequency higher than 10 Hz. The effect of the oscillation amplitude on the void conditions was reported for the amplitude less than 4 mm and thus the effect of small amplitude on the void fraction is not clear.

• In relatively old BWR, fuel assembly gap near control rod is different from the one near the upper grid plate. This means that the reactor bypass flow channel gap is different between fuel assemblies. When the earthquake shaken the fuel assembly gap, the gap (coolant amount) was changed resulting in the increase of the reactivity. In Onagawa BWR, C lattice fuel bundles were loaded, and the water gap between fuel and inner wall was changed due to the earthquake. Then, the reactor scrammed due to high neutron flux (120

%). However, this phenomenon did not occur in a reactor with D lattice bundle. It should be noted here that in D lattice bundle, water gap was uniform in horizontal bundle plane but in C lattice bundle, water gap was not uniform. Based on the shaking test of the fuel assembly, the gap displacement was measured.

Then, the reactivity analysis was performed with this measured gap change as an input condition. Based on the analysis, they concluded that the major reason for the reactor scram would be the gap change [10].

Then, they modified the lattice and the issue was solved.

RCE 001061, Rev. 1 Page 127 of 130

ATTACHMENT 11 Purdue Assessment A

FR.UFOUENCY ,Hz Figure Al Effect of Simulated Seismic Vibration Frequency on Void Fraction.

U. I FREWPIiICYm 8.75 '-14z 2

qw 0.81 MW/m 0.( t.ch-2 C-)

0A 2h-6-A

)2 .... .. ..... ... .........

I LN A.-

0 60 120 180 240 300 360 Figure A2 Effect of Simulated Seismic Vibration Direction on Local Void Fraction.

A fuel bundle in a PWR is more rigid than a BWR fuel bundle and the change in the coolant density flowing through a PWR core is smaller than the density change in a BWR core. It may be difficult to consider that the trip that occurred at North Anna Power Station is due to the same phenomenon as the trip that occurred in the Japanese BWRs.

RCE 001061, Rev. 1 Page 128 of 130

ATTACHMENT 11 Purdue Assessment According to Tohoku Electric Company, the BWR fuel assemblies can undergo a deflection up to 4 mm

[11].

7. Reference

[1] http ://dom.com/about/stations/nuclear/north-anna/index.j sp

[2] http://en.wikipedia.org/wiki/NorthAnnaNuclearGeneratingStation

[3] R. T. Lahey, Jr. and F. J. Moody. The Thermal-Hydraulics of a Boiling Water Nuclear Reactor, American Nuclear Society (1993).

[4] R. Situ, T. Hibiki, X. Sun, Y. Mi and M. Ishii, "Flow Structure of Subcooled Boiling Flow in an Internally Heated Annulus," International Journal of Heat and Mass Transfer, 47, 5351-564 (2004).

[5] L. S. Tong and Y. S. Tang, Boiling Heat Transfer and Two-Phase Flow, Taylor & Francis (1997).

[6] H. S. Park, T. H. Lee, T. Hibiki, W. P. Baek and M. Ishii, "Modeling of the Condensation Sink Term in an Interfacial Area Transport Equation," International Journal of Heat and Mass Transfer, 50, 5041-5053 (2007).

[7] R. Situ, M. Ishii, T. Hibiki, J. Y. Tu, G. H. Yeoh and M. Mori, "Bubble Departure Frequency in Forced Convective Subcooled Boiling Flow," International Journal of Heat and Mass Transfer, 51, 6268-6282 (2008).

[8] S. Kawamura, A. Orii, H. Karasawa, K. Nishida and H. Soneda, "Effect of Horizontal Excitation on Bubble Behaior at Subcooled Temperature," Proceedings of 1996 Spring Annual Meeting, Osaka University in Japan, Paper No. A49 (1996) (in Japanese).

[9] H. Nariai and T. Tanaka, "Void fraction of Subcooled Flow Boiling around Oscillating Heater Rod,"

Proceedings of 1994 Spring Annual Meeting, Tsukuba University in Japan, Paper No. J36 (1994) (in Japanese).

[10] S. Kawamura, Y. Koshi, K. Hattori, H. Katayama, S. Fujimoto and Y. Kudo, "A Study on Neutron Flux Transient in Oscillating Fuel Assemblies," Proceedings of 1996 Spring Annual Meeting, Osaka University in Japan, Paper No. A44 (1996) (in Japanese).

[11] Nihon Keizai Shinbun (Japan Economics Newspaper), August 21 in 1995 (

http://e.nikkei.com/e/fr/freetop.aspx)

RCE 001061, Rev. 1 Page 129 of 130

ATTACHMENT 13 1H Bus & Seismic Activity 6000 0.5 0.45 0.4 0.35 5000 0.3 0.25 0.2 4000 0.15 0.1 0.05 3000 0 I 5 I I II -0.05

- 1H Bus Voltage

- Seismic Transverse

-0.1 2000 I I -0.15

-0.2

-0.25

-0.3 1000

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-0.4

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ri Time ,14i 7Hz Hz I 1-Ln tn LA LA LA LA LA LA RCE 001061, Rev. 1 Page 130 of 130