NL-10-0403, Response to Apparent Violation in Inspection Report 05000366-09-005

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Response to Apparent Violation in Inspection Report 05000366-09-005
ML101190363
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 03/19/2010
From: Madison D
Southern Nuclear Operating Co
To:
NRC/RGN-II/DCI
References
IR-09-005, NL-10-0403
Download: ML101190363 (188)
v • d • e


Text

D. R. Madison IDennis) Southern Nuclear Vice President - Hatch Operating Company, Inc.

Plant Edwin I. Hatch 11028 Hatch Parkway, North Baxley, Georgia 31513 Tel 912.537.5859 Fax 912.366.2077 March 19,2010 SOUTHERN'\'

COMPANY Energy to Serve Your World'"

Docket Nos.: 50-366 NL-10-0403 U. S. Nuclear Regulatory Commission, Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Edwin I. Hatch Nuclear Plant, Unit 2 Response to Apparent Violation in Inspection Report 2009005 Ladies and Gentlemen:

On February 12, 2010, the U.S. Nuclear Regulatory Commission (NRC) issued Inspection Report 2009005 to Edwin I. Hatch Nuclear Plant. This inspection report identified a preliminary finding defined as an Apparent Violation (A V) of Technical Specification 5.4, Procedures, for failing to establish and perform preventive maintenance activities to replace electrolytic capacitors prior to their failure. In addition, this finding had a cross-cutting aspect in the Operating Experience component of the Problem Identification and Resolution area, P.2(b),

because pertinent industry operating experience was not effectively incorporated into the preventive maintenance program. This inspection report requested that Southern Nuclear Operating Company (SNC) submit a response to the NRC within 30 days of receipt. A phone call between Mr. Mark Ajluni, SNC Licensing Manager, and Mr. Scott Shaeffer, NRC Branch Chief, took place on March 2, 2010. During this call, Mr. Ajluni requested and was granted a five-business-day extension to the requested 30-day response.

SNC chose to respond in writing to the aforementioned AV and declined the opportunity for a pre-deciSional meeting with the NRC. SNC does not dispute the AV or the cross-cutting issue stated in this inspection report. However, SNC assessed the safety significance of the subject AV, as described in Enclosure 1, and concluded that this finding is of very low safety significance (Green) by a wide margin. The decreased safety significance is based upon additional plant-specific factors beyond those included in the NRC's assessment that take into account the real-world operational response to the event. These factors are listed below:

1. Plant staff actions, which would be taken in an actual event, to shut down the 2A Emergency Diesel Generator (EDG) prior to damage, repair and recover the 2A Plant Service Water pump, and re-start the 2A EDG.

U. S. Nuclear Regulatory Commission NL-10-0403 Page 2 The time-frame required to complete this recovery action without incurring damage to the 2A EDG is supported by a study conducted by Fairbanks Morse, the manufacturer of the EDG.

2. Fire modeling conducted by SNC demonstrates that fire scenarios, which take place in the 4160VAC emergency switchgear rooms, do not cause a LOSP.
3. Re-calculation of the Large Early Release Frequency (LERF) with SNC's LERF model results in a value of Green.
4. The exact time at which the 2A EDG LOSP timer card failed is unknown. The exact time at which the FW controller failed was known immediately, and it occurred four months after the discovery of the timer card failure. Therefore, they were never failed at the same time.

SNC requests that the NRC re-evaluate the safety significance of this preliminary finding considering the above information. This approach takes into account the real-world operational response to the event. contains Security Related Information as related to 10 CFR 2.390, the disclosure of which to unauthorized individuals could present a security vulnerability. Therefore, in accordance with 10 CFR 2.390(d)(1), SNC requests that Enclosure 2 not be made available electronically for public inspection in the NRC Public Document Room or from the NRC's electronic document system (ADAMS) accessible from the NRC Web site at htto:/Iwww.nrc.gov/readingrm/adams.html.

This letter contains no NRC commitments. If you have any questions, please advise.

Respectfully submitted, 1/1 \

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D. R. Madison Vice President - Hatch DRM/EGAI<>

Enclosures:

1. Response to Apparent Violation in Inspection Report 2009005
2. HNP Significance Determination Discussion for Two Incidents as per NRC report SRA HAT 0905 SOP Phase 111- on Compact Disc (Security Related Information - Withhold from public disclosure under 10 CFR 2.390)
3. Simulated ScenariOS
4. 34AB-P41-001-2, Loss of Plant Service Water

U. S. Nuclear Regulatory Commission NL-10-0403 Page 3

5. 34S0-R43-001-2, Diesel Generator Standby AC System
6. 34AB-R22-003-2, Statton Blackout
7. 73EP-EIP-001-0, Emergency Classification and Initial Actions
8. DOEJ*HR2100455301-M001, LOSP Evaluation of Fire Areas 240812409
9. Fairbanks Morse Report 5.08-6.06-0260, Genset Operation without Service Water
10. RCIC Calculation 1 (Unit 2 Battery Life)
11. RCIC Calculation 2 (Core Cooling from AC-Independent Systems Until Battery Depletion)
12. United Controls International Repair Report 3697, Failed LOCA CARD Test Report cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President - Letter with Enclosure 1 Only Mr. M. J. Ajluni, Manager, Nuclear Licensing - Letter with Enclosure 1 Only Ms. P. M. Marino, Vice President, Engineering - Letter with Enclosure 1 Only RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Document Control Desk Mr. L. A. Reyes, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch Mr. R. P. Zimmerman, Director, Office of Enforcement

Edwin I. Hatch Nuclear Plant Enclosure 1 Response to Apparent Violation in Inspection Report 2009005

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 Table of Contents I. Abstract II. Background III. SNC Risk Assessment IV. Conclusion V. References E1-i

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 I. Abstract Southern Nuclear Operating Company (SNC) submits the following analysis in response to Inspection Report 2009005, which was issued on February 12, 2010 by the U.S. Nuclear Regulatory Commission (NRC). This analysis concerns an Apparent Violation (AV) of Technical Specification (TS) 5.4, Procedures, for failing to establish and perform preventive maintenance activities to replace electrolytic capacitors prior to their failure, which was preliminarily determined by the NRC to be of greater than very low safety significance. The following information outlines SNC's position on the issue as well as the safety significance determination that SNC performed.

SNC does not dispute the AV or the cross-cutting issue stated in this inspection report. However, SNC assessed the safety significance of the subject AV, as described in this enclosure, and concludes that this finding is of very low safety significance (Green) by a wide margin. The decreased safety significance is based upon additional plant-specific factors beyond those included in the NRC's assessment; these factors are listed below:

1. Plant staff actions, which would be taken in an actual event, to shut down the 2A Emergency Diesel Generator (EDG) prior to damage, repair and recover the 2A Plant Service Water pump, and re-start the 2A EDG.

The time-frame required to complete this recovery action without incurring damage to the 2A EDG is supported by a study conducted by Fairbanks Morse, the manufacturer of the EDG.

2. Fire modeling conducted by SNC demonstrates that fire scenarios, which take place in the 4160VAC emergency switchgear rooms, do not cause a LOSP.
3. Re-calculation of the Large Early Release Frequency (LERF) with SNC's LERF model results in a value of Green.
4. The exact time at which the 2A EDG LOSP timer card failed is unknown.

The failure of the FW controller was known immediately, and it occurred four months after the discovery of the timer card failure. Therefore, they were never failed at the same time.

SNC requests that the preliminarily greater than Green finding be reduced to a finding of very low (Green) safety significance considering the above information.

II. Background On February 12, 2009, the Unit 2A EDG LOSP timer card was found in a failed state during the performance of Logic System Functional Test (LSFT) 42SV-R43-018-2 on the 2A EDG. This failure prevented the 2A PSW pump from loading; therefore, the pump was unable to provide normal 2A EDG cooling. If called upon in an actual event, the 2C PSW pump was available to E1-1

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 function and provide a backup source of cooling water for the 2A EDG.

Investigation revealed that the input/output (I/O) error light was illuminated on the timer card. This light indicates that the I/O self-check program in the timer identified an anomaly with an input or an output, and, as designed, the processor shut down and locked up the timer. The timer card did not annunciate in the Main Control Room (MCR). During later investigations, the annunciator relay and the timer card were tested and found to function correctly. The timer card was tested a total of 10 times and functioned correctly each time. A visual examination was also completed, but no apparent defects were identified. Finally, the timer card was sent to United Controls International for testing (Enclosure 12). This testing concluded that "there is no apparent reason for the card to have failed the timing sequence, and the failure could not be duplicated even when stressed to help reveal a problem with the card." Ultimately, no conclusive cause was found for the relay failure or the timer card failure. The relay and the timer card were replaced.

On June 23, 2009, Unit 2 experienced an automatic scram resulting from main turbine and reactor feed pump turbine trips due to high Reactor Water Level (RWL), which was caused when 2C32-K648, the main FW median level controller, failed. This Yokogawa controller monitors inputs from three RWL sensing devices and transmits the median level indication to the 2C32-R600 Master Controller. The direct cause of the event was the failure of an electrolytic capacitor in the power supply for 2C32-K648. Analysis of the failed controller circuit board revealed a bulged capacitor and heat indications on the circuit board. The failure of the capacitor was attributed to age-related degradation. 2C32-K648 was installed in 1997 per Design Change Request (DCR)95-054. This and four other controllers of similar age were replaced prior to the Unit 2 start-up following this scram. The failed power supply was in service for approximately 12 years.

III. SNC Risk Assessment The risk analysis conducted by SNC regarding the 2A EDG LOSP timer card failure and the FW median level controller failure demonstrates a GREEN risk worth. This risk worth includes plant staff actions, effects of fire in the 4160VAC emergency switchgear rooms, more accurate LERF modeling, and realistic failure probabilities.

The risk evaluation follows the description supplied by the NRC Region II Senior Reactor Analysts, as far as scenarios are concerned, as provided to SNC in HAT 0905, SDP Phase III.

The SNC risk analysis considers the worst-case scenario, which assumes a Unit 2 station blackout resulting in no Unit 2 available diesel generator. This analysis includes restoration of one Unit 2 EDG to supply power to safety-related loads.

E1-2

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 2A EDG LOSP Timer Card Failure Scenario The first issue that this analysis addresses is the 2A EDG LOSP timer card.

During testing, it was discovered that this card failed in such a manner that the normal 2A PSW pump did not load to the 2A EDG; therefore, the 2A EDG did not have cooling water. If called upon in an actual event, the 2C PSW pump was available to function to provide a backup source of cooling water for the 2A EDG. This would allow the 2A EDG to have cooling water for the event. The exposure time that is considered for this case, and the corresponding fire cases, is 345 days. This time-frame is based on the T/2 approach because the time at which the actual failure occurred is unknown.

The risk issue arises when the power source for the 2C PSW pump (1 B EDG) fails in conjunction with the 2C EDG. This leaves the 2A EDG running (lightly loaded because it did not load a PSW pump) with no cooling water. The failure of the 1B EDG and 2C EDG completes the case for a station blackout.

HNP Unit 2 and Unit 1 4160VAC buses are depicted in Figure 1 of this enclosure. It is noted that the 1B EDG can supply the Unit 1 1F 4160VAC bus or the Unit 2 2F 4160VAC bus one at a time but not simultaneously. It is also noted that the Startup Transformers (SUT), SUT 2D and SUT 2C are supplied from separate transmission lines in the high voltage switchyard.

In order to account for the above scenario and attempt recovery, power must be restored. The normal accounts for offsite power restoration are modeled in a similar fashion to the GEMS modeling used by the NRC. However, SNC evaluates the potential recovery of the 2A EDG. This approach to the scenario takes into account the real-world operational response to the event.

This diesel is not broken nor is its cooling water supply (the 2A PSW pump).

Operator action, accounting for shutting down the 2A EDG prior to damage, is used in the risk evaluation. This HRA is based on a calculation performed for HNP by FAIRBANKS MORSE (Enclosure 9), which supplies both actual and calculated data for diesel operation, while unloaded and without cooling water. The values calculated are 38.94 minutes to the first oil temperature alarm and 26.4 minutes until the first cooling water alarm. Actual test data shows a much longer time-frame (up to 270 minutes for the lube oil system and up to 148 minutes for the cooling water system). Therefore, a reasonable value of 35 minutes for operation without cooling water is chosen for the HRA calculation. This time-frame is the assumed amount of time that the operators have to secure the 2A EDG prior to damage from lack of cooling water.

Lack of EDG cooling is alarmed at the individual EDG panels and at the alarm station above the controls for the PSW pumps in the MCR. The PSW PRESS LOW alarm on panel 2H11-P652-1 (see picture 3(a) in Enclosure 3) and the PSW MAIN HDR DIVI PRESS LOW alarm on panel 2H11-P650 are some of these indications. The actions for these annunciators lead the Operator to the ABNORMAL procedure for Loss of PSW (34AB-P41-001-2, Enclosure 4),

which routes the operator to the procedure for shutting down the EDG from E1-3

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 the MCR with an auto start signal present (34S0-R43-001-2, Enclosure 5).

This method is used for this analysis, but an Operator in the EDG Building at the EDG can also trip the fuel racks for a successful shut down.

These procedures have been in place for some time. Verification of Operator actions regarding this scenario response shows that approximately 7 to 10 minutes is all the time needed to perform the shut down action after the event is diagnosed.

The modeled scenario now requires plant staff to recover the 2A PSW pump.

This recovery will be accomplished with the aid of the Technical Support Center (TSC) because the Emergency Response Facilities are active at this time (73EP-EIP-001-0, Enclosure 7). Plant simulation of this recovery action reveals a time-frame of approximately 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This is a rather stringent evaluation with conservative assumptions, and in an actual event, the PSW recovery could be more likely accomplished within two hours when considering the on-call Engineering staff and the expertise SNC has available to apply in such a situation.

A distribution is plotted based on the 3.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> mean for time to repair and a length of allowance for six hours based upon successful operation of Reactor Core Isolation Cooling (RCIC) on battery power. The HRA and this repair are only applied to risk cutsets where RCIC is successful. SNC has calculations to show that RCIC will operate at least five hours on available battery power during a station blackout (see Enclosures 10 and 11). This length of time also provides a decrease in decay heat. No core damage is expected during this time-frame.

Fire Considerations for Scenarios In addition to the normal core damage calculations, select Fire PRA scenarios are evaluated for risk. These scenarios consider fires in individual 4160VAC emergency switchgear rooms that are relevant to the scenarios discussed herein. The historic IPEEE Fire PRA for HNP provides a conservative approach to estimating the risk associated with fire scenarios. This version is considered out-dated when compared to more modern fire modeling techniques. SNC performed fire modeling techniques to address the pertinent modeling scenarios. These results are much improved over the present Fire PRA, which is over 15 years old, and they demonstrate that no fire scenarios exist within the rooms in question that can cause a LOSP event.

Taking into account that there are no fire scenarios that cause a LOSP (DOEJ-HR21 00455301-M001, Enclosure 8), as assumed by the NRC Analysts, SNC could omit fire considerations from this analysis. However, SNC does consider the fire analysis, with a revised fire frequency (see ), for the worst-case subscenario in the switchgear rooms with the intent of demonstrating the over-conservatism of the Fire PRA used by the E1-4

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 NRC and showing that even with fire risk addition, the end result is a very low safety significance.

LERF Considerations for Scenarios SNC has a LERF model that is considered accurate because it meets current ASME standards and is based on a model that was peer reviewed. There is a large difference between the SNC calculations and NRC calculations. This is attributed to the significant differences between the SNC model and the NRC LERF process. The SNC model is more realistic than the NRC's process used to calculate a Large Early Release. SNC's risk worth is calculated to be Green, and the risk worth of LERF for the overall analysis is negligible within SNC calculations.

In most SNC cases, some credit is given to primary containment or (in some cases) secondary containment to function. In all considerations, absolute assurance of success is not necessarily required; instead, the possibility of success is considered. Pathways of containment failure cover the possibility of an existing containment breach, which would open a path to the environment. In addition, in common with the NRC provided cutsets, both high and low vessel pressure failure scenarios are considered. Secondary containment is used to retard radioactive release for the low pressure cases; this is not used in the NRC cutsets. This containment credit is where the difference between the NRC cutsets for LERF and the SNC calculations exists; early containment failure is assumed to occur in the NRC cases 100%

of the time. The approach of 100% containment failure with no defense assumed to lessen the severity of release is over-conservative and does not reflect industry practice.

In addition, the following allowances are considered by SNC for Large Early Release but not by the NRC because the NRC process does not consider (1) late depressurization, (2) use of Severe Accident Management Guideline (SAMG) water sources, and (3) consideration of building holdup. Item 1 is fully supported considering several hours exist until vessel failure, in all cases considered. This time increases if Reactor Core Isolation Cooling (RCIC) is in operation. In addition, it is directed by Emergency Operating Procedures (EOPs). Item 2 is proceduralized for the use of fire-water, and time exists to make the installation. Item 3 has been discussed with the Region II SRA and his counterpart at NRC Headquarters; this concept is recognized by the NRC.

FW Median Level Controller Failure Scenario In the SRA report, HAT 0905, a FW turbine trip on Unit 2, occurring June 23, 2009, is included in the overall risk analysis. The time-frame, or exposure time, for this event is given to be less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The risk numbers the NRC provides are not well understood because they are so high. SNC re-ran this case under the conditions used by the NRC with a more conservative approach to the Anticipated Transient without Scram (ATWS) initiators and E1-5

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 calculates risk values in the E-9 range. These values are included in this report.

SNC has concluded that it is not appropriate to add the risk from the above described FW controller failure to the risk for the 2A EDG LOSP timer card failure. However, as a conservative measure SNC has added the risk associated with the FW controller failure in this analysis. The reasons that this inclusion is inappropriate are as follows:

1) The 2A EDG LOSP timer card failure cause is unknown whereas the FW controller failure cause is known.
2) The 2A EDG LOSP timer card fault exposure was calculated using the T/2 approach because the actual failure time is unknown. The failure of the FW controller was known immediately and it occurred four months after the discovery of the timer card failure. Therefore, they were never failed at the same time.
3) No precedent exists to add the risk of components from independent events in different systems with different safety classifications simply to address the two events as a common issue.

Calculation Results:

The overall analysis is divided into four (4) parts, or scenarios, that also include a sensitivity study. The 2A EDG LOSP timer card failure is considered from both a dual-unit LOSP (1) occurrence and a single-unit LOSP (2) occurrence. This consideration allows SNC to account for the 1B EDG and its ability to transition from one plant unit to another (Le. 1F 4160VAC or 2F 4160VAC emergency bus) during a LOSP condition. The condition where the 1B EDG is available to Unit 2 only, which is considered to be a LOSP on Unit 2 only, is also reviewed. In addition, the case of the dual-unit LOSP, which allows assignment to either unit, is reviewed.

The fire scenario cases (3) are considered separately as dual-unit LOSP scenarios. The FW median level controller failure, or Loss of FW scenario (4), referenced in HAT 0905, is considered separately as well.

In all scenarios, except for Scenario 4, only a LOSP is evaluated. Scenario 4 evaluates total core damage frequency (CDF) with the SNC loss of FW and ATWS loss of FW initiating events set to a value of 1.0. Setting these events to total failure is per the NRC analysis.

Scenarios Evaluated:

Scenario 1: Dual-Unit LOSP Scenario 2: Single-Unit LOSP Scenario 3: Fire Scenarios Scenario 4: Loss of FW Scenario E1-6

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 CDF =Core Damage Frequency LERF =Large Early Release Frequency Note: These are the two metrics used to calculate risk in this analysis.

CCDP = Conditional Core Damage Probability CLERP = Conditional Large Early Release Probability SCENARIO CCDP CLERP 1 - Dual-Unit LOSP 3.12E-7 2.46E-9 2 - Single-Unit LOSP 3.10E-7 2.46E-9 3 - Fire in 2F Switchgear Room 8.06E-9 1.085E-10 3 - Fire in 2G Switchgear Room 2.92E-8 1.93E-10 4 - Loss of FW Scenario 4.57E-9 1.17E-9 These numbers are summed to provide total risk. Scenarios 1, 3, and 4 are one total, and Scenarios 2, 3, and 4 are another total. This allows SNC to account for the two overall scenarios of (1) a dual-unit LOSP and (2) a single-unit LOSP.

SCENARIO TOTALS CCDP CLERP Scenarios 1, 3, and 4 3.54E-7 3.93E-9 Scenarios 2, 3, and 4 3.52E-7 3.93E-9 The CCDP and CLERP account for the exposure-time (NRC RASP Handbook), or time out-of-service, of the equipment. Less than 1E-6 for CCDP and less than 1E-7 for CLERP is considered GREEN for this condition.

IV. Conclusion The risk analysis conducted by SNC regarding the 2A EDG LOSP timer card failure and the FW median level controller failure demonstrates a large difference from the NRC value calculated during the Significance Determination Process (SDP). SNC has determined that the subject AV demonstrates a Green risk worth. This risk worth includes plant staff actions, effects of fire in the 4160VAC emergency switchgear rooms, more accurate LERF modeling, and realistic failure probabilities.

The plant staff actions depend on the ability of the Operations staff to shut down the 2A EDG prior to damage, the Engineering and Maintenance staffs to diagnose the problem with the 2A PSW pump and re-establish cooling water flow to the 2A EDG, and the Operations staff to re-start the 2A EDG.

The two plant staff actions of greatest significance are the ability to shut down the 2A EDG prior to damage and establish cooling water flow with the 2A PSW pump. These actions were successfully simulated with plant staff and are described in Enclosure 3, Simulated Scenarios.

To ensure that the risk worth is indeed Green, SNC uses conservatisms in the analysis. These conservatisms involve the fire PRA and the failure probability. Taking into account that there are no fire scenarios that cause a E1-7

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 LOSP, as assumed by the NRC Analysts, SNC could omit fire considerations from this analysis. However, SNC does consider the fire analysis, with a revised fire frequency (see Enclosure 2) for the worst-case subscenario in the switchgear rooms, with the intent of smphasizing the over-conservatism of the Fire PRA used by the NRC and showing that even with fire risk addition, the end result is a very low safety significance. SNC also considers the fact that the 2A EDG LOSP timer card was tested 10 times by plant staff after the initial failure. This failure probability could have been used by SNC instead of the 100% failure probability that was assumed by the NRC. In any event, when the analysis is conducted with a 100% failure probability, the result still demonstrates a Green risk worth.

As discussed previously, SNC has a LERF model that is considered accurate because it meets current ASME standards and is based on a model that has been peer reviewed. There is a large difference between the SNC calculations and NRC calculations because of the significant differences between the SNC LERF model and the NRC LERF process. The SNC model is much more realistic than the NRC's process used to calculate a Large Early Release. SNC's risk worth is calculated to be Green, and the risk worth of LERF for the overall analysis is negligible within the SNC calculations and demonstrates nearly two decades of margin.

SNC also concludes that it is not appropriate to add the risk from the FW controller failure to the risk for the 2A EDG LOSP timer card failure. The reasons that this inclusion is inappropriate are as follows:

1) The 2A EDG LOSP timer card failure cause is unknown whereas the FW controller failure cause is known.
2) The 2A EDG LOSP timer card fault exposure was calculated using the T/2 approach because the actual failure time is unknown. The failure of the FW controller was known immediately and it occurred four months after the discovery of the timer card failure. Therefore, they were never failed at the same time.
3) No precedent exists to add the risk of components from independent events in different systems with different safety classifications simply to address the two events as a common issue.

SNC has calculated that the subject performance deficiency has a very low (GREEN) safety significance by a wide margin and requests that the safety significance of this preliminary finding be re-evaluated. The decreased safety significance is based upon reconsideration of several assumptions, which are provided above and incorporated into SNC's risk evaluation with the supporting documentation in the following enclosures. Therefore, SNC requests that the preliminarily greater than Green finding be reduced to a finding of GREEN safety significance.

E1-8

Enclosure 1 Response to Apparent Violation in Inspection Report 2009005 V. References

1. CR2009102221
2. CR 2009106352 E1-9

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Edwin I. Hatch Nuclear Plant Enclosure 2 HNP Significance Determination Discussion for Two Incidents as per NRC report SRA HAT 0905 SOP Phase III on Compact Disc

Edwin I. Hatch Nuclear Plant Enclosure 3 Simulated Scenarios

Enclosure 3 Simulated Scenarios

1. Shutting Down the 2A EDG The below described scenario was performed by an on-shift Licensed Reactor Operator (RO) in the Unit 2 Main Control Room in a simulated transient on March 2, 2010. This exercise was conducted in a format similar to a Job Performance Measure (JPM). A timeline was established by timing the actions performed by the RO during this exercise.

The RO was allowed to read the initial conditions (listed under Conditions of Scenario) and ask questions for clarification. He was also allowed to review the Station Blackout procedure (34AB-R22-003-2) and the Loss of Plant Service Water procedure (34AB-P41-001-2). Once the RO understood the plant conditions and that he was the operator responsible for monitoring Emergency Diesel Generator (EDG) operation, he was informed by the Shift Supervisor (SS) that a high temperature alarm was assumed received on the 2A EDG and was given the order to trip the diesel. He was also informed that the System Operator (SO) dispatched to the diesel building had not yet arrived at the diesel and should be considered unavailable to assist in securing the engine.

Conditions of Scenario:

Unit 1 is experiencing a Loss of Offsite Power Condition, and the reactor is shut down. The 1A EDG is out of service for major maintenance, which requires the 1B EDG to be dedicated to Unit 1; therefore, it is unavailable for transfer to Unit 2. The 1C EDG is energizing the 1G 4kV switchgear.

Unit 2 is experiencing a LOSP Condition, and the reactor is shut down. The 2A EDG is energizing the 2E 4kV switchgear without cooling water from the 2A Plant Service Water (PSW) pump because of an unknown failure of the LOSP timer card. The 2C EDG has failed to start and is unavailable for use.

Timeline:

T =0 Initial Condition - Described above T= 2 min RO instructed by SS to secure the 2A EDG due to an assumed high temperature condition coincident with a loss of cooling water (Procedure 34AB-P41-001-2, Section 4.6.3, Figure 1, picture 3(a)).

T= 2 min 54 sec RO obtains the required procedure to secure the diesel from the main control room with a sealed-in start signal (34S0-R43-001-2, Section 7.3.2, Figure 2).

T= 3 min 32 sec RO obtains key to the MCR jumper cabinet (see picture 1).

T= 4 min 06 sec RO obtains jumper from cabinet (see picture 2).

E3-1

Enclosure 3 Simulated Scenarios T= 6 min 26 sec Operator installs the jumper in the back of the MCR panel 2H 11-P652 (see picture 4).

T= 7 min 42 sec RO trips the 2A EDG using the control switch on the front of MCR pane12H11-P652 (see pictures 3, 3(b),

and 3(c)).

Within Enclosure 1,Section III, SNC Risk Assessment, the high temperature alarms are discussed. This discussion makes the point that the time-frame for the first oil temperature alarm is 38.94 minutes, and the time-frame for the first cooling water alarm is 26.4 minutes. Even so, actual test data shows much longer time-frames (up to 270 minutes for the lube oil system and up to 148 minutes for the cooling water system). When conducting the HRA calculation, 35 minutes for operation without cooling water is the time-frame assumed for the operators to secure the 2A EDG prior to damage from lack of cooling water. As stated in the above scenario, the simulation assumes that the first high temperature alarm is received in the control room at the two-minute mark. This scenario was run prior to the receipt of the Fairbanks Morse report (Enclosure 9). In an actual event, if Operations waited for the first high temperature alarm to come in, as calculated by Fairbanks Morse, the 2A EDG would be shut down prior to the 35 minutes assumed for the HRA because Operators required only 5 minutes and 42 seconds in the scenario to shut down the 2A EDG after the high temperature alarm was

=

assumed (26.4 minutes + 5.7 minutes 32.1 minutes).

2. Recovering the 2A PSW Pump The below described scenario was performed by qualified System Engineers in a simulated transient in early January 2010. This exercise was conducted in a format similar to a JPM. A timeline for recovering the 2A PSW pump was established by timing the actions performed by the involved personnel during this exercise.

Conditions of Scenario:

The annunciator for low PSW pressure was received, and it is observed that the 2A and 2C PSW pumps are not running. The 2A EDG has been shut down by Operations because cooling water could not be established. There is an unknown failure of the LOSP timer card associated with the 2A PSW pump. The 1B EDG is dedicated to Unit 1; therefore, it is unavailable for transfer to Unit 2, so it cannot provide an alternate source of cooling water to the 2A EDG.

This condition resulted in placing Unit 2 in an emergency condition with emergency response facilities activating. The Emergency Director gave direction to recover the 2A PSW pump to allow the 2A diesel generator to be restarted. In order to recover the 2A PSW pump, link TB 1-18 (Figures 5 and 6, Picture 6) in the 2E Switchgear Room Panel 2H21-P230 must be opened in order to remove the breaker trip signal. A jumper must then be installed from TB1-15 (4C) to TB1-16 (6C1) in order to restore remote manual breaker E3-2

Enclosure 3 Simulated Scenarios operation from the MCR (Figures 5 and 6, Picture 6). The timeline below demonstrates the time-frame required for Engineering personnel to diagnose the issue and Maintenance personnel to follow instructions given by Engineering.

Timeline:

T=O Initial Condition - Described above T= 60 min Emergency Response Facility Activated (Procedure 73EP-EIP-001-0, Enclosure 7)

T= 2 hr 20 min Actions given to Maintenance to install jumper in 2E Switchgear Room Pane12H21-P230 so Operations can start the 2A PSW pump from the MCR (Figure 4, Figure 5, Figure 6, Picture 6)

T= 2 hr40 min Maintenance Briefed and Dispatched to 2E Switchgear Room Pane12H21-P230 T= 2 hr 55 min Maintenance installs the jumper in Panel 2H21-P230; control switch is moved to the ON position; 2A PSW pump breaker closes and subsequently trips T= 3 hr 3 min Engineering determines the breaker trip needs to be jumpered; gives Maintenance the needed instructions T= 3 hr 23 min Maintenance Briefed and Dispatched to 2E Switchgear Room Panel 2H21-P230 (Figure 4, Figure 5, Figure 6, Picture 6)

T= 3 hr 38 min Maintenance installs another jumper in Panel 2H21-P230; control switch is moved to the ON position; 2A PSW pump breaker closes; pump is available for service E3-3

Enclosure 3 Simulated Scenarios Step 4.6.3 from Procedure 34AB-P41-001-2 Loss of Plant Service Water TRIP Diesel Generator a DIG is running PSW CANNOT be restored Locally IF AND AND AND THEN OR PSW IS LOST EITHER From CIR per 34S0-R43-001-2 Sect. 7.3.2 Lube Oil Temp ~ 220°F OR Jacket Coolant Temp ~ 195°F OR It is evident that continued operation of the DIG will NOT assist recovery actions (4160V Bus damage, PSW still not

~vailable, available, etc) ...... ______

Figure 1 E3-4

Enclosure 3 Simulated Scenarios Section 7.3.2, Diesel Generator 2A (2C) Manual Shutdown, from 34S0-R43-001-2, Diesel Generator Standby AC System 7.3.2.1 At the direction of the SS, if desired to shut down the DG with an auto start signal present, which cannot be reset, THEN install jumpers per Attachment 8.

7.3.2.2 CONFIRM/PLACE the Diesel Gen 2A (2C) Mode Select switch in TEST.

7.3.2.3 Using the Diesel Gen 2A (2C) Speed Adjust switch, DECREASE the load on diesel to between 400 and 500 KW, WHILE maintaining diesel reactive load between 400 and 500 KVAR using Diesel Gen 2A (2C) Voltage Adjust switch.

7.3.2.4 TAKE ACB 135530 (135540), Diesel Gen 2A (2C) Emergency Supply, control switch to TRI P.

7.3.2.5 Confirm Diesel Gen 2A (2C) Emergency Supply breaker green (OPEN) light is ILLUMINATED.

7.3.2.6 If this is an emergency (no cooling water, fire at DG, etc.),

TAKE the Diesel Gen 2A (2C) Start switch to STOP.

Figure 2 E3-5

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Enclosure 3 Simulated Scenarios 1

Cabinet SOS Office Where RO Obtains Jumper Cabinet Picture 1 E3-10

Enclosure 3 Simulated Scenarios 2

Jumper Cabinet Where RO Obtains Jumper Picture 2 E3-11

Enclosure 3 Simulated Scenarios 3

Panel 2H 11-P652 Where RO Observes Condition Annunciator Received at PaneI2H11-P652-1 Picture 3 E3-12

Enclosure 3 Simulated Scenarios 3(a)

Annunciation Received at Panel 2H 11-P652-1 High Annunciator Temperature Received at Alarm T=35 seconds Picture 3(a)

E3-13

Enclosure 3 Simulated Scenarios 3(b)

Procedure 34S0-R43-001-2, Step 7.3.2.2 RO Places 2A EDG in Test Mode EDG Mode Select Switch Picture 3(b)

E3-14

Enclosure 3 Simulated Scenarios 3(c)

Procedure 34S0-R43-001-2, Steps 7.3.2.3,7.3.2.4,7.3.2.5, and 7.3.2.6 7.3.2.4/7.3.2.5 Picture 3(c)

E3-15

Enclosure 3 Simulated Scenarios 4

Back of Panel 2H11-P652 Procedure 34S0-R43-001-2, Step 7.3.2.2 Picture 4 E3-16

Enclosure 3 Simulated Scenarios 5

Panels 2H21-P200 and 2H21-P230 In the 2E Switchgear Room 2H21-P230 Picture 5 2H21-P200 E3-17

Enclosure 3 Simulated Scenarios 6

Back of Panel 2H21-P203 Where Maintenance Installs Jumper

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Edwin I. Hatch Nuclear Plant Enclosure 4 34AB-P41-001-2 Loss of Plant Service Water

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ABNORMAL OPERATING PROCEDURE 1 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. R. Brinson DATE 2-24-10 DATE:

N/A DATE 2-24-10 SSM I PM N/A N/A 1.0 CONDITIONS 1.1 ANNUNCIATORS 650-112 RFP/COND BRG METAL TEMP HIGH 650-165 TURB GEN/CWPS BRG TEMP HIGH 650-210 PSW PUMP 2A OVLD/LOCKOUT RELAY TRIP 650-220 PSW PUMP 2B OVLD/LOCKOUT RELAY TRIP 650-230 PSW PUMP 2C OVLD/LOCKOUT RELAY TRIP 650-240 PSW PUMP 2D OVLD/LOCKOUT RELAY TRIP 650-238 HX PSW/RBCCW DIFF PRESS LOW 650-103 PSW MAIN HDR DIV I PRESS LOW 650-156 PSW MAIN HDR DIV II PRESS LOW 650-137 4160V STA SVC FDR BRKR TRIP 650-215 TURB BLDG PSW FLOW HIGH 650-164 CONDENSER ROOM FLOODING 650-229 PSW INTK STRUCT WATER LEVEL LOW 1.2 Rapidly increasing Drywell temperature AND pressure.

1.3 PSW related temperature control valves indicate increasing equipment temperatures.

1.4 Low OR no PSW system pressure indications.

1.5 PSW system piping failure or leakage observed 1.6 Low OR no PSW flow due to:

  • TWS or PSW strainer blockage (Attachment 4, INDICATIONS OF PSW CLOGGING IN BUILDINGS, may help diagnose possible cause)
  • Trip of PSW pumps
  • Low river level MGR-0002 Rev. 8.1

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 TABLE OF CONTENTS Section 2.0 AUTOMATIC ACTIONS ......................................................................................................... 2 3.0 IMMEDIATE OPERATOR ACTIONS ..................................................................................... 2 4.0 SUBSEQUENT OPERATOR ACTIONS ................................................................................ 3 4.2 Initial Conditions .............................................................................................................. 3 4.13 Loss of Plant Service Water Pumps ................................................................................ 5 4.13.4 Loss Of Plant Service Water Pumps Due To Loading Timer Failure ....................... 5 4.14 Traveling Water Screen Clogging .................................................................................... 9 4.15 Plant Service Water Strainer Clogging ............................................................................ 9 4.16 Plant Service Water Piping Failure .................................................................................. 9 4.17 Removal of Equipment From Service ............................................................................ 11 4.18 Crosstie Cooling Sources .............................................................................................. 12 4.19 Re-opening Turbine Bldg Isolation Valves ..................................................................... 13 4.20 Operating Strategy For Event Mitigation ........................................................................ 15 Attachments 1 EQUIPMENT COOLED BY PSW ........................................................................................ 16 2 ALTERNATE COOLING TO CONDENSATE PUMPS ........................................................ 17 3 ALTERNATE COOLING TO THE MECHANICAL VACUUM PUMP ................................... 19 4 PSW IN BUILDING RESPONSE TO CLOGGING ............................................................... 21 2.0 AUTOMATIC ACTIONS 2.1 PSW Pumps in STANDBY will start on low system header pressure of 95 PSIG.

2.2 Equipment cooled by PSW will trip OR give annunciation on high temperature.

3.0 IMMEDIATE OPERATOR ACTIONS NONE MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 Attempt to start PSW Pumps. D 4.2 Initial Conditions IF EITHER

1. A total loss of PSW exists 1. Enter 34AB-C71-001-2, Scram Procedure, AND SCRAM the reactor
2. Turb Bldg PSW Inlet pressure on 2P41-R607 is less than 30 PSIG 2. Confirm CLOSED/CLOSE 2P41-F316A, 2P41-F316B, 2P41-F316C, 2P41-F316D valves.
3. There is an indication of a major pipe break in the Turbine 3. If either PSW division header pressure remains Building.

Bui/ding. below 50 PSIG, TRIP and PULL TO LOCK the PSW pumps in that division.

(associated green lights will go out)

4. Either PSW division header pressure is less than 50 PSIG
4. Go to section 4.18 AND re-evaluate whether it is permissible to re-open the opposite division isolation valves to establish flow to the Turb. Bldg .
  • As a result of OCR 98-011 & 012, when PSW Division header pressure is below the Auto Start pressure of a PSW pump, and any running PSW pump is placed in PTL, the green light will extinguish (no light indication). The pump breaker will close and pump will restart when the control switch is taken out of PTL to AUTO or RUN.

NOTES:

  • Portions of the Subsequent Operator Actions section of this procedure may be performed out of order based on the Shift Supervisor's prioritization. An example of this would be that the Operating Strategy For Event Mitigation section can be entered and performed concurrently with the other sections of this procedure.

4.3 For a partial loss of PSW, reduce Reactor Power per 34GO-OPS-005-2, Power Changes, to prevent overheating of plant equipment. D 4.4 Perform required actions per applicable Annunciator Response procedures when alarm setpoints are reached. D MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 4 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.5 If river level is low, enter and perform concurrently 34AB-Y22-002-0, Naturally Occurring Phenomenon. o 4.6 IF any diesels are running, perform the following:

4.6.1 Send Operator to monitor jacket water temperature for running diesels. o 4.6.2 Reduce diesel loading as allowed by plant conditions. o I NOTE: Per the FSAR a DG is capable of running fully loaded for> 3 minutes with no cooling W'.:la water.tL°t;r,1. I

.. 3 4.6 463 TRIP Diesel a DIG is running PSW CANNOT be restored Generator Locally IF AND AND AND THEN OR From CIR per PSW IS LOST EITHER 34S0-R43-001-1 Sect. 7.3.2 Lube Oil Temp 2: 220°F OR Jacket Coolant Temp 2: 195°F OR It is evident that continued operation of the DIG will NOT assist recovery actions (4160V Bus damage, PSW still not available etc) 4.7 At Shift Supervisor's direction, remove applicable equipment from service to prevent overheating dependent upon power level and plant status per the Removal of Equipment From Service section of this procedure. o 4.8 Trip AND isolate cooling water to equipment as necessary to reduce heat load on PSW.

Refer to Attachment 1 for a partial list of equipment cooled by PSW. o MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.9 Maintain Drywell pressure by performing the following:

4.9.1 Vent per 34S0-T48-002-2, Containment Atmospheric Control and Dilution Systems. 0 4.9.2 Crosstie Drywell Chilled Water to Rx Bldg Chilled Water per 34S0-P64-001-2. 0 4.10 IF RBCCW suction temperature reaches 105°F, as indicated by RBCCW Pump Suction Temp indicator 2P42-R600, panel 2H11-P650, refer to 34AB-P42-001-2, Loss of Reactor Building Closed Cooling Water. 0 4.11 Refer to 34AB-T41-001-2, Loss Of Area Ventilation System, due to loss of PSW affecting ECCS & MCREC coolers. 0 4.12 Address the remaining sections of this procedure and perform as applicable. 0 4.13 Loss of Plant Service Water Pumps 4.13.1 Confirm the following 4160V busses are energized:

  • 2R22-S007 (Bus 2G)PSW Pump 2P41-C001B o 4.13.2 Contact Electrical Maintenance to assist with checking the PSW Pump breaker(s) for the following trip conditions:
  • load shed 0
  • 86 device lockout 0
  • PSW 20 (2R25-S006) associated BKRs o 4.13.3 IF the breaker trip condition can be corrected, perform the following:

4.13.3.1 RESET the breaker(s) in accordance with 30AC-OPS-003-0, Plant Operations and 31 GO-OPS-021-0, Manipulation Of Controls And Equipment. 0 4.13.3.2 START at least one PSW Pump per division in accordance with 34S0-P41-001-2, Plant Service Water System. 0 4.13.4 Loss Of Plant Service Water Pumps Due To Loading Timer Failure CONTINUOUS I MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0

  • Indications of a loading timer failure:
  • receiving the annunciator "Diesel Gen 2A Loading Timer Failure",

NOTE:

  • Red LED light illuminated on the timer card,
  • counter in the timer panel NOT indicating 60.
  • Ref. drawings: H23609, H23804, SX27174, H23599, SX27173 (sht 3 for paneI2H21-P230) 4.13.4.1 IF 2P41-C001A, PSW Pump, failed to start on LOCA, due to 2R43-N781A, Loading Timer, failure, perform the following:

4.13.4.1.1 If 2E 4160v Bus is energized, manually START 2P41-C001A, PSW Pump, paneI2H11-P651. D 4.13.4.2 IF 2P41-C001A, PSW Pump, failed to start on LOSP, due to 2R43-N782A, Loading Timer, failure, perform the following:

4.13.4.2.1 Place 2P41-C001A in OFF, PULL-TO-LOCK, to prevent the PSW pump breaker from closing on low PSW low pressure condition. D 4.13.4.2.2 In paneI2H21-P230, open link TB2-6 (ETR) to disable ETR-1A Loadshed Relay. D NOTE: The pump breaker will close and pump will restart when the control switch is taken out of OFF, PULL-TO-LOCK to RUN 4.13.4.2.3 If 2E 4160v Bus is energized, confirm Diesel Loading will allow start of 2A PSW pump AND manually START 2P41-C001A, PSW Pump, paneI2H11-P651. D 4.13.4.2.4 If 2E 4160v Bus is NOT energized, Start and tie the 2A Diesel Generator per approved plant procedures (34AB-R43-001-2, DIG Recovery, 34AB-R22-003-2, Station Blackout, 34S0-R43-001-2, DIG Standby AC System). D 4.13.4.2.4.1 Once the 2A Diesel Generator has been started and 2R22-S005, 2E 4160V bus, energized, confirm Diesel Loading will allow start of 2A PSW pump AND manually START 2P41-C001A, PSW Pump, paneI2H11-P651. D MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 CONTINUOUS I

  • Indications of a loading timer failure:
  • receiving the annunciator "Diesel Gen 2C Loading Timer Failure.",

NOTE:

  • Red LED light illuminated on the timer card,
  • counter in the timer panel NOT indicating 60.
  • Ref. drawings: H23609, H23814, SX27243, H23601, SX27242 (sheet 3 for 2H21-P232) 4.13.4.3 IF 2P41-C001B, PSW Pump, failed to start on LOCA, due to 2R43-N781C, Loading Timer, failure, perform the following:

4.13.4.3.1 IF 2G 4160v Bus is energized, manually START 2P41-C001B, PSW Pump, paneI2H11-P651. o 4.13.4.4 IF 2P41-C001B, PSW Pump, failed to start on LOSP, due to 2R43-N782C, Loading Timer, failure, perform the following:

4.13.4.4.1 Place 2P41-C001B in OFF, PULL-TO-LOCK, to prevent the PSW pump breaker from closing on low PSW low pressure condition. 0 4.13.4.4.2 In paneI2H21-P232, open link TB2-6 (ETR) to disable ETR-1 A Loadshed Relay. o NOTE: The pump breaker will close and pump will restart when the control switch is taken out of OFF, PULL-TO-LOCK to RUN 4.13.4.4.3 If2G 4160v Bus is energized, confirm Diesel Loading will allow start of 2B PSW pump AND manually START 2P41-C001B, PSW Pump, paneI2H11-P651. o 4.13.4.4.4 If 2G 4160v Bus is NOT energized, Start and tie the 2C Diesel Generator per approved plant procedures (34AB-R43-001-2, DIG Recovery, 34AB-R22-003-2, Station Blackout, 34S0-R43-001-2, DIG Standby AC System). o 4.13.4.4.4.1 Once the 2C Diesel Generator has been started and 2R22-S007, 2G 4160V bus, energized, confirm Diesel Loading will allow start of 2B PSW pump AND manually START 2P41-C001B, PSW Pump, paneI2H11-P651. o MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 8 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 CONTINUOUS I

  • Indications of a loading timer failure:
  • receiving the annunciator "Diesel Gen 2B Loading Timer Failure.",

NOTE:

  • Red LED light illuminated on the timer card,
  • counter in the timer panel NOT indicating 60.
  • Ref. drawings: H23609, H23776, SX27244, SX27245, H23600 4.13.4.5 IF 2P41-C001C ANDIOR 2P41-C001D, PSW Pump, failed to start on LOCA, due to 2R43-N781 B, Loading Timer, failures, perform the following:

4.13.4.5.1 If 2F 4160v Bus is energized, manually START 2P41-C001C ANDIOR 2P41-C001D, PSW Pump, 2H11-P651. D 4.13.4.6 IF 2P41-C001C ANDIOR 2P41-C001D, PSW Pump, failed to start on LOSP, due to 2R43-N782B, Loading Timer, failure, perform the following:

4.13.4.6.1 Place 2P41-C001C AND/OR 2P41-C001D in OFF, PULL-TO-LOCK, to prevent the PSW pump breaker from closing on low PSW low pressure condition. D 4.13.4.6.2 In paneI2H21-P231, open link TB1-47 (ETR) to disable ETR-4B and ETR-1B, Loadshed Relays. D NOTE: The pump breakers will close and pump will restart when the control switch is taken out of OFF, PULL-TO-LOCK to RUN 4.13.4.6.3 If 2F 4160v Bus is energized, confirm Diesel Loading will allow start of 2C ANDIOR 2D PSW pump AND manually START 2P41-C001C ANDIOR 2P41-C001D, PSW Pump, 2H11-P651. D 4.13.4.6.4 If 2F 4160v Bus is NOT energized, Start and tie the 2B Diesel Generator per approved plant procedures (34AB-R43-001-2, DIG Recovery, 34AB-R22-003-2, Station Blackout, 34S0-R43-001-2, DIG Standby AC System). D 4.13.4.6.4.1 Once the 2B Diesel Generator has been started and 2R22-S006, 2F 4160V bus, is energized, confirm Diesel Loading will allow start of 2C ANDIOR 2D PSW pump AND manually START 2P41-C001C ANDIOR 2P41-C001D, PSW Pump, 2H11-P651.D MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.14 Traveling Water Screen Clogging 4.14.1 Ensure proper operation of the TWS per 34S0-W33-001-0. 0 4.14.2 If necessary, manually clean the screens. 0 4.14.3 If necessary, run the trash rakes. 0 4.14.4 If TWS clogging is of such magnitude that water flow into the intake pit is limited, then evaluate which and how many PSW and RHRSW pumps can be run without draining the intake pit. 0 4.14.5 If Intake Level drops to <60.7 ft, then refer to Tech Spec 3.7.2 and review for possible Emergency Classification per 73EP-EIP-001-0. 0 4.15 Plant Service Water Strainer Clogging 4.15.1 Refer to Attachment 4, PSW IN BUILDING RESPONSE TO CLOGGING. o 4.15.2 Ensure proper operation of the strainers per 34S0-P41-001-2. o 4.15.3 If necessary, manually backwash the strainers. o 4.16 Plant Service Water Piping Failure 4.16.1 Refer to Attachment 4, PSW IN BUILDING RESPONSE TO CLOGGING. 0 4.16.2 Dispatch personnel to localize the break and isolate locally by closing nearest isolation w~. 0 4.16.3 If the break cannot be isolated locally, THEN perform the following as appropriate:

4.16.3.1 If the break is upstream of 2P41-F316A and 2P41-F316C, THEN CLOSE these w~. 0 4.16.3.2 If the break is upstream of 2P41-F316B and 2P41-F316D, THEN CLOSE these w~. 0 4.16.3.3 If the break is in the Turb Bldg, THEN isolate by closing 2P41-F316A, 2P41-F316B, 2P41-F316C, 2P41-F316D. 0 4.16.3.4 If the break is in the EDG Bldg, Div I, THEN isolate by closing 2P41-F312A. 0 4.16.3.5 If the break is in the EDG Bldg, Div II, THEN isolate by closing 2P41-F312B. 0 MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.16.3.6 If the break is in the Rx Bldg, Div I, THEN isolate by closing 2P41-F315A. 0 4.16.3.7 If the break is in the Rx Bldg, Div II, THEN isolate by closing 2P41-F315B. 0 4.16.4 If the break can be subsequently isolated closer to the break, THEN determine if the Division isolation valves can be re-opened to resume cooling to unaffected systems. 0 4.16.5 If the break is on Division 1 and can not be isolated, confirm CLOSED/CLOSE 2P41-F109 AND 2T43-F156, Fire Main Crosstie to Division 1 Valves, at Torus Area Bay 1 Outer Catwalk. 0 4.16.6 If the break is on Division 2 and can not be isolated, confirm CLOSED/CLOSE 2P41-F108 AND 2T43-F157, Fire Main Crosstie to Division 2 Valves, at Torus Area Bay 1 Outer Catwalk. 0 MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.17 Removal of Equipment From Service

  • EQUIPMENT DEEMED ABSOLUTELY NECESSARY FOR MITIGATION OF EMERGENCY CONDITIONS MAY BE OPERATED BEYOND PROCEDURAL LIMITATIONS WITH APPROVAL OF THE SHIFT SUPERVISOR.
  • IN ORDER TO MAINTAIN THE MAIN CONDENSER AS A HEAT SINK, IT IS IMPERATIVE THAT CONDENSER VACUUM BE MAINTAINED. THEREFORE, IT IS IMPORTANT THAT STEAM SEALS, A SJAE (OR THE MVP), AN EHC PUMP, AND A CONDENSATE PUMP REMAIN IN OPERATION.
  • THE EHC SYSTEM MAY BE OPERATED WITHOUT COOLING FOR AN EXTENDED TIME.

CAUTIONS:

  • IF CONDENSATE ANDIOR CONDENSATE BOOSTER PUMPS ARE LEFT IN SERVICE, THE TEMPERATURES OF THE RUNNING PUMPS WILL INCREASE.

PUMPS CAN BE ROTATED WHEN A HIGH BEARING TEMPERATURE ALARM IS RECEIVED ON THE RUNNING PUMP. ALTERNATE COOLING FOR A CONDENSATE PUMP FROM FIRE WATER CAN BE OBTAINED BY USING ATTACHMENT 2.

  • THE MVP CAN BE OPERATED FOR UP TO 2 HOURS WITHOUT PSW COOLING WATER. ALTERNATE COOLING FROM FIRE WATER CAN BE OBTAINED BY USING ATTACHMENT 3.

4.17.1 IfTB Service Water Inlet Press is < 30 PSIG (2P41-R607),

OR 2P41-F316A, 2P41-F316B, 2p41-F316C, and 2P41-F316D, PSW To Turb Bldg valves, are CLOSED, THEN perform the following:

4.17.1.1 With the Shift Supervisor's concurrence, TRIP the following equipment:

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.18 Crosstie Cooling Sources 4.18.1 Perform the following actions to crosstie various cooling water sources:

4.18.1.1 If a pipe break has been isolated by the closure of 2P41-F316A, 2P41-F316B, 2P41-F316C,2P41-F316D, then Rx Bldg Div IfDiv II PSW may be crosstied by OPENING valves:

SUPPLY, THE PUMPS WILL SEIZE DUE TO MOTOR THRUST BEARING FAILURE.

4.18.1.2 IF RHRSW pumps are running, AND the corresponding PSW Division pumps are lost, supply cooling water to the RHRSW pumps from the opposite PSW division, IF available, by opening 2P41-F925 and 2P41-F926 at the Intake Structure. D 4.18.1.3 IF Div I PSW is lost, THEN align screen wash water to the Upstream Traveling Water Screen as follows:

4.18.1.3.1 Confirm CLOSED/CLOSE 2P41-F1500A Unit 2 Div I supply MOV, paneI2H11-P652. D 4.18.1.3.2 Confirm OPEN/OPEN 1P41-F1500B Unit 1 Div II supply MOV, paneI1H11-P652. D 4.18.1.4 IF Div. II PSW is lost, THEN align screen wash water to the Downstream Traveling Water Screen as follows:

4.18.1.4.1 Confirm CLOSED/CLOSE 2P41-F1500B, Unit 2 Div II supply MOV, paneI2H11-P652. D 4.18.1.4.2 Confirm OPEN/OPEN 1P41-F1500A, Unit 1 Div I supply MOV, paneI1H11-P652. D MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 13 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.19 Re-opening Turbine Bldg Isolation Valves 4.19.1 IF 2P41-F316A, 2P41-F316B, 2P41-F316C, 2P41-F316D, PSW To Turb Bldg, are CLOSED AND there is no physical indication of a PSW Line Break DOWNSTREAM of 2P41-F316A, 2P41-F316B, 2P41-F316C, 2P41-F316D, OR when directed by the EOPsto maintain 2P41-F316A, 2P41-F316B, 2P41-F316C, and 2P41-F316D open with no physical indication ofa PSW Line Break, THEN perform the following:

4.19.1.1 !E PSW Division 1 Header Pressure is above 80 PSIG, THEN:

4.19.1.1.1 PLACE the following keylock switches to OVERRIDE:

4.19.1.1.2 OPEN 2P41-F316A or 2P41-F316 C.

Critical 4.19.1.1.3 THROTTLE OPEN 2P41-F316C or 2P41-F316A maintaining Division 1 Header Pressure above 80 PSIG.

4.19.1.2 IF PSW Division 2 Header Pressure is above 80 PSIG, THEN:

4.19.1.2.1 PLACE the following keylock switches to OVERRIDE:

4.19.1.2.2 OPEN 2P41-F316B or 2P41-F316D Critical 4.19.1.2.3 THROTTLE OPEN 2P41-F316D or2P41-F316 B maintaining Division 2 Header Pressure above 80 PSIG.

4.19.1.3 IF 2P41-F316A, 2P41-F316B, 2P41-F316C, or 2P41-F316D must be left partially closed, THEN continue in this procedure.

4.19.1.4 IF all valves 2P41-F316A, 2P41-F316B, 2P41-F316C, 2P41-F316D can be fully opened AND PSW Division 1 AND 2 Header Pressures maintained above 80 PSIG, THEN exit this procedure.

POSTED at 2H 11-P652 Ref. 34AB-P41-001-2 MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 14 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.19.2 IF 2P41-F316A, 2P41-F316B, 2P41-F316C, and 2P41-F316D, PSW To Turb Bldg, are OPEN AND there is no physical indication of a PSW Line Break DOWNSTREAM of 2P41-F316A, 2P41-F316B, 2P41-F316C, and 2P41-F316D, and it is desired to maintain the valves OPEN, THEN perform the following:

4.19.2.1 IF PSW Division 1 Header Pressure is above 80 PSIG, THEN place Override Keylock Switch for 2P41-F316A & 2P41-F316D to OVERRIDE AND confirm 2P41-F316A & 2P41-F316D remain open. 0 4.19.2.2 IF PSW Division 2 Header Pressure is above 80 PSIG, THEN place Override Keylock Switch for 2P41-F316B & 2P41-F316C to OVERRIDE AND confirm 2P41-F316B & 2P41-F316C remain open. 0 MGR-0001 Rev. 4.0

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 15 OF 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

LOSS OF PLANT SERVICE WATER 34AB-P41-001-2 10.0 4.20 Operating Strategy For Event Mitigation 4.20.1 If a significant loss of PSW exists and can not be promptly restored, the following operating strategy will be employed:

4.20.1.1 Maintain the condenser as a heat sink by maintaining condenser vacuum.

This will require Steam Seals and a SJAE with one condensate pump.

  • The SJAE operation will require sufficient reactor pressure to operate.
  • Alternate cooling for the condensate pump from Fire Water can be obtained using Attachment 2.
  • The Mechanical Vacuum Pump can operate up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> without cooling water.

Alternate cooling from fire water can be obtained using Attachment 3.

4.20.1.2 Once the MVP is available, initiate an aggressive cooldown to -150 psig.

This provides several advantages:

  • maintains availability of RCIC
  • allows RPV level control via condensate if needed.

4.20.1.3 IF at any time, the Main Condenser is about to be unavailable as a heat sink, THEN address the RC flowchart actions for anticipating an Emergency Depress and rapidly depressurizing the Reactor to the condenser through the bypass valves using the BPV jack.

4.20.1.4 Eventually as decay heat dies, the BPV's will close and reactor pressure will begin to fall below 150 psig.

4.20.1.5 Once the suppression pool becomes the primary heat sink:

  • The torus could heat up -10 degrees per hour just from reactor decay heat.
  • An increase of up to 35 to 40 degrees F would be seen if the reactor is completely depressurized into the torus from> 900 psig.

4.20.1.6 In consultation with the TSC, proceed to the Severe Accident Management Technical Support Guideline J for additional guidance.

MGR-0001 Rev. 4.0

SNC PLANT E. I. HATCH I Pg 16 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER I 34AB-P41-001-2 10.0 ATTACHMENT ...1. Att. Pg.

TITLE: EQUIPMENT COOLED BY PSW 1 of 1 Division 1 Loads (Upstream of 2P41-F316A and 2P41-F316C)

Diesel Generator A RHR and CS Pump Room Coolers 2T41-B003A and 2T41-B003B (See Note 1)

RHR Pump Coolers A and C Jockey Pump Coolers 2A and 3A

  • RCIC Pump Room Coolers
  • Drywell Chiller Condenser
  • Hot Machine Shop Support Systems
  • Control Room HVAC Units (See Note 1)

Diesel Generator C RHR and CS Pump Room Coolers 2T41-B002A and 2T41-B002B (See Note 1)

RHR Pump Coolers Band D Jockey Pump Coolers 2B and 3B HPCI Pump Room Coolers

  • CRD Pump Room Coolers
  • Control Room HVAC Units (See Note 1)

Vacuum Pump Heat Exchanger Condensate Booster Pump Oil Coolers Condensate Pump Motors Generator Bus Heat Exchanger Hydrogen Coolers Alternator Cooler Turbine Building Chiller Condensers Stator Coolers Waste Gas Treatment Support Systems Radwaste Building Support Systems Reactor Feedpump Turbine Oil Coolers RBCCW Heat Exchangers EHC Coolers Main Turbine Lube Oil Coolers

  • Can be supplied by opposite division Note 1 Refer to 34AB-T41-001-2. Loss of Area Ventilation System.

MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 17 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER ATTACHMENT 2-I 34AB-P41-00 1-2 10.0 Att. Pg.

TITLE: ALTERNATE COOLING TO CONDENSATE PUMPS 1 of 2

  • A condensate pump requires 6 GPM for cooling.
  • The following equipment will be needed to perform this section:
  • 50 ft of fire hose from hose station 2HS-T04 NOTE:
  • One 12" pipe wrench or larger at 130 RB in EOP Gang Box
  • One 1/2" to 11/ 2" pipe to fire hose adapter at 130 RB in EOP Gang Box
  • One 10ft ladder 1.0 To provide alternate cooling (fire water) to the condensate pumps, perform the following:

1.1 CLOSE 2P41-F349. (135' elev. across from Cond Demin Panel near equipment hatch) 1.2 CLOSE 2P41-F350. (135' elev. across from Cond Demin Panel near equipment hatch) 1.3 CLOSE 2P41-F363A, 119' elev. North end 2A Condensate Booster pump.

1.4 CLOSE 2P41-F363B, 119' elev. North end 2B Condensate Booster pump.

1.5 CLOSE 2P41-F363C, 119' elev. North end 2C Condensate Booster pump.

1.6 CLOSE 2P41-F928, 119' elev. at column between Condensate and Cond Booster Pumps.

1.7 REMOVE pipe cap at 2P41-F928.

1.8 INSTALL 1 inch pipe to 1-1/2 fire hose adapter on 2P41-F928.

1.9 CONNECT 1-1/2 fire hose from 2HS-T04 hose station, West of 2C Condensate Pump, to 2P41-F928.

1.10 THROTTLE OPEN 2U43-F009, fire hose station valve, to charge Fire Hose.

The following step will supply cooling to all 3 condensate pumps.

1.11 THROTTLE OPEN 2P41-F928 to provide cooling to Condensate Pumps.

1.12 Notify the Control Room that alternate cooling is in service for the Unit 2 Condensate Pumps.

MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 18 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER I 34AB-P41-001-2 10.0 ATTACHMENT .£ Att. Pg.

TITLE: ALTERNATE COOLING TO CONDENSATE PUMPS 20f2 2.0 When no longer necessary to utilize alternate cooling, perform the following:

2.1 Notify the SS that alternate cooling water is being removed from the Condensate Pumps.

2.2 If directed by the SS, shut down running Condensate Pumps.

2.3 CLOSE 2P41-F928.

2.4 CLOSE 2U43-F009, Fire Hose Station Valve.

2.5 Remove adapter at 2P41-F928.

2.6 Install pipe cap at 2P41-F928.

2.7 OPEN 2P41-F349.

2.8 OPEN 2P41-F350.

2.9 OPEN 2P41-F363A.

2.10 OPEN 2P41-F363B.

2.11 OPEN 2P41-F363C.

MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 19 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER ATTACHMENT ~

I 34AB-P41-001-2 10.0 Att. Pg.

TITLE: ALTERNATE COOLING TO THE MECHANICAL VACUUM PUMP 1 of 2

  • The Mechanical Vacuum Pump requires 125 gpm at <95 F for cooling water flow.
  • The following equipment will be needed to perform this section:
  • Keys to door in locked high rad area at HP Office
  • 150 ft of 11/2" fire hose at 164 U2 RW Fire Equipment Area NOTE:
  • One 2 1/ 2" fire hose to 11/2" fire hose adapter at 130 RB EOP Gang Box
  • One 4 inch flange to 1 1/2" fire hose at 130 RB EOP Gang Box
  • One 8 ft ladder
  • Hose Station, 2HS-T04, located on east wall of stairwell 1.0 To provide alternate cooling (fire water) to the Mechanical Vacuum Pump, perform the following:

1.1 Place the Unit 2 Mechanical Vacuum Pump Control switch to OFF 1.2 CLOSE 2P41-F352. (112 TIB SE at CBWRT (chain operated))

1.3 CLOSE 2P41-F356. (112' elev. in Mechanical Vacuum pump room) 1.4 OPEN 2P41-F391 drain. (147 elev. Cond Bay NE corner on floor) 1.5 CLOSE 2P41-F406. (140' elev. above PSW strainer to condensate pump) 1.6 CLOSE 2P41-F365. (Control Switch at 2H21-P275, 130' elev. across from exciter swgr, HW at 130TFT21 in overhead near exciter swgr) 1.7 CLOSE 2P41-F375. (130' elev. TIB SW above stator exciter swgr) 1.8 CLOSE 2P41-F510. (142' at east end of Stator Cooling skid) 1.9 CLOSE 2P41-F383. (Control Switch at 2H21-P275, 130' elev. across from exciter swgr, HW at 130TGT21 in overhead near Isophase Bus PT's) 1.10 CLOSE 2P41-F377A. (164' elev Turbine Bldg. Chiller) 1.11 CLOSE 2P41-F377B. (164' elev. Left of Turbine Bldg. Chiller 2B) 1.12 REMOVE 4" flange near 2P41-F356 (MVP room) and INSTALL flange to 1-1/2" fire hose adapter.

1.13 INSTALL 2-1/2" to 1-1/2" reducer on 2U43-F012, the 2-1/2 valve at 2HS-T04.

(West of 2C Condensate Pump) 1.14 Install 1-1/2 Fire Hose from 2HS-T04 (West of 2C Condensate Pump) to adapter flange in mechanical vacuum pump room.

MGR-0009 Rev. 5.0

SNC PLANT E. I. HATCH I Pg 20 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER ATTACHMENT ~

I

\ 34AB-P41-00 1-2 10.0 Att. Pg.

TITLE: ALTERNATE COOLING TO THE MECHANICAL VACUUM PUMP 2 of2 1.15 CLOSE 2P41-F391, drain valve, 147' elev. (Cond Bay NE corner on floor) 1.16 OPEN 2P41-F356. (112' elev. Mechanical Vacuum pump room) 1.17 SLOWLY OPEN 2U43-F012, 2-1/2 in. valve at 2HS-T04.

1.18 Notify the Control Room that alternate fire water cooling is in service for the Unit 2 Mechanical Vacuum Pump.

2.0 When alternate cooling water to the MVP is no longer required, perform the following:

2.1 Notify the SS that alternate cooling water is being removed from the MVP.

2.2 If directed by the SS, shut down the MVP.

2.3 CLOSE 2P41-F356.

2.4 CLOSE 2U43-F012.

2.5 Remove adapter at 2P41-F356.

2.6 Install 4 inch flange at 2P41-F356.

2.7 OPEN 2P41-F377A.

2.8 OPEN 2P41-F377B.

2.9 OPEN 2P41-F383.

2.10 OPEN 2P41-F510.

2.11 OPEN 2P41-F375.

2.12 OPEN 2P41-F365.

2.13 OPEN 2P41-F406.

2.14 OPEN 2P41-F356.

2.15 OPEN 2P41-F352.

MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg 21 of 21 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

LOSS OF PLANT SERVICE WATER ATTACHMENT ..!.

I 34AB-P41-001-2 10.0 Att. Pg.

TITLE: PSW IN BUILDING RESPONSE TO CLOGGING 1 of 1 Division PSWBldg Possible Cause Pressure Pressure i  ! Clogged PSW Strainer Reactor Building

!  ! Clogged Trash Rakes

!  ! Clogged Traveling Water Screen Division PSWBldg Possible Cause Pressure Pressure Diesel Generator i  ! Clogged PSW Strainer Building

!  ! Clogged Trash Rakes

!  ! Clogged Traveling Water Screen Division PSWBldg Possible Cause Pressure Pressure i  ! Clogged PSW Strainer Turbine Building

!  ! Clogged Trash Rakes

!  ! Clogged Traveling Water Screen EDG PSW Possible Cause Standby PSW Pressure Pump

! Clogged Standby PSW Strainer MGR-0009 Rev. 5.0

Edwin I. Hatch Nuclear Plant Enclosure 5 34S0-R43-001-2 Diesel Generator Standby AC System

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH SYSTEM OPERATING PROCEDURE 1 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR G. L. Johnson DATE 11/10/99 DATE:

N/A SSM / PM 2-10-09 N/A DATE N/A 1.0 OBJECTIVE This procedure provides instructions for manual startup, shutdown, and operation of the Diesel Generator Standby AC System and Diesel Generator Auxiliary Systems.

NOTE: A partial listing of the TABLE OF CONTENTS [sections 7.2.2, 7.3.1, 7.4.3, & 7.4.4, and their respective titles] is posted at 2H11-P651 & 2H11-P652.

TABLE OF CONTENTS Section 2.0 APPLiCABILITy ..................................................................................................................... 2

3.0 REFERENCES

......................................................................................................................2 4.0 REQUiREMENTS .................................................................................................................. 3 5.0 PRECAUTIONS/LIMITATIONS ............................................................................................. 4 6.0 PREREQUiSiTES .................................................................................................................. 8 7.0 PROCEDURE ........................................................................................................................ 9 7.1 STANDBy ........................................................................................................................ 9 7.2 DIESEL GENERATOR STARTUP ................................................................................ 12 7.2.1 REMOTE/LOCAL MANUAL DIESEL STARTUP .................................................... 12 7.2.2 SYNCHRONIZING DIESEL GENERATOR TO AN ENERGIZED BUS ................. 17 7.2.3 REMOTE MANUAL TRANSFER OF DIESEL GENERATOR TO DE-ENERGIZED BUS ............................................................................................ 19 7.3 SHUTDOWN .................................................................................................................. 23 7.3.1 TRANSFERRING POWER FROM DIESEL GENERATOR 2A (2C) TO NORMAL OR ALTERNATE POWER ..................................................................... 23 7.3.2 DIESEL GENERATOR 2A (2C) MANUAL SHUTDOWN ....................................... 25 7.3.2.24 BARRING OF 2A12C DIESEL GENERATOR ............................................. 27 7.3.3 DIESEL GENERATOR B MANUAL SHUTDOWN FROM UNIT 2 ......................... 28 7.4 INFREQUENT OPERATIONS ....................................................................................... 29 7.4.1 TRANSFER OF FUEL OIL BETWEEN STORAGE TANKS ................................... 29 7.4.2 MANUAL TRANSFER OF FUEL OIL FROM MAIN STORAGE TANKS TO DAY TANKS ........................................................................................................... 34 7.4.3 TRANSFERRING DIESEL GENERATOR 1B FROM BUS 1F TO BUS 2F ........... 38 7.4.4 TRANSFERRING 4160V BUS 2F FROM 1B DIESEL TO NORMAL OR ALTERNATE POWER ............................................................................................ 42 7.4.5 DETERMINING EDG LUBE OIL FILTER OR STRAINER DIFFERENTIAL PRESSURE(DP) .................................................................................................... 44 MGR-0002 Ver. 8.1

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 2 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 Attachments 1 DIESEL GENERATOR STANDBY AC SYSTEM RESTORATION ..................................... .47 2 DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP .......................... .49 3 DIESEL GENERATOR 2A AND 2C VALVE LINEUP .......................................................... 62 4 DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP .................................. 70 5 DIESEL GENERATOR FUEL OIL SYSTEM RESTORATION ............................................. 76 6 DIESEL GENERATOR FUEL QUANTITY CHECK ............................................................. 80 7 DG LOSP LOCA LOGIC DIAGRAM .................................................................................... 85 8 JUMPER INSTALLATION FOR DG SHUTDOWN WITH AUTOSTART SIGNAL PRESENT .............................................................................................................. 86 9

SUMMARY

OF SPEED & VOLTAGE ADJUSTMENTS ...................................................... 87 10 REPLACEMENT OF MAIN STORAGE TANK(S) PIPE CAP AFTER FUEL OIL TRANSFER ........................................................................................................ 88 2.0 APPLICABILITY This procedure applies to the diesel generator standby AC system and the auxiliary systems which support the emergency diesel generators.

3.0 REFERENCES

3.1 FSAR, Unit 2, Sections 8.3.1.1.3, Standby AC Power; 9.5.4, Diesel Generator Fuel Oil Storage and Transfer System; 9.5.5, Diesel Generator Cooling Water System; 9.5.6, Diesel Generator Starting System; and 9.5.7, DIG Lubrication System 3.2 Technical Specifications, Unit:2, Sections 3.8.1, A.C. Sources: Operating and 3.8.2, A.C. Sources - Shutdown 3.3 H-11037, P&ID - Fuel Oil Diesel Oil System 3.4 H-11038, P&ID Demineralized Water 3.5 H-13413, Elementary Diagram Diesel Generator 1B 3.6 H-13587 and H-13588, Contact tabulation Sheet 1 and 2 of 2, Diesel Generators Controls 3.7 H-14239, Elementary Diagram Diesel Building 4160 & 600V Emerg. Sta. Ser. FOR's 3.8 H-21074, Diesel Engine & Fuel Oil Sys. P&ID 3.9 H-23352, Single Line Diagram Emergency Station Service MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 3.10 H-23371, Single Line Diagram 125V DC Emergency Station Service System 2R42B and 2R43B, Sheet 1 MPL's 2R25-S004 and 2R25-S005 and Sheet 2 MPL 2R25-S006 3.11 H-23587 and H-23588, Contact Tabulation Sht 1 and 2 of 2, Diesel Gen. Controls 2R43 Sys.

3.12 H-23669 and H-23670, Emergency Sta Ser Sys 2R20L & 2R20M Elem. Diag., Shts 2 & 3 of 3 3.13 H-23773 thru H-23779, Elementary Diag. - 2R43B Diesel Generator 2B, Shs. 1 thru 7 of 7 3.14 H-23801 thru H-23807, Elementary Diag. - 2R43A Diesel Generator 2A, Shs. 1 thru 7 of 7 3.15 H-23811 thru H-23817, Elementary Diag. - 2R43C Diesel Generator 2C, Shs. 1 thru 7 of 7 3.16 S-13638, DSL Gen - Schem - Jacket Coolant Sys 3.17 S-13639, DSL Gen - Schem Lube Oil Sys 3.18 S-13640, DSL Gen - Schem - Fuel Oil Sys 3.19 S-13641, DSL Gen - Schem - Air Coolant Sys 3.20 SX-28733, Service Manual - Model 3800TD Diesel Engine & Generator 3.21 RER C060097901 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS The number AND qualification level of Operations personnel performing the sections in this procedure will be determined by the Unit Shift Supervisor.

4.2 MATERIAL AND EQUIPMENT 4.2.1 Dipstick 4.2.2 Tank access keys 4.2.3 Phone headset AND phone extension cord or equivalent, OR radio lAW 20AC-IRS-001-0.

4.3 SPECIAL REQUIREMENTS Independent verification as described in 10AC-MGR-019-0, Procedure Use and Adherence, is required WHEN returning the Diesel Generator Standby AC System to service following maintenance OR testing.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40F 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS 5.1.1 Observe safety rules in the Southern Nuclear Safety and Health Manual and PPE requirements as provided in NMP-SH-003, Electrical Work Practices.

5.1.2 Operating the diesel at near zero load OR zero output amperage may result in a diesel trip due to reverse power.

5.1.3 Operation of the diesel at low loads for extended periods of time may result in oil accumUlation in the exhaust manifold due to insufficient gas flows and temperature to vaporize the oil.

5.1.4 It is the responsibility of the operator at the diesel generator to ensure safe operation of the diesel. Investigate any abnormal operating conditions such as smoke, fire, oil leaks, OR fuel oil leaks. Special attention is mandatory to any change in the sound of the diesel, particularly the beginning of a tapping OR knocking sound. IF conditions warrant, the local operator will NOT hesitate to shut down the diesel and notify the Control Room. If the lube oil strainer DP exceeds 5 PSID, the operator wi" notify Maintenance immediately.

5.1.5 The local diesel operator wi" be alert for a day tank vent overflow spill OR a day tank high level alarm. IF a high level alarm OR an oil spill occurs, both transfer pump control switches wi" be positioned to OFF. Immediate measures wi" be taken to prevent fuel oil from entering the Diesel Building drains.

5.1.6 WHEN entering the accesses to the Diesel Fuel Oil Storage Tanks, the guidelines for working in confined spaces found in NMP-SH-005, Confine Space Procedure must be observed.

5.1.7 Due to limitations of the level instrumentation for fuel oil storage tanks 1A, 1B, and 1C, a check of the fuel oil level with the measuring stick is required prior to addition of fuel oil to these tanks AND whenever decreasing tank level approaches the Administrative limit of 35,000 gallons. These actions wi" be required until the existing level transmitters are replaced with instruments not sensitive to the variations in specific gravity of the received fuel oil.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 50F 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 5.2 LIMITATIONS 5.2.1 Voltage must NOT exceed 4400 volts on any diesel generator phase.

5.2.2 The diesel generator will be run for one (1) hour after cranking for any reason, except WHEN requested to satisfy a vendor OR due to equipment malfunction.

5.2.3 A LOCA OR LOSP signal will deenergize the diesel generator Test Relays.

5.2.4 Voltage must NOT exceed 605 volts on any phase on 600V Bus 2C or 2D.

5.2.5 Diesel Generator frequency must be maintained between 59 and 61 Hertz.

5.2.6 IF EITHER of the following conditions cannot be met, the Diesel Generator must be considered INOPERABLE:

  • IF three or more main louver sections are failed closed OR more than 50% of the louver area is failed closed or blocked. This does not apply to automatic cycling of the louvers.
  • IF the double door cannot be blocked open, with room temperature> 40° F during the diesel(s) run, and with an assigned fire watch.

(Engineering Evaluation Number 968) 5.2.7 If a diesel generator trips, the cooling water outlet valve will remain OPEN until the start/stop switch on 2H11-P652 is taken to the STOP position.

5.2.8 The following conditions will trip a diesel generator:

PARAMETER SETPOINT Differential Current N/A Reverse Current N/A*

Lube Oil Pressure Low 21 PSIG Lube Oil Temperature High 230°F*

Jacket Coolant Temperature High 205°F*

Jacket Coolant Pressure Low 9 PSIG*

Crankcase Pressure High 0.5 inches H20*

Engine Overspeed 1000 RPM Start Failure < 250 RPM and < 6 psig oil pressure, 7 seconds after diesel is started

  • These trips are only applicable in the TEST mode MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 60F 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 5.2.9 Energizing a diesel generator's Test Relays results in the following:

  • Locks out associated diesel generator emergency start
  • Prevents AUTO closure of associated diesel generator output breaker
  • Allows paralleling of associated diesel generator and either its normal or alternate power supply
  • Prevents MANUAL closure of Start-up Transformer supply breakers to 4160V Busses 2A, 2B, 2C, and 2D. (AUTO fast transfer will still occur)
  • Prevents AUTO transfer of associated emergency 4160V bus to its alternate supply
  • Arms additional diesel generator trips as stated in 5.2.8.

5.2.10 An Emergency Diesel Generator can be considered operable provided:

  • Its Jacket Cooling Water temperature is at or above 80'F (with intake air temperature at or above 40°F) and Lube Oil Temperature has to be at or above 90' F
  • IF intake air (outside air) temperature decreases below 40°F, jacket coolant temperature must be maintained greater than or equal to 90°F for DG operability.

5.2.11 During receipt of diesel fuel oil, Operations will normally offload fuel oil into storage tanks 2A or 2C due to their more accurate level indication.

IF addition to the 1A, 1B, or 1C fuel oil storage tanks is required by the Shift Supervisor, the respective tank level will be determined with the appropriate measuring stick prior to addition. These actions will minimize the possibility of tank overflow.

5.2.12 During conditions of decreasing tank level for 1A, 1B, or 1C fuel oil storage tanks due to diesel operation or fuel oil transfers, tank level will be taken with the appropriate measuring stick when approaching the Administrative limit of 35,000 gallons in order to confirm the accuracy of the level indicators.

5.2.13 The following is a list of normal operating ranges for the DG parameters.

PARAMETER OPERAliNG RANGE 2R43-R018A(C), Coolant Pump Discharge Pressure ~ 30 PSIG 2R43-R001A(C), Fuel Oil Pressure to Filter (Black Hand) 20 to 35 PSIG 2R43-R001A(C), F.O. Pressure Filter to Engine (Red Hand) 20 to 35 PSIG 2R43-R010A(C), Raw Water Pressure ~20 PSIG 2R43-R015A(C), Crankcase Pressure > 1.5 in vacuum 2R43-R006A(C), Scavenging Air Pressure 0.0 - 15.0 PSIG 2R43-R016A(C), Lube Oil Pressure 30 to 40 PSIG 2R43-R017A(C), Jacket Coolant Temperature < 185°F 2R43-R011A(C), Lube Oil Temp From Engine < 215°F 2R43-R024A(C), Engine RPM (900 is normal) 810 - 1000 RPM MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 70F 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 Plant Hatch administrative limit is a total fuel supply of 175,000 gallons (T.S. 166,600 gallons) for the Emergency Diesel Generators. This volume is administratively controlled by maintaining a NOTE: fuel supply of 35,000 gallons (TS =:: 33,320 gallons) in the 1A, 1B, 1C, 2A, and 2C Diesel Generator Fuel Oil Storage Tanks. Reducing the volume below the Admin limit is not permitted for the receipt of fuel oil.

5.2.14 A minimum of 35,000 gallons of fuel oil in each Diesel Fuel Oil Storage Tank.

5.2.15 There shall be a minimum of 175,000 gallons of acceptable diesel fuel in the Unit 1 and Unit 2 Diesel Fuel Oil Storage Tanks.

5.2.16 Anytime a combination of Diesel Generators is operated> 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, Environmental Affairs must be notified. This requirement is to ensure compliance with EPA Regulations. Notification is to be made by calling the Environmental Affairs On Call Pager Number, 1-800-522-2246 and when asked, give Pin Number 0444071.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 80F 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 6.0 PREREQUISITES 6.1 AC electrical power available 6.2 DC electrical power available NOTE: System valve, electrical, and instrument valve lineups provided on the attachments are to be completed as required by 34GO-OPS-003-2, Startup System Status Checklist.

6.3 Diesel Generators 2A and 2C electrical lineup is complete per Attachment 2 of this procedure.

6.4 Diesel Generators 2A and 2C valve lineup is complete per Attachment 3 of this procedure.

6.5 Diesel Generators 2A and 2C instrument valve lineup is complete per Attachment 4 of this procedure.

6.6 Diesel Generator 1B valve lineup is complete per Attachment 5 of 34S0-R43-001-1, Diesel Generator Standby AC System (Unit 1).

6.7 Diesel Generator 1B instrument valve lineup is complete per Attachment 6 of 34S0-R43-001-1, Diesel Generator Standby AC System (Unit 1).

The following step provides the SM the ability to allow running a diesel for testing NOTE: following such things as maintenance. If the diesel is being run from the opposite division of PSW the diesel CAN NOT be considered operable.

6.8 Plant Service Water System is in operation per 34S0-P41-001-2, Plant Service Water System as required to support the applicable diesel as determined by the SM.

6.9 Service and Instrument Air are available per 34S0-P51-002-2, Instrument and Service Air System.

6.10 Demineralized Water Transfer System operating per 34S0-P21-002-2, Demineralizer Water Transfer System.

6.11 The area in the vicinity of the diesel generators has no evidence of fluid leaks.

6.12 Diesel Building Ventilation is operating per 34S0-X41-001-2, Diesel Generator Building Ventilation System.

6.13 Diesel Generator Building C02 System is operable per 34S0-X43-005-0, Diesel Generator Building Carbon Dioxide System.

6.14 Diesel generator air intake filters are cleaned AND installed.

6.15 Diesel Generator 125V DC Battery System is operable.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 9 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.0 PROCEDURE 7.1 STANDBY CONTINUOUS I WHEN performing this subsection to place Diesel Generator Standby AC System in NOTE: Standby following maintenance, testing OR operation, Attachment 1 is used to document independent verification.

7.1.1 Confirm the Diesel Fuel Oil System is in a Standby lineup by completing Attachment 5.

  • Placing all three Diesel Generators in Standby requires performing the following steps three times (Attachment 5 only performed once).

The following step is only completed WHEN placing Diesel Generator 2A OR 2C in NOTES:

  • Standby. EITHER of the valves listed confirms that the fuses in 2R25-S025 Frame 1A AND 2R20M-P001 are intact (required for Diesel Generator 2A). EITHER of the valves in parentheses indicates that the fuses in 2R25-S027 Frame 1A AND 2R20M-P002 are intact (required for Diesel 2C).

7.1.2 Confirm that either 2P41-F316A (2P41-F316C) OR 2P41-F316D (2P41-F316B), PSW To Turb Bldg Div I OR Div II, are energized [red OR green light ILLUMINATED].

7.1.3 Confirm the Diesel Gen 2A (2C. B) Mode Select switch is in NORM.

7.1.4 Confirm Diesel Gen 2A (2C. B) Diesel Test SAT 2C Out Of Svc Interlock switch is in NORM.

7.1.5 Confirm Diesel Gen 2A (2C. B) Voltage Reg Transfer switch is in AUTO.

7.1.6 Confirm the Diesel Gen 2A (2C. B) Shutdown System Operative red light is EXTINGUISHED.

7.1.7 Confirm the Diesel Gen 2A (2C. B) Start red light is EXTINGUISHED.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 10 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 There is a 100 second time delay from the time the Shutdown Relay pushbutton is NOTE: depressed UNTIL the Diesel Gen 2A (2C, 18) Auto Start Sys Operative light is ILLUMINATED.

7.1.8 IF Diesel Gen 2A (2C, B) Auto Start Sys Operative light is EXTINGUISHED, DEPRESS Diesel Gen 2A (2C, B) Shutdown Relay pushbutton AND confirm the Diesel Gen 2A (2C, B) Auto Start Sys Operative clear light is ILLUMINATED.

7.1.9 IF Diesel Gen 2A (2C, B) has been run, fuel oil has been transferred OR at the discretion of the Shift Supervisor, confirm Diesel2A (2C, B) Fuel Oil Storage Tanks (Main AND Day Tanks) inventories are correct per 34SV-R43-001 (003,002)-2, Diesel Generator 2A (2C, B) Monthly Test.

7.1.10 Confirm the following Diesel Air Compressor control switches are in AUTO:

CONTROL SWITCH DESCRIPTION

7.1.12 Locally, confirm Diesel Generator 2A (2C, 1B) governor lube oil sight gauge indicates approximately 50% in the sight glass.

7.1.13 Locally at the Diesel Generator 2A (2C, 1B) Woodward Governor, confirm the following:

  • Speed Droop control knob is set at 0 (fully counter-clockwise).
  • Load Limit control knob is set at 10 (fully clockwise).

INOTE: The following step only applies WHEN placing Diesel 2A OR 2C in Standby.

7.1.14 Confirm Diesel Generator 2A (2C) local mode selector switch is in REMOTE, panel 2R43-P003A(C).

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 11 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 NOTE: The following step only applies WHEN placing Diesel 1B in Standby.

7.1.15 Confirm Diesel Generator 1B Keylock control switch is in REMOTE UNIT 1 OR REMOTE UNIT 2 at panel 2R43-M01, as directed by the Shift Supervisor.

7.1.16 Locally, confirm Diesel Generator 2A (2C, 1B) Jacket Coolant expansion tank indicates.::: 3/4 full.

NOTE: Mark on sight glass indicates correct oil level for non-operating generator add oil if below mark.

OIL LEVEL READINGS WILL BE TAKEN AT EYE LEVEL WITH SIGHT GLASS.

READINGS TAKEN AT AN ANGLE COULD GIVE AN INACCURATE LEVEL NOTES:

INDICATION.

Posted locally at Diesel Generator 2A, 2C 7.1.17 Locally, confirm Diesel Generator 2A (2C, B) front AND rear bearing lube oil sight gauges indicate level at the proper level mark.

7.1.18 Locally, confirm t Diesel Generator 2A (2C, 1B) lube oil level (dipstick) is correct.

7.1.19 Confirm Battery Charger 2G, 2J, 1H, (2H, 1N, 2N) is in service as follows:

7.1.19.1 AC Input Switch is ON 7.1.19.2 DC Output Switch is ON 7.1.19.3 Throwover switches are in NORMAL UP position 7.1.19.4 Battery charger output voltage is between 125V and 140V DC.

7.1.20 IF Diesel Generator 2A (2C, B) has been run, bar the diesel per appropriate procedure subsection.

7.1.21 IF Diesel Generator 2A (2C, B) was barred, THEN independently verify that 2R43-F181A(2R43-F181C) or 1R43-F181 B is OPEN.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 12 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2 DIESEL GENERATOR STARTUP 7.2.1 REMOTE/LOCAL MANUAL DIESEL STARTUP CONTINUOUS I

  • All operations in this subsection are performed from pane12H11-P652 UNLESS otherwise specified. This procedure is designed to start up Diesel Generator 2A from EITHER the Main Control Room OR locally. To start up Diesel 2C, use this subsection AND substitute the component designations in parentheses. Start up NOTES: Diesel Generator 1B per 34S0-R43-001-1, Diesel Generator Standby AC System.
  • Plant Hatch has committed to maintain a total fuel supply of 175,000 gallons for the Emergency Diesel Generators. This volume is administratively controlled by maintaining a total fuel supply of 35,000 gallons in each of the Diesel Generator fuel oil storage tanks.

DIESEL FUEL OIL INVENTORIES MUST BE MAINTAINED AT OR ABOVE CAUTION: LEVELS SPECIFIED IN UNIT 2 TECHNICAL SPECIFICATIONS, SECTION 3.8.3.

7.2.1.1 Confirm the Diesel Gen 2A (2C) Shutdown System Operative red light is EXTINGUISHED.

7.2.1.2 Confirm the Diesel Gen 2A (2C) Start red light is EXTINGUISHED.

7.2.1.3 At the Diesel Gen 2A (2C) Voltage Reg Transfer switch, confirm the following:

  • Voltage Reg Transfer switch is in AUTO
  • AUTO red light is ILLUMINATED
  • MANUAL green light is EXTINGUISHED.

7.2.1.4 At Diesel Gen 2A (2C) Voltage Adjust switch, confirm the following:

  • RAISE red light is EXTINGUISHED
  • LOWER green light is EXTINGUISHED.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 13 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 There is a 100 second time delay from the time the Diesel Gen 2A Shutdown Relay NOTE: pushbutton is depressed UNTIL the Diesel Gen 2A (2C) Auto Start Sys Operative light is ILLUMINATED.

7.2.1.5 IF Diesel Gen 2A (2C) Auto Start Sys Operative light is EXTINGUISHED, DEPRESS Diesel Gen 2A (2C) Shutdown Relay pushbutton.

7.2.1.6 Confirm the Diesel Gen 2A (2C) Auto Start Sys Operative clear light is ILLUMINATED.

7.2.1.7 Confirm ACB 135530 (135540), Diesel Gen 2A (2C) Emergency Supply, indicates OPEN.

  • WITH THE DIESEL GENERATOR IN THE TEST MODE, THE EMERGENCY BUS WILL NOT AUTO TRANSFER TO THE DG OR THE ALTERNATE SUPPLY (2C SAT) UPON LOSS OF THE NORMAL SUPPLY (2D SAT). THE DG MODE SWITCH MUST BE PLACED IN NORM. TAKING THE DG OUT OF THE TEST MODE WILL CAUSE THE OUTPUT BRKR TO OPEN, !E IT WAS CLOSED.

CAUTION:

  • THE FOLLOWING CONDITIONS WILL TAKE A DG OUT OF THE TEST MODE:

(1) LOCA (2) LOSS OF 2C SAT (3) LOSS OF 2D SAT!E SAT IS DE-ENERGIZED WITH THE SAT 2C OUT OF SVC INTERLOCK SWITCH IN THE MAINTENANCE POSITION.

7.2.1.8 IF SUT 2C is de-energized, PLACE the Diesel Gen 2A (2C) Test SAT 2C Out of SVC Interlock switch in TEST.

7.2.1.9 PLACE the Diesel Gen 2A (2C) Mode Select switch in the TEST position.

7.2.1.10 Confirm annunciator 652-105(305), DIESEL 2A (2C) IN TEST MODE, ALARMS.

7.2.1.11 Establish communications between operator at the diesel generator AND the Main Control Room.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 14 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2.1.12 In Diesel Generator 2A (2C) Room, perform the following:

7.2.1.12.1 IF starting diesel generator from Control Room, confirm the AT ENGINE-REMOTE control switch is in the REMOTE position.

7.2.1.12.2 IF starting diesel generator from panel 2R43-P003A(C),

PLACE the AT ENGINE-REMOTE control switch in the AT-ENG position.

7.2.1.12.3 Confirm the Diesel Gen 2A (2C) Auto Start Sys Operative clear light is EXTINGUISHED.

CAUTION: IF LOSP AND / OR LOCA OCCURS, THEN SPEED DROOP CONTROL KNOB MUST BE POSITIONED TO ZERO (FULLY COUNTERCLOCKWISE) 7.2.1.12.4 At Diesel Generator 2A (2C) Woodward Governor Control, POSITION the Speed Droop control knob to 50.

7.2.1.12.5 At Diesel Generator 2A (2C) Woodward Governor Control, confirm the Load Limit control knob is at 10 (fully clockwise).

IF diesel lube oil level is NOT within W' of the upper FULL mark in the following step, the surveillance may continue provided the oil level is between the lower and upper NOTE:

FULL marks AND Maintenance is notified to add oil following the surveillance to bring the level to within W' of the upper FULL mark.

7.2.1.12.6 Confirm the diesel lube oil level is within W' below the upper FULL mark on the dipstick.

7.2.1.12.7 OPERATE the fuel oil hand pump UNTIL 2R43-R001A(C), Fuel Oil Pressure gauge, momentarily indicates> 15 PSIG (red hand pointer), on panel 2R43-P003A(C).

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 15 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 Prelubing must be performed WITHIN 15 minutes of starting the diesel engine. Delays NOTE: between prelubing AND engine starting will negate the effects of prelubing due to system drain down. The Prelube Pump will operate for approximately 3 minutes WHEN started.

7.2.1.13 lE starting the diesel from panel 2R43-P003A (C), perform the following:

7.2.1.13.1 TAKE the Pre Lube Pump switch to ON AND IF no emergency exists, prelube the diesel generator for three (3) minutes, panel 2R43-P003A(2R43-P003C).

7.2.1.13.2 DEPRESS AND HOLD the Engine START pushbutton UNTIL the engine STARTS at pane12R43-P003A (2R43-P003C).

7.2.1.14 IF starting the diesel from the Control Room, perform the following:

7.2.1.14.1 TAKE Diesel Gen 2A (2C) Prelube Pump switch to ON AND IF no emergency exists, prelube the diesel generator for three (3) minutes, paneI2H11-P652.

7.2.1.14.2 TAKE the Diesel Gen 2A (2C) Start switch to START position.

7.2.1.15 Confirm Diesel Generator 2A (2C) reaches synchronous speed (60 Hertz).

7.2.1.16 Locally, confirm 2P41-F339A(B), Generator 2A (2C) Cooling Water Outlet AOV, OPENS.

7.2.1.17 Confirm the Diesel2A (2C) Start red light is ILLUMINATED.

(810 RPM or 21 PSIG oil pressure) 7.2.1.18 Confirm the Diesel Gen 2A (2C) Shutdown Sys Operative light is ILLUMINATED.

(approximately 20 second time delay).

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 16 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.2.1.19 After the diesel generator has been running for at least 15 seconds.

confirm the following parameters are WITHIN their specified ranges to confirm proper operation on panel 2R43-P003A(2R43-P003C): (can confirm in any order)

PARAMETER OPERATING RANGE

PLACE the AT ENGINE-REMOTE control switch in REMOTE to flash the generator field.

7.2.1.21 Confirm the diesel generator output voltage is 4160V AC.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 17 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2.2 SYNCHRONIZING DIESEL GENERATOR TO AN ENERGIZED BUS CONTINUOUS I 7.2.2.1 After starting up the diesel generator, PLACE the Diesel Gen 2A (2C) Voltage Reg Transfer switch in MANUAL.

7.2.2.2 Confirm the Diesel Gen 2A (2C) Voltage Reg Transfer Auto red light is EXTINGUISHED.

7.2.2.3 Confirm the Diesel Gen 2A (2C) Voltage Reg Transfer Manual green light is ILLUMINATED.

IIIIIIi ~:AY.!!llQ{)t!N:

1111 CAUTION: DO NOT EXCEED 4400 VOLTS ON ANY PHASE OF THE DIESEL GENERATOR. GENERA I UK.

7.2.2.4 ADJUST the Diesel Gen 2A (2C) Voltage Adjust switch UNTIL diesel output voltage is equal to 4160V Bus 2E (2G) voltage.

7.2.2.5 PLACE Diesel Gen 2A (2C) Synch Switch (SSW) for ACB 135530 (135540), in ON.

7.2.2.6 Using Diesel Gen 2A (2C) Speed Adjust switch, adjust diesel speed to attain a slow synchroscope rotation in the clockwise (fast) direction (1 to 3 RPM).

7.2.2.7 With Diesel Gen 2A (2C) Voltage Adjust switch, ADJUST diesel output voltage to match the highest phase voltage on 4160V Bus 2E (4160V Bus 2G).

CRITICAL 7.2.2.8 WHEN the synchroscope indicates 2 minutes to 12 AND WHEN the synchroscope lights approach the dimmest point, I

TAKE Diesel Gen 2A (2C) Emergency Supply ACB 135530 (135540),

to the CLOSE position.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 18 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7

  • LIMIT TOTAL LOADING OF UNIT ONE AND UNIT TWO DIESELS TO 13,000 KW UNLESS NECESSARY TO PROTECT THE CORE OR CONTAINMENT OR TO PREVENT OFFSITE RADIOLOGICAL RELEASES. THIS WILL ENSURE AN ADEQUATE SUPPLY OF FUEL OIL FOR THE 7 DAY PERIOD FOLLOWING AN LOSP. CONTINUE LIMITING TOTAL LOAD UNTIL AN CAUTIONS:

CAUTIONS: ADDITIONAL MINIMUM OF 41,000 GALLONS OF FUEL OIL HAS BEEN ADDED TO THE STORAGE TANKS.

GENERATOR LOADING MUST NOT EXCEED THE FOLLOWING

  • DIESEL GENERATOR RATINGS TO AVOID DIESEL OVERLOAD:

CONTINUOUS 2850 KW HOUR}

12.5 DAY (300 HOUR) 3250 KW 7.2.2.9 Using the Diesel Gen 2A (2C) Speed Adjust switch, INCREASE the load on diesel to 500 KW.

7.2.2.10 Using the Diesel Gen 2A (2C) Voltage Adjust switch, ADJUST the reactive load to approximately 500 to 1000 KVAR.

7.2.2.11 IF no emergency exists, maintain diesel load at 500 KW UNTIL exhaust temperatures stabilize.

7.2.2.12 Increase load to desired load using Diesel Gen 2A (2C) Speed Adjust switch.

7.2.2.13 MAINTAIN the reactive load to between 200 and 1000 KVAR.

7.2.2.14 PLACE Diesel Gen 2A (2C) Synch Switch (SSW) for ACB 135530 (135540) in OFF.

7.2.2.15 Monitor running diesel generator oil levels AND make arrangements in advance to add oil, IF needed.

7.2.2.16 Monitor running diesel generator fuel quantity AND make arrangements in advance to transfer ANDIOR add oil, IF needed.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 19 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2.3 REMOTE MANUAL TRANSFER OF DIESEL GENERATOR TO DE-ENERGIZED BUS CONTINUOUS I The diesel generator output breaker closes automatically WHEN diesel generator NOTE: output voltage AND frequency are normal, the associated 4160V normal AND alternate supply breakers are open AND the bus undervoltage trip device is tripped.

7.2.3.1 After starting up the diesel generator, confirm the Diesel Gen 2A (2C) Voltage Reg Transfer Auto red light is ILLUMINATED.

7.2.3.2 PLACE Diesel Gen 2A (2C) Mode Select switch in NORM.

7.2.3.3 Confirm the Diesel Gen 2A (2C) Voltage Reg Transfer Manual green light is EXTINGUISHED.

7.2.3.4 Using Diesel Gen 2A (2C) Speed Adjust switch, ADJUST diesel generator frequency to 60 Hertz.

CAUTION: DO NOT EXCEED 4400 VOLTS ON ANY PHASE OF THE DIESEL GENERATOR.

7.2.3.5 IF ACB 135530 (135540), Diesel Gen 2A (2C) Emergency Supply, fails to close automatically, perform the following:

7.2.3.5.1 Confirm the normal and alternate supply breakers to 4160V Bus 2E (2G) are open:

  • 4160V BUS 2E (2G) NORMAL SUPPLY ACB 135554 (135594)
  • 4160V BUS 2E (2G) ALTERNATE SUPPLY ACB 135544 (135584) 7.2.3.5.2 IF necessary, ADJUST diesel output voltage to 4160V AC (Auto Voltage Adjust switch inside panel 2R43-P001 A(2R43-P001 C)).

7.2.3.5.3 Using Diesel Gen 2A (2C) Speed Adjust, LOWER frequency to 57 Hertz, THEN RAISE to 60 Hertz to force breaker closure.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 20 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2.3.6 IF Diesel 2A (2C) Speed AND Voltage are correct AND ACB 135530 (135540) is still NOT CLOSED, perform the following:

The terminal numbers are also the internal wire numbers, so the tags on the internal NOTE:

wires can be used to confirm that the jumper is on the correct terminals.

7.2.3.6.1 PLACE a jumper on TB-1 (right side) between terminals 30C2 (40C2) and 530C (540C) per 40AC-ENG-018-0, Temporary Modification Control to bypass Synch Acceptor Relay to allow DIG 2A (2C) output breaker closure, panel 2H21-P230 (2H21-P232).

7.2.3.6.2 PLACE Diesel Gen 2A (2C) Synch Switch (SSW) for ACB 135530 (135540) in ON, paneI2H11-P652.

7.2.3.6.3 Manually CLOSE ACB 135530, Diesel Gen 2A (2C) Emergency Supply, (135540),

panel 2H11-P652.

7.2.3.6.4 REMOVE jumper installed in step 7.2.3.6.1 per 40AC-ENG-018-0 in pane12H21-P230 (2H21-P232).

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 21 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7

  • DIESEL GENERATOR LOADING MUST NOT EXCEED THE FOLLOWING RATINGS TO AVOID DIESEL OVERLOAD:

CONTINUOUS 2850 KW 12.5 DAY (300 HOUR) 3250 KW

  • LIMIT TOTAL LOADING OF UNIT ONE AND UNIT TWO DIESELS TO 13,000 KW UNLESS NECESSARY TO PROTECT THE CORE OR CONTAINMENT OR TO PREVENT OFFSITE RADIOLOGICAL RELEASES.

THIS WILL ENSURE AN ADEQUATE SUPPLY OF FUEL OIL FOR THE 7 DAY PERIOD FOLLOWING AN LOSP. CONTINUE LIMITING TOTAL LOAD UNTIL AN ADDITIONAL MINIMUM OF 41,000 GALLONS OF FUEL OIL HAS BEEN ADDED TO THE STORAGE TANKS.

  • DIESEL GENERATOR FREQUENCY MUST BE MAINTAINED BETWEEN 59 AND 61 HERTZ.
  • PRIOR TO STARTING ANY OF THE 4160V LOADS LISTED BELOW, CAUTIONS: DIESEL LOAD MUST BE BELOW THE MAXIMUM VALUE LISTED BELOW.

MAX DIESEL LOAD LOAD FOR STARTING RHRSW PUMP (2C) 2050 KW RHRSW PUMP (2A, 2B, 20) 2150 KW CORE SPRAY PUMP 2400 KW RHR PUMP (2A, 2B) 2350 KW RHR PUMP (2C, 20) 2200 KW PSW PUMP (2A, 2B) 2650 KW

  • 2750 KW PSW PUMP (2C, 20) 2600 KW OW CHILLER 2600 KW
  • INDICATES PRELOAD VALUE FOR PSW PUMP WHEN TURBINE BUILDING IS ISOLATED.

7.2.3.7 Load the Diesel Generator 2A (2C) per the following:

7.2.3.7.1 Monitor AND ADJUST Diesel Generator 2A (2C) frequency as required to maintain it between 59 and 61 Hz, with the Diesel Generator 2A (2C) Speed Adjust switch.

7.2.3.7.2 Confirm Diesel Generator 2A (2C) KW reading is BELOW the maximum value in the above CAUTION.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 22 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.2.3.7.3 IF the Diesel Generator KW reading needs to be reduced, THEN perform any OR all of the following:

7.2.3.7.3.1 REDUCE the flow rates on the associated Diesel loads (pumps), as plant conditions allow.

7.2.3.7.3.2 REMOVE Diesel loads (pumps, etc.), as plant conditions allow.

7.2.3.7.3.3 To remove additional loads, refer to the following procedures electrical load list for the associated busses;

  • 34S0-R23-001-2, 600 VAC SYSTEM 7.2.3.7.4 PLACE loads on the diesel generator as required.

7.2.3.7.5 Monitor running diesel generator oil levels AND make arrangements in advance to add oil, IF needed.

7.2.3.7.6 Monitor running diesel generator fuel quantity AND make arrangements in advance to transfer AND/OR add oil, IF needed.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 23 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.3 SHUTDOWN 7.3.1 TRANSFERRING POWER FROM DIESEL GENERATOR 2A (2C) TO NORMAL OR ALTERNATE POWER CONTINUOUS I

  • All operations in this subsection are performed from panel 2H11-P652 UNLESS otherwise specified. This procedure is designed to shut down Diesel Generator 2A from EITHER the Main Control Room OR locally. To shut down Diesel 2C, use this NOTES: subsection AND substitute the component designations in parentheses.
  • lE a LOCA signal is present, this transfer cannot be performed since the diesel test logic cannot be energized to allow paralleling the diesel generator to the normal OR alternate supply.

7.3.1.1 Confirm power has been restored to Startup Transformers 2C OR 2D (potential lights, pane12H11-P651, for Startup Aux Xfmr 2C OR 2D ILLUMINATED).

7.3.1.2 IF ONLY SAT 2D is available, PLACE the Diesel Gen 2A (2C) Test SAT 2C Out of SVC Interlock switch in TEST.

7.3.1.3 RESET 4160V Bus 2E (2G) LOSP Lockout Relay in accordance with 30AC-OPS-003-0, Plant Operations, and 31 GO-OPS-021-0, Manipulation of Controls and Equipment, AND confirm annunciator 652-102(302), LOSS OF OFF SITE POWER, is CLEAR.

7.3.1.4 PLACE Diesel Gen 2A (2C) Mode Select switch in TEST.

CAUTION: lE LOSP AND / OR LOCA OCCURS, THEN SPEED DROOP CONTROL KNOB MUST BE POSITIONED TO ZERO (FULLY COUNTERCLOCKWISE) 7.3.1.5 WHILE maintaining Diesel2A (2C) frequency at desired level using the Diesel Gen 2A (2C) Speed Adjust switch, slowly ADJUST the speed droop knob at the Woodward Governor to 50 (in preparation for parallel operations).

7.3.1.6 PLACE OR confirm Diesel Gen 2A (2C) Voltage Reg Transfer switch in MANUAL.

7.3.1.7 PLACE Synch Switch (SSW) for ACB 135554 (135594), 4160V Bus 2E (2G) Normal Supply, OR ACB 135544 (135584), Alternate Supply, in ON.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 24 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7

  • IF THE DIESEL GENERATOR LOAD IS LESS THAN 500 KW, IT IS DESIRABLE TO HAVE THE SYNCHROSCOPE ROTATING IN THE COUNTERCLOCKWISE DIRECTION TO AVOID OPERATING THE DIESEL AT LOW LOADS WHEN PARALLELED TO THE GRID.

CAUTIONS:

  • IE, HOWEVER, THE DIESEL GENERATOR LOAD IS GREATER THAN 500 KW, IT IS NECESSARY TO HAVE THE SYNCHROSCOPE ROTATING IN THE CLOCKWISE DIRECTION TO PREVENT THE DIESEL FROM PICKING UP ADDITIONAL LOAD WHICH MAY OVERLOAD THE DIESEL.

7.3.1.8 Using Diesel Gen 2A (2C) Speed Adjust switch, adjust diesel speed to attain a slow synchroscope rotation in the desired direction (1 to 3 RPM).

CAUTION: DO NOT EXCEED 4400 VOLTS ON ANY PHASE OF THE DIESEL GENERATOR.

7.3.1.9 Using Diesel Gen 2A (2C) Voltage Adjust, adjust diesel output voltage to match the highest phase of the incoming source (Startup Transformer 2D (normal) OR 2C (alternate)).

CRITICAL 7.3.1.10 WHEN the synchroscope indicates 2 minutes to 12 AND WHEN the synchroscope lights approach the dimmest point, TAKE ACB 135554 (135594), 4160V 2E (2G) Normal Supply, OR ACB 135544 (135584), Alternate Supply, to the CLOSE position.

7.3.1.11 PLACE Synch Switch (SSW) for ACB 135554 (135594), 4160V Bus 2E (2G) Normal Supply, OR ACB 135544 (135584), Alternate Supply, in OFF.

7.3.1.12 Proceed to Diesel Generator 2A (2C) Manual Shutdown subsection to remove the Diesel Gen 2A (2C) from parallel operation.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 25 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.3.2 DIESEL GENERATOR 2A (2C) MANUAL SHUTDOWN CONTINUOUS I All operations in this subsection are performed from panel 2H11-P652 UNLESS otherwise specified. This procedure subsection is designed to shut down Diesel NOTE:

Generator 2A from EITHER the Main Control Room OR locally. To shut down Diesel 2C, use this subsection AND substitute the component designations in parentheses.

7.3.2.1 At the direction of the SS, if desired to shut down the DG with an auto start signal present which can not be reset, THEN install jumpers per Attachment 8.

7.3.2.2 CONFIRM/PLACE the Diesel Gen 2A (2C) Mode Select switch in TEST.

7.3.2.3 Using the Diesel Gen 2A (2C) Speed Adjust switch, DECREASE the load on diesel to between 400 and 500 KW, WHILE maintaining diesel reactive load between 400 and 500 KVAR using Diesel Gen 2A (2C) Voltage Adjust switch.

7.3.2.4 TAKE ACB 135530 (135540), Diesel Gen 2A (2C) Emergency Supply, control switch to TRIP.

7.3.2.5 Confirm Diesel Gen 2A (2C) Emergency Supply breaker green (OPEN) light is ILLUMINATED.

7.3.2.6 If this is an emergency (no cooling water, fire at DG, etc.),

TAKE the Diesel Gen 2A (2C) Start switch to STOP.

7.3.2.7 At Diesel Generator 2A (2C) Woodward Governor, perform the following:

  • POSITION the Speed Droop control knob to 0 (fully counter-clockwise)
  • Confirm that Load Limit control knob is at 10 (fully clockwise) 7.3.2.8 IF necessary, ADJUST frequency to 60 Hertz, with Diesel Gen 2A (2C) Speed Adjust switch. (N/A if already shut down) 7.3.2.9 IF necessary, ADJUST voltage to 4160V, with Diesel Gen 2A (2C) Voltage Adjust switch.

(N/A if already shut down) 7.3.2.10 PLACE Diesel Gen 2A (2C) Voltage Reg Transfer switch in AUTO.

7.3.2.11 Confirm Diesel Gen 2A (2C) Voltage Reg Transfer Auto red light is ILLUMINATED.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 26 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.3.2.12 IF necessary, ADJUST voltage to 4160V, with Diesel Gen 2A (2C) Auto Voltage Adjust switch inside pane12R43-P001A (2R43-P001 C). (N/A if already shut down)

OPERATION OF THE DIESEL AT LOW LOADS MUST BE MINIMIZED TO CAUTION: REDUCE OIL ACCUMULATION IN THE EXHAUST MANIFOLD CAUSED BY INSUFFICIENT GAS FLOWS AND TEMPERATURE TO VAPORIZE THE OIL.

7.3.2.13 After allowing the diesel to run for five (5) minutes unloaded (for even cooling), TAKE the Diesel Gen 2A (2C) Start switch to STOP. (N/A if already shut down) 7.3.2.14 Confirm the:

Diesel Gen 2A (2C) Start AND Diesel Gen 2A (2C) Shutdown System Operative red lights, are EXTINGUISHED.

7.3.2.15 If the DG was shut down with an auto start signal present, stop here until the auto start signal has been cleared and reset.

7.3.2.15.1 When the auto start signal has been reset, at the direction of the SS, remove jumper(s) installed per Attachment 8.

There is a 100 second time delay from the time the diesel is shut down UNTIL the NOTE:

Diesel Gen 2A (2C) Auto Start Sys Operative light is ILLUMINATED.

7.3.2.16 IF Diesel Gen 2A (2C) Auto Start Sys Operative light is EXTINGUISHED, DEPRESS Diesel Gen 2A (2C) Shutdown Relay pushbutton.

7.3.2.17 Confirm the Diesel Gen 2A (2C) Auto Start Sys Operative light is ILLUMINATED.

7.3.2.18 PLACE the Diesel Gen 2A (2C) Mode Select switch in NORM.

7.3.2.19 IF Diesel Gen 2A (2C) Test SAT 2C Out Of Svc Interlock switch is in TEST, PLACE switch in NORM.

7.3.2.20 Confirm annunciator 652-105 (305), DIESEL 2A (2C) IN TEST MODE, is CLEAR.

7.3.2.21 Locally, confirm 2P41-F339A (B), Diesel Generator 2A (2C) Cooling Water Outlet AOV, has CLOSED.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 27 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.3.2.22 Approximately ten minutes after Diesel Generator 2A (2C) is shutdown, confirm 2R43-C008A (C), standby lube oil pump, starts.

7.3.2.23 IF shutting down Diesel Generator 2A OR 2C, in Diesel Generator Room 2A (2C),

perform the following subsection:

7.3.2.24 BARRING OF 2A12C DIESEL GENERATOR Constant communications are required between the local operator at the diesel AND NOTE:

the Main Control Room whenever the Control Switch is in the AT ENG position.

WHILE BARRING THE ENGINE, TAKE CAUTION TO ENSURE THAT NO CAUTION: MAINTENANCE ACTIVITIES ARE BEING PERFORMED ON 4160 V BREAKERS BY PERSONNEL IN THE SWITCHGEAR ROOM 2E (2G).

7.3.2.24.1 After Diesel Generator 2A (2C) has cooled down for at least 15 minutes, establish constant communications between the operator at Diesel Generator 2A (2C)

AND the Main Control Room.

  • STEPS 7.3.2.24.2 THRU 7.3.2.24.15 CONTAIN ACTIONS WHICH WILL MAKE THE DIESEL GENERATOR INCAPABLE OF AUTOMATIC OR MANUAL STARTING AND AUTOMATIC TYING TO IT'S EMERGENCY BUS.
  • IN THE EVENT OF AN EMERGENCY, 2R43-F181A(C) MUST BE OPENED, CAUTIONS: THE ENGINE OVER SPEED MUST BE RESET AND CONTROL OF THE ENGINE MUST BE RETURNED TO THE CONTROL ROOM BY PLACING THE CONTROL SWITCH IN THE REMOTE POSITION BY THE LOCAL OPERATOR.
  • THE FOLLOWING STEPS MUST BE PERFORMED WITHOUT DELAY TO REDUCE THE TIME THAT CONTROL OF THE DIESEL IS AT THE ENGINE.

7.3.2.24.2 Notify the Shift Supervisor to ensure Tech Specs are complied with PRIOR to performing the following step.

7.3.2.24.3 PLACE the Diesel Generator 2A (2C) control switch in AT ENG, panel 2R43-P003A(2R43-P003C).

7.3.2.24.4 Confirm the CONTROL AT ENGINE, R43-114(314), annunciator ALARMS.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 28 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.3.2.24.5 DEPRESS the emergency stop pushbutton (above the Woodward Governor).

7.3.2.24.6 Confirm the ENGINE OVERSPEED annunciator ALARMS at 2H11-P652 and 2R43-P001A (2R43-P001 C).

7.3.2.24.7 CLOSE 2R43-F181A(C), Diesel Generator 2A (2C) Upper and Lower Main Bearing Accumulator/Booster Isolation Valve.

7.3.2.24.8 Rotate the engine at least one full revolution by momentarily depressing the engine START pushbutton on paneI2R43-P003A(2R43-P003C).

7.3.2.24.9 OPEN 2R43-F181A(2R43-F181C), Diesel Generator 2A (2C) Upper and Lower Main Bearing Accumulator/Booster Isolation Valve.

7.3.2.24.10 At the engine control side above the Woodward Governor, PULL the overspeed reset lever in a counterclockwise (left) direction to reset the trip.

7.3.2.24.11 REPEAT the above step AND confirm the reset lever subsequently moves freely indicating the reset latch is fully engaged.

7.3.2.24.12 DEPRESS the RESET pushbutton on panel 2R43-P003A(C).

7.3.2.24.13 Confirm the ENGINE OVERSPEED annunciator CLEARS.

7.3.2.24.14 PLACE the Diesel Generator 2A (2C) control switch in REMOTE (constant communications no longer required), panel 2R43-P003A (2R43-P003C).

7.3.2.24.15 Confirm the CONTROL AT ENGINE annunciator CLEARS.

7.3.2.24.16 Confirm the Diesel Gen 2A (2C) Auto Start Sys Operative light is ILLUMINATED.

7.3.2.24.17 Return the diesel generator to Standby per appropriate subsection of this procedure.

7.3.3 DIESEL GENERATOR B MANUAL SHUTDOWN FROM UNIT 2 Refer to 34S0-R43-001-1, subsection 7.3.3.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 29 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4 INFREQUENT OPERATIONS 7.4.1 TRANSFER OF FUEL OIL BETWEEN STORAGE TANKS CONTINUOUS I

  • Figure 1, at the end of this procedure, shows a simplified diagram of the fuel oil system in standby for reference.
  • Plant Hatch administrative limit is a total fuel supply of 175,000 gallons (T.S. 166,600 NOTES: gallons) for the Emergency Diesel Generators. This volume is administratively controlled by maintaining a fuel supply of 35,000 gallons (TS ?: 33,320 gallons) in the 1A, 1B, 1C, 2A, and 2C Diesel Generator Fuel Oil Storage Tanks. Reducing the volume below the Admin limit is not permitted for the receipt of fuel oil.
  • THE FOLLOWING STEP REMOVES THE AUTOMATIC ABILITY OF ONE TRANSFER PUMP TO FILL ITS DAY TANK. THE OPERABILITY OF THE REDUNDANT TRANSFER PUMP MUST BE CONFIRMED PRIOR TO PERFORMING THIS TRANSFER.
  • CONTINUOUS AIR MONITORING IS REQUIRED IF TRANSFERRING /

CAUTIONS: RECEIVING FUEL OIL BETWEEN TANKS.

  • PERFORM CONFINED SPACE ENTRY PER NMP-SH-005.
  • MOVEMENT OF THE FUEL OIL CAN PUSH THE ATMOSPHERE FROM INSIDE THE STORAGE TANK INTO THE TANK ACCESS AREA, CHANGING THE WORKING ENVIRONMENT WITHIN THE ACCESS AREA AND RESULT IN A SAFETY HAZARD FOR CONFINED SPACE ENTRANTS.

7.4.1.1 A minimum of 35,000 gallons of fuel oil in each Diesel Fuel Oil Storage Tank.

7.4.1.2 There shall be a minimum of 175,000 gallons of acceptable diesel fuel in the Unit 1 and Unit 2 Diesel Fuel Oil Storage Tanks.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 30 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 IF the Shift Supervisor requires the transfer of D/G fuel oil to 1A, 1B, or 1C fuel oil storage tank, NOTE: the respective tank level will be determined with the appropriate measuring stick prior to addition.

This action wi" minimize the possibility of tank overflow during the transfer of diesel fuel oil.

BEFORE ENTERING THE ACCESSES TO THE DIESEL FUEL OIL STORAGE TANKS, CAUTION: ENSURE THE REQUIREMENTS HAVE BEEN PERFORMED FOR NMP-SH-005, CONFINED SPACE PROCEDURE.

7.4.1.3 Select AND circle the tank to which oil wi" be transferred AND confirm OPEN/OPEN the valve designated on Table 1.

On Attachment 5, circle corresponding tank enclosure for tank selected.

TABLE 1 TO STORAGE TANK OPEN VALVE 1R43-A002C 1R43-F004C 1R43-A002B 1R43-F005C 1R43-A002A 1R43-F006C 2Y52-A001A 2R43-F008C 2Y52-A001C 2R43-F007C MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 31 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7

    • Two possible lineups are available to transfer from each Storage Tank Tank.. Only one of these lineups (see Table 2) may be used at a time.

NOTES:

NOTES: ** During transfer of DIG fuel oil FROM 1A, 1B, or 1C storage tanks, tank level will be determined using the appropriate measuring stick when approaching the Administrative limit of 35,000 gallons in order to confirm accuracy of the level indicators.

7.4.1.4 Select AND circle the tank from which the oil will be transferred AND confirm OR complete the valve lineup indicated in Table 2.

On Attachment 5, circle corresponding tank enclosure for tank selected.

TABLE 2 FROM STORAGE TANK OPEN VALVE CLOSE VALVE TRANSFER PUMP 1R43-F004A 1R43-F004B 1Y52-C101C 1R43-A002C 1R43-F004E 1R43-F004D 1Y52-C001C 1R43-F005A 1R43-F005B 1Y52-C101B 1R43-A002B 1R43-F005E 1R43-F005D 1Y52-C001B 1R43-F006A 1R43-F006B 1Y52-C101A 1R43-A002A 1R43-F006E 1R43-F006D 1Y52-C001A 2R43-FOOBA 2R43-FOOBB 2Y52-C101A 2Y52-A001A 2R43-F008E 2R43-FOOBD 2Y52-C001A 2R43-F007A 2R43-F007B 2Y52-C101C 2Y52-A001C 2R43-F007E 2R43-F007D 2Y52-C001C IF OTHER WORK ACTIVITIES (WITH THE POTENTIAL TO INTRODUCE FOREIGN MATERIAL INTO EDG FUEL OIL MAIN TANK RECEIVING THE TRANSFER) ARE IN CAUTION: PROGRESS WHILE THE 2W' PIPECAP IS REMOVED, THEN REQUIREMENTS OF NMP-MA-009, FOREIGN MATERIAL EXCLUSION PROGRAM, MUST BE CONSIDERED.

7.4.1.5 Remove 2W' pipe cap from Main Storage Tank to which the fuel oil will be transferred, to allow tank to vent while filling (located inside the tank enclosure).

7.4.1.5.1 Remove 2W' pipe cap from Main Storage Tank FROM WHICH the fuel oil will be transferred, allow tank to vent while filling (located inside the tank enclosure).

7.4.1.6 Establish communications between the Main Control Room AND the Fuel Transfer Pump Operator (Diesel Building).

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 32 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.4.1.7 At the appropriate panel in the Diesel Building (listed below),

PLACE the control switch for the fuel transfer pump indicated in Table 2 in HAND:

Fuel Transfer Pump Panel Number Location (Swgr Room) 1Y52-C101C 1H21-P256 1G 1Y52-C001C 1H21-P255 1E 1Y52-C101B 1H21-P256 1G 1Y52-C001B 1H21-P255 1E 1Y52-C101A 1H21-P256 1G 1Y52-C001A 1H21-P255 1E 2Y52-C101A 2H21-P256 2G 2Y52-C001A 2H21-P255 2E 2Y52-C101C 2H21-P256 2G 2Y52-C001C 2H21-P255 2E Operable control room level instruments may be used for level determination or trending during transfer of DIG fuel oil; however, final level determination for 1A, 1B, or 1C fuel oil storage NOTE: tanks must be determined by measuring stick when adding oil to these tanks OR whenever approaching the lower Administrative limit of 35,000 gallons.

7.4.1.8 WHEN the transfer is complete as indicated by desired level on the Control Room gauge, PLACE the transfer pump control switch in OFF, confirm the transfer pump stops, return the control switch to AUTO AND confirm the pump remains OFF.

CAUTION: FAILURE TO RESTORE VALVES TO STANDBY CAN RESULT IN OVER FILLING DAY TANKS OR MAIN TANKS DURING SUBSEQUENT TRANSFERS OR FUEL RECEIPT.

7.4.1.9 IF additional transfers are to be made, perform the following:

7.4.1.9.1 IF a different TO storage tank is to be used, restore valve to standby for the previously selected TO STOR TANK" (one valve).

7.4.1.9.2 REPLACE and VERIFY 2%" pipe cap on Main Storage Tank to which the fuel oil was transferred. (RECORD on Attachment 10)

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 33 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.1.9.3 IF a different FROM storage tank is to be used, restore valves to standby for the previously selected "FROM STOR TANK" (two valves).

7.4.1.9.4 Restart this subsection.

7.4.1.10 REPLACE and VERIFY 2112" pipe cap on Main Storage Tank to which the fuel oil was transferred. (RECORD on Attachment 10) 7.4.1.11 Restore valves to standby for the "TO STOR TANK" (one valve) AND the "FROM STOR TANK" (two valves).

7.4.1.12 Determine the Main Fuel Oil Tank Inventories are correct by performing the appropriate subsection of Attachment 6.

7.4.1.13 Confirm the Fuel Oil Transfer System is returned to Standby, by completing Attachment 5 for all valves in each tank enclosure identified as being selected AND associated pump control switches.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 34 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.2 MANUAL TRANSFER OF FUEL OIL FROM MAIN STORAGE TANKS TO DAY TANKS CONTINUOUS I 7.4.2.1 IF transferring fuel oil from Main Storage tank to its respective Day tank, THEN perform the following:

7.4.2.1.1 Select AND circle the day tank to which oil will be transferred AND confirm OPEN/OPEN one of the valves designated on Table 3.

NOTE: Two possible lineups are available to transfer from each storage tank. Only one of these lineups (See Table 3) may be used at time for manual transfer.

TABLE 3 TO DAY TANK FROM STORAGE TANK TRANSFER PUMP OPEN VALVE 1Y52-C101C 1R43-FOO4B 1R43-AOO1C 1R43-AOO2C 1Y52-COO1C 1R43-FOO4D 1Y52-C101B 1R43-FOO5B 1R43-AOO1B 1R43-AOO2B 1Y52-COO1B 1R43-FOO5D 1Y52-C101A 1R43-FOO6B 1R43-AOO1A 1R43-AOO2A 1Y52-COO1A 1R43-FOO6D 2Y52-C101A 2R43-FOO8B 2Y52-A101A 2Y52-AOO1A 2Y52-COO1A 2R43-FOO8D 2Y52-C101C 2R43-FOO7B 2Y52-A101C 2Y52-AOO1C 2Y52-COO1C 2R43-FOO7D 7.4.2.1.2 Proceed to step 7.4.2.4.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 35 OF BB DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.2.2 Select AND circle the day tank to which oil will be transferred AND OPEN one of the valves designated on Table 4.

On Attachment 5, circle corresponding tank enclosure for tank selected.

TABLE 4 TO DAY TANK OPEN VALVE 1R43-AOO1C 1R43-F004A OR FOO4E 1R43-AOO1B 1R43-F005A OR FOO5E 1R43-AOO1A 1R43-F006A OR FOO6E 2Y52-A101A 2R43-FOOBA OR FOOBE 2Y52-A101C 2R43-F007A OR FOO7E THE FOLLOWING STEP REMOVES THE AUTOMATIC ABILITY OF ONE TRANSFER CAUTION: PUMP TO FILL ITS DAY TANK. THE OPERABILITY OF THE REDUNDANT TRANSFER PUMP MUST BE CONFIRMED PRIOR TO PERFORMING THIS TRANSFER.

7.4.2.3 Select the storage tank from which the oil will be transferred AND complete the valve lineup indicated in Table 5.

On Attachment 5, circle corresponding tank enclosure for tank selected.

NOTE: Two possible lineups are available to transfer from each storage tank.

Only one of these lineups (See Table 5) may be used at a time.

TABLE 5 FROM STORAGE TANK OPEN VALVE CLOSE VALVE TRANSFER PUMP 1R43-FOO4A 1R43-FOO4B 1Y52-C101C 1R43-AOO2C 1R43-FOO4E 1R43-FOO4D 1Y52-COO1C 1R43-FOO5A 1R43-FOO5B 1Y52-C101 B 1R43-AOO2B 1R43-FOO5E 1R43-FOO5D 1Y52-COO1B 1R43-FOO6A 1R43-FOO6B 1Y52-C101A 1R43-AOO2A 1R43-FOO6E 1R43-FOO6D 1Y52-COO1A 2R43-FOOBA 2R43-FOOBB 2Y52-C101A 2Y52-AOO1A 2R43-FOOBE 2R43-FOOBD 2Y52-COO1A 2R43-FOO7A 2R43-FOO7B 2Y52-C101C 2Y52-AOO1C 2R43-FOO7E 2R43-FOO7D 2Y52-COO1C MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 36 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.2.4 Establish communications between the Main Control Room AND the Fuel Transfer Pump Operator (Diesel Building).

7.4.2.5 At the appropriate panel in the Diesel Building (listed below),

PLACE the control switch for the fuel transfer pump indicated in Table 5, (OR Table 3, IF step 7.4.2.1 was performed), in HAND:

Fuel Transfer Pum~ Panel Number Location {Swgr Room}

  • 2Y52-COO1C 2H21-P255 2E 7.4.2.6 Monitor the level in all Day AND Main Storage Tanks, ensure the level changes ONLY occur in the Day Tank being pumped into AND the Main Storage Tank being pumped from.

7.4.2.7 IF an undesirable level increase in the Main Storage Tank is detected AND the Day Tank requires filling, perform the following:

7.4.2.7.1 STOP the transfer.

7.4.2.7.2 Restore valve to standby for the "TO DAY TANK" (one valve).

7.4.2.7.3 Line up the other path to fill that Day Tank AND restart the transfer.

7.4.2.7.4 Request maintenance to repair the leaking Transfer Pump Discharge Check Valve at the earliest opportunity.

7.4.2.8 WHEN the transfer is complete as indicated by proper level in the selected day tank (Control Room gauge),

PLACE the transfer pump control switch in OFF, confirm the pump STOPS, return the control switch to AUTO and confirm the pump remains OFF.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 37 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.4.2.9 Determine that the Tank Inventories are correct by performing Attachment 6.

7.4.2.10 Return the Fuel Oil Transfer System to Standby by completing Attachment 5 for all valves in each tank enclosure identified as being selected AND associated pump control switches.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 38 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7 7.4.3 TRANSFERRING DIESEL GENERATOR 1B FROM BUS 1F TO BUS 2F CONTINUOUS I

  • This subsection is used to transfer power to the 4160V Bus 2F after normal AN D alternate have been lost to that bus. 4160 V Bus 1F may OR may NOT have normal power, but the decision has been made that powering up Bus 2F is required. This subsection is NOT applicable IF a LOCAILOSP exists on Unit 1 OR Unit 2. Diesel 1B will stay with the unit that NOTES: has the LOCAILOSP UNLESS both units have LOCAILOSP, in which case both 1F AND 2F buses will be deenergized.
  • With 4160V Bus 1F energized from EITHER its normal OR alternate supply, WHEN the Diesel 1B output breaker to 4160V Bus 1F is opened, the Diesel 1B output breaker to 4160V Bus 2F wi" close automatically IF 2F is deenergized.

7.4.3.1 IF Normal OR Alternate power can be restored to 4160V Bus 1F, perform the following:

7.4.3.1.1 Transfer 4160V Bus 1F to Normal OR Alternate power per 34S0-R43-001-1 ,

Diesel Generator Standby AC System.

7.4.3.1.2 PLACE Diesel Generator 1B Keylock control switch in REMOTE UNIT 2 locally at panel 2R43-M01.

7.4.3.1.3 Confirm the Diesel Gen B Voltage Reg Transfer Auto red light is ILLUMINATED, panel 2H 11-P652.

7.4.3.1.4 Confirm the Diesel Gen B Voltage Reg Transfer Manual green light is EXTINGUISHED, paneI2H11-P652.

7.4.3.1.5 Using Diesel Gen B Speed Adjust switch, adjust diesel generator frequency to 60 Hertz, panel 2H11-P652.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 39 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.3.2 IF neither Normal nor Alternate power is available to 4160V Bus 1F, perform the following:

NOTE: 2P41-C002, Stby PSW Pump, will be lost due to de-energizing 1R24-S026 UNTIL Diesel Gen B is supplying power to 4160V Bus 2F.

7.4.3.2.1 Line up Cooling Water to Diesel Generator 1B from Unit 1 divisional Plant Service Water per 34S0-P41-005-2, Standby Diesel Service Water System.

7.4.3.2.2 STOP 2P41-C002, Stby PSW Pump, per 34S0-P41-005-2, Standby Diesel Service Water Pump.

7.4.3.2.3 PLACE Diesel Generator 1B Keylock control switch in REMOTE UNIT 2 locally, at panel 2R43-M01.

7.4.3.2.4 TAKE and HOLD ACB 135912, Diesel Gen B Emergency Supply to 4160V Bus 1F, control switch to TRIP UNTIL ACB 135570, Diesel Gen B Emergency Supply, CLOSES, panel 2H11-P652.

7.4.3.3 IF 4160V Bus 1F is de-energized, confirm that Feeder To SST 2F1 2S11-S009 Alt Sply 1R24-S026 Diesel Bldg MCC 1B is CLOSED to supply power to 1R24-S026.

7.4.3.4 RESET 4160V Bus 2F LOSP Lockout Relay in accordance with 30AC-OPS-003-0, Plant Operations, and 31 GO-OPS-021-0, Manipulation of Controls and Equipment, AND confirm annunciator 652-202, LOSS OF OFF SITE POWER, is CLEAR.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7

  • DIESEL GENERATOR FREQUENCY MUST BE MAINTAINED BETWEEN 59 AND 61 HERTZ.
  • TO AVOID DIESEL OVERLOAD FOLLOWING A LOSS OF OFF SITE POWER, DIESEL GENERATOR LOADING MUST NOT EXCEED THE FOLLOWING RATINGS.

1000 HOUR 2850 KW 7 DAY (168 HOUR) 3250 KW

  • LIMIT TOTAL LOADING OF UNIT ONE AND UNIT TWO DIESELS TO 13,000 KW UNLESS NECESSARY TO PROTECT THE CORE OR CONTAINMENT OR TO CAUTIONS: PREVENT OFFSITE RADIOLOGICAL RELEASES. THIS WILL ENSURE AN ADEQUATE SUPPLY OF FUEL OIL FOR THE 7 DAY PERIOD FOLLOWING AN LOSP. CONTINUE LIMITING TOTAL LOAD UNTIL AN ADDITIONAL MINIMUM OF 41,000 GALLONS OF FUEL OIL HAS BEEN ADDED TO THE STORAGE TANKS.
  • PRIOR TO STARTING ANY OF THE 4160V LOADS LISTED BELOW, DIESEL LOAD MUST BE BELOW THE MAXIMUM VALUE LISTED BELOW.

LOAD MAX DIESEL LOAD FOR STARTING RHRSW PUMP (2C) 2050 KW RHR PUMP (2C, 20) 2D) 2200 KW PSW PUMP (2C (2C, 2D)

20) 2600 KW 7.4.3.5 Load Diesel Generator 1B as necessary to support Unit 2 activities per the following:

7.4.3.5.1 Monitor AND ADJUST Diesel Generator 1B frequency as required, to maintain it between 59 and 61 Hz, with the Diesel Generator 1B Speed Adjust switch.

7.4.3.5.2 Confirm Diesel Generator 1B KW reading is BELOW the maximum value in the above CAUTION.

7.4.3.5.3 IF the Diesel Generator 1B KW reading needs to be reduced, THEN perform any OR all of the following:

7.4.3.5.3.1 REDUCE the flow rates on the associated Diesel loads (pumps),

as plant conditions allow.

7.4.3.5.3.2 REMOVE Diesel loads (pumps, etc.), as plant conditions allow.

7.4.3.5.3.3 To remove additional loads, refer to the following procedures electrical load list for the associated busses:

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 41 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.3.5.4 PLACE loads on Diesel Generator 1B as required.

7.4.3.6 WHEN power is available to bus 1R24-S026, START 2P41-C002, Stby PSW Pump, AND return Unit 1 PSW lineup to normal per 34S0-P41-005-2, Standby Service Water System.

7.4.3.7 WHEN normal OR alternate power is available to 4160V Bus 2F, transfer power from the 1B diesel to normal power per the appropriate subsection of this procedure.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 42 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.4 TRANSFERRING 4160V BUS 2F FROM 1B DIESEL TO NORMAL OR ALTERNATE POWER CONTINUOUS I NOTE: !E a LOCA signal is present, this transfer cannot be performed since the diesel test logic cannot be energized to allow paralleling the diesel generator to the normal OR alternate supply.

7.4.4.1 Confirm power has been restored to Startup Transformers 2C OR 20.

(potential lights for Startup Aux Xfmr 2C OR 20 ILLUMINATED, panel 2H11-P651).

7.4.4.2 RESET 4160V Bus 2F LOSP Lockout Relay in accordance with 30AC-OPS-003-0, Plant Operations, and 31GO-OPS-021-0, Manipulation of Controls and Equipment, AND confirm annunciator 652-202, LOSS OF OFF SITE POWER, is clear.

7.4.4.3 PLACE Diesel Gen B Mode Select switch in TEST.

7.4.4.4 PLACE Diesel 1B keylock switch at panel 2R43-M01 to "REMOTE UNIT 2".

CAUTION: IF LOSP AND / OR LOCA OCCURS, THEN SPEED DROOP CONTROL KNOB MUST BE POSITIONED TO ZERO (FULLY COUNTERCLOCKWISE) 7.4.4.5 WHILE maintaining Diesel1B frequency at desired level, using the Diesel 1B Speed Adjust switch, slowly ADJUST the speed droop knob at the Woodward Governor to 50 (in preparation for parallel operations).

7.4.4.6 PLACE OR confirm Diesel Gen B Voltage Reg Transfer switch in MANUAL, panel 1H11-P652.

7.4.4.7 PLACE OR confirm Diesel Gen B Voltage Reg Transfer switch in MANUAL, pane12H11-P652 7.4.4.8 PLACE Synch Switch (SSW) for ACB 135574, 4160V Bus 2F Normal Supply, OR ACB 135564, Alternate Supply, in ON.

7.4.4.9 ADJUST the Diesel Gen B Voltage Adjust switch UNTIL diesel output voltage is equal to 4160V.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 43 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-00 1-2 24.7

  • IF THE DIESEL GENERATOR LOAD IS LESS THAN 500 KW, IT IS DESIRABLE TO HAVE THE SYNCHROSCOPE ROTATING IN THE COUNTERCLOCKWISE DIRECTION TO AVOID OPERATING THE DIESEL AT LOW LOADS WHEN PARALLELED TO THE GRID.

CAUTION:

  • lE, HOWEVER, THE DIESEL GENERATOR LOAD IS GREATER THAN 500 KW, IT IS NECESSARY TO HAVE THE SYNCHROSCOPE ROTATING IN THE CLOCKWISE DIRECTION TO PREVENT THE DIESEL FROM PICKING UP ADDITIONAL LOAD WHICH MAY OVERLOAD THE DIESEL.

7.4.4.10 Using Diesel Gen B Speed Adjust switch, ADJUST diesel speed to attain a slow synchroscope rotation in the desired direction (1 to 3 RPM).

CAUTION: DO NOT EXCEED 4400 VOLTS ON ANY PHASE OF THE DIESEL GENERATOR.

7.4.4.11 Using Diesel Gen B Voltage Adjust, adjust diesel output voltage to match the highest phase of the incoming source (Startup Transformer 2D (normal) OR 2C (alternate)).

CRITICAL 7.4.4.12 WHEN the synchroscope indicates 2 minutes to 12 AND I

WHEN the synchroscope lights approach the dimmest point, TAKE ACB 135574, 4160V 2F Normal Supply OR ACB 135564, Alternate Supply to CLOSE.

7.4.4.13 Using the Diesel Gen B Speed Adjust switch, DECREASE the load on diesel to between 400 and 500 KW WHILE maintaining diesel reactive load between 400 and 500 KVAR, using Diesel Gen B Voltage Adjust switch.

7.4.4.14 TAKE ACB 135570, Diesel Gen B Emergency Supply, control switch to TRIP.

7.4.4.15 Confirm ACB 135570, Diesel Gen B Emergency Supply, green (OPEN) light is ILLUMINATED.

7.4.4.16 PLACE Synch Switch (SSW) for ACB 135574, 4160V Bus 2F Normal Supply, OR ACB 135564, Alternate Supply, in OFF.

7.4.4.17 SHUT DOWN the Diesel Generator 1B in accordance with 34S0-R43-001-1, Diesel Generator Standby AC System.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 44 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 7.4.5 DETERMINING EDG LUBE OIL FILTER OR STRAINER DIFFERENTIAL PRESSURE(DP)

CONTINUOUS I NOTE: The following directions for determining lube oil filter DP for the 2A and 2C EDG are posted locally on the 2R43-D012A & C filters.

7.4.5.1 To determine the lube oil filter DP for the 2A EDG, perform the following at the local EDG skid:

7.4.5.1.1 OPEN one of 2R43-D012A, lube oil filter, instrument isolation valves.

7.4.5.1.2 Read pressure on the lube oil filter pressure indicator.

7.4.5.1.3 CLOSE the valve opened in step 7.4.5.1.1.

7.4.5.1.4 OPEN the other 2R43-D012A, lube oil filter, instrument isolation valve.

7.4.5.1.5 Read pressure on the lube oil filter pressure indicator.

7.4.5.1.6 CLOSE the valve opened in step 7.4.5.1.4.

7.4.5.1.7 Filter DP is the absolute difference in the two pressure readings.

7.4.5.2 To determine the 2R43-D012C, 2C EDG Lube Oil Filter, differential pressure, perform the following at the local EDG skid:

7.4.5.2.1 Read pressure on the lube oil filter pressure indicator.

7.4.5.2.2 Rotate the 3-way instrument isolation valve to read pressure from the other side of the filter.

7.4.5.2.3 Read pressure on the lube oil filter pressure indicator.

7.4.5.2.4 Filter DP is the absolute difference in the two pressure readings.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 45 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 NOTE: The following directions for determining lube oil strainer DP for the 2A and 2C EDG are posted locally on the 2R43-D013A & 2R43-D013C Strainers.

7.4.5.3 To determine 2R43-D013A (2R43-D013C), 2A (2C) EDG Lube Oil Strainer, differential pressure, perform the following at the local EDG skid:

7.4.5.3.1 Read pressure on the lube oil strainer pressure indicator.

7.4.5.3.2 Rotate the 3-way instrument isolation valve to read pressure from the other side of the strainer.

7.4.5.3.3 Read pressure on the lube oil strainer pressure indicator.

7.4.5.3.4 Strainer DP is the absolute difference in the two pressure readings.

MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 46 OF 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 Figure 1 Simplified Fuel Oil System Diagram (66 JA 1R43-F029 DAY TANK Simplified diagram of fuel oil diesel oil system. Reference drawings HII037 and H21074.

DAY TANK DAY TANK DAY TANK DAY TANK Be MGR-0001 Ver. 3

SNC PLANT E. I. HATCH I Pg 47 of 88 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT -L Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM RESTORATION 1 of 2 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ I DATE _ _ __

_ _ _ _~I1 Date ____

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time ______ _ _ ___

COMMENTS: ______________________________________

OPS-0592 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 48 of 88 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ..L Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM RESTORATION 2 of 2 Diesel Generator STEP DESCRIPTION CHECKED VERIFIED NUMBER PSW To Turb Bldg Div I, 2P41-F316A (C) energized OR PSW To Turb 7.1.2 Bldg Div II, 2P41-F316D (B) energized 7.1.3 Diesel Gen 2A (2C, B) Mode Select switch is in NORM Diesel Gen 2A(2C, 1B) Diesel Test SAT 2C Out Of Svc Interlock switch in 7.1.4 NORM 7.1.5 Diesel Gen 2A(2C, 1B) Voltage Reg Transfer switch in AUTO Diesel Gen 2A (2C, B) Shutdown System Operative red light is 7.1.6 EXTINGUISHED 7.1.7 Diesel Gen 2A (2C, B) Start red light is EXTINGUISHED Diesel Gen 2A (2C, B) Auto Start System Operative clear light is 7.1.8 ILLUMINATED Diesel 2A(2C, 1B) Fuel Oil Storage Tanks (Main and Day Tanks) 7.1.9 Inventories are correct Diesel Gen 2A(2C, 1B) Air Compressor 2A1(2C1, 1B1), 2R43-COO5A 2R43-C005A 7.1.10 (C, 1R43-C001 B) control switches in AUTO Diesel Gen 2A (2C, B) Air Compressor 2A2(2C2, 1B2), 2R43-COO6A 2R43-C006A 7.1.10 (C, 1R43-C010B) control switches in AUTO Diesel Generator 2A (2C, B) cooling water outlet AOV, 2P41-F339A 7.1.11 (B, 2P41-F340) is CLOSED 7.1.12 Diesel Generator 2A(2C) governor lube oil level proper Diesel Generator 2A (2C, B) Woodward Governor Speed Droop is set at 0 7.1.13 (fully counter-clockwise)

Diesel Generator 2A (2C, B) Woodward Governor Load Limit is set at 10 7.1.13 (fully clockwise) 7.1.14 Diesel Generator 2A (2C) local mode selector switch in REMOTE Diesel Generator 1B local mode selector switch in REMOTE AT UNIT 1 or 7.1.15 REMOTE AT UNIT 2 7.1.16 Diesel Generator 2A (2C, B) Jacket Coolant Expansion Tank >3/4 full.

7.1.17 Diesel Generator 2A (2C, B) bearing lube levels proper 7.1.18 Diesel Generator 2A (2C, B) lube oil sump level (dipstick) proper 7.1.19 Battery Charger 2G, 2J, 1H,(2H, 2N, 1N) is in service IF Diesel Generator 2A (2C, B) has been run, the DIG has been barred per 7.1.20 NA appropriate procedure barring subsection.

IF Diesel Generator 2A (2C, B) was barred, THEN independently verify that 7.1.21 2R43-F181A(C) or 1R43-F181B is OPEN OPS-0592 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 49 of 88 DOCUMENT TITLE: jDOCUMENT

\ DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT 2.. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 1 of 13 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ _ _ 1 / DATE _ _ _ _ __

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ ____ __ _ .....:/I Date _ _ _ ___

COMMENTS: _________________________________

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 50 of 88 P1l50 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ..£. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 2 of 13 SWITCH/BRK NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2H11*P652 Main Control Room Panel 43A1 Diesel Gen 2A Mode Select switch NORM 86/E 4160V Bus 2E LOSP Lockout Relay RESET 43B1 Diesel Gen B Mode Select switch NORM 86F 4160V Bus 2F LOSP Lockout Relay RESET 43C1 Diesel Gen 2C Mode Select switch NORM 86/G 4160V Bus 2G LOSP Lockout Relay RESET Diesel (2A) Test SAT 2C Out Of Svc Interlock NORM Diesel Gen 2A Prelube Pump OFF Diesel (B) Test SAT 2C Out Of Svc Interlock NORM Diesel Gen B Prelube Pump OFF Diesel (2C) Test SAT 2C Out Of Svc Interlock NORM Diesel Gen 2C Prelube Pump OFF SS 530 Diesel Gen 2A Synch Switch (SSW) ACB 135530 OFF SS 570 Diesel Gen B Synch Switch (SSW) ACB 135570 OFF SS 540 Diesel Gen 2C Synch Switch (SSW) ACB 135540 OFF Diesel Gen 2A Voltage Reg Transfer switch AUTO Diesel Gen B Voltage Reg Transfer switch AUTO Diesel Gen 2C Voltage Reg Transfer switch AUTO 1F Switchgear Room, Diesel Building 1R24-M032B 125V DC Throw-Over Sw (near battery charger Note 2 Note 2 - Battery Charger Switches are normally in the UP position, but IF alternate battery charger is used, switch must be in the DOWN position.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 51 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .1.. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 3 of 13 SWITCH! BRKR NORMAL DESCRIPTION POSITION CHECKED VERIFIED NUMBER POSITION CHECKED VERIFIED 1R24-S026 600/208V MCC-1B (1F Switchgear Room, Diesel Building) 208V Section of 1R24-S026 Frames 6 thru 8 Via R25-S030 Frame 2A ON (Transformer 1E, 1 R11-S005)

R11-S005)

Frame 2CR 125V Battery Charger 1H, 1R42-S032B ON To Stby Serv Water Pmp 2P41-C002 60 HP Local Starter, Frame 2DR ON 2R27-S037 Frame 2E Diesel Generator 1B Jacket Water Heater, 1R43-B001 B ON Frame 3AR 125V Battery Charger 1N, 1R42-S032D ON Frame 3B DG 1B Lube Oil Heater, 1R43-B002B ON Motor Control Switch (toggle switch) for IT Feeder From Frame 5C ON Sta. Serv Xfmr 2FI, 2S11-S009 Diesel Generator 1B Skid Mounted Panel Board, Frame 6A ON 1R43-S001B 1R25-S030 120/208V AC Cab. 1K (1F Switchgear Room) 1H21-P231 DG 1B Gen Brg & Stator Temp Mon, 14 ON 1R43-R772B 15 1R43-N516 DG 1B Fuel Oil Day Tk Wtr Detector ON 20 1R43-P001 B DG 1B WH Meter, 1R43-M004 ON 2JESB02, Elect. Sup. to 2P41-ES-K602 and Diesel1B Fuel 22 ON Tank Levellndic. 2R43-R607B & R608B 27 1R24-S026, Del Bldg MCC 1B 1R24-S026 Space Heaters ON 61 1R25-S 113 208V Section of MCC-1 B, 1R24-S026 ON 1R25-S005125V DC Cab 1E (1F Switchgear Room) 1R22-S006 FR 5 4160V Swgr Bus 1F Fdr ACB's (FR 5, 6, 1 ON 10,12, & 13) & UNDV Relay 2 1R43-P001 B, Generator 1B Field Flashing ON 3 1R22-S006 FR1 4160V Swgr Bus 1F ACB #135713 ON 4 1H21-P201 Generator 1B Trip Relaying ON 5 1R22-S006 FR 11 4160V Swgr Bus 1F ACB #135714 ON OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I ~g 52 of 88 Pg DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .l.. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 4 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 6 1R43-S00 1B Diesel 1B Start Control Circuit A ON 7 1 R22-S006 Fr 6 Diesel Generator 1B ACB #135912 ON 9 1R43-P001 B Diesel Generator 1B Annunciators ON 10 1R43-S001 B Diesel 1B Stop And Shutdown Control ON 13 1H21-P231 4160V Bus 1F Loading LOCAILOSP Timers ON 14 1R43-S001 B Diesel 1B Governor Control ON 15 1R43-P001 B, Diesel Generator 1B Manual M. O. Volt Adj ON 16 1H21-P231, Diesel Generator 1B Test & LOSP Lockout Relays ON 17 1H21-P231, Generator 1B Voltage Regulator Transfer ON 18 1R43-S001 B Diesel 1B Start Control Circuit B ON 19 1H21-P231, Diesel Generator 1B Anti-Parallel Circuit ON 20 1R43-P001 B, Diesel Generator 1B Auto M. O. Volt Adj ON 21 1R52 Emergency Lighting ON 1R43-S001 B, Diesel Generator 1B Jacket Coolant Level 22 ON Control Sol Vlv 2H21-P256 Switchgear 2G Room Diesel Building Diesel2C Fuel Pump 2C2, 2Y52-C101C AUTO Diesel2A Fuel Pump 2A2, 2Y52-C101A AUTO 2R26-M032C 125V DC Throw-Over Sw. (Next to Battery Charger) NOTE 2 2R24-S027 600/208V MCC 2C (2G Swgr Room)

Frame 1A 2R25-S031/208V Section of 2R24-S027 Via Xfmr 2R11-S006 ON Frame 1BR 125V STBY Battery Charger 2N, 2R42-S032E ON Frame 1CL 125V Battery Charger 2J. 2R42-S032C ON Note 2 - Battery Charger Switches are normally in UP position, but!E alternate battery charger is used, switch must be in DOWN position.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 53 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .£ Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 5 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Frame 1D Diesel Gen 2C Jacket Water Htr. 2R43-B007C ON Frame 1E Diesel Gen 2C lube Oil Heater, 2R43-B006C ON Frame 2E Diesel Gen 2A Air Compressor 2A2, 2R43-C006A ON Frame 2F Diesel Gen 2C Air Compressor 2C2, 2R43-C006C ON Frame 7E Diesel Gen 2A Fuel Oil Pump 2A2, 2Y52-C101A ON Frame 7F Diesel Gen 2C Fuel Oil Pump 2C2, 2Y52-C101C ON Frame 8Cl DG 2C 208V Skid MTD Ckt Brkr Panel Board ON 2R25-S006125V DC Dist. Pnl2F (2G Swgr Room) 2R22-S007 FR 6 4160V Swgr Bus 2G for ACB's (FR 2,3,4,5,6, Brkr 1 ON 8 9 & 11) & Undy Relayinq Undv Relavinq Brkr2 2R43-P001 C Generator 2C Field Flashing ON Brkr 3 2R22-S007 FR 10 4160V Swgr Bus 2G ACB # 135584 ON Brkr4 2H21-P202 Generator 2C Trip Relaying ON Brkr 5 2R22-S007 FR 1 4160V Swgr Bus 2G ACB # 135594 ON Brkr 6 2R43-S001 C Diesel 2C Start Control Circuit A ON Brkr 7 2R22-S007 FR 7 Diesel Genrtr 2C ACB # 135540 ON 2H21-P202, Emerg Relaying For 4160/600V Station Service Brkr 8 ON Xfmr 2D Brkr 9 2R43-P001 C Diesel Genrtr 2C Annunciators ON Brkr 10 2R43-S001 C Diesel 2C Stop And Shutdown Control ON 2H21-P232, 4160V Bus 2G loading lOCAllOSP Timers & PSW Brkr 11 ON Valve Interlocks Brkr 13 2R43-P001 C, Diesel Genrtr 2C Manual M.O. Volt Adj ON Brkr 14 2R43-S001 C Diesel 2C Governor Control ON OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 54 of 88 PlJ54 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT 2... Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 6 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Brkr 15 2R43-S001 C Diesel 2C Start Control Circuit B ON Brkr 16 2H21-P232, Diesel Genrtr 2C Test & LOSP Lockout Relay ON Brkr 17 2H21-P232, Generator 2C Voltage Regulator Transfer ON Brkr 18 2R43-P001 C, Diesel Generator 2C Auto M. O. Volt Adj ON Brkr 19 2H21-P232, Diesel Generator 2C Anti-Parallel Circuit ON Brkr 20 2R43-S001 C, Diesel Genrtr 2C Jacket Coolant Level Control Sol Vlv ON Brkr 21 2X51-RM-2G Emergency Lighting ON Brkr 22 2H21-P256, PSW Div 2 MOVs Emergency Relaying ON Brkr 23 2R22-S024D 4KV Test Cabinet For Bus 2G ON 2H21-P256, Div 2 diesel Abnormal Flow & Standby Pump Start Brkr 25 ON RelayinQ Relavinq Brkr 27 2H21-P232, Diesel 2C Synchro Acceptor 2R43C-D2C-SA ON 2R22-S006 FR 3 4160V Swgr Bus 2F For Fdr ACB's (FR 3,9, &

Brkr 29 ON 10)(Div 2) 2H11-P199 Battery Charger 2J, 2R42-S032C Standby Bat Chgr Brkr 30 ON 2N 2R42-S032E 2R25-S031 120/208V Dist. Pnl. 2L (2G Swgr Room) 2H21-P232, Diesel Genrtr. 2C Genrtr Bearing & Stator Temp Brkr 14 ON Indication Brkr 15 2R43-N009C Diesel Gen Fuel Oil Day Tank 2C Water Detector ON FDR 1, 2R43-N007C Main Control Room Day Tank 2C Level Brkr 24 Transmitter FDR 2, 2R43-N010C Main Control Room Main Fuel ON Oil Tank Transmitter Brkr 29 2R24-S027 FR 1 Diesel Building MCC-2C Space Heaters ON Brkr 33 2R22-S007 FR 6 4KV Swgr Bus 2G Space Heaters All Frames ON OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 55 of 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM I 34S0-R43-001-2 24.7 ATTACHMENT .£ Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 7 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Brkr 36 2R22-S007 FR 6 4KV Swgr Bus 2G Motor Heaters All Frames ON Brkr 63* 2R20M-P002 208V Section of 2R24-S027 MCC 2C ON 2R22-S007 4160V Sta. Servo Swgr 2G R'CKD IN Supply From Start-Up Aux Transf 20 (Winding #2) ACB 135594 CLOSED Frame 1 30 Amp DC Control Power Breaker CLOSED 70 Amp AC Heater Power Breaker CLOSED 30 Amp AC Heater Power Breaker CLOSED Frame 6 70 Amp DC Control Power and Undervoltage Relaying Breaker CLOSED 25 Amp DC Undervoltage Relaying Breaker CLOSED R'CKD IN Supply From DIG 2C, 2R43-S001 C ACB 135540 OPEN Frame 7 30 Amp DC Control Power Breaker CLOSED R'CKD IN Supply From Start-Up Aux Transf 2C ACB 135584 OPEN Frame 10 30 Amp DC Control Power Breaker CLOSED Local Air Compressor Switches, 2C Diesel Generator Room Diesel 2C Air Compr 2C1, 2R43-C005C, Control Switch AUTO Diesel 2C Air Compr 2C2, 2R43-C006C, Control Switch AUTO 2R43-P001C Inside lower right cover (2C Diesel Room)

S61 Diesel 2C Unit-Parallel Switch UNIT INSTALL Diesel 2C Ammeter Shorting Block (if cover in place, cannot be SB COVER, aborted position)

NORMAL

  • This switch sWitch has no automatic trip triP functions. Overcurrent protection is IS provided by fuses In in fuse panel 2R20M-P002.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 56 of 88 DOCUMENT TITLE: IDOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ..£. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 8 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R43-P003C Diesel Gen. 2C Cont. Pnl. (Indicating Lights)

BRKR 1 (Note 1) Control Power On Circuit - 1 (Coolant Pump 2R43-C012C) ON Control Power On Circuit - 2 (Stby Lube Oil Pump BRKR 2 (Note 1 ON 2R43-C008C)

BRKR 3 (Note 1) Control Power On Circuit - 3 (Prelube Pump 2R43-C009C) ON BRKR 4 (Note 1 Control Power On Circuit - 4 (Generator Heater) ON 2R22-S006 4160V Sta. Servo Swgr 2F Supply From Start-Up Aux Transf 2D (Winding #2) ACB R'CKD IN 135574 CLOSED Frame 1 30 Amp DC Control Power Breaker CLOSED 70 Amp AC Heater Power Breaker CLOSED 30 Amp AC Heater Power Breaker CLOSED Frame 6 70 Amp DC Control Power and Undervoltage Relaying CLOSED Breaker 25 Amp DC Undervoltage Relaying Breaker CLOSED R'CKD IN Supply From DIG 1B, 1R43-S001 B ACB 135570 OPEN Frame 7 30 Amp DC Control Power Breaker CLOSED R'CKD IN Supply From Start-Up Aux Transf 2C ACB 135564 OPEN Frame 11 30 Amp DC Control Power Breaker CLOSED 2R25-S035 120/208V Dist. Pnl. 2K (2F Swgr Room) 13 2R24-S048 FR 1 Diesel Building MCC 2D Fr 1 Space Heater ON 21 2R22-S006 FR 6 4KV Swgr 2F Space Heaters All Frames ON 24 2R22-S006 FR 6 4KV Swgr 2F Motor Space Heaters ON 33 2R24-S048 FR 2208V Section of 2R24-S048 MCC-2D ON Note 1 - Circuit Breakers are located in panel 2R43-S001 C(A) Breaker Box at the West end of the Diesel Generator Skid.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH 1 I P~ 57 of 88 Pg DOCUMENT TITLE: /DOCUMENT NUMBER:

JDOCUMENT VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT -.£. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 9 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R25-S005125V DC Dist Pnl2E (2F Switchgear Room) 1R24-S026 FR 5 DSL Bldg 1B 600/208V MCC-1B 600V Emerg Sply 2 ON Brkr Cntl 2R22-S006 FR 6 4160V Swgr Bus 2F FOR ACB's (FR 2 & 6) &

3 ON UNDV Relavinq Relaying 2R22-S006 FR 12 Dsl Bldg 1B 600/208V MCC-1B 4160V Emerg 4 ON Sply Brkr Cntl Spiv 5 2R22-S006 Fr 7 Diesel Generator 2B ACB #135570 ON 6 2R22-S024C 4KV Test Cabinet for Bus 2F ON 7 2H21-P231 4160V Bus 2F Loading LOCAILOSP Timers ON 8 2X51-RM-2F Emergency Lighting ON 2H21-P231 Diesel Genrtr 2B Test, LOSP Lockout Relays & Anti-9 ON Relav Parallel Relay 10 2H21-P201 Emerg relaying for 4160/600V Station Service Xfmr 2CD ON 11 1H21-P231 Genrtr 2B Voltage Reg Transfer ON 12 2H21-P258 Diesel Genrtr 2B Annunciators ON 2H21-P231 Diesel2A Synchro Acceptor Relay 2R43B-D2B-SA 13 ON (for ACB #135570) 2R22-S006 FR 13 600/208V MCC-2D, 2R24-S048, 4160V Breaker 14 ON Control 15 2R22-S006 FR 1 4160V Swgr Bus 2F ACB #135574 ON 16 2R22-S006 FR 11 4160V Swgr Bus 2F ACB #135564 ON Local Air Compressor Switches, 2A Diesel Generator Room Diesel 2A Air Compr 2A 1, 2R43-C005A, Control Switch AUTO Diesel 2A Air Compr 2A2, 2R43-C006A, Control Switch AUTO OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 58 of 88 P1l58 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT 2. Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 10 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R43-P001A Inside lower right cover (2A Diesel Room)

S61 Diesel 2A Unit-Parallel Switch UNIT INSTALL Diesel 2A Ammeter Shorting Block (if cover in place, cannot be SB COVER, in shorted position)

NORMAL 2R43-P003A Diesel Gen. 2A Cont. Pnl. (Indicating Lights)

BRKR 1 Control Power On Circuit - 1 (Coolant Pump 2R43-C012A) ON (Note 1)

BRKR2 Control Power On Circuit - 2 (Stby Lube Oil Pump 2R43-C008A) ON (Note 1)

BRKR3 Control Power On Circuit - 3 (Prelube Pump 2R43-C009A) ON (Note 1)

BRKR4 Control Power On Circuit - 4 (Generator Heater) ON (Note 1 2H21-P255 Switchgear 2E Room Diesel Building Diesel2C Fuel Pump 2C1, 2Y52-C001C AUTO Diesel2A Fuel Pump 2A1, 2Y52-C001A AUTO 2R26-M032A 125V DC Throw-Over Sw (next to battery charger) NOTE 2 2R22-S005 4160V Sta. Servo Swgr. 2E R'CKD IN Supply From Start-Up Aux Transf 2D (Winding #1) ACB 135554 CLOSED Frame 1 30 Amp DC Control Power Breaker CLOSED 70 Amp AC Heater Power Breaker CLOSED 30 Amp AC Heater Power Breaker CLOSED Frame 5 70 Amp DC Control Power and Undervoltage Relaying Breaker CLOSED 25 Amp DC Undervoltage Relaying Breaker CLOSED R'CKD IN Supply From Diesel Gen 2A, 2R43-S001A, ACB 135530 OPEN Frame 6 30 Amp DC Control Power Breaker CLOSED Note 1 - Circuit Breakers are located in panel 2R43-S001 C(A) Breaker Box at the West end of the Diesel Generator Skid.

Note 2 - Battery Charger Switches are normally in UP position, but IF alternate battery charger is used, switch must be in DOWN position.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 59 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM I

/DOCUMENT DOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .£ Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 11 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION R'CKD IN Supply From Start-Up Aux Transf 2C ACB 135544 OPEN Frame 10 30 Amp DC Control Power Breaker CLOSED 2R24-S02S 600/208V MCC 2A ESS. DIV. 1 (2E Swgr Room)

Frame 1A 2R25-S029 208V Section of 2R24-S025 Via Xfmr 2R11-S004 ON Frame 1BR 125V STBY Battery Charger 2H, 2R42-S032B ON Frame 1CR 125V Battery Charger 2G, 2R42-S032A ON Frame 1D Diesel Gen 2A Jacket Water Htr. 2R43-B007A ON Frame 1E Diesel Gen 2A Lube Oil Heater, 2R43-B006A ON Frame 2E Diesel Gen 2A Air Compressor 2A 1, 2R43-C005A ON Frame 2F Diesel Gen 2C Air Compressor 2C1, 2R43-C005C ON Frame 7E Diesel Gen 2A Fuel Oil Pump 2A1, 2Y52-C001A ON Frame 7F Diesel Gen 2C Fuel Oil Pump 2C1, 2Y52-C001C ON Frame 8C Diesel Gen 2A 208V Skid MTD Ckt Brkr panel Board ON 2R2S-S00412SV DC Dist. Pnl. 2D (2E Swgr Room) 2R22-S005 FR 5, 4160V Swgr Bus 2E FDR ACB's Brkr 1 ON (FR 2 3 4 5 7 8 9 & 11 Undy Undv Relay Relav Brkr 2 2R43-P001A Generator 2A Field Flashing ON Brkr 3 2R22-S005 FR 1 4160V Swgr 2E ACB #135554 ON Brkr4 2H21-P200 Generator 2A Trip Relaying ON Brkr 5 2R22-S005 FR 10 4160V Swgr 2D ACB #135544 ON Brkr 6 24R3-S001A Diesel 2A Start Control Circuit A ON Brkr 7 2R22-S005 FR 6 Diesel Genrtr 2A ACB #135530 ON 2H21-P200, Emerg Relaying for 4160/600V Station Service Brkr 8 ON Xfmr2C OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 60 of 88 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM I 34S0-R43-001-2 24.7 ATTACHMENT .£ Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 12 of 13 BREAKER NORMAL DESCRIPTION ~HECKED CHECKED VERIFIED NUMBER POSITION Brkr 9 2R43-P001A Diesel Genrtr 2A Annunciators ON Brkr 10 2R43-S001A Diesel 2A Stop & Shutdown Control ON 2H21-P230, 4160V Bus 2E Loading LOCAILOSP Timers & PSW Brkr 13 ON Valve Interlocks Brkr 14 2R43-S001A Diesel2A Governor Control ON Brkr 15 2R43-P001A, Diesel Generator 2A Manual M. O. Volt Adj ON Brkr 16 2H21-P230, Diesel Generator 2A Test & LOSP Lockout Relays ON Brkr 17 2H21-P230, Generator 2A Voltage Regulator Transfer ON Brkr 18 2R43-S001A Diesel 2A Start Control Circuit B ON Brkr 19 2H21-P230, Diesel Generator 2A Anti-Parallel Circuit ON Brkr 20 2R43-P001A, Diesel Generator 2A Auto M. O. Volt Adj ON Brkr 21 2X51-RM-2E Emergency Lighting ON 2R43-S001 A, Diesel Generator 2A Jacket Coolant Level Control Brkr 22 ON SolVlv Brkr 23 2R22-S024B 4KV Test Cabinet for Bus 2E ON Brkr 24 2H21-P255, PSW Division 1 MOV's Emergency Relaying ON 2H21-P255, Div 1 Diesel Abnormal Flow & Standby Pump Start Brkr 25 ON Relavinq Relaying 2H21-P230, Diesel2A Synchro Acceptor 2R43A-D2A-SA Brkr 27 ON (For ACB #135530) 2H21-P198, Battery Charger 2G, 2R42-S032A, STBY Bat Chr 2H, Brkr 29 ON 2R42-S032B Brkr 30 2R22-S006 FR 4 4160V Swgr Bus 2F FDR ACB's (FR 4,5, & 8) ON 2R25-S029 120/208V Dist. Pnl. 2J (2E Swgr Room) 2H21-P230, Diesel Generator 2A Genrtr Bearing & Stator Temp Brkr 14 ON Indication Brkr 23 2R24-N009A Diesel Gen Fuel Oil Day Tank 2A Water Detector ON OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 61 of 88 DOCUMENT TITLE: IDOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT .£ Att. Pg.

TITLE: DIESEL GENERATOR STANDBY AC SYSTEM ELECTRICAL LINEUP 13 of 13 BREAKER NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION FDR 1, 2R43-N007A Main Control Room Day Tank 2A Level Brkr 24 xsmitter FOR 2, 2R43-N010A Main Control Room Main Fuel Oil ON Tank Transmitter Brkr 29 2R24-S025 FR 1, Diesel Bldg MCC 2A Space Heaters ON Brkr 33 2R22-S005 FR 5, 4KV Swgr Bus 2E Space Heaters- All Frames ON Brkr 36 2R22-S005 FR 5, 4KV Swgr Bus 2E Motor Heaters ON Brkr 63* 2R20M-P001 208V Section of 2R24-S025 MCC-2A ON 2R25-S001125V DC Cabinet 2A (Control Building, 130' elevation)

Brkr 7 Unit Aux Sta Svc And Cooling Twr Xfmrs Diff Aux Relaying ON Brkr 28 LOCA Aux Relay Logic ON 2R25-S002 125V DC Cabinet 2B (Control Building, 130' elevation)

Brkr 1 Diesel Test Interlock 1H11-P653 ON Brkr 26 2H 11-P627 LOCA Aux Relay Logic ON 2R25-S036 120/208V Ess. Cab 2A (Control Building 130' el.)

2H11-P652, Dsl Gen. 2A Cooler Serv Wtr Disch Vlv, 2P41-F339A, Brkr 2 ON and Fuel Tank Lvi Indicators 2R43-R607A & R608A 2R25-S037 120/208V Ess. Cab 2B (Control Building 130' el.)

2H11-P652, Dsl Gen. 2C Cooler Serv Wtr Disch Vlv, 2P41-F339B, Brkr 31 ON and Fuel Tank Lvi Indicators 2R43-R607C & R608C

  • This switch has no automatic trip functions. Overcurrent protection is provided by fuses in Fuse paneI2R20M-P001.

OPS-0593 Ver. 4 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 62 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 1 of 8 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ I/ DATE ____

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ ___ ____ _I,/ Date _____

COM MEN TS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 63 of 88 PR63 DOCUMENT TITLE: \ DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT -2 Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 2 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Diesel Generator 2C Day Tank Room DIG 2C Day Tank 2C Inlet From Pump 2Y52-C101C Strainer, 2R43-F151C CLOSED 2R43-D020C Drain Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C101C Isolation 2R43-F017B OPEN Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C001C Isolation 2R43-F016B OPEN Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C001C Strainer, 2R43-F152C CLOSED 2R43-D021C Drain Valve DIG Day Tank Moisture Detector, 2R43-A007C, Isolation 2R43-F040 OPEN Valve 2R43-F3005C DIG Day Tank Moisture Detector, 2R43-A007C, Drain Valve CLOSED 1P52-F187 Inst. Air Isolation to DIG 2C Day Tank Room CLOSED Diesel Generator 2C Diesel Room DIG 2C Starting Air Receiver, 2R43-A005C, Pressure Switch, 2R43-F163C- OPEN 2R43-N014C Root Valve DIG 2C Starting Air Receiver, 2R43-A006C, Pressure Switch, 2R43-F162C OPEN 2R43-N013C Root Valve DIG 2C Starting Air Compressors, 2R43-C005C and COO6C, 2R43-F129C OPEN Press Switch N015C and N016C Root Valve LOCKED 2R43-F035C DIG 2C Starting Air Receiver, 2R43-A005C, Discharge Valve OPEN 2R43-F034C DIG 2C Starting Air Receiver, 2 R43-A006C , Discharge Valve LOCKED OPEN 2R43-F094C DIG 2C Starting Air Receiver, 2R43-A005C, Drain Valve CLOSED 2R43-F026C DIG 2C Starting Air Receiver, 2R43-A006C, Drain Valve CLOSED 2R43-F031C DIG 2C Starting Air Receiver, 2R43-A005C, Inlet Valve OPEN 2R43-F030C DIG 2C Starting Air Receiver, 2R43-A006C, Inlet Valve OPEN DIG 2C Starting Air Compressor, 2R43-C005C, Discharge 2R43-F090C OPEN Valve DIG 2C Starting Air Compressor, 2R43-C006C, Discharge 2R43-F091C OPEN Valve LOCKED 2R43-F037C DIG 2C Starting Air Manual Isolation Valve OPEN 2R43-F099C DIG 2C Starting Air Manual Override Valve CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 64 of 88 DOCUMENT TITLE: IDOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 3 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Diesel Generator 2C Diesel Room (Continued)

CAPPED, 2R43-F3006C DIG 2C Lube Oil Strainer, 2R43-D013C, Drain Valve CLOSED 2R43-F103C DIG 2C Lube Oil Strainer, 2R43-D013C, Drain to Sump Valve CLOSED DIG 2C Starting Air Receiver 2R43-A005C and 2R43-F171C CLOSED 2R43-A006C Test Valve DIG 2C Cooling Water Return From Turbo-Supercharger 2R43-F110C OPEN Isolation Valve DIG 2C Jacket Water Coolant Supply to Turbo-Supercharger 2R43-F106C OPEN

  1. 1 Isolation Valve 2R43-F101C DIG 2C Prelube Pump Sump Suction Isolation Valve OPEN 2R43-F078C DIG 2C Jacket Water Coolant Heat Exchanger Drain Valve CLOSED 2R43-F079C DIG 2C Jacket Water Coolant System Drain Valve CLOSED 2R43-F104C DIG 2C Lube Oil Heat Exchanger Drain Valve CLOSED DIG 2C Raw Water to Air Coolant Heat Exchanger Drain 2R43-F150C CLOSED Valve 2R43-F3002C DIG 2C Lube Oil Sump Drain Valve CLOSED DIG 2C Fuel Oil Supply Line Strainer, 2R43-D005C, Drain 2R43-F153C CLOSED Valve DIG 2C Clean Fuel Drain Tank, 2R43-A008C, Level Gauge, 2R43-F154C OPEN Uj:>Qer Isolation Valve 2R43-D001 C Upper DIG 2C Clean Fuel Drain Tank, 2R43-A008C, Level Gauge, 2R43-F155C OPEN 2R43-D001 C Lower Isolation Valve 2R43-F001C DIG 2C Clean Fuel Drain Tank, 2R43-A008C, Outlet Valve CLOSED DIG 2C Dirty Fuel Drain Tank, 2R43-A003C, Level Gauge, 2R43-F156C OPEN 2R43-D004C Upper Isolation Valve DIG 2C Dirty Fuel Drain Tank, 2R43-A003C, Level Gauge, 2R43-F157C OPEN 2R43-D004C Lower Isolation Valve 2R43-F105C DIG 2C Dirty Fuel Drain Tank, 2R43-A003C, Drain Valve CLOSED DIG 2C Jacket Water Coolant Heat Exchanger Filling Bypass 2R43-F075C CLOSED Valve 2R43-F102C DIG 2C Lube Oil Filter, 2R43-D012C, Drain Valve CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 65 of 88 PJl65 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 4 of8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION DIG 2C Jacket Water Coolant Circulating Pump, 2R43-C012C, 2R43-F069C OPEN Inlet Valve DIG 2C Jacket Water Coolant Circulating Pump, 2R43-C012C, 2R43-F070C OPEN Discharqe DischarQe Valve 2R43-F057C DIG 2C Prelube Pump, 2R43-C009C, Suction Supply Valve OPEN 2R43-F3003C DIG 2C Jacket Water Coolant Heat Exchanger Air Vent Valve CLOSED 2R43-F052C DIG 2C Air Coolant Heat Exchanger Air Vent Valve CLOSED 2R43-F051C DIG 2C Air Coolant Heat Exchanger Drain Valve CLOSED 2R43-F3004C DIG 2C Jacket Water Coolant System Drain Valve CLOSED 2R43-F049C DIG 2C Air Coolant System Drain Valve CLOSED DIG 2C Jacket Water Coolant Supply to Turbo-Supercharge 2R43-F107C OPEN

  1. 2 Isolation Valve DIG 2C Standby Lube Oil Pump, 2R43-C008C, Discharge 2R43-F055C OPEN Valve 2R43-F054C DIG 2C Standby Lube Oil Pump, 2R43-C008C, Inlet Valve OPEN DIG 2C Demineralized Water (Jacket Water Coolant) 2R43-F089C CLOSED Expansion Tank Drain Valve DIG 2C Demineralized Water (Jacket Water Coolant) 2R43-F082C CLOSED Expansion Tank Manual Fill Valve DIG 2C Upper and Lower Main Bearing Accumulator/Booster 2R43-F181C OPEN Isolation Valve Diesel Generator 2A Day Tank Room DIG 2A Day Tank 2A Inlet From Pump 2Y52-C101A Strainer, 2R43-F151A CLOSED 2R43-D020A Drain Valve DIG 2A Day Tank 2A Inlet From Pump 2Y52-C101A Isolation 2R43-F017A OPEN Valve DIG 2A Day Tank 2A Inlet From Pump 2Y52-C001A Isolation 2R43-F016A OPEN Valve DIG 2A Day Tank 2A Inlet From Pump 2Y52-C001A Strainer, 2R43-F152A CLOSED 2R43-D021A Drain Valve DIG 2A Day Tank Moisture Detector, 2R43-A007 A, Isolation 2R43-F041 OPEN Valve 2R43-F3005A DIG 2A Day Tank Moisture Detector, 2R43-A007A, Drain Valve CLOSED 1P52-F185 Inst. Air Isolation to DIG 2A Day Tank Room CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg66 Pg 66 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM I

!DOCUMENT DOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 5 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Diesel Generator 2A Room DIG 2A Starting Air Receiver, 2R43-A005A, Pressure Switch, 2R43-F163A OPEN 2R43-N014A Root Valve DIG 2A Starting Air Receiver, 2R43-A006A, Pressure Switch, 2R43-F162A OPEN 2R43-N013A Root Valve DIG 2A Starting Air Compressors, 2R43-C005A and COO6A, 2R43-F129A OPEN Press Switch N015A and N016A Root Valve LOCKED 2R43-F035A DIG 2A Starting Air Receiver, 2R43-A005A, Discharge Valve OPEN 2R43-F034A DIG 2A Starting Air Receiver, 2R43-A006A, Discharge Valve LOCKED OPEN DPEN 2R43-F094A DIG 2A Starting Air Receiver, 2R43-A005A, Drain Valve CLOSED 2R43-F026A DIG 2A Starting Air Receiver, 2R43-A006A, Drain Valve CLOSED 2R43-F031A DIG 2A Starting Air Receiver, 2R43-A005A, Inlet Valve OPEN 2R43-F030A DIG 2A Starting Air Receiver, 2R43-A006A, Inlet Valve OPEN DIG 2A Starting Air Compressor, 2R43-C005A, Discharge 2R43-F090A OPEN Valve DIG 2A Starting Air Compressor, 2R43-C006A, Discharge 2R43-F091A OPEN Valve 2R43-F037A DIG 2A Starting Air Manual Isolation Valve LOCKED OPEN DEEN 2R43-F099A DIG 2A Starting Air Manual Override Valve CLOSED 2R43-F3006A DIG 2A Lube Oil Strainer, 2R43-D013A, Drain Valve CAPPED, CLOSED 2R43-F103A DIG 2A Lube Oil Strainer, 2R43-D013A, Drain to Sump Valve CLOSED DIG 2A Cooling Water Return From Turbo-Supercharger 2R43-F110A OPEN Isolation Valve DIG 2A Jacket Water Coolant Supply to Turbo-Supercharger 2R43-F106A OPEN

  1. 1 Isolation Valve 2R43-F101A DIG 2A Prelube Pump Sump Suction Isolation Valve OPEN 2R43-F078A DIG 2A Jacket Water Coolant Heat Exchanger Drain Valve CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 67 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .l.. Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 6 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION Diesel Generator 2A Room (Continued) 2R43-F079A DIG 2A Jacket Water Coolant System Drain Valve CLOSED 2R43-F104A DIG 2A Lube Oil Heat Exchanger Drain Valve CLOSED DIG 2A Raw Water to Air Coolant Heat Exchanger Drain 2R43-F150A CLOSED Valve 2R43-F3002A DIG 2A Lube Oil Sump Drain Valve CLOSED DIG 2A Fuel Oil Supply Line Strainer, 2R43-DOOSA, Drain 2R43-F153A CLOSED Valve DIG 2A Clean Fuel Drain Tank, 2R43-AOOBA, Level Gauge, 2R43-F154A OPEN 2R43-D001A Upper Isolation Valve DIG 2A Dirty Fuel Drain Tank, 2R43-AOOBA, Level Gauge, 2R43-F155A OPEN 2R43-D001A Lower Isolation Valve 2R43-FOO1A DIG 2A Clean Fuel Drain Tank, 2R43-AOOBA, Outlet Valve CLOSED DIG 2A Starting Air Receiver 1R43-AOOSA and 2R43-AOO6A 2R43-F171A CLOSED Test Valve DIG 2A Dirty Fuel Drain Tank, 2R43-A003A, Level Gauge, 2R43-F156A OPEN 2R43-D004A Upper Isolation Valve DIG 2A Dirty Fuel Drain Tank, 2R43-A003A, Level Gauge, 2R43-F157A OPEN 2R43-D004A Lower Isolation Valve 2R43-F105A DIG 2A Dirty Fuel Drain Tank, 2R43-A003A, Drain Valve CLOSED DIG 2A Jacket Water Coolant Heat Exchanger Filling Bypass 2R43-F075A CLOSED Valve 2R43-F102A DIG 2A Lube Oil Filter, 2R43-D012A, Drain Valve CLOSED DIG 2A Jacket Water Coolant Circulating Pump, 2R43-F069A OPEN 2R43-C012A Inlet Valve DIG 2A Jacket Water Coolant Circulating Pump, 2R43-F070A OPEN DischarQe Valve 2R43-C012A Discharoe 2R43-F057A DIG 2A Prelube Pump, 2R43-C009A, Suction Supply Valve OPEN DIG 2A Jacket Water Coolant Heat Exchanger Air Vent 2R43-F3003A CLOSED Valve 2R43-F052A DIG 2A Air Coolant Heater Exchanger Air Vent Valve CLOSED 2R43-F051A DIG 2A Air Coolant Heat Exchanger Drain Valve CLOSED 2R43-F3004A DIG 2A Jacket Water Coolant System Drain Valve CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 68 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 7 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R43-F049A DIG 2A Air Coolant System Drain Valve CLOSED DIG 2A Jacket Water Coolant Supply to Turbo-Supercharge 2R43-F107A OPEN

  1. 2 Isolation Valve DIG 2A Standby Lube Oil Pump, 2R43-COOBA, Discharge 2R43-F055A OPEN Valve 2R43-F054A DIG 2A Standby Lube Oil Pump, 2R43-COOBA, Inlet Valve OPEN DIG 2A Demineralized Water (Jacket Water Coolant) 2R43-F089A CLOSED Expansion Tank Drain Valve DIG 2A Demineralized Water (Jacket Water Coolant) 2R43-F082A CLOSED Expansion Tank Manual Fill Valve DIG 2A Upper and Lower Main Bearing Accumulatorl 2R43-F181A OPEN Booster Isolation Valve Diesel Generator Building East Corridor DIG 2A Day Tank 2A Inlet From Pump 2Y52-C101A 2R43-F023 OPEN Isolation Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C101C 2R43-F022 OPEN Isolation Valve DIG 2A Day Tank 2A Inlet From Pump 2Y52-C101A Vent 2Y52-F3001 CLOSED Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C101C Vent 2Y52-F3000 CLOSED Valve DIG 2A Day Tank 2A Inlet From Pump 2Y52-COO1A 2R43-F028 OPEN Isolation Valve DIG 2C Day Tank 2C Inlet From Pump 2Y52-C001 C 2R43-F027 OPEN Isolation Valve 1P21-F141 DIG Building Demin Water Isolation Valve OPEN 2R43-A001C Dsl. FO Stor Tnk. 2C

SNC PLANT E. I. HATCH I Pg 69 of 88 Pg69 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT -l. Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C VALVE LINEUP 8 of 8 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R43-A001A Os!. FO Stor Tnk. 2A - North of Diesel Building 2R43-FOO8D Pump 2Y52-C001A Discharge to Day Tank OPEN 2R43-FOO8E Pump 2Y52-C001A Discharge to Common Header CLOSED Fuel Oil Common Header Discharge to Main Storage Tank, 2R43-FOO8C CLOSED 2Y52-A001A, Isolation Valve 2R43-FOO8B Pump 2Y52-C101A Discharge to Day Tank OPEN 2R43-FOO8A Pump 2Y52-C101A Discharge to Common Header CLOSED 1P52-F2274B Inst. Air Isolation to DIG 2A Fuel Oil Storage Tank CLOSED OPS-0594 Ver. 6.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I ~a70 Pg 70 of 88 DOCUMENT TITLE: IDOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ..1:.. Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 1 of 6 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ I/ DATE _ _ _ ___

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ _ _ _ _./,/ Date _ _ __

COM ME NTS: _

OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 71 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ..1... Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 2of6 NORMAL VALVE NUMBER DESCRIPTION CHECKED VERIFIED POSITION Diesel Generator 2C Day Tank Room DIG 2C Fuel Oil Day Tank, 2R43-A101C, Level Alarm Switch, CAPPED, 2R43-F146C 2R43-NOOSC Vent Valve CLOSED DIG 2C Fuel Oil Day Tank, 2R43-A101C, Level Alarm Switch, 2R43-F142C OPEN 2R43-NOOSC Isolation Valve DIG 2C Fuel Oil Day Tank, 2R43-A101C, Level Alarm Switch, 2R43-F143C OPEN 2R43-NOOSC Isolation Valve DIG 2C Fuel Oil Day Tank, 2R43-A101C, Level Alarm Switch, CAPPED, 2R43-F149C 2R43-NOOSC Drain Valve CLOSED DIG 2C Fuel Oil Pump, 2YS2-C101C, OnlOff Level Switch, CAPPED, 2R43-F14SC 2R43-N003C Vent Valve CLOSED DIG 2C Fuel Oil Pump, 2YS2-C101C, OnlOff Level Switch, 2R43-F140C OPEN 2R43-N003C Isolation Valve DIG 2C Fuel Oil Pump, 2YS2-C001C, OnlOff Level Switch, 2R43-F138C OPEN 2R43-N001 C Isolation Valve DIG 2C Fuel Oil Pump, 2YS2-C001 C, OnlOff Level Switch, CAPPED, 2R43-F144C 2R43-N001 C Vent Valve CLOSED DIG 2C Fuel Oil Pump, 2YS2-C001 C, OnlOff Level Switch, CAPPED, 2R43-F147C 2R43-N001 C Drain Valve CLOSED DIG 2C Fuel Oil Pump, 2YS2-C001 C, OnlOff Level Switch, 2R43-F139C OPEN 2R43-N001 C Isolation Valve DIG 2C Fuel Oil Pump, 2YS2-C1 01 C, OnlOff Level Switch, 2R43-F141C OPEN 2R43-N003C Isolation Valve DIG 2C Fuel Oil Pump, 2YS2-C101C, OnlOff Level Switch, CAPPED, 2R43-F148C 2R43-N003C Drain Valve CLOSED Diesel Generator 2C Room DIG 2C Starting Air Compressor, 2R43-COOSC, Pressure 2R43-F1S8C OPEN Switch 2R43-N01SC Isolation Valve DIG 2C Starting Air Compressor, 2R43-COOSC, Pressure 2R43-F1S9C CLOSED Switch 2R43-N01SC Vent Valve DIG 2C Starting Air Compressor, 2R43-C006C, Pressure 2R43-F160C OPEN Switch 2R43-N016C Isolation Valve DIG 2C Starting Air Compressor, 2R43-C006C, Pressure 2R43-F161C CLOSED Switch 2R43-N016C Vent Valve DIG 2C Starting Air Receiver, 2R43-AOOSC, Pressure Switch, 2R43-N014C-IV1 OPEN 2R43-N014C Isolation Valve DIG 2C Starting Air Receiver, 2R43-AOOSC, Pressure Switch, 2R43-N014C-W1 CLOSED 2R43-N014C Vent Valve DIG 2C Starting Air Receiver, 2R43-AOOSC, Pressure 2R43-ROO4C-IV1 OPEN Indicator 2R43-R004C Isolation Valve DIG 2C Starting Air Receiver, 2R43-AOOSC, Pressure 2R43-ROO4C-W1 CLOSED Indicator 2R43-R004C Vent Valve OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 72 of 88 DOCUMENT TITLE: IDOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT .+/-- Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 30f6 NORMAL VALVE NUMBER DESCRIPTION CHECKED VERIFIED POSITION DIG 2C Starting Air Receiver, 2R43-A006C, Pressure Switch, 2R43-N013C-IV1 OPEN 2R43-N013C Isolation Valve DIG 2C Starting Air Receiver, 2R43-A006C, Pressure Switch,

~R43-N013C-W1 2R43-N013C-W1 CLOSED 2R43-N013C Vent Valve DIG 2C Starting Air Receiver, 2R43-A006C, Pressure Indicator, 2R43-ROO3C-IV1 OPEN 2R43-R003C Isolation Valve DIG 2C Starting Air Receiver, 2R43-A006C, Pressure Indicator, 2R43-ROO3C-W1 CLOSED 2R43-R003C Vent Valve DIG 2C Lube Oil Pressure Backup Start Failure Pressure Switch, 2R43-N019C-IV1 OPEN 2R43-N019C Isolation Valve DIG 2C Lube Oil Pressure Backup Start Failure Pressure Switch, 2R43-N020C-IV1 OPEN 2R43-N020C Isolation Valve DIG 2C Low Lube Oil Pressure Alarm Pressure Switch, 2R43-N021 C-IV1 OPEN 2R43-PS-N021C Isolation Valve DIG 2C Low Lube Oil Pressure Shutdown Pressure Switch, 2R43-N022C-IV1 OPEN 2R43-PS-N022C Isolation Valve DIG 2C Low Lube Oil Pressure Shutdown Pressure Switch, 2R43-N035C-IV1 OPEN 2R43-PS-N035C Isolation Valve DIG 2C Lube Oil Strainer, 2R43-D013C, Pressure Indicator, 2R43-F166C OPEN 2R43-R014C Isolation Valve DIG 2C Lube Oil Filter, 2R43-D012C, Pressure Indicator, 2R43-F165C OPEN 2R43-R013C Isolation Valve DIG 2C Lube Oil Pressure Indicator, 2R43-R016C (2R43-POO3C),

2R43-F114C OPEN Isolation Valve DIG 2C Crankcase Pressure Switch, 2R43-N027C, and Indicator, 2R43-F164C OPEN 2R43-R015C Isolation Valve DIG 2C Raw Water Press Low Aim Sw, 2R43-N017C, and 2R43-F053C OPEN Indicator 2R43-R010C Isolation Valve DIG 2C Fuel Oil Pressure to Filter Black Hand (BH) Indicator, 2R43-F021C OPEN 2R43-R001 C Isolation Valve DIG 2C Fuel Oil Pressure to Engine Red Hand(RH) 2R43-F020C Indicator, 2R43-R001 C, and Low Pressure Switch, OPEN 2R43-N012C, Isolation Valve DIG 2C (Jacket) Coolant Press Shutdown Pressure Switch, 2R43-F074C OPEN 2R43-N032C Isolation Valve DIG 2C Jacket Water Pressure Indicator, 2R43-R018C, Isolation 2R43-F073C OPEN Valve DIG 2C Scavenging Air Pressure Indicator, 2R43-ROO6C, 2R43-ROO6C-IV1 OPEN Isolation Valve DIG 2C Starting Air Pressure Indicator, 2R43-R002C, Isolation 2R43-F167C OPEN Valve OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 73 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .+/-. Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 4of6 VALVE NORMAL DESCRIPTION CHECKED VERIFIED NUMBER POSITION 2R43-A001 C Dsl FO Stor Tnk. 2C-North of Diesel Building DIG Fuel Oil Transfer Pump, 2YS2-C001 C, Discharge 2Y52-F061 CLOSED Pressure Gauoe GauQe 2YS2-R012 Isolation DIG Fuel Oil Transfer Pump, 2YS2-C1 01 C, Discharge 2Y52-F063 CLOSED Pressure GauQe 2YS2-R014 Isolation Gauoe Diesel Generator 2A Day Tank Room DIG 2A Fuel Oil Day Tank, 2R43-A101A, Level Alarm Switch, CAPPED, 2R43-F146A 2R43-NOOSA Vent Valve ClOSED CLOSED DIG 2A Fuel Oil Day Tank, 2R43-A101A, Level Alarm Switch, 2R43-F142A OPEN 2R43-NOOSA Isolation Valve DIG 2A Fuel Oil Day Tank, 2R43-A101A, Level Alarm Switch, 2R43-F143A OPEN 2R43-NOOSA Isolation Valve DIG 2A Fuel Oil Day Tank, 2R43-A101A, Level Alarm Switch, CAPPED, 2R43-F149A 2R43-NOOSA Drain Valve ClOSED CLOSED DIG 2A Fuel Oil Pump, 2YS2-C101A, OnlOff Level Switch, CAPPED, 2R43-F145A 2R43-N003A Vent Valve r.IO~FD ClOSED DIG 2A Fuel Oil Pump, 2YS2-C101A, OnlOff Level Switch, 2R43-F140A OPEN 2R43-N003A Isolation Valve DIG 2A Fuel Oil Pump, 2YS2-C001A, OnlOff Level Switch, 2R43-F138A OPEN 2R43-N001A Isolation Valve DIG 2A Fuel Oil Pump, 2YS2-C001A, OnlOff Level Switch, CAPPED, 2R43-F144A 2R43-N001A Vent Valve r.IO~F.D ClOSED DIG 2A Fuel Oil Pump, 2YS2-C001A, OnlOff Level Switch, 2R43-F139A OPEN 2R43-N001A Isolation Valve DIG 2A Fuel Oil Pump, 2YS2-C001A, OnlOff Level Switch, CAPPED, 2R43-F147A 2R43-N001A Drain Valve CLOSED DIG 2A Fuel Oil Pump, 2YS2-C101A, OnlOff Level Switch, 2R43-F141A OPEN 2R43-N003A Isolation Valve DIG 2A Fuel Oil Pump, 2YS2-C101A, OnlOff Level Switch, CAPPED, 2R43-F148A r.1()~I=D rl(,)<:::'~D 2R43-N003A Drain Valve Diesel Generator 2A Room DIG 2A Starting Air Compressor, 2R43-COOSA, Pressure 2R43-F158A OPEN Switch 2R43-N01SA Isolation Valve DIG 2A Starting Air Compressor, 2R43-COOSA, Pressure 2R43-F159A CLOSED Switch 2R43-N01SA Vent Valve DIG 2A Starting Air Compressor, 2R43-C006A, Pressure 2R43-F160A OPEN Switch 2R43-N016A Isolation Valve DIG 2A Starting Air Compressor, 2R43-C006A, Pressure 2R43-F161A CLOSED Switch 2R43-N016A Vent Valve OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E.I. HATCH I Pg 74 of 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM ATTACHMENT A.

I 34S0-R43-00 1-2 24.7 Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 50f6 NORMAL VALVE NUMBER DESCRIPTION CHECKED VERIFIED POSITION DIG 2A Starting Air Receiver, 2R43-AOOSA, Pressure 2R43-N014A-IV1 OPEN Switch 2R43-N014A Isolation Valve DIG 2A Starting Air Receiver, 2R43-AOOSA, Pressure 2R43-N014A-W1 CLOSED Switch 2R43-N014A Vent Valve DIG 2A Starting Air Receiver, 2R43-AOOSA, Pressure 2R43-ROO4A-IV1 OPEN Indicator 2R43-R004A Isolation Valve DIG 2A Starting Air Receiver, 2R43-AOOSA, Pressure 2R43-ROO4A-W1 CLOSED Indicator 2R43-R004A Vent Valve DIG 2A Starting Air Receiver, 2R43-A006A, Pressure 2R43-N013A-IV1 OPEN Switch 2R43-N013A Isolation Valve DIG 2A Starting Air Receiver, 2R43-A006A, Pressure 2R43-N013A-W1 CLOSED Switch 2R43-N013A Vent Valve DIG 2A Starting Air Receiver, 2R43-A006A, Pressure 2R43-ROO3A-IV1 OPEN Indicator 2R43-R003A Isolation Valve DIG 2A Starting Air Receiver, 2R43-A006A, Pressure 2R43-ROO3A-W1 CLOSED Indicator 2R43-R003A Vent Valve DIG 2A Lube Oil Pressure Backup Start Failure 2R43-N019A-IV1 OPEN Pressure Switch 2R43-N019A Isolation Valve DIG 2A Lube Oil Pressure Backup Start Failure 2R43-N020A-IV1 OPEN Pressure Switch 2R43-N020A Isolation Valve DIG 2A Low Lube Oil Pressure Alarm Pressure Switch, 2R43-N021A-IV1 OPEN 2R43-PS-N021A Isolation Valve DIG 2A Low Lube Oil Pressure Shutdown Pressure 2R43-N022A-IV1 OPEN Switch 2R43-PS-N022A Isolation Valve DIG 2A Low Lube Oil Pressure Shutdown Pressure 2R43-N03SA-IV1 OPEN Switch 2R43-PS-N03SA Isolation Valve DIG 2A Lube Oil Strainer, 2R43-D013A, Pressure 2R43-F166A OPEN Indicator 2R43-R014A Isolation Valve DIG 2A Lube Oil Filter, 2R43-D012A, Pressure Indicator, 2R43-F173A CLOSED 2R43-R013A Isolation Valve DIG 2A Lube Oil Filter, 2R43-D012A, Pressure Indicator, 2R43-F16SA CLOSED 2R43-R013A Isolation Valve DIG 2A Lube Oil Pressure Indicator, 2R43-R016A 2R43-F114A OPEN (2R43-P003A), Isolation Valve DIG 2A Crankcase Pressure Switch, 2R43-N027A, and 2R43-F164A OPEN Indicator 2R43-R01SA Isolation Valve DIG 2A Raw Water Press Low Aim Sw, 2R43-N017A, 2R43-FOS3A OPEN and Indicator 2R43-R010A Isolation Valve DIG 2A Fuel Oil Pressure to Filter Black Hand(BH) 2R43-F021A OPEN Indicator 2R43-R001A Isolation Valve OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 75 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM 1IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT A. Att. Pg.

TITLE: DIESEL GENERATOR 2A AND 2C INSTRUMENT VALVE LINEUP 6of6 NORMAL VALVE NUMBER DESCRIPTION CHECKED VERIFIED POSITION DIG 2A Fuel Oil Pressure to Engine Red Hand(RH) Indicator, 2R43-F020A 2R43-R001A, and Low Pressure Switch, 2R43-N012A, Isolation OPEN Valve DIG 2A (Jacket) Coolant Press Shutdown Pressure Switch, 2R43-F074A OPEN 2R43-N032A, Isolation Valve DIG 2A Jacket Water Pressure Indicator, 2R43-R018A, 2R43-F073A OPEN Isolation Valve DIG 2A Scavenging Air Pressure Indicator, 2R43-ROO6A, 2R43-ROO6A-IV1 OPEN Isolation Valve DIG 2A Starting Air Pressure Indicator, 2R43-R002A, Isolation 2R43-F167A OPEN Valve 2R43-A001A Os!. FO Stor Tnk. 2A-North of Diesel Building DIG Fuel Oil Transfer Pump, 2Y52-C001A, Discharge Pressure 2Y52-F060 CLOSED Gauge, 2Y52-R011, Isolation DIG Fuel Oil Transfer Pump, 2Y2-C101A, Discharge Pressure 2Y52-F062 CLOSED Gauge, 2Y52-R013, Isolation OPS-0595 Ver. 6 G16.030 MGR-0009 Ver. 4

HATCHLI SNC PLANT E. I. HATCH Pg 76 of 88 DOCUMENT TITLE: jDOCUMENT

/DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ...Q.. Att. Pg.

TITLE: DIESEL GENERATOR FUEL OIL SYSTEM RESTORATION 1 of 4 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ I DATE _ _ __

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ _ _ _....:/ -'1 Date _____

COMMENTS: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __

OPS-0598 Ver. 1.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 77 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT 2 Att. Pg.

TITLE: DIESEL GENERATOR FUEL OIL SYSTEM RESTORATION 2of4 Circle corresponding tank enclosure for tank selected.

1R43-A002A 1R43-A002B 1R43-A002C 2Y52-A001A 2Y52-A001C NORMAL MPL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 1H21-P256 Switchgear 1G Room Diesel Building 1Y52-C101A Diesel Gen 1A Fuel Pump 1A2 AUTO 1Y52-C101B Diesel Gen 1B Fuel Pump 1B2 AUTO 1Y52-C101C Diesel Gen 1C Fuel Pump 1C2 AUTO 1H21-P255 Switchgear 1E Room Diesel Building 1Y52-COO1A Diesel Gen 1A Fuel Pump 1A1 AUTO 1Y52-COO1B Diesel Gen 1B Fuel Pump 1B1 AUTO 1Y52-COO1C Diesel Gen 1C Fuel Pump 1C1 AUTO 2H21-P256 Switchgear 2G Room Diesel Building 2Y52-C101A Diesel 2A Fuel Pump 2A2 AUTO 2Y52-C101C Diesel 2C Fuel Pump 2C2 AUTO 2H21-P255 Switchgear 2E Room Diesel Building 2Y52-COO1A Diesel 2A Fuel Pump 2A 1 AUTO 2Y52-COO1C Diesel 2C Fuel Pump 2C1 AUTO 1R43-A002A Dsl FO Stor Tnk 1A - North of Diesel Building 1R43-FOO6A Pump 1Y52-C101A Discharge to Common Header CLOSED 1R43-FOO6B Pump 1Y52-C101A Discharge to Day Tank OPEN Fuel Oil Pumps, 1Y52-C101A and 1Y52-C001A, Return to 1R43-FOO6C CLOSED Storaqe Tank Isolation Valve 1R43-FOO6D Pump 1Y52-C001A Discharge to Day Tank OPEN 1R43-FOO6E Pump 1Y52-C001A Discharge to Common Header CLOSED OPS-0598 Ver. 1.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 78 of 88 Pg78 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT Ji. Att. Pg.

TITLE: DIESEL GENERATOR FUEL OIL SYSTEM RESTORATION 30f4 NORMAL MPL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 1R43-A002B Dsl FO Stor Tnk 1B - North of Diesel Building 1R43-FOOSA Pump 1Y52-C101B Discharge to Common Header CLOSED 1R43-FOOSB Pump 1Y52-C1 01 B Discharge to Day Tank OPEN Fuel Oil Pumps, 1Y52-C101B and 1Y52-C001B, Return to 1R43-FOOSC CLOSED Storaae Tank Isolation Valve 1R43-FOOSD Pump 1Y52-C001 B Discharge to Day Tank OPEN 1R43-FOOSE Pump 1Y52-C001 B Discharge to Common Header CLOSED 1R43- A002C Dsl FO Stor Tnk 1C - North of Diesel Building I

1R43-F004A Pump 1Y52-C101C Discharge to Common Header CLOSED 1R43-F004B Pump 1Y52-C101C Discharge to Day Tank OPEN Fuel Oil Pumps, 1Y52-C101C and 1Y52-C001C, Return to 1R43-F004C CLOSED Storaae Tank Isolation Valve 1R43-F004D Pump 1Y52-C001C Discharge to Day Tank OPEN 1R43-F004E Pump 1Y52-C001 C Discharge to Common Header CLOSED 2Y52-A001 C Dsl FO Stor Tnk 2C - North of Diesel Building 2R43-F007A Pump 2Y52-C101C Discharge to Common Header CLOSED 2R43-F007B Pump 2Y52-C1 01 C Discharge to Day Tank OPEN Fuel Oil Common Header Discharge to Main Storage Tank, 2R43-F007C CLOSED 2Y52-A001 C Isolation Valve 2R43-F007D Pump 2Y52-C001 C Discharge to Day Tank OPEN 2R43-F007E Pump 2Y52-C001 C Discharge to Common Header CLOSED OPS-OS98 Ver. 1.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 79 of 88 DOCUMENT TITLE: I DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT 2 Att. Pg.

TITLE: DIESEL GENERATOR FUEL OIL SYSTEM RESTORATION 4of4 NORMAL MPL NUMBER DESCRIPTION CHECKED VERIFIED POSITION 2Y52-A001A Dsl FO Stor Tnk 2A - North of Diesel Building 2R43-F008A Pump 2Y52-C101A Discharge to Common Header CLOSED 2R43-F008B Pump 2Y52-C101A Discharge to Day Tank OPEN Fuel Oil Common Header Discharge to Main Storage Tank, 2R43-F008C CLOSED 2Y52-A001A, Isolation Valve 2R43-F008D Pump 2Y52-C001A Discharge to Day Tank OPEN 2R43-F008E Pump 2Y52-C001A Discharge to Common Header CLOSED OPS-0598 Ver. 1.1 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 80 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR FUEL QUANTITY CHECK 1 of 5 PERSON(S) PERFORMING OR VERIFYING LINEUP (PRINT NAME) INITIALS LINEUP COMPLETED: TIME _ _ _ _ I DATE _ _ __

REVIEWED BY: SS _ _ _ _ _ _ _ _ _ _ _ _ Time _ _ _ _I Date _ _ __

COMMENTS: __ __ __

OPS-1379 Ver. 0.3 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 81 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT ..Q.. Att. Pg.

TITLE: DIESEL GENERATOR FUEL QUANTITY CHECK 2 of 5 1.0 DIESEL GENERATOR MAIN FUEL OIL STORAGE TANK QUANTITY 1.1 IF 2R43-R608A,C, Control Room Main Tk Indicators, are operable, record its reading on step 1.3.

1.2 IF Control Room Indicators are inop, THEN at Diesel Generator 2A, ANDIOR 2C Storage Tanks, with a dipstick, measure the level of the tank and then record the level.

Diesel Generator 2A Storage Tank Level _ _ _ _ _ _ inches Diesel Generator 2C Storage Tank Level inches 1.3 IF step 1.2 was performed convert the storage tank level from inches to gallons utilizing Table 1, Diesel Generator Main Fuel Oil Storage Tank; AND record the quantities:

Diesel Generator 2A Storage Tank Level _ _ _ _ _ _ gallons Diesel Generator 2C Storage Tank Level gallons

  • Even with control room level indicators operable, tank level determination for 1A, 1B, or 1C fuel oil storage tanks will be determined using the appropriate measuring stick whenever addition of fuel oil involving these storage tanks is required or whenever approaching the lower Administrative limit of 35,000 gallons. These actions will minimize the possibility of tank overflow and ensure the accuracy of the storage tank quantity when approaching respective limits.

NOTES:

  • Operable Control Room Main Tank indicators for 1A, 1B, or 1C DIG's may be used for tank level determination OR level trending during DIG surveillances OR fuel oil transfers provided:
  • Tank level is not approaching the lower Admin limit of 35,000 gallons.
  • No addition of fuel oil to storage tank has occurred.

1.4 IF 1R43-R901A, 1R43-R901B, 1R43-R901C, Control Room Main Tk Indicators are operable AND tank level is not approaching the lower Administrative limit of 35,000 gallons OR increased due to fuel oil addition, record its reading on step 1.6.

OPS-1379 Ver. 0.3 G16.030 MGR-0009 Ver. 4

HATCH I SNC PLANT E. I. HATCHl Pg 82 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT .&... Att. Pg.

TITLE: DIESEL GENERATOR FUEL QUANTITY CHECK 3 of 5 1.5 IF Control Room Indicators are inoperable OR otherwise required, THEN at Diesel Generator 1A, 1B, AND/OR 1C Storage Tanks, with a dipstick, measure the level of the tank and then record the level.

Diesel Generator 1A Storage Tank Level _ _ _ _ _ _ inches Diesel Generator 1B Storage Tank Level inches Diesel Generator 1C Storage Tank Level inches 1.6 IF step 1.5 was performed convert the storage tank level from inches to gallons utilizing Table 1, Diesel Generator Main Fuel Oil Storage Tank; AND record the quantities:

Diesel Generator 1A Storage Tank Level _ _ _ _ _ _ gallons Diesel Generator 1B Storage Tank Level gallons Diesel Generator 1C Storage Tank Level gallons 1.7 Add Diesel Generator Storage Tanks 1A, 1B, 1C, 2A, and 2C quantities AND record the total inventory.

Total Inventory _ _ _ _ _ _ _ _ _ _ _ _ gallons 1.8 Verify the conversion of level to quantities AND the computation of total inventory.

LlC OPER 1.9 Confirm the total inventory of Fuel Oil Storage Tanks 1A, 1B, 1C, 2A, and 2C is >175,000 gallons.

1.10 Confirm the level in each of the following Fuel Oil Storage Tanks is > 35,000 gallons (TS.::: 33,320 gallons).

TANK MPL 2Y52-A001A 2Y52-A001C 1R43-A002B 1R43-A002A 1R43-A002C 1.11 IF any Main Fuel Oil Storage Tank level is less than 35,000 gallons (TS .::: 33,320 gallons), AND/OR the total inventory of fuel oil is< 175,000 gallons, THEN notify the Unit 1 Shift Supervisor.

OPS-1379 Ver. 0.3 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 83 of 88 DOCUMENT TITLE: /DOCUMENT NUMBER:

jDOCUMENT VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT ~ Att. Pg.

TITLE: DIESEL GENERATOR FUEL QUANTITY CHECK 4of5 40fS 2.0 DIESEL GENERATOR DAY TANK LEVEL QUANTITY 2.1 RECORD Diesel Generator Day Tank levels at paneI2H11-P652.

paneI2H11-P6S2.

DIG, DAY TANK 2R43-R607A, 2Y52-A 101 A, Level. _ _ _ _ _ _ gallons 2R43-R607C, 2Y52-A 101 C, Level,_ _ _ _ _ _ gallons 2.2 RECORD Diesel Generator Day Tank levels at paneI1H11-P652.

DIG. DAY TANK 1R43-R900A, 1R43-A001A, Level _ _ _ _ _ gallons 1R43-R900B, 1R43-A001 B, Level _ _ _ _ _ gallons 1R43-R900C, 1R43-A001 C, Level _ _ _ _ _ gallons 2.3 Confirm that the level in each of the following Fuel Oil Day Tanks is greater than 900 gallons (TS ~ 500 SOO gallons).

DIG, DAY TANK 2R43-R607A,2Y52-A101A 2R43-R607C, 2Y52-A 101 C 1R43-R900A, 1R43-A001A 1R43-R900B, 1R43-A001 B 1R43-R900C, 1R43-A001 C 2.4 IF any Diesel Generator Day Tank level is less than 900 gallons THEN notify the Unit 1 Shift Supervisor.

OPS-1379 Ver. 0.3 G16.030 MGR-0009 Ver. 4

SNC PLANT E.I. HATCH I Pg 84 of 88 DOCUMENT TITLE: /DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM 34S0-R43-001-2 24.7 ATTACHMENT .Q.. Att. Pg.

TITLE: DIESEL GENERATOR FUEL QUANTITY CHECK 5 of 5 TABLE 1 DEPTH GALLONS DEPTH GALLONS DEPTH GALLONS DEPTH GALLONS (INCHES) (INCHES) (INCHES) (INCHES) 0.00 -- - -- 40.00 8073.17 80.00 22861.01 120.00 36413.75 1.00 -- - -- 41.00 8411.77 81.00 23238.94 121.00 36683.49 2.00 -- - -- 42.00 8753.13 82.00 23616.04 122.00 36948.04 3.00 -- - -- 43.00 9097.77 83.00 23992.20 123.00 37207.22 4.00 -- - -- 44.00 9443.73 84.00 24367.34 124.00 37460.81 5.00 -- - -- 45.00 9792.77 85.00 24741.37 125.00 37708.61 6.00 -- - -- 46.00 10144.19 86.00 25114.21 126.00 37950.36 7.00 -- - -- 47.00 10497.89 87.00 25485.75 127.00 38185.83 8.00 -- - -- 48.00 10853.77 88.00 25855.92 128.00 38414.73 9.00 -- - -- 49.00 11211.75 89.00 26224.62 129.00 38636.75 10.00 -- - -- 50.00 11571.73 90.00 26591.77 130.00 38851.57 11.00 0 51.00 11933.63 91.00 26957.28 131.00 39058.81 12.00 199.01 52.00 12297.36 92.00 27321.05 132.00 39258.07 13.00 406.02 53.00 12662.82 93.00 27682.99 133.00 39448.88 14.00 620.63 54.00 13029.92 94.00 28043.03 134.00 39630.67 15.00 842.47 55.00 13398.59 95.00 28401.06 134.5 39717.98 16.00 1071.19 56.00 13768.71 96.00 28756.99 135.00 39735.83 17.00 1306.51 57.00 14140.22 97.00 29110.73 136.00 39769.95 18.00 1548.12 58.00 14513.02 98.00 29462.20 137.00 39801.82 19.00 1795.79 59.00 14887.01 99.00 29811.29 138.00 39831.24 20.00 2049.27 60.00 15262.12 100.00 30157.92 139.00 39857.93 21.00 2308.34 61.00 15638.25 101.00 30501.98 140.00 39881.54 22.00 2572.80 62.00 16015.32 102.00 30843.39 141.00 39901.53 23.00 2842.44 63.00 16393.23 103.00 31182.05 142.00 39917.00 24.00 3117.10 64.00 16771.90 104.00 31517.85 143.00 39925.44 25.00 3396.59 65.00 17151.23 105.00 31850.71 26.00 3680.75 66.00 17531.15 106.00 32180.51 27.00 3969.43 67.00 17911.55 107.00 32507.15 28.00 4262.48 68.00 18292.36 108.00 32830.53 29.00 4559.75 69.00 18673.48 109.00 33150.54 30.00 4861.12 70.00 19054.82 110.00 33467.07 31.00 5166.45 71.00 19436.30 111.00 33780.00 32.00 5475.62 72.00 19817.83 112.00 34089.22 33.00 5788.50 73.00 20199.31 113.00 34394.60 34.00 6104.98 74.00 20580.66 114.00 34696.03 35.00 6424.93 75.00 20961.79 115.00 34993.36 36.00 6748.26 76.00 21342.60 116.00 35286.47 37.00 7074.86 77.00 21723.03 117.00 35575.22 38.00 7404.61 78.00 22102.96 118.00 35859.45 39.00 7737.41 79.00 22482.31 119.00 36139.02 OPS-1379 Ver. 0.3 G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 85 of 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM ATTACHMENT .L I 34S0-R43-001-2 24.7 Att. Pg.

TITLE: DG LOSP LOCA LOGIC DIAGRAM 1 of 1

.1~1--------lMwOde 1 Relay:y..asL-_ _ _...II.~

. .------'MiILIOde Select Switch Aux Relay:,..sL-_

... ----i..... ___--'-'LOSP

..-__ ~LOSP Lock Out Rela...,YJ--_

RelaQ,YJ--_ _-+l~1

~~1

~loseonBu~

~ Undervoltage ~

ClosedWhe 27EAX2 27EAX3 Closed When SUT2D 230KV TX I 27EBX2 I

27EBX3

/ ' SUT 2C """ Closed

'"""--- Close on Bus ~

/' Energized "'" --..undervoltag~

27CAX2 27CAX1 DG Test, SUT2C,OOS Interlock Switch T6A T6B TDC TDC 1 sec A1X2} Opens 125VDC on a Cab2D LOCA A1X3 2R25-S004 Closed DG Mode Switch in TEST 43A1 86 } Normally Closed When 86 Reset LOSP Lockout Relay (Normally De-energized) 43A1X 1. Allows closure of normal supply breaker if DG supply breaker is closed and no LOSP is sensed.

2. Locks out emergency start ckt. "A".
3. Prevents auto closure of DG supply breaker.
4. Prevents auto closure of alternate supply breaker.
5. Allows auto or manual closure of alternate supply when DG breaker is closed if no LOSP is sensed.

43A1X1 1. Locks out emergency start ckt. "B".

2. Locks out anti-parallel relay trip.
3. Allows manual closure of DG supply breaker.
4. Arms additional DG trips (OTH, CTH, CPL, CCPH).

43A1X2 1. Prevents load shed when DG supply breaker is closed.

43A1X3 1. Cannot manually close SUT supplies for 4160V A, B, C, & D busses if any DG is in "TEST" mode.

2. Alternate supply for 4160V A, B, C, & D buses will not fast transfer if alternate supply to emergency bus is closed.

G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 86 of 88 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION No:

DIESEL GENERATOR STANDBY AC SYSTEM ATTACHMENT JL I 34S0-R43-001-2 24.7 Att. Pg.

TITLE: JUMPER INSTALLATION FOR DG SHUTDOWN WITH AUTOSTART 1 of 1 SIGNAL PRESENT CAUTION: The installation of the jumper(s) in the following step removes auto start capability for the affected DG.

1.0 For the DG (2A, 2C) to be shut down, install a jumper at the listed location between the listed links:

Diesel Generator Place jumper between links 2A 2H11-P652 TB10 P16 (pt. 18) to A1X1 (pt. 14) 2C 2H11-P652 TB4 P16 (pt. 8) to C1X1 (pt. 5) 1.1 Return to subsection 7.3.2.

1.2 When directed by the SS, remove the jumper(s) for the affected DG(s).

Diesel Generator Remove jumper between links 2A 2H11-P652 TB10 P16 (pt. 18) to A1X1 (pt. 14)

Ind. Verified 2C 2H11-P652 TB4 P16 (pt. 8) to C1X1 (pt. 5)

Ind. Verified 1.3 Return to subsection 7.3.2.

OPS-1812 Ver. N/A G16.030 MGR-0009 Ver. 4

HATCH I SNC PLANT E. I. HATCH] Pg 87 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM I

!DOCUMENT DOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT JL Att. Pg.

TITLE:

SUMMARY

OF SPEED & VOLTAGE ADJUSTMENTS 1 of 1 Summary of Speed & Voltage Adjustments Speed Increase Voltage Increase DG KW ~ KW ~

Alone (Supplying VARs ~ VARs ~

the Bus) Bus Freq t Bus Volt t DG KW t KW ~

Paralleled VARs t VARs t Bus Freq ~ Bus Volt ~

G16.030 MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 88 of 88 DOCUMENT TITLE:

DIESEL GENERATOR STANDBY AC SYSTEM IDOCUMENT NUMBER:

34S0-R43-001-2 VERSION No:

24.7 ATTACHMENT J..Q.. Att. Pg.

TITLE: REPLACEMENT OF MAIN STORAGE TANK(S) PIPE CAP AFTER 1 of 1 FUEL OIL TRANSFER PROCEDURE DESCRIPTION INITIAL VERIF.

STEP REPLACE AND VERIFY 2W' pipe cap(s) on Main Storage Tank(s) to which fuel oil was transferred. Indicate associated Main Storage Tank(s) below:

0 1R43-A002C 0 1R43-A002B 7.4.1.9.2 OR 7.4.1.10 0 1R43-A002A o 2Y52-A001A o 2Y52-A001 C PERSONS PERFORMING I VERIFYING PIPE CAP REPLACEMENT (PRINT NAME) INITIALS OPS-1924 Ver. N/A G16.030 MGR-0009 Ver. 4

Edwin I. Hatch Nuclear Plant Enclosure 6 34AB-R22-003-2 Station Blackout

SOUTHERN NUCLEAR DOCUMENT TYPE: PAGE PLANT E. I. HATCH ABNORMAL OPERATING PROCEDURE 1 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 EXPIRATION APPROVALS: EFFECTIVE DATE: DEPARTMENT MGR J. I. Hammonds DATE 10/17/03 DATE:

N/A NPGM/POAGM/PSAGM 8-26-08 N/A DATE N/A TABLE OF CONTENTS Section Page UNIT 2 STATION BLACKOUT - - UNIT 2 ACTIONS .................................................................... 2 1.0 CONDITIONS ........................................................................................................................ 2 2.0 AUTOMATIC ACTIONS ......................................................................................................... 2 3.0 IMMEDIATE OPERATOR ACTIONS ..................................................................................... 2 4.0 SUBSEQUENT OPERATOR ACTIONS ................................................................................ 3 UNIT 1 STATION BLACKOUT - - UNIT 2 ACTIONS .................................................................... 7 1.0 CONDITIONS ........................................................................................................................ 7 2.0 AUTOMATIC ACTIONS ......................................................................................................... 7 3.0 IMMEDIATE OPERATOR ACTIONS ..................................................................................... 7 4.0 SUBSEQUENT OPERATOR ACTIONS ................................................................................ 7 5.0 RESTORATION ..................................................................................................................... 7 Attachments 1 ADDITIONAL ANNUNCIATORS ........................................................................................... 8 2 600V NON-ESSENTIAL LOADS ........................................................................................... 9 3 DIESEL GENERATOR 2A ................................................................................................... 10 4 DIESEL GENERATOR 1B ................................................................................................... 12 5 DIESEL GENERATOR 2C ................................................................................................... 16 6 ELECTRICAL LOADS FROM 600V BUS 2C ...................................................................... 18 7 ADDITIONAL ELECTRICAL LOADS FROM 600V BUS 2C .............................................. ,. 19 8 ELECTRICAL LOADS FROM 600V BUS 2D ...................................................................... 20 9 ADDITIONAL ELECTRICAL LOADS FROM 600V BUS 2D ................................................ 21 10 LOCAL MANUAL OPERATION OF RCIC ........................................................................... 22 MGR-0002 Ver. 8

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 20F24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 I. UNIT 2 STATION BLACKOUT - - UNIT 2 ACTIONS 1.0 CONDITIONS 1.1 At least two of the three Unit 2 4160V AC Emergency busses 2E, 2F, AND 2G are de-energized, as indicated by TRIPPED breaker indications AND extinguished bus pot lights on 2H11-P652.

1.2 ANNUNCIATORS

LOSS OF OFFSITE POWER (4160V AC 2E), 652-102 LOSS OF OFFSITE POWER (4160V AC 2F), 652-202 LOSS OF OFFSITE POWER (4160V AC 2G), 652-302 4160V BUS 2E VOLTAGE LOW, 652-122 4160V BUS 2F VOLTAGE LOW, 652-222 4160V BUS 2G VOLTAGE LOW, 652-322 1.2.1 See Attachment 1 for additional annunciators.

2.0 AUTOMATIC ACTIONS 2.1 Unit 2 reactor will SCRAM, IF both Reactor Protection System (RPS) busses are de-energized.

2.2 Unit 1 reactor will SCRAM, IF Loss of Offsite Power extends to Unit 1.

2.3 MSIV's will CLOSE, lE both RPS busses are de-energized.

2.4 Diesel Generator for affected bus will AUTO start.

2.5 HPCI AND RCIC will AUTO start on Low-Low reactor water level.

2.6 Diesel MCC 1B (1R24-S026) will align to 4160V AC Bus 2F, WHEN Unit 2 DIG output Breaker CLOSES.

2.7 The loads listed on Attachment 2 for 600V busses 2C AND 2D are locked out following an undervoltage condition UNTIL the 600V AC Non-Essential load lockout is reset.

3.0 IMMEDIATE OPERATOR ACTIONS NONE.

G16.030 MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 3 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 4.0 SUBSEQUENT OPERATOR ACTIONS 4.1 CONFIRM appropriate Diesel Generator response to the situation This procedure is intended to maintain the plant in a safe condition assuming:

  • A loss of all AC power exists for up to one hour,
  • Only one 4160V AC emergency bus is energized for the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, NOTE: AND
  • No other accident occurs during this time.

UNDER the above circumstances, actions in this procedure take precedence over actions in any other procedure.

IF any condition exists requiring entry into the EOPs WHICH is NOT a direct result of the SBO OR of actions taken per this procedure, THEN EOP actions must take precedence, AND this procedure is to be used for guidance ONLY.

4.2 Enter 34AB-P41-001-2, Loss of Plant Service Water, and perform concurrently.

4.3 43 a DIG is running TRIP Diesel Generator PSW CANNOT be restored Locally IF AND AND THEN OR AND from CIR per PSW IS LOST 34S0-R43-001-2 Lube Oil Temp ~ 220°F OR Jacket Coolant Temp ~ 195°F OR It is evident that continued operation of the DIG will NOT assist recovery actions (4160V Bus damage, PSW still not available, etc)

G16.030 MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 40F24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 4.4 GENERAL ACTIONS 4.4.1 Attempt to restore power on at least one 4160V Emergency Bus using the appropriate procedure:

34AB-R22-002-2, Loss of 4160V Emergency Bus 34AB-R43-001-2, Diesel Generator Recovery 4.4.2 IF power is restored to more than one of the 4160V AC Emergency busses, THEN exit this procedure.

4.4.3 IF power is NOT restored to a 4160V AC Emergency Bus, perform the following to prolong the duty of the 125/250V DC Station Service batteries:

THE 125/250V BATTERIES MAY ONLY LAST 5 TO 6 HOURS, DEPENDING ON EQUIPMENT DEMAND. THE SS NEEDS TO BE AWARE THAT ONCE THE CAUTION:

BATTERIES ARE DEPLETED, THERE WILL BE NO POWER TO ELECTRICALLY OPEN THE SRVS.

4.4.3.1 Operate RCIC to control reactor water level, but minimize activities that involve operation of system valves (e.g., starting and stopping RCIC).

4.4.3.2 Secure HPCI as soon as possible.

4.4.3.3 As soon as possible, AND within 30 minutes of the loss of the battery chargers, remove the following from service:

DO NOT SECURE THE MAIN TURBINE EBOP UNTIL AFTER THE TURBINE CAUTION: SHAFT HAS STOPPED ROTATING, SPEED ZERO ALARM RECEIVED, OR SPEED ZERO INDICATION OBSERVED.

  • RFPT Emergency Oil Pumps G16.030 MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 5 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 4.4.3.4 Secure DC loads that are not critically needed, especially large DC motors on 2R22-S016.

FRAME NO. FUNCTION 1T 125V DC Cab 2A, Control Building (2R25-S001) 1B Station battery test bank (2R42-S063) 2T 125/250V Battery Charger 2A + 2B Throwover Switch (2R26-M031A,B) 2M 125/250V DC Switchgear 2C feeder (2R22-S018) 2B 250V DC MCC 2A, Reactor Building Feeder (2R24-S021) 3T 125V DC Distribution Cab 20 feeder (2R25-S129) 3M Emergency Lighting Cabinet (2R25-S077) 3B 125V DC Cabinet 2C, Turbine Bldg., (2R25-S003) 4.4.3.5 Secure DC loads that are not critically needed, especially large DC motors on 2R22-S017.

FRAME NO. FUNCTION 1T 125V DC Cabinet 2B, Control Building (2R25-S002) 1B Station Battery Test Bank (2R22-S063) 2T 125/250V DC Battery Charger 20 + 2E Throwover Switch (2R26-M031 C, D) 2M Emergency Lighting Cabinet (2R25-S069), Emerg. Lighting Cntrl Bldg (2R24-S094) 2B Spare 3T 125/250V DC Switchgear 20 Feeder (2R22-S019) 3M 250V DC MCC-2B Reactor Bldg. Feeder (2R24-S022) 3B 125V DC Distribution Cab 2E feeder (2R25-S130) 4.4.3.6 CONFIRM alarm "125/250 BATERRY VOLTS LOW", 651-125 does not exist but if this alarm is in, determine the cause and correct as soon as possible.

4.4.3.7 Avoid use of equipment that requires DC power.

4.4.3.8 Consider requesting the TSC to evaluate restoring the battery chargers by connection of temporary power to the 600V Emergency Buses as per TSG JD2.

4.4.4 IF DC AND AC power supplies are lost AND no source of high pressure feedwater is available, THEN operate RCIC in local/manual per Attachment 10, Local Manual Operation of RCIC.

4.4.5 Notify the Unit 1 Shift Supervisor that Unit 2 is in a station blackout condition AND to enter 34AB-R22-003-1, Station Blackout, procedure,Section II.

G16.030 MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 6 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 4.4.6 Perform the actions in 34AB-T41-001-2, Loss Of ECCS, MCREC, Or Area Ventilation System, for a Loss of Control Building Ventilation.

4.4.7 !E RHR was NOT in STANDBY, locally confirm OR position the following containment isolation valves given BELOW:

Valve Location Position 2E11-F004A 087RLR14 (N.E. Diagonal) Open 2E11-F004B 087RLR24 (S.E. Diagonal) Open 2E11-F004C 087RLR14 (N.E. Diagonal) Open 2E11-F004D 087RLR24 (S.E. Diagonal) Open 2E11-F011A 087RLR14 (N.E. Diagonal) Closed 2E11-F011 B 087RLR24 (S.E. Diagonal) Closed 2E11-F026A 087RLR14 (N.E. Diagonal) Closed 2E11-F026B 087RLR24 (S.E. Diagonal) Closed 2E11-F028A 118RHR14 (N.E. Torus Room) Closed 2E11-F028B 118RHR24 (S.E. Torus Room) Closed 4.4.8 Determine WHICH 4160V Emergency Bus has been energized by its Diesel Generator AND perform the appropriate Attachment.

IF: THEN:

Emergency Bus 2E is energized Perform Attachment 3 Emergency Bus 2F is energized Perform Attachment 4 Emergency Bus 2G is energized Perform Attachment 5 G16.030 MGR-0001 Ver. 3

SOUTHERN NUCLEAR PAGE PLANT E. I. HATCH 7 OF 24 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION NO:

STATION BLACKOUT 34AB-R22-003-2 3.6 II. UNIT 1 STATION BLACKOUT - - UNIT 2 ACTIONS 1.0 CONDITIONS 1.1 Notification has been received that Unit 1 is in a station blackout condition.

2.0 AUTOMATIC ACTIONS NIA, Not applicable to this procedure.

3.0 IMMEDIATE OPERATOR ACTIONS None 4.0 SUBSEQUENT OPERATOR ACTIONS IF UNIT 2 DRYWELL COOLING CANNOT BE MAINTAINED, THEN STEP 4.2 MUST BE ACCOMPLISHED PROMPTLY CAUTION: OR THE 1B DIESEL GENERATOR WILL TRANSFER TO UNIT 2 ON A UNIT 2 HIGH DRYWELL PRESSURE SIGNAL.

4.1 Confirm OR operate all Unit 2 Drywell Cooling Fans AND confirm cooling water (Drywell Chilled Water System) is available to the Drywell Coolers.

4.2 IF Unit 2 Drywell cooling can NOT be maintained OR Unit 1 is using the 1B DIG THEN OPEN link TB6-26(12T2), in PaneI2H11-P652, Bay B, to REMOVE Unit 2 high drywell pressure signal.

5.0 RESTORATION 5.1 CLOSE link TB6-26(12T2) in Panel 2H11-P652, Bay B.

Verified G16.030 MGR-0001 Ver. 3

SNC PLANT E. I. HATCH I 8 of 24 Pg 8of24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT ..1..

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: ADDITIONAL ANNUNCIATORS 2of2 ANNUNCIATORS 4160V BUS 2E BRKR 135544 TRIPPED/LKDOUT, 652-117 4160V BUS 2F BRKR 135564 TRIPPED/LKDOUT, 652-217 4160V BUS 2G BRKR 135584 TRIPPED/LKDOUT, 652-317 BATTERY CHARGER MALFUNCTION, 652-120 BATTERY CHARGER MALFUNCTION, 652-220 BATTERY CHARGER MALFUNCTION, 652-320 DIESEL MCC 1B UNDERVOLTAGE/DEENERGIZED, 652-223 600V BUS 2C BREAKER TRIPPED, 652-118 600V BUS 20 BREAKER TRIPPED, 652-318 600V BUS 2C UNDERVOLTAGE, 652-123 600V BUS 20 UNDERVOLTAGE, 652-323 4160V STA SVC FOR BRKR TRIP, 650-137 MGR-0009 Ver. 4

SNC PLANT E.I. HATCH I Pg 9 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT I 34AB-R22-003-2 3.6 ATTACHMENT 2. Att. Pg.

TITLE: 600V NON-ESSENTIAL LOADS 1 of 1 600 Volt Bus 2C Loads Locked Out 2R11-S043 Vital AC Transformer (alt. supply for Vital AC) 2P42-C001A RBCCW pump 2A 2P42-C001C RBCCW pump 2C 2N39-C001 Turning Gear Motor 2R42-S026 Station Battery Charger 2A 2R42-S027 Station Battery Charger 2B 2R42-S028 Station Battery Charger 2C 2N34-C005 Turning Gear Oil Pump and Turning Gear Motor 2N34-C003 Motor Suction Pump 600 Volt Bus 20 Loads Locked Out 2P51-C001B Air Compressor 2B 2R44-S001 Vital AC Battery Charger 2P42-C001B RBCCW Pump 2B 2R42-S029 Station Battery Charger 2D 2R42-S030 Station Battery Charger 2E 2R42-S031 Station Battery Charger 2F MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg10of24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT ~

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 2A 1 of 2 1.0 Confirm running OR START the 2A PSW Pump, 2P41-C001A.

2.0 Confirm CLOSED OR CLOSE Division I Turbine Building PSW Isolation Valve, 2P41-F316A.

3.0 IF RCIC is running AND capable of controlling reactor water level, THEN secure HPCI.

4.0 TRIP the following equipment control switches:

2C11-C001A, CRD Pump 2A on Pane12H11-P603 (Norm).

2E21-C001A, Core Spray Pump 2A on Pane12H11-P601 (Norm).

2P64-B006A, Drywell Chiller 2A on Panel 2H 11-P700 (Off).

5.0 At Panel 2H11-P652, RESET the 600V AC Bus 2C Nonessential Load lockout pushbutton.

NOTE: Station Service Battery Chargers must be PLACED in service WITHIN the first hour of the station blackout condition.

6.0 Place 125/250VDC Station Service Battery Chargers, 2R42-S026, S027, AND S028 (2A, 2B & 2C) in service per 34S0-R42-001-2.

7.0 At Panel 2H 11-P657, confirm running OR START Station Service Battery Room 2A Exhaust Fan,2Z41-C014.

NOTE: The suppression pool temperature is NOT expected to reach the HCTL, during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> blackout event duration.

8.0 BEFORE the suppression pool temperature reaches the Heat Capacity Temperature Limit, perform the following:

8.1 Turn off electrical loads listed in Attachment 6.

BEFORE STARTING ANY 4160V PUMPS, ENSURE DIESEL 2A LOAD IS BELOW CAUTION:

THE MAXIMUM VALUE LISTED IN 34AB-R22-002-2 AND MAINTAIN FREQUENCY BETWEEN 59 AND 61 HERTZ.

8.2 Initiate Suppression Pool Cooling Mode of RHR using RHRSW Pump 2A, 2E11-C001A AND RHR Pump 2A, 2E11-C002A, per 34S0-E11-010-2, Residual Heat Removal System.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 11 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT ~

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 2A 20f2 9.0 Monitor AND ADJUST Diesel Generator 2A frequency as required to maintain it between 59 AND 61 Hz, with the Diesel Generator 2A Speed Adjust switch.

10.0 Monitor the Diesel Generator 2A load ensuring that it does NOT exceed 3250 KW.

10.1 IF required, locally remove the additional loads listed in Attachment 7 to meet this requirement.

The following coolers will have power, BUT will be inoperable due to lack of Service Water:

NOTE: 2T41-B001A CRD Pump Room Cooler A 2T41-B002A SE Diagonal Room Cooler A 2T41-B005A HPCI Pump Room Cooler A 11.0 IF the Diesel Generator 2A load limit (3250 KW) will NOT be exceeded, RETURN to service, room coolers AND other loads removed earlier as required to support operating equipment.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 12 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT .+/--

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 1B 1 of 4 1.0 Confirm running OR start the Standby PSW pump, 2P41-C002.

2.0 Notify the Unit 1 Shift Supervisor that Unit 2 will be using Diesel Generator 1B.

3.0 Place the Unit Select switch "SB" for Diesel Generator 1B on Local Panel 1R43-S001 B in the Unit 2 position. (key required, obtain from CR or local DG bldg gangbox.)

4.0 Determine WHICH Plant Service Water Division is available AND perform the appropriate action(s).

IF THEN 2C PSW Pump, 2P41-C001 C, is running. - Proceed to step 5.0 2C PSW Pump, 2P41-C001C, is NOT running - START the 2C PSW pump, 2P41-C001C

- Proceed to step 5.0.

2C PSW Pump, 2P41-C001C, will NOT start. - Confirm running OR START 20 PSW Pump, 2P41-C001 D.

- Proceed to step 6.0.

NEITHER PSW Pumps will start. - Proceed to step 5.0.

5.0 IF RCIC is running AND capable of controlling reactor water level, THEN secure HPCI.

5.1 TRIP AND Pull to Lock (IF possible) the following pump control switches:

2P41-C001D, PSW Pump 20, on Pane12H11-P650 (PTL) 2E11-C001D, RHR Pump 20, on Pane12H11-P601 (Norm) 2C11-C001B, CRD Pump 2B, on Pane12H11-P603 (Norm) 5.2 Perform actions in 34S0-R23-001-2, 600 AC System to ENERGIZE 2C 600V AC Bus from 2F 4160V AC Bus.

5.3 WHEN 2C 600V AC Bus is ENERGIZED, THEN perform the following:

5.3.1 Confirm CLOSED OR CLOSE the Division 1 Turbine Building PSW Isolation Valve, 2P41-F316A.

5.3.2 At Panel 2H11-P652, RESET the 600V AC Bus 2C Nonessential Load lockout push button.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 13 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT ...1..

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 1B 20f4 NOTE: Station Service battery chargers must be PLACED in service WITHIN the first hour of the station blackout condition.

5.3.3 PLACE 125/250VDC Station Service Battery Chargers, 2R42-S026, S027 AND S028 (2A, 2B & 2C) in service per 34S0-R42-001-2.

5.3.4 At Panel 2H11-P657, confirm running OR START the Station Battery Room 2A Exhaust Fan, 2Z41-C014.

NOTE: The suppression pool temperature is NOT expected to reach the HCTL, during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> blackout event duration.

5.4 BEFORE the suppression pool temperature reaches the Heat Capacity Temperature Limit (HCTL) perform the following:

5.4.1 Turn off the electrical loads listed in Attachment 6.

BEFORE STARTING ANY 4160V PUMPS, ENSURE DIESEL 1B LOAD IS BELOW CAUTION: THE MAXIMUM VALUE LISTED IN 34AB-R22-002-2 AND MAINTAIN FREQUENCY BETWEEN 59 AND 61 HERTZ.

5.4.2 Initiate Suppression Pool Cooling Mode of RHR using RHRSW Pump 2C, 2E11-C001 C, AND RHR Pump 2C, 2E11-C002C per 34S0-E11-01 0-2, Residual Heat Removal System.

5.5 Monitor AND ADJUST Diesel generator 1B frequency as required to maintain it between 59 AND 61 Hz, with the Diesel Generator 1B Speed Adjust switch.

5.6 Monitor the Diesel Generator 1B load, ensuring that is does NOT exceed 3250 KW.

5.6.1 IF required, locally, REMOVE the additional loads listed in Attachment 7 to meet this requirement.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH JI Pg 14 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT A:..

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 1B 30f4 The following coolers will have power, BUT will be inoperable due to lack of Service Water:

NOTE: 2T41-B001A CRD Pump Room Cooler A 2T41-B002A SE Diagonal Room Cooler A 2T41-B005A HPCI Pump Room Cooler A 5.7 IF the Diesel Generator 1B load limit (3250 KW) WILL NOT be exceeded, RETURN to service, room coolers AND other loads removed earlier as required to support operating equipment.

NOTE: Step 6.0 will NOT be performed, UNLESS DIRECTED by Decision Table in step 4.0.

6.0 IF HPCI is running AND capable of controlling reactor water level THEN secure RCIC.

6.1 TRIP AND Pull to Lock (IF possible) the following pump control switches:

2P41-C001C, PSW Pump 2C, on Pane12H11-P650 (PTL) 2E11-C002C, RHR Pump 2C, on Pane12H11-P601 (Norm) 2C11-C001B, CRD Pump 2B, on Pane12H11-P603 (Norm) 6.2 Perform actions in 34S0-R23-001-2, 600V/480V AC System to ENERGIZE 20 600V AC Bus from 2F 4160V AC Bus.

6.3 WHEN 20 600V Bus is ENERGIZED, THEN perform the following:

6.3.1 Confirm CLOSED OR CLOSE the Division II Turbine Building PSW Isolation Valve, 2P41-F316B.

6.3.2 At PaneI2H11-P652, RESET the 600V AC Bus 20 undervoltage lockout push button.

NOTE: Station Service Battery Chargers must be PLACED in service WITHIN the first hour of the station blackout condition.

6.3.3 PLACE 125/250V DC Station Service Battery Chargers, 2R42-S029, S030 AND S031 (20, 2E & 2F) in service per 34S0-R42-001-2.

6.3.4 At PaneI2H11-P654, confirm running OR START the Station Service Battery Room 2B Exhaust Fan, 2Z41-C015.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 15 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT -+/--

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: DIESEL GENERATOR 1B 4of4 NOTE: The suppression pool is NOT expected to reach the HCTL during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> blackout event duration.

6.4 BEFORE the suppression pool temperature reaches the Heat Capacity Temperature Limit, perform the following:

6.4.1 Turn off the electrical loads listed in Attachment 8.

6.4.2 Manually OPEN RHRSW CROSSTIE VALVES 2E11-F119A AND 2E11-F119B (BAY 1 EAST OUTER CATWALK).

6.4.3 At Intake Structure, OPEN 2P41-F925 AND 2P41-F926 to supply cooling water to RHRSW Pump 2C from PSW Dlvison 1.

BEFORE STARTING ANY 4160V PUMPS, ENSURE DIESEL 1B LOAD IS BELOW CAUTION: THE MAXIMUM VALUE LISTED IN 34AB-R22-002-2 AND MAINTAIN FREQUENCY BETWEEN 59 AND 61 HERTZ.

6.4.4 Initiate Suppression Pool Cooling Mode of RHR using RHRSW Pump 2C, 2E11-C001C, AND RHR Pump 2D, 2E11-C002D, per 34S0-E11-010-2, Residual Heat Removal System.

6.5 Monitor AND ADJUST Diesel Generator 1B frequency as required to maintain it between 59 AND 61 Hz, with the Diesel Generator 1B Speed Adjust switch.

6.6 Monitor the Diesel Generator 1B load, ensuring that is does NOT exceed 3250 KW.

6.6.1 IF required, locally REMOVE the additional loads listed in Attachment 9 to meet this requirement.

The following coolers will have power, BUT will be inoperable due to lack of Service Water:

NOTE: 2P64-B006B Drywell Chiller B 2T41-B003B NE Diagonal Room Cooler B 2T41-B004B RCIC Pump Room Cooler B 6.7 IF the Diesel Generator 1B load limit (3250 KW) will NOT be exceeded, RETURN to service, room coolers AND other loads removed earlier as required to support operating equipment.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 16 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT I 34AB-R22-003-2 3.6 ATTACHMENT ..§.. Att. Pg.

TITLE: DIESEL GENERATOR 2C 1 of 2 1.0 Confirm running OR START the 2B PSW Pump, 2P41-C001 B.

2.0 Confirm CLOSED OR CLOSE Division II Turbine Building PSW Isolation Valve, 2P41-F316B 3.0 IF HPCI is running AND capable of controlling reactor water level, THEN secure RCIC.

4.0 TRIP the following pump control switches:

2P64-B006B, Drywell Chiller B on Pane12H11-P700 2E21-C001B, Core Spray Pump 2B on Pane12H11-P601 5.0 At PaneI2H11-P652, RESET the 600V Bus 20 undervoltage lockout pushbutton.

NOTE: Station Service Battery Chargers must be PLACED in service WITHIN the first hour of the station blackout condition.

6.0 Place 125/250 VDC Station Service Battery Chargers, 2R42-S029, S030, AND S031, (20, 2E & 2F) in service per 34S0-R42-001-2.

7.0 At PaneI2H11-P654, confirm running OR START Station Service Battery Room 2B Exhaust Fan, 2Z41-C015.

NOTE: The suppression pool temperature is NOT expected to reach the HCTL, during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> blackout event duration.

8.0 BEFORE the suppression pool temperature reaches the Heat Capacity Temperature Limit, perform the following:

8.1 Turn off the electrical loads listed in Attachment 8.

BEFORE STARTING ANY 4160V PUMPS, INSURE DIESEL 2C LOAD IS BELOW CAUTION: THE MAXIMUM VALUE LISTED IN 34AB-R22-002-2 AND MAINTAIN FREQUENCY BETWEEN 59 AND 61 HERTZ.

8.2 Initiate Suppression Pool Cooling mode of RHR using RHRSW Pump 2B, 2E11-C001 B, AND RHR Pump 2B, 2E11-C002B per 34S0-E11-010-2, Residual Heat Removal System.

9.0 Monitor AND ADJUST Diesel Generator 2C frequency as required to maintain it between 59 AND 61 Hz, with the Diesel Generator 2C Speed Adjust switch.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 17 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT I 34AB-R22-003-2 3.6 ATTACHMENT 2- Att. Pg.

TITLE: DIESEL GENERATOR 2C 2of2 10.0 Monitor the Diesel Generator 2C load, ensuring that it does NOT exceed 3250 KW.

10.1 IF required, locally REMOVE the additional loads listed in Attachment 9 to meet this requirement.

The following coolers will have power, BUT will be inoperable due to lack of Service Water:

NOTE: 2P64-B006B Drywell Chiller B 2T41-B003B NE Diagonal Room Cooler B 2T41-B004B RCIC Pump Room Cooler B 11.0 IF the Diesel Generator 2C load limit (3250 KW) will NOT be exceeded, RETURN to service, room coolers AND other loads removed earlier, as required to support operating equipment.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH IJ Pg 18 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT .Q..

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: ELECTRICAL LOADS FROM 600V BUS 2C 1 of 1 TABLE 1 Loads which must be removed from 600V bus 2C:

MPL No. Description Switch Panel 2N34-C005 Turning Gear Oil Pump 2H11-P650 2N34-C003 Motor Suction Pump 2H11-P650 2N39-C001 Turning Gear Motor 2H11-P650 2P42-C001A RBCCW Pump A 2H11-P650 2P42-C001C RBCCW Pump C 2H11-P650 2P51-C001A Air Compressor 1A 2H11-P650 2T41-B001A CRD Pump Room Cooler A 2H11-P657 2T41-B002A SE Diagonal Room Cooler A 2H11-P657 2T41-B005A HPCI Pump Room Cooler A 2H11-P657 2T47-C001A Drywell Cooling Return Fan 2H11-P657*

2T47-B007A Drywell Area Cooling Unit 2H11-P657*

2T47-B008A Drywell Area Cooling Unit 2H11-P657*

2T47-B009A Drywell Area Cooling Unit 2H11-P657*

NOTE: MG set 2C71-S001A will trip on the loss of power, and must NOT be reloaded onto 600V load center bus 2C.

  • May be re-energized later to control Drywell AND/OR local temperature IF the diesel has available load capacity.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 19 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT .L I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: ADDITIONAL ELECTRICAL LOADS FROM 600V BUS 2C 1 of 1 TABLE 2 Additional Unit 2 loads which may be removed from Diesel Generator 2A or 1B NOTE: The following loads are applicable to both Diesel Generator 2A AND 1B:

MPL No. Description Switch Location 2X41-B003A Diesel 2A Room Heater 2A 1 2R24-S025, Frame 2B 2X41-B003B Diesel2A Room Heater 2A2 2R24-S025, Frame 2C 2X41-B003C Diesel 2A Room Heater 2A3 2R24-S025, Frame 20 2X41-B004A SWGR 2E Room Heater 2A 1 2R24-S025, Frame 3B 2X41-B004B SWGR 2E Room Heater 2A2 2R24-S025, Frame 3C 2X41-B004C SWGR 2E Room Heater 2A3 2R24-S025, Frame 3D NOTE: The following loads are applicable to Diesel Generator 1B only:

1X41-B001D Diesel 1B Room Heater 1A 1R24-S026, Frame 3E 1X41-B001E Diesel 1B Room Heater 1B 1R24-S026, Frame 3F 1X41-B001F Diesel 1B Room Heater 1C 1R24-S026, Frame 4F 2X41-B004D SWGR 2F Room Heater 2A 2R24-S048, Frame 1C 2X41-B004E SWGR 2F Room Heater 2B 2R24-S048, Frame 10 2X41-B004F SWGR 2F Room Heater 2C 2R24-S048, Frame 1E 2E21-C002A Jockey Pump 2H11-P601*

2E21-C002B Jockey Pump 2H11-P601

  • May be removed after the start of Suppression Pool Cooling.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I P~20 Pg 20 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT .1L I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: ELECTRICAL LOADS FROM 600V BUS 2D 1 of 1 TABLE 3 Loads which must be removed from 600V bus 2D:

MPL No. Description Switch Panel 2P42-C001B RBCCW Pump B 2H11-P650 2P51-C001B Air Compressor B 2H11-P650 2T41-B001B CRD Pump Room Cooler B 2H11-P654 2T41-B002B NE Diagonal Room Cooler B 2H11-P654 2T41-B004B RCIC Pump Room Cooler B 2H11-P654 2T47-C001B Drywe" Cooling Return Fan 2H11-P654 2T47-B007B Drywe" Area Cooling Fan 2H11-P654 2T47-BOOBB Drywe" Area Cooling Fan 2H11-P654 2T47-B009B Drywe" Area Cooling Fan 2H11-P654 NOTE: MG set 2C71-S001B wi" trip on the loss of power, and must NOT be reloaded onto 600V load center bus 2D.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 21 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT JL I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: ADDITIONAL ELECTRICAL LOADS FROM 600V BUS 2D 1 of 1 TABLE 4 Additional Unit 2 loads which may be removed from Diesel Generator 2C or 1B if supplying 600V 2D:

MPL No. Description Switch Location 2X41-B004G SWGR Room 2G Heater 2A 2R24-S027, Frame 3B 2X41-B004H SWGR Room 2G Heater 2B 2R24-S027, Frame 3C 2X41-B004J SWGR Room 2G Heater 2C 2R24-S027, Frame 3D 2X41-B003D DIG 2C Room Heater 2A 2R24-S027, Frame 2B 2X41-B003E DIG 2C Room Heater 2B 2R24-S027, Frame 2C 2X41-B003F DIG 2C Room Heater 2C 2R24-S027, Frame 2D 2E21-C003A Jockey Pump 2H11-P601 2E21-C003B Jockey Pump 2H11-P601*

  • May be removed after the start of Suppression Pool Cooling.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 22 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATIACHMENT JJL I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: LOCAL MANUAL OPERATION OF RCIC 1 of 3

-It will be necessary to adjust RCIC Flow by throttling 2E51-F524, Turbine Trip & Throttle Vlv.

Maintain approximately 4000 RPM as read on a hand held tach. Hand held Tachs can be located at the I&C toolroom, Water Buffalo Shed and the EP Closet by the NRC Conference Room in the Simulator Bldg. 1st Floor.

NOTES: -This SUbsection provides guidance for RCIC operation with no AC or DC power. This mode of RCIC operation will only be allowed with SS permission during extremely adverse emergency conditions.

-IF RCIC is already running, verify proper operation (local gauges in RCIC diagonal) and injection is occurring by checking RPV level changes on the Reactor Building 158 elevation, THEN perform steps in this section to maintain proper operation.

1.0 Locally, confirm that a suction source is aligned by performing one of the following:

1.1 On RCIC Inst. Rack, 2H21-P017, confirm that 2E51-R002, RCIC pump suction pressure indicator, indicates ~ 14 PSIG.

1.2 If unable to perform 1.1, perform the following:

1.2.1 CONFIRM/OPEN 2E51-F010, CST Suction Vlv. (CLOSED IF CST unavailable) 1.2.2 CONFIRM/CLOSE 2E51-F029, Suppression Pool Suction Vlv. (OPEN IF CST is unavailable) 1.2.3 CONFIRM/CLOSE 2E51-F031, Suppression Pool Suction Vlv. (OPEN IF CST is unavailable) 2.0 IF RCIC operation is for RPV pressure control, THEN locally perform the following:

2.1 OPEN and RACK OUT Breaker 8A on MCC 2R24-S021, for 2E51-F022, RCIC Test Line to CSTVlv.

2.2 OPEN and RACK OUT Breaker 3C on MCC 2R24-S022, for 2E41-F011, HPCI Test Line to CSTVlv.

2.3 OPEN 2E51-F022, RCIC Test Line To CST Vlv.

2.4 OPEN 2E41-F011, HPCI Test Line to CST Vlv.

MGR-0009 Ver. 4

SNC PLANT E. I. HATCH I Pg 23 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT ATTACHMENT ~

I 34AB-R22-003-2 3.6 Att. Pg.

TITLE: LOCAL MANUAL OPERATION OF RCIC 2 of 3 3.0 Locally, OPEN the following breakers:

3.1 OPEN and RACK OUT Breaker 9B on MCC 2R24-S021 , for 2E51-F045, Steam to Turb Vlv.

3.2 OPEN and RACK OUT Breaker 1A on MCC 2R24-S021, for 2E51-F046, Turbine Cooling Water Vlv.

3.3 OPEN and RACK OUT Breaker 3B on MCC 2R24-S021, for 2E51-F013, RCIC Pump Discharge Vlv.

3.4 OPEN and RACK OUT Breaker 1B on MCC 2R24-S021, for 2E51-F524, RCIC Trip and Throttle Vlv.

4.0 Locally ALIGN/CONFIRM the following valves:

4.1 CONFIRM OPEN/OPEN 2E51-F012, Pump Discharge Vlv.

4.2 CLOSE 2E51-F524, RCIC Trip and Throttle Vlv.

4.3 OPEN 2E51-F045, Stm to Turb Vlv.

4.4 OPEN 2E51-F046, Turb Clg Wtr Vlv.

5.0 IF RCIC operation is for RPV level control, perform the following:

5.1 OPEN 2E51-F013, RCIC Pump Disch Vlv.

5.2 Confirm CLOSED/CLOSE 2E51-F022, RCIC Test Line To CST Vlv.

6.0 Confirm adequate steam supply by performing one of the following:

6.1 On RCIC Inst. Rack, 2H21-P017, confirm that 2E51-R003, RCIC steam supply pressure indication corresponds to the expected reactor pressure.

6.2 If unable to perform 6.1, perform the following:

6.2.1 IF possible, CONFIRM OPEN 2E51-F008, Stm Supply Outbd Vlv.

6.2.2 IF possible, CONFIRM OPEN 2E51-F007, Stm Supply Inbd Vlv.

7.0 Slowly OPEN 2E51-F524, Trip and Throttle Vlv, to obtain approximately 4000 RPM on a hand held Tach.

MGR-0009 Ver. 4

SNC PLANT E.I. HATCH I Pg 24 of 24 DOCUMENT TITLE: DOCUMENT NUMBER: Version No:

STATION BLACKOUT I 34AB-R22-003-2 3.6 ATTACHMENT J.Q.. Att. Pg.

TITLE: LOCAL MANUAL OPERATION OF RCIC 3 of 3 8.0 Confirm 2E51-F019, Minimum Flow VI v, is closed.

9.0 Locally, monitor system parameters in the RCIC diagonal and RPV level changes on the Reactor Building 158 elevation.

10.0 When it is desired to shut down from local manual operation, perform the following:

10.1 Confirm that RCIC system operation is no longer required for reactor water level control, OR reactor pressure control.

PROLONGED OPERATION OF RCIC TURBINE BELOW 2000 RPM IS TO BE AVOIDED TO ENSURE ADEQUATE OIL PRESSURE FOR PROPER TURBINE CAUTION:

GOVERNOR OPERATION, BEARING LUBRICATION, AND TO PREVENT TURBINE EXHAUST VALVE CHATTER.

10.2 Use Turbine Trip & Throttle Vlv to REDUCE turbine speed to slightly greater than 2000 RPM (using hand held Tach).

10.3 CLOSE 2E51-F045, Stm To Turb Vlv.

10.4 IF RCIC was injecting to Reactor Vessel, THEN CLOSE 2E51-F013, RCIC Pump Disch Vlv.

10.5 IF RCIC was discharging to the CST, THEN perform the following:

10.5.1 CLOSE 2E51-F022, Test to CST Vlv.

10.5.2 CLOSE 2E41-F011, HPCI Test to CST.

10.6 Locally, confirm CLOSED/CLOSE 2E51-F019, Minimum Flow Vlv.

10.7 CLOSE 2E51-F046, Turb Clg Wtr Vlv.

10.8 CLOSE 2E51-F524, Trip and Throttle Vlv.

10.9 When electrical power is restored, return the RCIC system to the standby lineup per 34S0-E51-00 1-2.

MGR-0009 Ver. 4

Edwin I. Hatch Nuclear Plant Enclosure 7 73EP-EIP-001-0 Emergency Classification and Initial Actions

SOUTHERN NUCLEAR PLANT E.I. HATCH I DOCUMENT TYPE:

EMERGENCY PREPAREDNESS PROCEDURE PAGE 1 OF 128 DOCUMENT TITLE: DOCUMENT NUMBER: REVISIONNERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:

17.0 EXPIRATION ~l\.PPROVALS:

APPROVALS: EFFECTIVE DATE: DEPARTMENT MANAGER Rand:t Ott DATE 03/10/09 DATE:

03/10/09 N/A NPGM/POAGM/PSAGM John Lewis DATE 03/10/09 1.0 OBJECTIVE This procedure establishes the methodology for emergency classification. Specific Emergency Action Levels (EALs) and minimum initial actions to respond to a given emergency are established in this procedure.

TABLE OF CONTENTS Section Title Page 2.0 APPLICABILITY 1

3.0 REFERENCES

2 4.0 REQUIREMENTS 3 5.0 PRECAUTIONS/LIMITATIONS 4 6.0 PREREQUISITES 4 7.0 PROCEDURE 5 Attachment 1 Fission Product Barrier Evaluation Chart - Modes 1, 2 & 3 8 "Hot" Initiating Condition Matrix Evaluation Chart -

Attachment 2 9 Modes 1,2 & 3 "Cold" Initiating Condition Matrix Evaluation Chart -

Attachment 3 10 Modes 4, 5 & Defueled HNP Emergency Action Levels - Initiating Conditions, Attachment 4 11 Threshold Values, and Basis Attachment 5 Classification Determination Worksheet 125 Unit 1 and Unit 2 Pretreatment Monitor (Flow vs. mR/hr)

Attachment 6 126 Graphs 2.0 APPLICABILITY This procedure applies to emergency classification determinations and associated initial responses. This procedure will be utilized for actual emergencies, emergency drills/exercises, or training as required.

MGR-0002 Rev. 8

SOUTHERN NUCLEAR PLANT E.I. HATCH I PAGE 2 OF 128 DOCUMENT TITLE: DOCUMENT NUMBER: VERSION EMERGENCY CLASSIFICATION AND INITIAL ACTIONS 73EP-EIP-001-0 NO:

17.0

3.0 REFERENCES

3.1 00AC-REG-001-0, Federal And State Reporting And Federal Document Posting Requirements 3.2 10AC-MGR-006-0, Hatch Emergency Plan 3.3 20AC-ADM-002-0, Quality Assurance Records Administration 3.4 31GO-OPS-013-0, Notification and Reports 3.5 73EP-ADM-001-0, Maintaining Emergency Preparedness 3.6 73EP-EIP-004-0, Duties of Emergency Director 3.7 73EP-EIP-005-0, On-Shift Operations Personnel Emergency Duties 3.8 73EP-EIP-015-0, Offsite Dose Assessment 3.9 73EP-EIP-018-0, Prompt Dose Assessment 3.10 73EP-EIP-073-0, Offsite Emergency Notifications 3.11 NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73 3.12 Hatch Unit 1 & 2 Technical Specifications (TS) 3.13 North American Electric Reliability Council/Southeastern Electric Reliability Council (NERC/SERC) Standard(s):

CIP-001-1 (R1), (R2), (R3), (R4) 3.14 SNC Letter Dated December 30,2005; NL-05-2236, Southern Nuclear Operating Company Emergency Plans Transition to NEI 99-01 Emergency Action Level Scheme 3.15 SNC Letter Dated October 13, 2006; NL-06-2177, Transition to NEI 99-01 Emergency Action Levels Response to Request for Additional Information 3.16 SNC Letter Dated April 1, 2007; NL-07-0522, Transition to NE199-01 Emergency Action Levels - Response to Request for Additional Information; 3.17 NRC Letter Dated April 30, 2007; LC # 14580 - Emergency Action Level Revisions for Southern Nuclear Operating Company, Inc. (Includes Safety Evaluation By The Office of Nuclear Reactor Regulation Related to Proposed Revisions to the Emergency Action Levels for The Edwin I. Hatch Nuclear Plant, Unit Nos. 1 and 2 (HNP).

MGR-0001 Rev. 4

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17.0 4.0 REQUIREMENTS 4.1 PERSONNEL REQUIREMENTS 4.1.1 The Emergency Director is responsible for initial classification of events. The Shift Manager (SM) shall initially assume the responsibilities of the Emergency Director (ED) until relieved.

4.1.1.1 If the SM is unavailable, then the affected unit's Shift Supervisor (SS) will become the Emergency Director.

4.1.1.2 IF the SM is unavailable and the event involves both units, the Unit 1 Shift Supervisor (SS) will become the Emergency Director.

4.1.1.3 Any of these persons will assume the position of Emergency Director in the Control Room until a qualified relief can arrive on site and receive an adequate turnover.

4.1.2 Anyone of the following persons may assume the position of Emergency Director after receipt of turnover information from the off going ED.

  • Plant Manager
  • Site Support Manager
  • Operations Manager
  • Maintenance Manager
  • Other qualified Emergency Director 4.1.3 After turnover of Emergency Director responsibilities, the Shift Manager then becomes responsible for recognizing changes in plant conditions and advising the ED concerning classification of events.

4.1.4 The Technical Support Center (TSC) and the Emergency Operations Facility (EO F)

Managers or designee are responsible for providing recommendations on emergency classifications to the Emergency Director.

4.2 MATERIAL AND EQUIPMENT N/A - Not applicable to this procedure 4.3 SPECIAL REQUIREMENTS 4.3.1 Portions of this procedure require the results from calculations of projected doses at or beyond the site boundary to determine the appropriate emergency classification.

4.3.2 Portions of this procedure will require actual dose measurements (onsite or off-site) to determine the appropriate emergency classification.

MGR-0001 Rev. 4

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17.0 5.0 PRECAUTIONS/LIMITATIONS 5.1 PRECAUTIONS 5.1.1 The Emergency Director may use judgment as the final criterion for determining the classification of off-normal events that are not included in this procedure.

5.1.2 The value of any emergency actions, which may require movement of plant personnel, must be judged against the danger to personnel or nuclear safety.

5.2 LIMITATIONS N/A - not applicable to this procedure.

6.0 PREREQUISITES 6.1 An off-normal event has occurred or is in progress (during actual emergencies, emergency drills/exercises, or training) requiring evaluation to determine the need for an emergency classification.

MGR-0001 Rev. 4

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17.0 REFERENCE I 7.0 PROCEDURE 7.1 EMERGENCY CLASSIFICATION AND INITIAL ACTIONS CAUTIONS

  • PERSONNEL AND PLANT SAFETY MUST BE ADDRESSED AS THE HIGHEST PRIORITY, IF NECESSARY, PRIOR TO AN EMERGENCY CLASSIFICATION.
  • CLASSIFICATION SHOULD NOT BE DELAYED IN ANTICIPATION OF EITHER EVENTS BEING TERMINATED OR THE THREAT TO SAFETY ENDING. THE EMERGENCY SHOULD BE ASSESSED AND CLASSIFIED WITHIN 15 MINUTES AFTER IT IS RECOGNIZED THAT THE EMERGENCY ACTION LEVEL HAS BEEN EXCEEDED.
  • EVENTS SHOULD BE CLASSIFIED BASED ON MEETING THE IC AND TV FOR AN EAL CONSIDERING EACH UNIT INDEPENDENTLY. IF BOTH UNITS ARE IN CONCURRENT EVENTS, THEN THE HIGHEST CLASSIFICATION MUST BE MADE AND USED FOR THE OFFSITE NOTIFICATIONS WITH THE OTHER UNIT EVENTS NOTED ON THE EMERGENCY NOTIFICATION FORM.

7.1.1 Upon notification of an abnormal condition or observation of abnormal instrument readings, assess and classify the event as follows:

7.1.1.1 Determine the affected unit's operating mode at the time of the event. Attachment 5 Emergency Classification Worksheet can be used to aid in determining the classification.

Select appropriate block on Line 1 to indicate the mode.

7.1.1.1.1 IF the affected unit is in Modes 1, 2 or 3, then go to step 7.1.2.

7.1.1.1.2 IF the affected unit is in Modes 4,5 or DEFUELED, then go to step 7.1.4.

7.1.2 Using Attachment 1, evaluate the status of the fission product barriers (FBR) and select the highest applicable emergency classification.

7.1.2.1 The appropriate block on line 2.a of Attachment 5 can be used to indicate the condition for each fission product barrier.

7.1.2.2 The appropriate block on line 2.b of Attachment 5 can be used to indicate the highest applicable emergency classification level based on this evaluation ..

7.1.3 Using Attachment 2, determine the highest emergency classification level based on the "HOT" Initiating Conditions (IC) and Threshold Values (TV). The IC can be recorded on line 3 of Attachment 5. Refer to Attachment 4, as needed, for clarification of the basis for the IC.

MGR-0001 Rev. 4

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17.0 7.1.3.1 Determine the highest emergency classification level from either the FPB or "HOT" IC.

7.1.3.2 The appropriate block on line 5 of Attachment 5 can be checked to indicate the highest emergency classification level and record the associated IC#.

7.1.3.3 Proceed to step 7.1.5.

7.1.4 Using Attachment 3, determine the highest applicable "COLD" IC and associated TV. The IC can be recorded on line 4 of Attachment 5. Refer to Attachment 4, as needed, for clarification of the basis for the IC.

7.1.4.1 The appropriate block on line 5 of Attachment 5 can be checked to indicate the highest emergency classification level and record the associated IC#.

7.1.4.2 Proceed to step 7.1.5.

7.1.5 Line #6 of Attachment 5 can be used to write a description of the Initiating Condition corresponding to the emergency declaration.

7.1.6 Line #7 of Attachment 5 can be used to record the date and time of the emergency declaration.

7.1.7 Use the Emergency Classification Level, IC#, and the description for the events to complete the Statellocal and NRC notification forms in accordance with procedures 73EP-EIP-073-0 and 34GO-OPS-013-0. Attachment 5 Emergency Classification Worksheet can be used to aid in completion of these forms.

7.1.8 Refer to procedure 73EP-EIP-004-0 for other Emergency Director duties.

Note: NUREG 1022, Rev 2, Section 3.1.1 states in part "Occasionally, a licensee may discover that an event or condition had existed which met the emergency plan criteria but that no emergency had been declared and the basis for the emergency class no longer exists at the time of this discovery. This may be due to a rapidly concluded event or an oversight in the emergency classification made during the event or it may be determined during a post-event review. Frequently, in cases of this nature, which were discovered after the fact, licensees have declared the emergency class, immediately terminated the emergency class and then made the appropriate notifications. However, the staff does not consider actual declaration of the emergency class to be necessary in these circumstances; an ENS notification (or an ENS update if the event was previously reported but misclassified) within one hour of the discovery of the undeclared (or misclassified) event will provide an acceptable alternative. (2)

2. Notification of the State and local emergency response organizations should be made in accordance with the arrangements made between the licensee and offsite organizations."

MGR-0001 Rev. 4

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17.0 7.1.9 If an event has occurred that meets an IC and TV for declaration but no emergency has been declared at the time of discovery AND the basis for the emergency class no longer exists, THEN the condition should be reported using the guidance of NUREG 1022, Section 3.1.1.

7.1.9.1 Use procedures 00AC-REG-001-0 and 31 GO-OPS-013-0 to make this notification.

7.1.9.2 Contact the site Emergency Preparedness Coordinator or designee to contact and inform the State and local emergency response agencies.

7.2 PERIODIC REVIEW OF THE CLASSIFICATION LEVEL 7.2.1 The Emergency Director shall periodically review current or projected plant conditions to determine if the emergency should be upgraded or terminated using guidance contained in section 7.1 of this procedure.

7.2.2 The TSC Manager shall periodically review plant conditions, determine if the emergency should be upgraded or terminated based on current or projected status, and make recommendations to the Emergency Director.

7.2.3 The EOF Manager shall periodically review offsite radiological conditions, determine if the emergency should be upgraded or terminated based on current field surveys or projected releases, and make recommendations to the Emergency Director.

7.3 TERMINATING THE EMERGENCY CLASSIFICATION 7.3.1 For a Notification of Unusual Event (NUE), the Emergency Director may terminate the Emergency when plant conditions have stabilized and the reason for the NUE has been corrected. An NUE can be terminated without coordination with offsite authorities.

7.3.2 For an ALERT, SITE-AREA or GENERAL Emergency, the Emergency Director may terminate the emergency after discussions with plant management, applicable members of the HNP emergency organization, the NRC, and officials from the Georgia Emergency Management Agency (GEMA), Appling, Jeff Davis, Tattnall and Toombs Counties does not result in the identification of a valid reason for not terminating the emergency.

7.3.3 After the decision has been made to terminate the emergency, the Emergency Director will take actions as described in procedure 73EP-EIP-004-0.

7.4 DOCUMENTATION AND RECORDS 7.4.1 All data and information generated during the emergency event will be maintained by applicable emergency response personnel in each facility. This information will be utilized to generate a written close-out report upon termination of the emergency event. The report will be prepared as described in procedure 73EP-ADM-001-0.

7.4.2 Records generated during actual emergencies will be maintained in accordance with 20AC-ADM-002-0.

MGR-0001 Rev. 4

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _1_ Attachment Page TITLE: Fission Product Barrier Evaluation Chart 1 Of 1 SEE THE "FISSION PRODUCT BARRIER" EVALUATION CHART MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _2_ Attachment Page TITLE: "Hot" Initiating Condition Matrix - Modes 1, 2 and 3 Evaluation Chart 1 Of 1 SEE THE "HOT" INITIATING CONDITION MATRIX EVALUATION CHART -

MODES 1,2 & 3 MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions ATTACHMENT _3_

I 73EP-EIP-001-0 17.0 Attachment Page TITLE: "Cold" Initiating Condition Matrix - Modes 4, 5 and Defueled Evaluation 1 Of 1 Chart SEE THE "COLD" INITIATING CONDITION MATRIX EVALUATION CHART ..

MODES 4, 5 & DEFUELED MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 1 Of 114 Hatch Nuclear Plant EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES, AND BASIS MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions Emergency_ II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 2 Of 114 TABLE OF CONTENTS ATTACHMENT PAGE 1.0 PURPOSE 3

2.0 BACKGROUND

4 3.0 ACRONYMS 7 4.0 DEFINITIONS 11 5.0 MODE DESCRIPTION 18 6.0 INITIATING CONDITIONS, THRESHOLD VALUES & BASIS 19 6.1 Category R - Abnormal Radiological 20 6.2 Category f - fission Product Barrier 35 6.3 Category S - ~stem Malfunctions - Hot Matrix 41 6.4 Category C -Cold Shutdown System Malfunctions 63 6.5 Category E - ISFSI Events 83 6.6 Category H - Hazards and Others 85 MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 3 Of 114 1.0 PURPOSE The purpose of this Attachment is to provide additional guidance and clarification to the EAL classification Initiating Conditions (IC) matrices in Attachments 1-3 of this procedure. They are utilized in the classification of off-normal events into one of four emergency classification levels.

This attachment provides the IC, Threshold Values (TV), and Basis for each Emergency Action Levels (EALs) grouped by their Recognition Categories. If after reviewing the classification IC matrices, the classification of an event or the determination if a Threshold Value is met or exceeded is unclear, the basis information provides additional clarification for each IC.

There are three considerations related to emergency classes. These are:

(1) The potential impact on radiological safety, either as now known or as can be reasonably projected; (2) How far the plant is beyond its predefined design, safety, and operating envelopes; and (3) Whether or not conditions that threaten health are expected to be confined to within the site boundary.

Although the majority of the EALs provide very specific thresholds, the Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classes (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classes. At all times, when conditions present themselves that are not explicitly provided in the EAL scheme the Emergency Director has discretion to declare an event based on his knowledge of the emergency classes and judgment of the situation or condition. Specific EALs (HU5, HA6, HS3, & HG2) are provided within the scheme to allow these discretionary classifications.

The classification procedure is written to classify events based on meeting the IC and a TV for an EAL considering each Unit independently. The IC Matrices are human factored to read from top to bottom (i.e., General Emergency down to Notification of Unusual Event) within a category or subcategory to eliminate the higher classifications before reaching a lower classification. This arrangement lessens the possibility of under-classifying a condition. During events, the ICs and TVs are monitored and if conditions meet another higher EAL, that higher emergency classification is declared and appropriate notifications made. The Notifications are made on a site basis, not a Unit (1 or 2) basis. If both Units are in concurrent classifications, the highest classification must be used for the notification and the other Unit events noted on the SNC Emergency Notification form.

The SNC policy is that once an emergency classification is made, it cannot be downgraded to a lower classification. Termination criteria contained in procedure 73EP-EIP-004-0, Duties of Emergency Director shall be completed for an event to be terminated. At termination, on an event specific basis, the site can either enter normal operating conditions or enter a recovery condition as described in procedure 73EP-EIP-002-0, Recovery.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 4 of 114

2.0 Background

HNP must respond to a formal set of threshold conditions that require plant personnel to take specific actions with regard to notifying state and local governments and the public when certain off-normal indicators or events are recognized. Emergency classes are defined in 10 CFR 50. The levels of response and conditions leading to those responses are defined in a joint NRC/FEMA guidelines contained in Appendix 1 of NUREG-0654/FEMA-REP-1, Rev. 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," October 1980. The nuclear industry has developed NEI 99-01, Revision 4, a set of generic EAL guidelines and supporting basis, to use to develop the site specific EALs, their Threshold Values and Basis. This generic guidance is intended to clearly define conditions that represent increasing risk to the public and can give consistent classifications when applied at different sites. It is a NRC endorsed acceptable alternative to the guidance in NUREG-0654.

This information is presented by Recognition Category:

  • R - Abnormal Rad Levels/Radiological Effluent
  • .Q - Cold Shutdown.! Refueling System Malfunction
  • E - Events Related to Independent Spent Fuel Storage Installations (ISFSI)
  • f - fission Product Barrier Degradation
  • H - Hazards and Other Conditions Affecting Plant Safety
  • §. - §.ystem Malfunction In this Attachment each of the EALs in Recognition Categories R, C, 0, E, .!::!, and §. are structured in the following way:
  • Recognition Category - As listed above.
  • Emergency Class - NUE, Alert, Site Area Emergency or General Emergency.
  • Initiating Condition - Symptom or Event-Based, Generic Identification and Title.
  • Operating Mode Applicability - Power Operation, Startup, Hot Shutdown, Cold Shutdown, Refueling, Defueled, or All.
  • Threshold Value(s) corresponding to the IC.
  • Basis information for plant-specific readings and factors that may relate to changing the generic IC or EAL to a different emergency class, such as for Loss of All AC Power.

For Recognition Category f, the EAL information is presented in a matrix format. The method was chosen to clearly show the synergism among the EALs and to support more accurate dynamic assessments. For category f, the EALs are arranged by fission product barrier.

Classifications are based on various combinations of barrier challenges.

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Emergency Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Initiatillg Conditions, Threshold Values and Basis 5 Of 114 2.0 Background (cont.)

Emergency classes were established by the Nuclear Regulatory Commission (NRC), for grouping off-normal nuclear power plant conditions according to (1) their relative radiological seriousness, and (2) the time-sensitive onsite and off-site radiological emergency preparedness actions necessary to respond to such conditions. The existing radiological emergency classes, in ascending order of seriousness, are called: Notification of Unusual Event, Alert, Site Area Emergency, and General Emergency.

With the emergency classes defined, the Initiating Conditions and Threshold Values that must be met for each EAL to be placed under the emergency class can be determined. There are two basic approaches to determining these EALs. EALs and emergency class boundaries coincide for those continuously measurable, instrumented ICs, such as radioactivity, core temperature, coolant levels, etc. For these ICs, the EAL will be the threshold reading that most closely corresponds to the emergency class description using the best available information. For discrete (discontinuous) events, the approach will have to be somewhat different. Typically, in this category are internal and external hazards such as fire or earthquake. The purpose for including hazards in EALs is to assure that station personnel and offsite emergency response organizations are prepared to deal with consequential damage these hazards may cause. If, indeed, hazards have caused damage to safety functions or fission product barriers, this should be confirmed by symptoms or by observation of such failures. Of course, security events must reflect potential for increasing security threat levels.

The EALs and ICs can be grouped in one of several schemes. The classification scheme incorporates symptom-based, event-based, and barrier-based EALs and ICs.

Symptom-based EALs and ICs refers to those indicators that are measurable over some continuous spectrum, such as core temperature, coolant levels, containment pressure, etc. The level of seriousness indicated by these symptoms depends on the degree to which they have exceeded technical specifications, the other symptoms or events that are occurring contemporaneously, and the capability of the licensed operators to gain control and bring the indicator back to safe levels.

Event-based EALs and ICs refer to occurrences with potential safety significance, such as the failure of a high-pressure safety injection pump, a safety relief valve failure, or a loss of electric power to some part of the plant. The range of seriousness of these "events" is dependent on the location, number of contemporaneous events, remaining plant safety margin, etc. Several categories of emergencies have no instrumentation to indicate a developing problem, or the event may be identified before any other indications are recognized. For emergencies related to the reactor system and safety systems, the ICs shift to an event based scheme as the plant mode moves toward cold shutdown and refueling modes. For non-radiological events, such as FIRE, floods, wind loads, etc. event-based ICs are the norm.

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Emergenc~ Classification and Initial Actions Emergenct I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 6 Of 114 2.0 Background (cont.)

Barrier-based EALs and ICs refer to the level of challenge to principal barriers used to assure containment of radioactive materials contained within a nuclear power plant. For radioactive materials that are contained within the reactor core, these barriers are: fuel cladding, reactor coolant system pressure boundary, and primary containment. The level of challenge to these barriers encompasses the extent of damage (loss or potential loss) and the number of barriers concurrently under challenge. The fission product barrier matrix is a hybrid approach that recognizes that some events may represent a challenge to more than one barrier, and that the containment barrier is weighted less than the reactor coolant system pressure boundary and the fuel clad barriers.

The most common bases for establishing Threshold Values are the technical specifications and setpoints for each plant that have been developed in the design basis calculations and the Final Safety Analysis Report (FSAR). For those conditions that are easily measurable and instrumented, the boundary is likely to be the EAL (observable by plant staff, instrument reading, alarm setpoint, etc.) that indicates entry into a particular emergency class. That radiation level also may be the setpoint that closes the main steam isolation valves (MSIV) and initiates the reactor scram. This same radiation level threshold, depending on plant-specific parameters, also may be the appropriate EAL for a direct entry into an emergency class.

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Emergency Classification and Initial Actions EmerQency I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAl - Initiating Conditions, Threshold Values and Basis 7 Of 114 3.0 Acronyms Acronyms, when used within an IC, Threshold Value or the basis, are defined within the body of the document. With this method, the user has a direct reference to the acronym's usage without having to go to this section to determine its particular contextual meaning. Some exceptions exist for common nuclear power applications. In general the following list is provided for review purposes:

IJCi/gm Micro-Curie per Gram IJCi/sec Micro-Curie per Second AC Alternating Current AOP Abnormal Operating Procedure ARI Alternate Rod Insertion ARM Area Radiation Monitor, Alarm Response Manual ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CAS Central Alarm Station CDE Committed Dose Equivalent CFR Code of Federal Regulations CMT, CNMT, CTMT Containment CO 2 Carbon Dioxide CPM Counts Per Minute CPS Counts Per Second CRD Control Rod Drive DBE Design Basis Earthquake DC Direct Current DEI, DEI 131 Dose equivalent Iodine 131 DW Drywell DWRRM Drywell Wide Range Radiation Monitor EAl Emergency Action level ECCS Emergency Core Cooling System ECl Emergency Classification level ED Emergency Director EDG Emergency Diesel Generator ENN Emergency Notification Network ENS Emergency Notification System EOF Emergency Operations Facility MGR-0009 Rev. 5.0

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Emergenc~ Classification and Initial Actions Emergency II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 8 Of 114 3.0 Acronyms (Continued)

EOP Emergency Operating Procedure EPA Environmental Protection Agency EPIP, EIP Emergency Plan Implementing Procedure ERG Emergency Response Guideline ERO Emergency Response Organization ESF Engineered Safeguards Feature FAA Federal Aviation Administration FBI Federal Bureau of Investigation FEMA Federal Emergency Management Agency FPB Fission Product Barrier FSAR Final Safety Analysis Report GE General Emergency, General Electric GDC General Design Criteria GPM Gallons per Minute Ga"ons HCTL Heat Capacity Temperature Limit HNP Hatch Nuclear Plant HPCI High Pressure Coolant Injection H2 Hydrogen IC Initiating Condition 10 Inside Diameter IDLH Immediately Dangerous to Life and Health IPEEE Individual Plant Examination of External Events (GL 88-20)

IRM Intermediate Range Monitor ISFSI Independent Spent Fuel Storage Installation Insta"ation LCO Limiting Condition of Operation LER Licensee Event Report LFL Lower Flammability Limit LOCA Loss of Coolant Accident LPSI Low Pressure Safety Injection MCR Main Control Room MSIV Main Steam Isolation Valve MSL Main Steam Line mR millirem mi"irem mR/hr millirem per Hour mi"irem Mw Megawatt MGR-0009 Rev. 5.0

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Emergency: Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 9 Of 114 3.0 Acronyms (Continued)

NEI Nuclear Energy Institute NORAD North American Aerospace Defense Command NPP Nuclear Power Plant NRC Nuclear Regulatory Commission NOUE, NUE Notification of Unusual Event OBE Operating Basis Earthquake OCA Owner Controlled Area ODCM Offsite Dose Calculation Manual OPX Off Premise Extension ORO Offsite Response Organization O2 Oxygen PA Protected Area PAG Protective Action Guide PAR Protective Action Recommendation PBX Private Business Exchange PRAIPSA Probabilistic Risk Assessment/Probabilistic Safety Assessment PSIG Pounds per Square Inch Gauge Rlhr Rem per Hour R Rem RAS Required Action Statement RB Reactor Building RECP Radiological Effluent Control Plan RETS Radiological Effluent Technical Specification RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System RPS Reactor Protection System RPV Reactor Pressure Vessel RWL Reactor Water Level Rx Reactor SAE Site Area Emergency SAS Secondary Alarm Station SAT Startup Auxiliary Transformer SBGTS Stand-By Gas Treatment System SFP Spent Fuel Pool SI Safety Injection MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions lI 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 10 Of 114 3.0 Acronyms (Continued)

SM Shift Manager SPDS Safety Parameter Display System SRM Source Range Monitor TEDE Total Effective Dose Equivalent TOAF, TAF Top of Active Fuel TS Technical Specification TSC Technical Support Center TV Threshold Value VA Vital Area VAC Volts Alternating Current VDC Volts Direct Current VOIP Voice Over Internet Protocol MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 11 Of 114 4.0 DEFINITIONS 4.1 ALERT The classification of Alert applies to situations in which events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of intentional malicious dedicated efforts of hostile action. Any releases of radioactive material for the Alert classification are expected to be limited to small fractions of the U.S. Environmental Protection Agency (EPA) Protective Action Guideline (PAG) exposure levelsDiscussion: Rather than discussing the distinguishing features of "potential degradation" and "potential substantial degradation," a comparative approach would be to determine whether increased monitoring of plant functions is warranted at the Alert level as a result of safety system degradation. This addresses the operations staff's need for help, independent of whether an actual decrease in plant safety is determined. This increased monitoring can then be used to better determine the actual plant safety state, whether escalation to a higher emergency class is warranted, or whether de-escalation or termination of the emergency class declaration is warranted. Dose consequences from these events are small fractions of the EPA PAG plume exposure levels, i.e., about 10 mR to 100 mR TEDE.

4.2 CIVIL DISTURBANCE An organized demonstration by an individual or group of unexpected, unidentified, or unauthorized people within the Owner Controlled Area (OCA) which is used to promote a political or social issue or belief.

4.3 CLOSED WINDOW A term indicating the position of the Beta radiation shield on a dose rate instrument or its probe. It allows the instrument to read Gamma radiation only for use in measuring/estimating the DOE.

4.4 COMMITTED DOSE EQUIVALENT (CDE)

This term means the dose equivalent to organs or tissues of reference that will be received from an intake of radioactive material by an individual during the 50-year period following the intake.

4.5 COMMITTED EFFECTIVE DOSE EQUIVALENT (CEDE)

The sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the committed dose equivalent to these organs or tissues.

4.6 CONFINEMENT BOUNDARY The barrier(s) between areas containing radioactive SUbstances sUbstances and the environment. This term is used in reference to ISFSI les.

4.7 CONTAINMENT INTEGRITY The Primary Containment is OPERABLE per Tech nical Specification 3.6.1.1. The Secondary Containment is OPERABLE per Technical Specification 3.6.4.1.

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Emergency Classification and Initial Actions EmemencY' II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 12 Of 114 4.0 DEFINITIONS (Continued) 4.8 CONTAINMENT BARRIER The Primary Containment Barrier includes the Drywell, the Torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

4.9 CREDIBLE THREAT A threat is considered credible through use of 82SS-SEC-051-0, Threat Assessment. A threat is credible when (1) physical evidence supporting the threat exists, or (2) information independent from the actual threat message exists that support the threat, or (3) a specific group or organization claims responsibility for the threat, or (4) a message (written or verbal) is received that contains specific information about plant locations, systems or device description an average person would most likely not know. The determination of credibility should be made by the Shift Manager with input from the Shift Captain or their designated representatives.

4.10 DEEP DOSE EQUIVALENT (DOE)

A term which applies to external whole body exposure, it is the dose equivalent at a tissue depth of 1 cm.

4.11 EMERGENCY ACTION LEVEL (EAL)

A pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency class. An EAL can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency class.

Discussion: There are times when an EAL will be a threshold point on a measurable continuous function, such as a primary system coolant leak that has exceeded technical specifications for a specific plant. At other times, the EAL and the IC will coincide, both identified by a discrete event that places the plant in a particular emergency class.

4.12 EXPLOSION A rapid, violent, unconfined combustion or catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

4.13 FUEL CLAD BARRIER The Fuel Clad barrier is the zircalloy tubes that contain the fuel pellets.

4.14 FIRE Combustion characterized by heat and light. Sources of smoke (Le. slipping belts, overheated electrical equipment) do not constitute FIREs. Observation of flame is preferred but NOT required if large quantities of smoke and heat are observed.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL -Initiating_ Conditions, Threshold Values and Basis 13 Of 114 4.0 DEFINITIONS (Continued) 4.15 GENERAL EMERGENCY (GE)

The classification of GE applies to those events which are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential loss of containment integrity or hostile action that results in an actual loss of physical control of the facility. Release of radioactive material for the GE classification can reasonably be expected to exceed EPA P AG exposure levels offsite for more than the immediate site area Discussion: The bottom line for the General Emergency is whether evacuation or sheltering of the general public is indicated based on EPA PAGs, and therefore should be interpreted to include radionuclide release regardless of cause. In addition, it should address concerns as to uncertainties in systems or structures (e.g. containment) response, and also events such as severe spent fuel pool events. To better assure timely notification, EALs in this category must primarily be expressed in terms of plant function status, with secondary reliance on dose projection. In terms of fission product barriers, loss of two barriers with loss or potential loss of the third barrier constitutes a General Emergency.

4.16 HOSTILE ACTION An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, takes hostages, and lor intimidates the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non- terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

4.17 HOSTILE FORCE One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

4.18 IDENTIFIED leakage Is defined as the measured leakage into the Drywell equipment drain system 4.19 IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH)

A condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

4.20 IMPEDE As used in this procedure, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Conditions, Threshold Values and Basis 14 Of 114 4.0 DEFINITIONS (Continued) 4.21 INITIATING CONDITION (IC)

One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency, or such an emergency has occurred. This term may be analogous to Emergency Action Level (EAL) is some cases. Discussion: Defined in this manner, an IC is an emergency condition which sets it apart from the broad class of conditions that mayor may not have the potential to escalate into a radiological emergency. It can be a continuous, measurable function that is outside technical specifications, such as elevated RCS temperature or falling reactor coolant level (asymp1om). It also encompasses occurrences such as FIRE (an event) or reactor coolant pipe failure (an event or a barrier breach).

4.22 LOWER FLAMMABILITY LIMIT (LFL)

The minimum concentration of a combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and a gaseous oxidizer.

4.23 NORMAL PLANT OPERATIONS Are activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONS.

4.24 NOTIFICATION OF UNUSUAL EVENT (NUE)

The classification of a NUE applies to situations in which events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Discussion: Potential degradation of the level of safety of the plant is indicated primarily by exceeding plant technical specification Limiting Condition of Operation a"owable action statement time for achieving required mode change. Precursors of (LCO) allowable more serious events should also be included because precursors do represent a potential degradation in the level of safety of the plant. Minor releases of radioactive materials are included. In this emergency class, however, releases do not require monitoring or offsite response (e.g., dose consequences of less than 10 mR).

4.25 OWNER CONTROLLED AREA The utility owned property around the plant where access is controlled during declared emergencies by the plant security force.

4.26 PROTECTED AREA (PA)

The area which normally encompasses all controlled areas within the security protected area fence. The IFSFI "protected area" is not included in this general definition because it has a separate category for its ICs.

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EmeCgenc~ Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 15 Of 114 4.0 DEFINITIONS (Continued) 4.27 REACTOR COOLANT SYSTEM (RCS) BARRIER The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

4.28 RECOGNITION CATEGORY A term that describes the broad categories that the Initiating Conditions (IC) (Le. Emergency Action Levels (EALs>> have been divided into to make the correct IC selection easier for the user. These categories have distinct letters associated with them which are used as the first letter of the IC.

These are: S- .§.ystem Malfunction, H- Hazards and Other, F- fission Product Barrier, R- Abnormal Radiation (Radiological),

C- Cold Shutdown System Malfunctions and E- ISFSI .Events.

4.29 SIGNIFICANT TRANSIENT An UNPLANNED event involving one or more of the following: (1) automatic runback >25%

thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Scram, (4) Safety System Injection Activation, or (5) thermal power oscillations >10%.

4.30 SITE AREA EMERGENCY (SAE)

The classification of a SAE applies to those events which are in progress or have occurred that involve actual or likely major failures of plant functions needed for protection of the public from radiation or contamination or hostile action that results in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) prevent effective access to, equipment needed for the protection of the public. Any releases of radioactive material for the SAE classification are not expected to exceed EPA PAG exposure levels except near the site boundary. Discussion: The discriminator (threshold) between Site Area Emergency and General Emergency is whether or not the EPA PAG plume exposure levels are expected to be exceeded outside the site boundary. This threshold, in addition to dynamic dose assessment considerations discussed in the EAL guidelines, clearly addresses NRC and offsite emergency response agency concerns as to timely declaration of a General Emergency.

4.31 STATION BLACKOUT A complete loss of offsite and onsite AC power, as indicated by failure to energize any 4160VAC Emergency bus.

4.32 STRIKE ACTION MGR-0009 Rev. 5.0

A work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONS.

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-Initiating Conditions, Threshold Values and Basis 16 Of 114 4.0 DEFINITIONS (Continued) 4.33 SUBCRITICL SUBCRITICAL: IRMs below Range 6 and Period is negative. Discussion: The definition of SUBCRITICAL (IRMs below Range 6 and Period is negative) is the point at which the EOP guidance directs the operators to exit the EOPs and continue shutdown activities with normal plant procedures.

4.34 THRESHOLD VALUE (TV)

Is a pre-determined, site-specific, observable threshold for a plant Initiating Condition (IC) that places the plant in a given emergency classification. A TV can be: an instrument reading; an equipment status indicator; a measurable parameter (onsite or offsite); a discrete, observable event; results of analyses; entry into specific emergency operating procedures; or another phenomenon which, if it occurs, indicates entry into a particular emergency class. This term is analogous to the term Emergency Action Level (EAL) is most cases.

4.35 TOTAL EFFECTIVE DOSE EQUIVALENT (TEDE)

Is a term that means the sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

4.36 UNIDENTIFIED leakage Is defined as the measured leakage into the Drywell floor drain system 4.37 UNPLANNED Is a parameter change or an event that is NOT the result of an intended evolution and requires corrective or mitigative actions. This includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit.

4.38 VALID An indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

4.39 VISIBLE DAMAGE Is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering.

Surface blemishes (e.g., paint chipping, scratches) should not be included.

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Emergency Classification and Initial Actions II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 17 Of 114 4.0 DEFINITIONS (Continued) 4.40 VITAL AREA Any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. This includes the Control Building, Reactor Building, Diesel Generator Building, Intake Structure and Primary containment.

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Emergency Classification and Initial Actions EmerQency I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 18 Of 114 5.0 MODE DESCRIPTION Hatch Units 1 and 2 Technical Specifications Table 1.1-1 provides the following mode definitions:

Mode Title Reactor Mode Switch Position Average ReS Temperature (OF) 1 Power Operation Run NA 2 Startup Refuel(a) or Startup/Hot Standby NA 3 Hot Shutdown (a) Shutdown Greater than 212 4 Cold Shutdown (a) Shutdown Less than or equal to 212 5 Refueling (b) Shutdown or Refueling NA (a) All reactor vessel head closure bolts fully tensioned.

(b) One or more reactor vessel head closure bolts less than fully tensioned These modes are used throughout the Hatch EALs with no modifications from NEI 99-01. For the condition when a unit is defueled, the Initiating Conditions designated as Mode Condition "ALL" or "Defueled" are applicable.

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Emer~ency Classification and Initial Actions Emergency I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 19 Of 114 6.0 HNP EALS - INITIATING CONDITIONS, THRESHOLD VALUES AND BASIS MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 21 Of 114 Initiating Condition RG1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE OR 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology.

Operating Mode Applicability: All Threshold Values: (1 OR 2 OR 3)

NOTES:

  • If dose assessment results are available at the time of declaration, the classification should be based on Threshold Value #2 instead of Threshold Value #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently escalated.
  • The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the duration has or will likely exceed 15 minutes.
1. VALID reading on either of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:

Reactor Building Vent Accident Range Monitor - 1/2D11-P005 2.9 IJCi/cc Main Stack Accident Range Monitor- 1D11-P006 8.8 x 103 IJCi/cc OR

2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE OR 5000 mR thyroid CDE at or beyond the site boundary.

OR

3. Field survey results indicate CLOSED WINDOW dose rates exceeding 1000 mRlhr expected to continue for more than one hour; OR analyses of field survey samples indicate thyroid CDE of 5000 mR for one hour of inhalation, at or beyond site boundary.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 22 Of 114 Basis: RG1 NOTE: The values used in TV# 1 are based on detailed calculations contained in Attachment 8-2 of SNC Letter (# NL-07-0522) to the NRC dated April 2, 2007,

["Response to Request for Additional Information"].

CLOSED WINDOW: is a term indicating the position of the Beta radiation shield on a dose rate instrument or its probe. It allows the instrument to read Gamma radiation only for use in measuring/estimating the Deep Dose Equivalent (DDE).

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. While these failures are addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone.

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The monitor list in Threshold Value #1 includes monitors on potential release pathways.

The monitor reading Threshold Values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term used are the same as those used for determining the monitor reading Threshold Values in ICs RU1 and RA1. Since doses are generally not monitored in real-time, a release duration of one hour is assumed, and that the Threshold Values are based on a site boundary (or beyond) dose of 1000 mRihour whole body or 5000 mR/hour thyroid, whichever is more limiting.

Since dose assessment (Utilizing procedure 73EP-EIP-015-0, Offsite Dose Assessment or 73EP-EIP-018-0, Prompt Dose Assessment) is based on actual meteorology, whereas the monitor reading Threshold Values are not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made, the dose assessment results override the monitor reading Threshold Values. Classification should NOT be delayed pending the results of these dose assessments.

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EmerQency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 23 Of 114 Initiating Condition RS1 Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR TEDE OR 500 mR Thyroid CDE for the Actual or Projected Duration of the Release.

Operating Mode Applicability: All Threshold Values: (1 OR 2 OR 3)

NOTES:

  • If dose assessment results are available at the time of declaration, the classification should be based on Threshold Value #2 instead of Threshold Value #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated I completed in order to determine if the classification should be subsequently escalated.
  • The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the duration has or will likely exceed 15 minutes.
1. VALID reading on either of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:

Reactor Building Vent Accident Range Monitor -1/2D11-P005 0.29 IJCi/cc Main Stack Accident Range Monitor - 1D11-P006 8.8 x 102 IJCi/cc OR

2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE OR 500 mR thyroid CDE at or beyond the site boundary.

OR

3. Field survey results indicate CLOSED WINDOW dose rates exceeding 100 mR/hr expected to continue for more than one hour; OR analyses of field survey samples indicate thyroid CDE of 500 mR for one hour of inhalation, at or beyond the site boundary.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 24 Of 114 Basis: RS1 NOTE: The values used in TV# 1 are based on detailed calculations contained in Attachment 8-2 of SNC Letter (# NL-07-0522) to the NRC dated April 2, 2007,

["Response to Request for Additional Information"].

VALID: An indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

CLOSED WINDOW: is a term indicating the position of the Beta radiation shield on a dose rate instrument or its probe. It allows the instrument to read Gamma radiation only for use in measuring/estimating the Deep Dose Equivalent (DOE).

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed a small fraction of the EPA Protective Action Guides (pAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. While these failures are addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone.

The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The monitor list in Threshold Value #1 includes monitors on potential release pathways.

The monitor reading Threshold Values are determined using a dose assessment method that back calculates from the dose values specified in the IC. The meteorology and source term used is the same as those used for determining the monitor reading Threshold Values in ICs RU1 and RA1. Since doses are generally not monitored in real-time, a release duration of one hour is assumed and the Threshold Values are based on a site boundary (or beyond) dose of 100 mRihour whole body or 500 mRihour thyroid, whichever is more limiting.

Since dose assessment (Utilizing procedure 73EP-EIP-015-0, Offsite Dose Assessment or 73EP-EIP-018-0, Prompt Dose Assessment) is based on actual meteorology, whereas the monitor reading Threshold Values are not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures call for the timely performance of dose assessments using actual meteorology and release information. If the results of these dose assessments are available when the classification is made, the dose assessment results override the monitor reading Threshold Values. Classification should not be delayed pending the results of these dose assessments.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 25 Of 114 Initiating Condition Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or Longer.

Operating Mode Applicability: All Threshold Values: (1 OR 2)

NOTES:

  • The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the duration has or will likely exceed 15 minutes.
1. VALID reading on any effluent monitor that exceeds 200 times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer.

Monitor 200 X Setpoint Value Liquid Radwaste Effluent Line Monitor - 1/2D11-K604 8.39 x 106 cpm Service Water System Effluent Line Monitor - 1/2D11-K605 2.02 x 106 cpm Reactor Building Vent Normal Range Monitor-1D11-K619 A(B) 1.52 x 104 cpm

& 2D11-K636 A(B)

Recombiner Building Vent Monitor -1 D11-R763 A(B) 3.16 x 107 cpm Main Stack Normal Range Monitor - 1 D11-K600 A(B) 6.08 x 103 cps OR

2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates in excess of 200 times the values indicated in the Offsite Dose Calculation Manual (ODCM), with a release duration of 15 minutes or longer.

Sample Source Release Rate Liquid Radwaste Effluent Line 5.1 x 10-1 IJCi/cc Service Water System Effluent Line 5.1 x 10-1 IJCi/cc Reactor Building Vent Stack 3.18 x 10-4 IJCi/cc Recombiner Building Vent 6.59 x 10-1 IJCi/cc Main Stack 1.22 x 10° IJCi/cc MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 26 Of 114 Basis:

NOTE: The values used in TV# 1 and 2 are based on detailed calculations contained in Attachment 8-2 of SNC Letter (# NL-07-0522) to the NRC dated April 2, 2007,

["Response to Request for Additional Information"].

UNPLANNED: a parameter change or an event that is NOT the result of an intended evolution and requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses a potential or actual decline in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Nuclear power plants incorporate features intended to control the release of radioactive effluents to the environment.

Administrative controls are established to prevent unintentional releases, and to control and monitor intentional releases. These controls are located in the ODCM. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

The Radiological Effluent Control Program (RECP) multiples are specified in ICs RU1 and RA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose or dose rate. Releases should not be prorated or averaged.

UNPLANNED, as used in this context, includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.

Threshold Value #1 addresses radioactivity releases that for whatever reason cause effluent radiation monitor readings that exceed two hundred times the alarm setpoint established by the radioactivity discharge permit. This alarm setpoint may be associated with a planned batch release, or a continuous release path. Indexing the Threshold Value to the ODCM setpoints in this manner ensures that the Threshold Value will never be less than the setpoint established by a specific discharge permit.

Setpoints are 100 times those of RU1.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 27 Of 114 Basis (cont.):

Threshold Value # 2 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.

Threshold Values #1 and #2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the RECP and is used in calculating the alarm setpoints. The fundamental basis of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant implied by the uncontrolled release.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 28 Of 114 Initiating Condition Damage to Irradiated Fuel OR Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.

Operating Mode Applicability: All Threshold Values: (1 OR 2)

1. UNPLANNED VALID alarm on any of the following radiation monitors:

Refuel Floor Area Radiation Monitors:

  • 1D21-K601 M - Spent Fuel Pool and New Fuel Storage area
  • 2D21-K611 L - RPV Refuel Floor 228' Refuel Floor Ventilation Monitors:
  • 1/2D11-K609 A-D - Rx Bldg. Potential Contaminate Area Vent Exhaust Rad Monitor
  • 1/2D11-K611 A-D - Refuel Floor Vent Exhaust
  • 2D11-K634 A-D - Refuel Floor Rx Well Vent. Exhaust
2. Loss of water level that has or will result in the uncovering of irradiated fuel outside the Reactor Vessel as indicated by ANY of the following:

Personnel report of irradiated fuel uncovered during fuel assembly movements.

Spent Fuel Pool- Less than 202' elevation (Top of Fuel Racks)

Fuel Transfer Canal - Less than 217' 6" elevation (Top of Fuel Bundle in Transit)

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 29 Of 114 Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses specific events that have resulted, or may result, in unexpected rises in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent degradation in the level of safety of the plant.

Threshold Value #1 addresses radiation monitor indications of fuel uncovery and/or fuel damage. Raised readings on ventilation monitors may be indication of a radioactivity release from the fuel, confirming that damage has occurred. Raised background at the monitor due to water level lowering may mask raised ventilation exhaust airborne activity and needs to be considered. While a radiation monitor could detect a rise in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered.

In Threshold Value #2, indications include water level and personnel reports. Visual observation will be the primary indicator for spent fuel pool and fuel movement activities. Personnel report of personnel during fuel assembly movements is included to ensure that reports of actual or potential fuel uncovery are classified. Depending on available level indication, the declaration threshold may need to be based on indications of water makeup rate or lowering in makeup tank level.

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Release of Radioactive Material or Rises in Radiation Levels Within the Facility That IMPEDES Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Operating Mode Applicability: All Threshold Values: (1 OR 2)

1. VALID radiation readings greater than 15 mRlhr in areas requiring continuous occupancy to maintain plant safety functions:

Control Room area radiation monitor 1D21-K600 B or C Central Alarm Station (by survey)

OR

2. UNPLANNED VALID ARM readings greater than 1000 mRlhr in areas requiring infrequent access to maintain plant safety functions.

U1 Reactor Building as indicated by 1D21-K601 C Spent Fuel Pool Demin Equipment ARMs: 1D21-K601 F Tip Area 1D21-K601 G 130' NE Working Area 1 D21-K601 H 130' SW Working Area 1 D21-K601 K 158' Working Area 1D21-K601 L Decant Pump and Equip Room 1D21-K601 N South CRD Hydraulic Units 1D21-K601 P North CRD Hydraulic Units 1 D21-K601 R South Core Spray and RHR Area 1 D21-K601 S Equipment Access Airlock 1D21-K601 T HPCI Turbine Area 1D21-K601 U Tip (Core) Probe Drive Area 1D21-K601 V RCIC Equipment Area SW 1D21-K601 W CRD Pump Room NW 1 D21-K601 X RB 203' Working Area 1D21-K601 Y North Core Spray and RHR Area U2 Reactor Building as indicated by 2D21-K601 B 158' Level SE ARMs: 2D21-K601 C 158' Level NE 2D21-K601 D 158' Level NW 2D21-K601 F Tip Area 2D21-K601 G 130' NE Working Area 2D21-K601 H 130' SW Working Area 2D21-K601 L Decant Pump & Equip Room 2D21-K601 N South CRD Hydraulic Units 2D21-K601 P Spent Fuel Pool Passageway 2D21-K601 R 185' Level Operating Floor 2D21-K601 S 185' Level Sample Panel 2D21-K601 T CRD Repair Area 2D21-K601 U 185' Level RWCU Control Panel 2D21-K601 V RCIC Equipment Area 2D21-K601 W CRD Pump Room SW 2D21-K601 X RHR & Core Spray Room NE 2D21-K601 Y RHR & Core Spray Room SE MGR-0009 Rev. 5.0

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DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating InitiatinQ Conditions, Threshold Values and Basis 31 Of 114 Basis:

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

IMPEDE: As used here, includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant.

This IC addresses raised radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually or that requires local monitoring, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the rise in radiation levels is not a concern of this IC. The Emergency Director must consider the source or cause of the raised radiation levels and determine if any other IC may be involved.

This IC is not meant to apply to raises in the containment radiation monitors as these are events which are addressed in the fission product barrier matrix ICs. Nor is it intended to apply to anticipated temporary rises due to planned events. (e.g., incore detector movement, Radwaste container movement, depleted resin transfers, etc.)

The monitored area requiring continuous occupancy is the Control Room. The value of 15 mRlhr is derived from the GDC 19 value of 5 Rem in 30 days with adjustment for expected occupancy times.

For areas requiring infrequent access, the 1000 mRlhr (Locked High Rad Area) is based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits, and in doing so, will IMPEDE necessary access.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 32 Of 114 Initiating Condition RU1 Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or Longer.

Operating Mode Applicability: All Threshold Values: (1 OR 2)

NOTE:

  • The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the duration has or will likely exceed 60 minutes.
1. VALID reading on any effluent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

Monitor 2 x Setpoint Value Liquid Radwaste Effluent Line Monitor - 1/2D11-K604 8.39 x 104 cpm Service Water System Effluent Line Monitor - 1/2D11-K605 2.02 x 104 cpm Reactor Building Vent Normal Range Monitor-1D11-K619 A(B) &

1.52 x 102 cpm 2D11-K636 A(B)

Recombiner Building Vent Monitor -1 D11-R763 A(B) 3.16 x 105 cpm Main Stack Normal Range Monitor- 1D11-K600 A(B) 6.08 x 101 cps OR

2. Confirmed sample analyses for UNPLANNED gaseous or liquid releases indicates concentrations with a release duration of 60 minutes or longer, in excess of two times Technical Specification 5.5.4 as confirmed by Offsite Dose Calculation Manual (ODCM).

Sample Source Release Rate Liquid Radwaste Effluent Line 5.1 x 10-3 IJCi/cc Service Water System Effluent Line 5.1 x 10-3 IJCi/cc Reactor Building Vent Stack 3.18 x 10-6 IJCi/cc Recombiner Building Vent 6.59 x 10-3 IJCi/cc Main Stack 1.22 x 10-2 IJCi/cc MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Initiating Conditions, Threshold Values and Basis 33 Of 114 Basis: RU1 NOTE: The values used in TV# 1 and 2 are based on detailed calculations contained in Attachment 8-2 of SNC Letter (# NL-07-0522) to the NRC dated April 2, 2007,

["Response to Request for Additional Information"].

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions. As used in this context, UNPLANNED includes any release for which a radioactivity discharge permit was not prepared, or a release that exceeds the conditions (e.g.,

minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the Emergency Director should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.

This IC addresses a potential or actual decline in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. Administrative controls are established to prevent unintentional releases, or control and monitor intentional releases.

These controls are located in the ODCM. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

Threshold Value #1 addresses radioactivity releases that, for whatever reason, cause effluent radiation monitor readings to exceed two times the Technical Specification limit and releases are not terminated within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the ODCM to warn of a release that is not in compliance with the Tech Spec 5.5.4. Indexing the Threshold Value to the ODCM setpoints in this manner ensures that the Threshold Value will never be less than the setpoint established by a specific discharge permit. 60 minutes is used is used because it signifies a degradation in the level of plant safety because of the ongoing release vs. a single discharge over the ODCM limits which has terminated.

Threshold Value 2 addresses UNPLANNED uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.

Thresholds #1 and #2 directly correlate with the IC since annual average meteorology is required to be used in showing compliance with the TS 5.5.4 and is used in calculating the alarm setpoints. However, the fundamental basis of this IC is NOT a dose or dose rate, but rather the degradation in the level of safety of the plant implied by the uncontrolled release.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 34 Of 114 Initiating Condition Unexpected Rise in Plant Radiation.

Operating Mode Applicability: All Threshold Values: [(1.a AND 1.b.) OR 2]

1. a. VALID indication of uncontrolled water level lowering in the reactor refueling cavity, spent fuel pool (SFP) or fuel transfer canal with all irradiated fuel assemblies remaining covered by water as indicated by any of the following:

Personnel report of low water level SFP low level alarm annunciator - Spent Fuel Storage Pool Level Low 654-022-1/2

b. UNPLANNED VALID Area Radiation Monitor (ARM) reading rise on any of the following:

1D21-K601 A - Rx Head Laydown Area 1D11-K601 0 - Refuel Floor 1D21-K601 E -Drywell Shield Plug 1D21-K601 M - Spent Fuel Pool and New Fuel Storage area 2D21-K601 A - Rx Head Laydown Area 2D21-K601 M - Spent Fuel/Fuel Pool Areas 2D21-K601 E - Dryer/Separator Pool 2D21-K611 K - RPV Refuel Floor 228' 2D21-K611 L - RPV Refuel Floor 228' NOTE:

  • Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.
2. UNPLANNED VALID ARM readings rise by a factor of 1000 over normal levels.

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UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC addresses raised radiation levels as a result of water level lowering above the RPV flange or events that have resulted, or may result, in unexpected rises in radiation dose rates within plant buildings.

These radiation rises represent a loss of control over radioactive material and may represent a potential degradation in the level of safety of the plant.

Threshold 1a limits: Personnel report of low water level is the primary indicator.

While other radiation monitors could detect a rise in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. Increased radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss.

Threshold Value #2 addresses UNPLANNED rises in in-plant radiation levels that represents a degradation in the control of radioactive material, and represent a potential degradation in the level of safety of the plant.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 36 Of 114 6.2 Category F .. Fission Product Barrier MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 37 Of 114 Fission Product Barrier Evaluation GENERAL EMERGENCY Initiating Condition FG1 - Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier SITE AREA EMERGENCY Initiating Condition FS1 - Loss or Potential Loss of ANY Two Barriers ALERT Initiating Condition I?& ~ - ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCS Barrier UNUSUAL EVENT Initiating Condition

=-..;;;::-::.. - ANY Loss or ANY Potential Loss of Containment Barrier Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

FUEL CLAD BARRIER THRESHOLD VALUES The Fuel Clad barrier is the zircalloy tubes that contain the fuel pellets.

1. Primary Coolant Activity Level The "Loss" value 300 IJCi/gm DEI 131 or 2.93E+03 IJCi/gm (2.06E+03 IJCi/gm non-gaseous) total RCS activity value corresponds to 300 IlCi/gm 1131 equivalent. Assessment indicates that this amount of coolant activity is well above that expected for iodine spikes and corresponds to less than 5% fuel clad damage. This amount of radioactivity indicates significant clad damage and thus the Fuel Clad Barrier is considered lost.

There is no "Potential Loss" Threshold associated with this item.

2. Reactor Vessel Water Level The "Loss" Threshold (less than -193 inches) is the minimum value to assure core cooling without further degradation of the clad. This value corresponds to 2/3 coverage of active fuel.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 38 Of 114 The "Potential Loss" Threshold (less than -155 inches) is the same as the RCS barrier "Loss" Threshold #2 below and is -155 inches which is used in the EOPs for operator actions. Thus, this Threshold indicates a "Loss" of RCS barrier and a "Potential Loss" of the Fuel Clad Barrier. This Threshold appropriately escalates the emergency class to a Site Area Emergency.

3. Drywell Radiation Monitoring The 7000 R/hr DWRRM reading is a value which indicates the release of Reactor coolant, with elevated activity indicative of fuel damage, into the Drywell. The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 /-lCi/gm dose equivalent 1131 or the calculated concentration equivalent to the clad damage used in Threshold #1 into the drywell atmosphere. Reactor coolant concentrations of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within technical specifications and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier Loss Threshold #4. Thus, this Threshold indicates a loss of both Fuel Clad barrier and RCS barrier.

There is no "Potential Loss" Threshold associated with this item.

4. Other Indications Offgas pre- and post-treatment monitors Offscale High and Drywell Post LOCA Monitor Offscale High detect the effluent of the Offgas system and therefore indicate fission products escaping the clad.

There is no "Potential Loss" Threshold associated with this item.

5. Emergency Director Judgment This Threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this Threshold as a factor in Emergency Director's judgment that the barrier may be considered lost or potentially lost. (See also IC# SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

Res BARRIER THRESHOLD VALUES The RCS Barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

1. Drywell Pressure The 1.85 psig drywell pressure is based on the drywell high pressure set point which indicates a LOGA by automatically initiating the EGGS or equivalent makeup system. Pressure increases requiring containment venting for temperature or pressure (Le. loss of chillers, Drywell Nitrogen fill event, etc.)

when not in an accident situation should not be considered.

There is no "Potential Loss" Threshold associated with this item.

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2. Reactor Vessel Water Level This "Loss" Threshold is the same as "Potential Loss" Fuel Clad Barrier Threshold #2. The

-155" water level corresponds to the level which is used in EOPs to indicate challenge of core cooling.

This Threshold appropriately escalates the emergency class to a Site Area Emergency. Thus, this Threshold indicates a loss of the RCS barrier and a Potential Loss of the Fuel Clad Barrier.

There is no "Potential Loss" Threshold associated with this item.

3. ReS Leak Rate An unisolable MSL break is a breach of the RCS barrier. Thus, this Threshold is included for consistency with the Alert emergency classification.

The potential loss of RCS based on leakage is set at a level indicative of a small breach of the RCS but which is well within the makeup capability of normal and emergency high pressure systems. Core uncovery is not a significant concern for a 50 gpm leak; however, break propagation leading to significantly larger loss of inventory is possible.

Potential loss of RCS based on primary system leakage outside the drywell is determined from temperature or area radiation alarms setpoint in the areas of the main steam line tunnel, main turbine generator, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment. The indicators should be confirmed to be caused by RCS leakage. The area temperature or radiation alarm setpoints are indicated for this example to enable an Alert classification.

4. Drywell Radiation Monitoring The 138 Rlhr reading is a value which indicates the release of reactor coolant to the drywell. The reading was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (Le., within TIS) into the drywell atmosphere. This reading will be less than that specified for Fuel Clad Barrier Threshold #3.

Thus, this Threshold would be indicative of a RCS leak only.

There is no "Potential Loss" Threshold associated with this item.

5. Other (Site-Specific) Indications For the Loss, Drywell Post LOCA Monitor 4.71 E+04 cpm, indicates a breach of the RCS as an effluent.

No radiation monitors capable of indicating a potential loss of the RCS barrier were identified.

There is no "Potential Loss" Threshold associated with this item.

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6. Emergency Director Judgment This Threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this Threshold as a factor in Emergency Director's judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

PRIMARY CONTAINMENT BARRIER THRESHOLD VALUES The Primary Containment Barrier includes the Drywell, the Torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

Containment Barrier Thresholds are used primarily as discriminators for escalation from an Alert to a Site Area Emergency or a General Emergency.

1. Drywell Pressure CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Rapid unexplained loss of pressure (Le., not attributable to Containment spray or condensation effects) following an initial pressure increase indicates a loss of Primary CONTAINMENT INTEGRITY. Drywell pressure should rise as a result of mass and energy release into containment from a LOCA. Thus, drywell pressure not rising under these conditions indicates a loss of Primary CONTAINMENT INTEGRITY. This indicator relies on the operator's recognition of an unexpected response for the condition and therefore does not have a specific value associated. The unexpected response is important because it is the indicator for a containment bypass condition.

The 56 PSIG for potential loss of containment is based on the containment drywell design pressure.

Existence of an explosive mixture means a Hydrogen and Oxygen concentration of at least the lower deflagration limit curve exists.

Explosive mixture inside containment ~6% Hydrogen and ~5% Oxygen.

2. Reactor Vessel Water Level There is no "Loss" Threshold associated with this item.

The entry into the Primary Containment Flooding emergency procedure indicates reactor vessel water level cannot be restored and that a core melt sequence is in progress. EOPs direct the operators to enter Containment Flooding when Reactor Vessel Level cannot be restored to greater than 2/3 core height or is unknown. Entry into Containment Flooding procedures is a logical escalation in response to the inability to maintain reactor vessel level. The conditions in this potential loss Threshold represents imminent core melt sequences which, if not corrected, could lead to vessel failure and increased potential for containment failure. If the emergency operating procedures have been ineffective in restoring reactor vessel level above the RCS and Fuel Clad Barrier Threshold Values, MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 41 Of 114 there is not a "success" path and a core melt sequence is in progress. Entry into Containment flooding procedures is a logical escalation in response to the inability to maintain reactor vessel level. Severe accident analysis (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation with the reactor vessel in a significant fraction of the core damage scenarios, and the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow emergency operating procedures to arrest the core melt sequence.

Whether or not the procedures will be effective should be apparent within the time provided. The Emergency Director should make the declaration as soon as it is determined that the procedures have been, or will be, ineffective.

3. Containment Isolation Failure or Bypass This Threshold is intended to cover the inability to isolate the containment when containment isolation is required. In addition, the presence of area radiation or temperature alarms high setpoint indicating unisolable primary system leakage outside the drywell are covered after a containment isolation. The indicators should be confirmed to be caused by RCS leakage. Also, an intentional venting of primary containment for pressure control per EOPs to the secondary containment and/or the environment is considered a loss of containment. Containment venting for temperature or pressure when not in an accident situation (Le. loss of chillers, Drywell Nitrogen fill event, etc.) should not be considered.

There is no "Potential Loss" Threshold associated with this item.

4. Significant Radioactive Inventory in Containment NOTE: TV#4 is based on detailed calculations contained in Attachment 8-2 of SNC Letter (#

NL-07-0522) to the NRC dated April 2, 2007, ["Response to Request for Additional Information"].

There is no "Loss" Threshold associated with this item.

The 1.1 x 105 Rlhr DWRRM reading is a value which indicates significant fuel damage well in excess of that required for loss of RCS and Fuel Clad. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant. Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. Such conditions do not exist when the amount of clad damage is less than 20%.

6. Emergency Director Judgment This Threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this Threshold as a factor in Emergency Director's judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 43 Of 114 Initiating Condition SG1 Prolonged Loss of All Offsite Power AND Prolonged Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: [(1.a AND 1.b.) AND (2 OR 3)]

1. Loss of all AC power indicated by:
a. Loss of power to or from Startup Auxiliary Transformers (SAT) 1/2C and 1/20 resulting in loss of all off-site electrical power to 4160 VAC Emergency Buses 1/2E, 1/2F, and 1/2G for greater than 15 minutes AND
b. Failure of emergency diesel generators to supply power to emergency busses.

AND EITHER

2. Restoration of at least one 4160 VAC Emergency Bus, 1/2E, 1/2F, or 1/2G, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of time of loss is NOT likely.

OR

3. Fuel Clad Barrier Evaluation indicates continuing degradation (Loss OR Potential Loss) of core cooling.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 44 Of 114 Basis: SG1 Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore AC power is based on a site blackout coping analysis performed in conformance with 10 CFR 50.63 and Regulatory Guide 1.155, "STATION BLACKOUT". Appropriate allowance for offsite emergency response including evacuation of surrounding areas should be considered. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response.

STATION BLACKOUT: A complete loss of offsite and onsite AC power, as indicated by failure to energize any 4160VAC Emergency bus.

This IC is specified to assure that in the unlikely event of a prolonged STATION BLACKOUT, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded.

Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Director a reasonable idea of how quickly (s)he may need to declare a General Emergency based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of Fission Product Barriers is imminent?
2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

Thus, indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with particular emphasis on Emergency Director's judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAl -Initiating Conditions, Threshold Values and Basis 45 Of 114 Initiating Condition SG2 Failure of the Reactor Protection System to Complete an Automatic Scram AND Manual Scram was NOT Successful AND there is Indication of an Extreme Challenge to the Ability to Cool the Core.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Threshold Value: [1 AND (2.a. OR 2.b.)]

1. Indications exist that a reactor protection system setpoint was exceeded and automatic scram did not occur, and a manual scram did not result in the reactor being made SUBCRITICAL (IRMs below Range 6 and Period is negative).

AND

2. EITHER of the following:
a. Core cooling is severely challenged as indicated by inability to restore and maintain Reactor Water level greater than -185" on the affected unit.

OR

b. Heat removal is extremely challenged as indicated by Exceeding the Heat Capacity Temperature Limit (HCTl) Curve (EOP Graph 2).

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Initiating Conditions, Threshold Values and Basis 46 Of 114 Basis: SG2 SUBCRITICAL: IRMs below Range 6 and Period is negative.

Automatic and manual SCRAM are NOT considered successful if action away from the reactor control console (1/2H11-P603Panel) is required to SCRAM the reactor.

Under the conditions of this IC and its associated Threshold Values, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities, such as standby liquid control, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence. The extreme challenge to the ability to cool the core is intended to mean that the reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level (-185")as described in the EOP bases.

Another consideration is the inability to initially remove heat during the early stages of this sequence.

Considerations include inability to remove heat via the main condenser, or via the torus (e.g., due to high water temperature).

In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (5% power) a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time.

The Reactor should be considered SUBCRITICAL when IRMs below Range 6 and Period is negative.

A failure of the RPS to shut down the reactor (as indicated by reactor NOT being SUBCRITICAL) is a degraded plant condition that may require the injection of boron to shut down the reactor per EOP guidance. The definition of SUBCRITICAL (IRMs below Range 6 and Period is negative) is the point at which the EOP guidance directs the operators to exit the EOPs and continue shutdown activities with normal plant procedures.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 47 Of 114 Initiating Condition SS1 Loss of All Offsite Power AND Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: (1.a AND 1.b AND 1.c)

1. Loss of all AC power indicated by:
a. Loss of power to or from Startup Auxiliary Transformers (SAT) 1/2C and 1/2D resulting in loss of all off-site electrical power to 4160 VAC Emergency Buses 1/2E, 1/2F, and 1/2G for greater than 15 minutes AND
b. Failure of emergency diesel generators to supply power to emergency busses AND
c. Restoration of at least one 4160 VAC Emergency bus, 1/2E, 1/2F, or 1/2G, has NOT occurred within 15 minutes of time of loss of all AC power Basis:

CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of CONTAINMENT INTEGRITY, thus this event can escalate to a General Emergency. The 15 minute time duration is to exclude transient or momentary power losses.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to emergency busses. Even though an emergency bus may be energized, if necessary loads (Le., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 48 Of 114 Initiating Condition SS2 Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded AND Manual Scram Was NOT Successful.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Threshold Value:

1. Indications exist that a reactor protection system setpoint was exceeded AND Automatic scram did NOT occur, AND A manual scram did NOT result in the reactor being made SUBCRITICAL (IRMs below Range 6 and Period is negative).

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 49 Of 114 Basis: SS2 SUBCRITICAL: IRMs below Range 6 and Period is negative Automatic and manual scram are not considered successful if action away from the reactor control console (1/2H11-P603Panel) was required to scram the reactor.

This Threshold Value is not applicable if no RPS set points are exceeded prior to initiating a successful manual scram. RPS set points may be exceeded following a successful SCRAM when the mode switch is taken to the shutdown position. This may cause an RPS set point to be exceeded due to the change in Nuclear Instrumentation Scram set point when the mode switch is taken out of the Run position. If the RPS then fails to initiate a scram, then this should be evaluated as an automatic RPS set point being exceeded.

The RPS is designed to function to shut down the reactor. The system is "fail safe," in that it de-energizes to function. An Anticipated Transient Without Scram (ATWS) event results from either a failure of RPS (electrical failure) or the Control Rod Drive system to permit the control rods to insert.

The Reactor should be considered SUBCRITICAL when IRMs below Range 6 and Period is negative.

A failure of the RPS to shut down the reactor (as indicated by reactor NOT being SUBCRITICAL) is a degraded plant condition that may require the injection of boron to shut down the reactor per EOP guidance. The definition of SUBCRITICAL IRMs below Range 6 and Period is negative is the point at which the EOP guidance directs the operators to exit the EOPs and continue shutdown activities with normal plant procedures.

Under these continued power operation conditions, the reactor may be producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that may lead to imminent loss or potential loss of both fuel clad and RCS. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Barrier Degradation or ED Judgment ICs.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 50 Of 114 Initiating Condition SS3 Loss of All Vital DC Power.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value:

1. Loss of DC power to 125/250 VDC Bus 1/2R22-S016 and 1/2R22-S017 indicated by bus voltage indications less than 105/210 VDC for greater than 15 minutes.

Basis:

CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Loss of DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of CONTAINMENT INTEGRITY when there is significant decay heat and sensible heat in the reactor system.

105/210 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Initiating Conditions, Threshold Values and Basis 51 Of 114 Initiating Condition SS4 Complete Loss of Heat Removal Capability.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value:

1. Heat Capacity Temperature Limit (HCTL) Curve (EOP Graph 2) CANNOT be maintained in the "Safe" region.

Basis:

This Threshold Value addresses a severe challenge to Primary Containment at pressure and temperature.

Under these conditions, there is an actual major failure of a system intended for protection of the public.

Thus, declaration of a Site Area Emergency is warranted.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 52 Of 114 Initiating Condition SS6 Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: (1.a. AND 1.b. AND 1.c. AND 1.d.)

1. a. A SIGNIFICANT TRANSIENT in progress AND
b. Loss of most or all (approximately 75% of annunciators on panels 601,602, & 603) annunciators or indicators associated with safety systems AND
c. Compensatory non-alarming indications are NOT available AND
d. Indications needed to monitor Safety System Functions and critical Reactor parameters are NOT available.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 53 Of 114 Basis: SS6 SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) automatic runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Scram, (4) Safety System Injection Activation, or (5) thermal power oscillations >10%.

This IC and its associated Threshold Value are intended to recognize the inability of the control room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.

The annunciators for this Threshold Value are limited to include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values.

"Compensatory non-alarming indications" in this context includes computer based information such as SPDS.

The indications needed to monitor safety functions necessary for protection of the public must include control room indications, computer generated indications and dedicated annunciation capability. The specific indications are those used to determine such functions as the ability to shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, and to maintain containment intact.

"Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Supervisor be tasked with making a judgment decision as to whether additional personnel are required to provide augmented monitoring of system operation.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 54 Of 114 Initiating Condition Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded AND Manual Scram Was Successful.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Threshold Value:

1. Indication(s) exist that a reactor protection setpoint was exceeded AND An automatic scram did NOT occur, AND A manual scram resulted in the reactor being SUBCRITICAL (IRMs below Range 6 and Period is negative).

Basis:

SUBCRITICAL: IRMs below Range 6 and Period is negative This condition indicates failure of the automatic protection system to SCRAM the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the automatic protection system is the issue. A manual scram is any set of actions by the reactor operator(s) at the reactor control console (1/2H11-P603Panel) which causes control rods to be rapidly inserted into the core and brings the reactor SUBCRITICAL (e.g., reactor trip button, Alternate Rod Insertion). Failure of manual scram would escalate the event to a Site Area Emergency.

The Reactor should be considered SUBCRITICAL when IRMs below Range 6 and Period is negative.

A failure of the RPS to shut down the reactor (as indicated by reactor NOT being SUBCRITICAL) is a degraded plant condition that may require the injection of boron to shut down the reactor per EOP guidance. The definition of SUBCRITICAL (IRMs below Range 6 and Period is negative) is the point at which the EOP guidance directs the operators to exit the EOPs and continue shutdown activities with normal plant procedures.

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Emergency Classification and Initial Actions Emergency.Classification II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 55 Of 114 Initiating Condition UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With EITHER (1) a SIGNIFICANT TRANSIENT in Progress, OR (2) Compensatory Non-Alarming Indicators are Unavailable.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: [1 AND (a. OR b.)]

1. UNPLANNED loss of most or all (approximately 75% of annunciators on panels 601, 602, & 603)

MCR annunciators or indicators associated with safety systems for greater than 15 minutes AND EITHER

a. A SIGNIFICANT TRANSIENT is in progress OR
b. Compensatory non-alarming indications are NOT available MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 56 Of 114 Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

SIGNIFICANT TRANSIENT: is an UNPLANNED event involving one or more of the following: (1) automatic runback >25% thermal reactor power, (2) electrical load rejection >25% full electrical load, (3) Reactor Scram, (4) Safety System Injection Activation, or (5) thermal power oscillations >10%.

This IC and its associated Threshold Values are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered.

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

The concern in this Threshold Value is the difficulty associated with assessment of plant conditions.

The annunciators or indicators for this Threshold Value include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values.

"Compensatory non-alarming indications" in this context includes computer based information such as SPDS. If both a major portion of the annunciation system and all computer monitoring are unavailable, the Alert is required.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 57 Of 114 Initiating Condition AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in STATION BLACKOUT.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: (1.a AND 1.b)

1. a. AC power capability to 4160 VAC Emergency Buses 1/2E, 1/2F and 1/2G reduced to a single power source for greater than 15 minutes AND
b. ANY additional single failure will result in STATION BLACKOUT.

Basis:

STATION BLACKOUT: is a complete loss of offsite and onsite emergency AC power, as indicated by failure to energize any 4160VAC Emergency bus ..

This IC and the associated Threshold Values are intended to provide an escalation from IC SU1, "Loss of All Offsite Power To Essential Busses for Greater Than 15 Minutes." The condition indicated by this IC is the degradation of the offsite and onsite power systems such that any additional single failure would result in a STATION BLACKOUT. This condition could occur due to a loss of offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses on the Unit NOT being supplied power by the "Swing" diesel generator or two DG on the other unit.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 58 Of 114 Initiating Condition 1 Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value: (1.a AND 1.b)

1. a. Loss of power to or from Startup Auxiliary Transformers (SAT) 1/2C and 1/20 resulting in loss of all off-site electrical power to 4160 VAC Emergency Buses 1/2E, 1/2F, and 1/2G for greater than 15 minutes AND
b. Emergency diesel generators supplying power to 1/2E, 1/2F, and 1/2G.

Basis:

STATION BLACKOUT: is a complete loss of offsite and onsite emergency AC power, as indicated by failure to energize any 4160VAC Emergency bus.

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power (e.g., STATION BLACKOUT). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

This condition could occur due to a loss of offsite power with a the emergency diesel generators supplying power to two emergency buses on the Unit NOT being supplied power by the "Swing" diesel generator and three emergency buses on the unit being supplied power by the "Swing" diesel generator.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 59 Of 114 Initiating Condition Inability to Reach Required Shutdown Within Technical Specification Limits.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value:

1. Plant is NOT brought to required operating mode within Technical Specifications Limiting Condition for Operation (LCO) Required Action Statement (RAS) Time limit.

Basis:

LCOs require the plant to be brought to a designed shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this mayor may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the Technical Specifications requires a one hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An NUE is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of a NUE is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 60 Of 114 Initiating Condition UNPLANNED Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Value:

1. UNPLANNED loss of most or all (approximately 75% of annunciators on panels 601, 602, & 603)

MCR annunciators or indicators associated with safety systems for greater than 15 minutes.

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC and its associated Threshold Value are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment. Recognition of the availability of computer based indication equipment is considered.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a greater risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

The concern in this Threshold Value is the difficulty associated with assessment of plant conditions.

The annunciators or indicators for this Threshold Value include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other Threshold Values.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indicated during these modes of operation.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 61 Of 114 Initiating Condition Fuel Clad Degradation.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Values: (1 OR 2)

NOTE:

  • Use the Unit 1 or Unit 2 Pretreatment (Flow vs. mRlhr) Graphs in 73EP-EIP-001-0, Attachment 6 to determine if the Pretreatment Radiation Monitor exceeds the TV of 240,000 IJCi/sec.
1. Valid reading on Pretreatment Radiation Monitor greater than 240,000 IJCi/sec for greater than 60 minutes.

OR

2. RCS coolant sample activity (DE1 131 ) greater than:

39 IJCi/gm (total activity)

OR 27 IJCi/gm (non gaseous activity).

Basis:

VALID: an indication, report, or condition, is considered to be VALID when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed. Implicit in this definition is the need for timely assessment.

This IC is included as a NUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. Threshold Value #1 addresses the specific failed fuel monitor that provides indication of fuel clad integrity.

Threshold Value #2 addresses coolant samples exceeding coolant technical specifications for iodine spike.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 62 Of 114 Initiating Condition RCS Leakage.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Values: (1 OR 2)

1. UNIDENTIFIED OR pressure boundary leakage greater than 10 gpm.

OR

2. IDENTIFIED leakage greater than 25 gpm.

Basis:

IDENTIFIED leakage: Is defined as the rate of leakage into the Drywell equipment drain system.

UNIDENTIFIED leakage: Is defined as the rate of leakage into the Drywell floor drain system.

This IC is included as a NUE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the UNIDENTIFIED and pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance tests. The Threshold Value for identified leakage is set at a higher value due to the lesser significance of IDENTIFIED leakage in comparison to UNIDENTIFIED or pressure boundary leakage.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 63 Of 114 Initiating Condition UNPLANNED Loss of All Onsite OR Offsite Communications Capabilities.

Operating Mode Applicability: Power Operation (Mode 1)

Startup (Mode 2)

Hot Shutdown (Mode 3)

Threshold Values: (1 OR 2)

1. UNPLANNED loss of ALL of the following on-site communications capability affecting the ability to perform routine operations:

In plant telephones (includes hardwired and wireless)

Plant Page Plant radio systems OR

2. UNPLANNED loss of ALL of the following off-site communications capability:

ENN (Emergency Notification Network)

ENS (Emergency Notification System)

Commercial phones (Radio, PBX, Satellite, Wireless)

VOIP (Voice Over Internet Protocol)

OPX (Off Premise Extension)

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities.

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant conditions. This Threshold Value is intended to be used only when extraordinary means are being used to make communications possible.

The list for onsite communications loss encompasses the loss of all means of routine communications.

The list for offsite communications loss encompasses the loss of all means of communications with offsite authorities.

MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg. 74 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 64 Of 114 Initiating Condition SU8 Inadvertent Criticality.

Operating Mode Applicability: Hot Shutdown (Mode 3)

Threshold Value:

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC addresses inadvertent criticality events. While the primary concern of this IC is criticality events that occur in Cold Shutdown or Refueling modes, the IC is applicable in other modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of the plant, warranting a NUE classification.

This condition is identified using the period monitors. The term "extended" is used in order to allow exclusion of expected short term positive period from planned control rod movements. These short term positive periods are the result of the rise in neutron population due to SUBCRITICAL multiplication.

MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg. 75 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 65 Of 114 6.4 Category C - Cold Shutdown System Malfunctions MGR-0009 Rev. 5.0

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Emergency EmerQency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 66 Of 114 Initiating Condition CG1 Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Values: [1 AND (2.a. OR 2.b.) AND 3]

1. Loss of RPV inventory as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks AND

2. RPV Level:
a. Less than -155" (TAF) for greater than 30 minutes OR
b. RPV level CANNOT be monitored WITH indication of core uncovery for greater than 30 minutes as evidenced by EITHER of the following:

DWRRM 1/2D11-K621A(B): greater than 9.5 R/hr Erratic Source Range Monitor Indication AND

3. Primary Containment challenged as indicated by any of the following:

Explosive mixture inside Primary Containment: H2 greater than or equal to 6%

AND O2 greater than or equal to 5%

Primary Containment Pressure: Greater than or equal to 56 psig Secondary CONTAINMENT INTEGRITY NOT established Secondary Containment radiation monitors: Greater than Max Safe values (SC EOP - Table 6)

MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg. 77 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 67 Of 114 Basis: CG1 CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

For Threshold Value 1 in the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed to assure that the ability to monitor level will not be interrupted.

For both cold shutdown and refueling modes sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

Threshold Value 2 represents the inability to restore and maintain RPV level to above the top of active fuel. Fuel damage is probable if RPV level cannot be restored, as available decay heat will cause boiling, further reducing the RPV level.

NRC analysis indicates that core damage may occur within an hour following continued core uncovery therefore, conservatively, 30 minutes was chosen.

As water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in elevated DWRRM indication and possible alarm. Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers.

Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the primary containment breached or challenged then the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive mixture of dissolved gasses in CONTAINMENT. However, CONTAINMENT monitoring and/or sampling should be performed to verify this assumption and a General Emergency declared if it is determined that an explosive mixture exists.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating

-lnitiatinQ Conditions, Threshold Values and Basis 68 Of 114 Initiating Condition CS1 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Threshold Values: [(1.a. OR 1.b.) OR (2.a. OR 2.b.)]

1. Loss of RPV inventory affecting core decay heat removal capability with Secondary CONTAINMENT INTEGRITY NOT established as indicated by:
a. RPV level less than - 41" (6" below the Level 2 actuation setpoint)

OR

b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks OR

2. Loss of RPV inventory affecting core decay heat removal capability with Secondary CONTAINMENT INTEGRITY established as indicated by:
a. RPV level less than -155" (TAF)

OR

b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by EITHER:

Unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks OR Erratic Source Range Monitor Indication MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg. 79 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 69 Of 114 Basis: CS1 CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Under the conditions specified by this IC, continued lowering in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV.

In cold shutdown the decay heat available to raise ReS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

The 30 minute duration allows sufficient time for actions to be performed to recover needed cooling equipment.

Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

MGR-0009 Rev. 5.0

~1-~~l(i!!;:!.\!.Lbe monitored WITH Indication. of SNC PLANT E.!. HATCH I Pg. 78 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emerqency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT 4*" Attachment Page TITLE: HNP EAL - Initiatinq Conditions, Threshold'Y~lues andBasis 68 Of 114 Initiating Condition CS1 Loss of RPV Inventory Affecting Core Decay tfeat Removal Capability.

, .'."::;::t'"

Operating Mode Applicability: Cold Shutdown (Mode 4)

Threshold Values: [(1.a. OR 1.b.) OR (2.a. OR 2.b.)]

1. Loss of RPV inventory affecting core decay heat removal capability with Secondary CONTAINMENT INTEGRITY NOT established as indicated by:
a. RPV level less than - 41" (6" below the Level 2 actuation setpoint)

OR

b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks OR

2. Loss of RPV inventory affecting core decay heat removal capability with Secondary CONTAINMENT INTEGRITY established as indicated by:
a. RPV level less than -155" (TAF)

OR

b. RPV level CANNOT be monitored for greater than 30 minutes with a possible loss of RPV inventory as indicated by EITHER:

Unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks OR Erratic Source Range Monitor Indication MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I Pg. 79 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 69 Of 114 Basis: CS1 CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Under the conditions specified by this IC, continued lowering in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach, pressure boundary leakage, or continued boiling in the RPV.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

The 30 minute duration allows sufficient time for actions to be performed to recover needed cooling equipment.

Post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

MGR-0009 Rev. 5.0

SNC PLANT E.!. HATCH I Pg. 80 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 70 Of 114 Initiating Condition CS2 Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling (Mode 5)

Threshold Values: [(1.a. OR 1.b.) OR (2.a. OR 2.b.)]

1. WITH Secondary CONTAINMENT INTEGRITY NOT established:
a. RPV level less than - 41" (6" below the Level 2 actuation setpoint)

OR

b. RPV level CANNOT be monitored WITH indication of core uncovery as evidenced by EITHER of the following:

Drywell Wide Range Radiation Monitor (DWRRM) 1/2D11-K621A(B) greater than 9.5 R/hr.

Erratic Source Range Monitor Indication OR

2. WITH Secondary CONTAINMENT INTEGRITY established
a. RPV level less than -155" (TAF)

OR

b. RPV level CANNOT be monitored WITH Indication of core uncovery as evidenced by EITHER of the following:

DWRRM 1/2D11-K621A(B) greater than 9.5 R/hr Erratic Source Range Monitor Indication MGR-0009 Rev. 5.0

SNC PLANT E.1. HATCH I Pg. 81 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions EmergencyClassification I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 71 Of 114 Basis: CS2 CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Under the conditions specified by this IC, continued lowering in RPV level is indicative of a loss of inventory control. Inventory loss may be due to an RPV breach or continued boiling in the RPV.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode.

As water level in the RPV lowers, the dose rate above the core will rise. The dose rate due to this core shine should result in elevated DWRRM indication and possible alarm. For Threshold Value 1.b and Threshold Value 2.b calculations were performed to conservatively estimate a site-specific dose rate setpoint indicative of core uncovery. Additionally, post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.

For Threshold Value 2 in the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed to assure that the ability to monitor level will not be interrupted.

This effluent release is not expected with Secondary CONTAINMENT INTEGRITY established.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating InitiatinQ Conditions, Threshold Values and Basis 72 Of 114 Initiating Condition Loss of RCS Inventory.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Threshold Values: [1 OR (2.a. AND 2.b.)]

1. Loss of RCS inventory as indicated by RPV level less than -35" (Level 2 actuation setpoint)

OR

2. a. RPV level CANNOT be monitored for greater than 15 minutes AND
b. A loss of RCS inventory may be occurring as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks Basis:

These Threshold Values serve as precursors to a loss of ability to adequately cool the fuel. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level lowering and potential core uncovery.

The Level 2 Actuation Setpoint was chosen because it is a standard setpoint at which High Pressure injection systems automatically start during normal operations (Modes 1-3). The inability to restore and maintain level after reaching this setpoint would therefore be indicative of a failure of the RCS barrier.

In the cold shutdown mode, normal RCS level and RPV level instrumentation systems will normally be available. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

MGR-0009 Rev. 5.0

SNC PLANT E.1. HATCH II Pg. 83 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 73 Of 114 Initiating Condition Loss of RPV Inventory with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling (Mode 5)

Threshold Values: [1 OR (2.a. AND 2.b.)]

1. Loss of RCS inventory as indicated by RPV level less than -35" (Level 2 actuation setpoint)

OR

2. a. RPV level CANNOT be monitored for greater than 15 minutes AND
b. A loss of RCS inventory may be occurring as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks Basis:

These Threshold Values serve as precursors to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV level lowering and potential core uncovery.

The Level 2 Actuation Setpoint was chosen because it is a standard setpoint at which High Pressure injection systems automatically start during normal operations (Modes 1-3). The inability to restore and maintain level after reaching this setpoint would therefore be indicative of a failure of the RCS barrier.

In the refueling mode, normal means of RPV level indication may not be available. Redundant means of RPV level indication will be normally installed to assure that the ability to monitor level will not be interrupted. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

MGR-0009 Rev. 5.0

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Emergency EmerQency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 74 Of 114 Initiating Condition Loss of All Offsite Power AND Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Defueled Threshold Value: (1.a AND 1.b AND 1.c)

1. a. Loss of power to or from Startup Auxiliary Transformers (SAT) 1/2C and 1/2D resulting in loss of all off-site electrical power to 4160 VAC Emergency Buses 1/2E, 1/2F, and 1/2G.

AND

b. Failure of emergency diesel generators to supply power to emergency busses.

AND

c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including RHR, ECCS, Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown, refueling, or defueled mode the event can be classified as an Alert, because of the significantly reduced decay heat, lower temperature and pressure, increasing the time available to restore one of the emergency busses, relative to that specified for the Site Area Emergency Threshold Value.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to emergency busses. Even though an emergency bus may be energized, if necessary loads are not operable on the energized bus then the bus should not be considered operable.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 75 Of 114 Initiating Condition Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Values: [1.a. AND 1.b.] OR [2.a. AND (2.b. OR 2.c.)] OR [3.a. OR 3.b.]

1. An UNPLANNED event results in RCS temperature exceeding 212°F with:
a. Secondary CONTAINMENT INTEGRITY NOT established AND
b. RCS integrity NOT established NOTES:
  • The Emergency Director should not wait until the indicated time of Threshold Values 2 or 3 has elapsed, but should declare the event as soon as it is determined that the duration that RCS temperature exceeds 212° F has or will likely be exceeded.
  • If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then Threshold Values 2 and 3 are not applicable.

OR

2. An UNPLANNED event results in RCS temperature exceeding 212°F for greater than 20 minutes with:
a. Secondary CONTAINMENT INTEGRITY established AND
b. RCS integrity NOT established OR
c. RCS inventory reduced.

OR

3. An UNPLANNED event results in:
a. RCS temperature exceeding 212°F for greater than 60 minutes OR
b. RPV pressure increasing greater than 10 psig MGR-0009 Rev. 5.0

SNC PLANT E.!. E.I. HATCH II Pg. 86 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 76 Of 114 Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

Threshold Value 1 addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither Secondary CONTAINMENT INTEGRITY nor RCS integrity are established. No delay time is allowed for Threshold Value1 because the evaporated reactor coolant that may be released into the Containment during this heatup condition could also be directly released to the environment. RCS integrity is in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

Threshold Value 2 addresses the complete loss of functions required for core cooling for GREATER THAN 20 minutes during refueling and cold shutdown modes when Secondary CONTAINMENT INTEGRITY is established but RCS integrity is not established or RCS inventory is reduced. The allowed 20 minute time frame was included to allow operator action to restore the heat removal function, if possible.

Threshold Value 3 addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in Threshold Value 1 and 2, RCS integrity should be considered to be in place when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation. The status of Secondary CONTAINMENT INTEGRITY in this Threshold Value is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60 minute time frame should allow sufficient time to restore cooling without there being a substantial degradation in plant safety.

The 10 psig pressure rise covers situations where, due to high decay heat loads, the time provided to restore temperature control, should be less than 60 minutes. The RCS pressure setpoint chosen is 10 psig and can be read on Control Board instrumentation. The Note indicates that Threshold Value 3 is not applicable if actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the 60 minute time frame assuming that the RCS pressure rise has remained less than 10 psig.

NRC analyses show that sequences can cause core uncovery in 15 to 20 minutes and severe core damage within an hour after decay heat removal is lost.

The Emergency Director must remain alert to events or conditions that lead to the conclusion that exceeding the Threshold Value is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 77 Of 114 Initiating Condition 1 RCS Leakage.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Threshold Values:

1. Unable to establish or maintain RPV level greater than +3".

Basis:

This IC is included as a NUE because it is considered to be a potential degradation of the level of safety of the plant. Prolonged loss of RCS Inventory may result in escalation to the Alert level via either IC CA 1 (Loss of RCS Inventory) or CA4 (Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV).

The difference between CU1 and CU2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and RCS inventory and level monitoring means are normally available. In the refueling mode the RCS is not intact and RPV level and inventory are monitored by different means.

Expanded basis for these assumptions is provided NEI 99-01, Rev. 4 (NUMARC/NESP-007),

"Methodology for Development of Emergency Action Levels", January 2003.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 78 Of 114 Initiating Condition UNPLANNED Loss of RCS Inventory with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling (Mode 5)

Threshold Values: [1 OR (2.a. AND 2.b.)]

1. UNPLANNED RWL lowering below the RPV flange for greater than or equal to 15 minutes OR
2. a. RPV level CANNOT be monitored AND
b. A loss of RPV inventory may be occurring as indicated by unexplained level rise in any of the following:

Drywell Floor Drain Sumps Drywell Equipment Drain Sumps Torus Torus Room Sumps Reactor Building Floor Drain Sumps Turbine Building Floor Drain Sumps Rad Waste Tanks MGR-0009 Rev. 5.0

SNC PLANT E.1. HATCH I Pg.89 of 127 DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 79 Of 114 Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC is included as a NUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RWL below the RPV flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level decreasing below the RPV flange warrants declaration of a NUE due to the reduced RCS inventory that is available to keep the core covered.

Threshold Value 1 involves a lowering in RWL below the top of the RPV flange that continues for 15 minutes due to an UNPLANNED event. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using any of the redundant means of refill that should be available.

In the refueling mode, normal means of core temperature indication and RWL indication may not be available. Redundant means of RPV level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RPV inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions / 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 80 Of 114 Initiating Condition Loss of All Offsite Power to Essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Value: (1.a. AND 1.b.)

1. a. Loss of power to or from Startup Auxiliary Transformers (SAT) 1/2C and 1/20 resulting in loss of all off-site electrical power to 4160 VAC Emergency Buses 1/2E, 112F, and 1/2G for greater than 15 minutes AND
b. One emergency diesel generator supplying power to 1/2E, 1/2F, or 1/2G.

Basis:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete Loss of AC Power. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 81 Of 114 Initiating Condition UNPLANNED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Values: (1 OR 2)

NOTE: The Emergency Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the duration of the loss of RCS temperature AND RPV level indication has or will likely exceed 15 minutes.

1. An UNPLANNED event results in RCS temperature exceeding 212°F.

OR

2. Loss of all RCS temperature AND RPV level indication for greater than 15 minutes.

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC is included as a NUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant.

During refueling the level in the RPV will normally be maintained above the RPV flange. Refueling evolutions that lower water level below the RPV flange are carefully planned and procedurally controlled. Loss of forced decay heat removal at reduced inventory may result in more rapid rises in RCS/RPV temperatures depending on the time since shutdown.

The Emergency Director must remain attentive to events or conditions that lead to the conclusion that exceeding the Threshold Value is imminent. If, in the judgment of the Emergency Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 82 Of 114 Initiating Condition UNPLANNED Loss of All Onsite OR Offsite Communications Capabilities.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Values: (1 OR 2)

1. UNPLANNED loss of ALL of the following on-site communications capability affecting the ability to perform routine operations:

Plant telephones (Includes hardwired and wireless)

Plant page Plant radio systems OR

2. UNPLANNED loss of ALL of the following off-site communications capability:

ENN (Emergency Notification Network)

ENS (Emergency Notification System)

Commercial phones (Radio, PBX, Satellite, Wireless)

VOIP (Voice Over Internet Protocol)

OPX (Off Premise Extension)

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values is to recognize a loss of communications capability that either defeats the plant operations staff ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities.

The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This Threshold Value is intended to be used only when extraordinary means are being utilized to make communications possible.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 83 Of 114 Initiating Condition UNPLANNED Loss of Required DC Power for Greater than 15 Minutes.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Value: (1.a. AND 1.b.)

1. a. UNPLANNED loss of DC power to 125/250 VDC Bus 1/2R22-S016 & 1/2R22-S017 indicated by bus voltage indications less than 105/210 VDC AND
b. Failure to restore power to at least one DC bus within 15 minutes from the time of loss.

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

The purpose of this IC and its associated Threshold Values is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during Cold Shutdown or Refueling operations. This Threshold Value is intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

UNPLANNED is included to preclude the declaration of an emergency as a result of planned maintenance activities.

105/210 VDC bus voltage is based on the minimum bus voltage necessary for the operation of safety related equipment. This voltage value incorporates a margin of at least 15 minutes of operation before the onset of inability to operate those loads.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -lnitiatinQ

-Initiating Conditions, Threshold Values and Basis 84 Of 114 Initiating Condition Inadvertent Criticality.

Operating Mode Applicability: Cold Shutdown (Mode 4)

Refueling (Mode 5)

Threshold Values:

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

Basis:

UNPLANNED: a parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

This IC indicates a potential degradation of the level of safety of the plant, warranting a NUE classification.

The term "extended" is used in order to allow exclusion of expected short term positive periods from planned fuel bundle or control rod movements during core alterations. These short term positive periods are the result of the rise in neutron population due to SUBCRITICAL multiplication.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 85 Of 114 6.5 Category E - ISFSI Events MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 86 Of 114 Initiating Condition 1 Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: ALL Threshold Value:

1. Damage to a loaded dry fuel storage cask CONFINEMENT BOUNDARY due to natural phenomena events, accident conditions or any condition in the opinion of the Emergency Director that affects or causes a loss of loaded dry fuel storage cask CONFINEMENT BOUNDARY.

Basis:

CONFINEMENT BOUNDARY: is the barrier(s) between areas containing radioactive substances and the environment.

A NUE in this IC is categorized on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

Any condition not explicitly detailed as a Threshold Value, which, in the judgment of the Emergency Director (ED), is a potential degradation in the level of safety of the ISFSI. ED judgment is to be based on known conditions and the expected response to mitigating activities within a short time period.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 87 Of 114 6.6 Category H .. Hazards and Others MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 88 Of 114 Initiating Condition HG1 Security Event Resulting in Loss Of Physical Control of the Facility.

Operating Mode Applicability: All Threshold Value:

1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions or spent fuel pool cooling.

Basis:

HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of VITAL AREAs (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. These safety functions are reactivity control, reactor water level, and decay heat removal. If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

This Threshold Value should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely.

Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se. Design of the remote shutdown capability and the location of the transfer switches should be taken into account.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 89 Of 114 Initiating Condition HG2 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency.

Operating Mode Applicability: All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of CONTAINMENT INTEGRITY or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Basis:

CONTAINMENT INTEGRITY: Primary Containment OPERABLE per Technical Specification 3.6.1.1.

Secondary Containment OPERABLE per Technical Specification 3.6.4.1.

HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, takes hostages, and lor intimidates the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non- terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the General Emergency class.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 90 Of 114 Initiating Condition HS1 Confirmed Security Event in a Plant VITAL AREA.

Operating Mode Applicability: All Threshold Values:

1. Confirmed security event in the VITAL AREA as determined from the security plan reported by the security Shift Captain or designee Basis:

VITAL AREA: any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. This includes the Control Building, Reactor Building, Diesel Generator Building, Intake Structure and Primary containment.

The security plan identifies numerous events/conditions that constitute a threat/compromise to a Station's security. Only those events that involve Actual or Likely Major failures of plant functions needed for protection of the public need to be considered. The following events would not normally meet this requirement: Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions.

Reference is made to security shift supervision because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant security plan.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 91 Of 114 Initiating Condition HS2 Control Room Evacuation Has Been Initiated AND Plant Control Cannot Be Established.

Operating Mode Applicability: All Threshold Value: (1.a. AND 1.b.)

1. a. Control Room evacuation has been initiated AND
b. Control of the plant can NOT be established per 31 RS-OPS-001-1/2 "Shutdown From Outside The Control Room", within 15 minutes.

Basis:

Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. The intent of this IC is to capture those events where control of the plant cannot be reestablished in a timely manner. The time for transfer is based on analysis or assessments as to how quickly control must be reestablished without core uncovering and/or core damage. The determination of whether or not control is established at the remote shutdown panel is based on Emergency Director (ED) judgment. The ED is expected to make a reasonable, informed judgment within the time for transfer that the operators have control of the plant.

The intent of the Threshold Value is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis should be placed on those components and instruments that supply protection for and information about safety functions. These safety functions are reactivity control (ability to shutdown the reactor and maintain it shutdown), reactor water level (ability to cool the core), and decay heat removal (ability to maintain a heat sink).

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 92 Of 114 Initiating Condition HS3 Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency.

Operating Mode Applicability: All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts (1) toward site personnel or equipment that could lead to the likely failure of, or (2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels that exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Basis:

HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and lor intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency class description for Site Area Emergency.

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EmerQency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 93 Of 114 Initiating Condition HS4 Site Attack Operating Mode Applicability: All Threshold Value:

1. A notification from the site security force that an armed attack, explosive attack, airliner impact, or other HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 94 Of 114 Basis: HS4 HOSTILE ACTION: An act toward a Nuclear Power Plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take hostages, and lor intimidate the licensee to achieve an end.

This includes attack by air, land, or water using weapons, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

PROTECTED AREA: the area which normally encompasses a" controlled areas within the security protected area fence.

Although security officers are we" trained and prepared to protect against HOSTILE ACTION, it is appropriate for Offsite Response Organizations (ORO) to be notified and encouraged to begin preparations for public protective actions to be better prepared should it be necessary to consider further actions.

This Threshold Value is intended to address the potential for a very rapid progression of events due to a dedicated attack. It is not intended to address incidents that are accidental or acts of civil disobedience, such as hunters or physical disputes between employees within the OCA or PA. That initiating condition is adequately addressed by other Threshold Values. HOSTILE ACTION identified above encompasses various acts including:

  • air attack (airliner impacting the protected area)
  • land-based attack (HOSTILE FORCE penetrating protected area)
  • waterborne attack (HOSTILE FORCE on water penetrating protected area)
  • BOMBs breeching the protected area This Threshold Value is intended to address the contingency for a very rapid progression of events due to an airborne hostile attack such as that experienced on September 11, 2001 and the possibility for additional attacking aircraft. This Threshold Value is not premised solely on the potential for a radiological release. Rather the issue includes the need for assistance due to the possibility for significant and indeterminate damage from additional attack elements.

Although vulnerability analyses show NPPs to be robust, it is appropriate for OROs to be notified and to activate in order to be better prepared to respond should protective actions become necessary. If not previously notified by NRC that the aircraft impact was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant. The status and size of the plane is provided by NORAD through the NRC.

This Threshold Value addresses the immediacy of a threat to impact site vital areas within a relatively short time. The fact that the site is under serious attack with minimal time available for additional assistance to arrive requires ORO readiness and preparation for the implementation of protective measures.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 95 Of 114 Initiating Condition Natural and Destructive Phenomena Affecting the Plant VITAL AREA.

Operating Mode Applicability: All Threshold Values: [(1.a. AND 1.b.) OR 2 OR 3 OR 4 OR 5 OR 6]

1. a. "Seismic Instrumentation Triggered" (Unit 2) alarm indicating horizontal acceleration greater than or equal to 0.08g Operating Basis Earthquake (OBE Level)

OR Any horizontal (N-S, E-W) peak shock annunciator 12.7 Hz amber light illuminated indicates 100% OBE actuated on Pane11H11-P701 AND

b. Seismic Instrumentation Triggered" (Unit 1) alarm indicating horizontal acceleration greater than 0.005g OR Unit 1 OR Unit 2 Seismic Peak Shock Recorder High "G" Alarm OR Unit 1 AND Unit 2 Time-History Recorders start OR
2. Tornado or high winds greater than 100 mph within the PROTECTED AREA boundary resulting in VISIBLE DAMAGE to any of the following plant structures/equipment OR the Control Room has indication of degraded performance of those systems:

Primary Containment Reactor Building Diesel Generator Building Intake Structure Control Building OR

3. Vehicle crash within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to any of the following plant structures or equipment therein OR Control Room indication of degraded performance of systems required for safe shutdown of the plant Primary Containment Reactor Building Diesel Generator Building Intake Structure Control Building Continued on the Next Page MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 96 Of 114 Initiating Condition C{]&~ (cont.)

Operating Mode Applicability: All Threshold Values: [(1.a. AND 1.b.) OR 2 OR 3 OR 4 OR 5 OR 6]

4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any area containing safety-related equipment, their controls or their power supplies.

Primary Containment Reactor Building Control Building Diesel Generator Building NOTE: Applicable Areas are:

Unit 1/2 Southeast Diagonal Rooms (RHR Diagonals)

Unit 1/2 Northeast Diagonal Rooms (RHR Diagonals)

Unit 1/2 HPCI Rooms Unit 1 Southwest Diagonal Room (RCIC Diagonal)

Unit 2 Northwest Diagonal Room (RCIC Diagonal)

The plant's IPE Internal Floods Analysis determined that the RHR systems in the Unit 1/2 Southeast and Northeast Diagonal Rooms are required for all modes of operation. The high pressure ECCS systems located in the HPCI Rooms adjacent to U1 Northeast and U2 Southeast diagonals and Unit 1 Southwest Diagonal and Unit 2 Northwest Diagonal Room (RCIC Diagonals) are only included in the list because they are required by Technical Specifications for modes 1 and 2 (Le., these rooms are only applicable for this threshold in modes 1 and 2).

5. Exceeding Max Safe Operating Values specified in EOP 31 EO-EOP-014-1 (2) SC - Secondary Containment Control/RR-Radioactivity Release Control Table 5 Secondary Containment Operating Water Levels OR Flooding that creates industrial safety hazards (e.g., electric shock) that preclude access necessary to operate or monitor safety equipment in the Reactor Buildings (see note), Control Building, Diesel Generator Building, or Intake Structure.

OR

6. Sustained hurricane winds greater than 74 mph onsite resulting in VISIBLE DAMAGE to plant structures within the PROTECTED AREA boundary VITAL AREAS containing equipment necessary for safe shutdown OR has caused damage as evidenced by control room indication of degraded performance of those systems.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 97 Of 114 Basis:

PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

VITAL AREA: Any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. This includes the Control Building, Reactor Building, Diesel Generator Building, Intake Structure and Primary containment.

The Threshold Values in this IC escalate from the NUE Threshold Values in HU1 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events.

The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this Threshold Value to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occur on the basis of other ICs (e.g., System Malfunction).

Threshold Value #1 is based on the OBE earthquake FSAR design basis. Seismic events of this magnitude can result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety systems.

Threshold Value #2 is based on the FSAR design basis. The high wind 100 mph value is based on FSAR design basis (105 mph design) and the highest meter reading available (100 mph). Wind loads greater than 115 mph can cause damage to safety functions.

Threshold Value #3 addresses crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant.

Threshold Value #4 addresses the threat to safety related equipment imposed by missiles generated by main turbine rotating component failures. This list of areas include safety-related equipment, their controls, and their power supplies.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 98 Of 114 Basis (cont.):

EAL #5 addresses the effect of internal flooding that has resulted in degraded performance of systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety of the plant. This flooding may have been caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant that are not designed to be wetted or submerged. The plant's IPE Flooding Analysis (Hatch U1 Internal Floods Analysis H96-0 and Hatch U2 Internal Floods Analysis H97-0) were used to determine the areas for the first condition when developing this EAL.

The second condition addresses the high pressure ECCS systems (HPCI & RCIC) required by Technical Specifications for modes 1 and 2 (Le., these rooms are only applicable for this threshold in modes 1 and 2). HPCI Room, Unit 1 Southwest Diagonal Room, and Unit 2 Northwest Diagonal Room (RCIC Diagonals) were included for completeness, even though not a part of the IPE Flooding Analysis.

Threshold Value #6 covers site-specific phenomena of a hurricane. The Threshold Value is based damage attributable to the wind.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 99 Of 114 Initiating Condition FIRE OR EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.

Operating Mode Applicability: All Threshold Value:

1. FIRE OR EXPLOSION AND Affected system parameter indications show degraded performance OR Plant personnel report VISIBLE DAMAGE to permanent structures or safety related equipment in any of the following Vital Areas:

Primary Containment Reactor Building Diesel Generator Building Intake Structure Control Building MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 100 Of 114 Basis:

FIRE: is combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIREs. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

VITAL AREA: any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. This includes the Control Building, Reactor Building, Diesel Generator Building, Intake Structure and Primary Containment.

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

Areas containing functions and systems required for the safe shutdown of the plant are specified to determine if the FIRE or EXPLOSION is potentially affecting any redundant trains of safety systems.

Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels / Radiological Effluent, or Emergency Director Judgment ICs.

This Threshold Value addresses a FIRE / EXPLOSION and not the degradation in performance of affected systems. System degradation is addressed in the System Malfunction Threshold Values. The reference to damage of systems is used to identify the magnitude of the FIRE / EXPLOSION and to discriminate against minor FIREs / EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs /

EXPLOSIONs in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact that the FIRE / EXPLOSION was large enough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE /

EXPLOSION is large enough to affect more than one component.

This situation is not the same as removing equipment for maintenance that is covered by a plant's Technical Specifications. Removal of equipment for maintenance is a planned activity controlled in accordance with procedures and, as such, does not constitute a substantial degradation in the level of safety of the plant. A FIRE / EXPLOSION is an UNPLANNED activity and, as such, does constitute a SUbstantial degradation in the level of safety of the plant. In this situation, an Alert classification is warranted.

The inclusion of a "report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this Threshold Value to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The declaration of an Alert and the activation of the Technical Support Center will provide the Emergency Director with the resources needed to perform these damage assessments. The Emergency Director also needs to consider any security aspects of the EXPLOSIONs, if applicable.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 101 Of 114 Initiating Condition Release of Toxic, Asphyxiant or Flammable Gases Within or Adjacent to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Operating Mode Applicability: All Threshold Values: (1 OR 2)

1. Report or detection of toxic or asphyxiant gas within or adjacent to a VITAL AREA in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).

OR

2. Report or detection of flammable gases in concentration greater than the LOWER FLAMMABILITY LIMIT within or adjacent to a VITAL AREA.

Basis:

VITAL AREA: any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation. This includes the Control Building, Reactor Building, Diesel Generator Building, Intake Structure and Primary containment.

IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (lDLH): A condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

LOWER FLAMMABILITY LIMIT (LFL): The minimum concentration of a combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and gaseous oxidizer.

This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas contiguous to plant VITAL AREAs or other significant buildings or areas.

Threshold Value #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH within a VITAL AREA or any area or building contiguous to VITAL AREA. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.

Threshold Value #2 is met when the flammable gas concentration in a VITAL AREA or any building or area contiguous to a VITAL AREA exceed the LOWER FLAMMABILITY LIMIT. This Threshold Value addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 102 Of 114 Initiating Condition Confirmed Security Event in a Plant PROTECTED AREA.

Operating Mode Applicability: All Threshold Values:

1. Confirmed security event in the PROTECTED AREA as determined from the security plan reported by Security Shift Captain or designee.

Basis:

PROTECTED AREA: the area which encompasses all controlled areas within the security protected area fence.

The security plan identifies numerous events/conditions that constitute a threat/compromise to a Station's security. Only those events that involve Actual or Potential Substantial degradation to the level of safety of the plant need to be considered. The following events would not normally meet this requirement: Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, or loss/compromise of safeguards materials.

Reference is made to the security shift supervision because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant security plan.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 103 Of 114 Initiating Condition Control Room Evacuation Has Been Initiated.

Operating Mode Applicability: All Threshold Value:

1. Entry into 31 RS-OPS-001-1/2 "Shutdown From Outside The Control Room" for Control Room evacuation.

Basis:

With the control room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facility is necessary. Inability to establish plant control from outside the control room will escalate this event to a Site Area Emergency.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 104 Of 114 Initiating Condition Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert.

Operating Mode Applicability: All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of intentional malicious dedicated efforts of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis:

HOSTILE ACTION: An act toward a Nuclear Power Plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take hostages, and lor intimidate the licensee to achieve an end.

This includes attack by air, land, or water using weapons, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

This Threshold Value is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the Alert emergency class.

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EmerQency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 105 Of 114 Initiating Condition Notification of an Airborne Attack Threat Operating Mode Applicability: All Threshold Value:

1. A validated notification from NRC of an airliner attack threat less than 30 minutes away.

Basis:

The intent of this Threshold Value is to ensure that notifications for the security threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. Only the plant to which the specific threat is made need declare the Alert. This Threshold Value is met when a plant receives information regarding an airliner attack threat from NRC and the airliner is less than 30 minutes away from the plant.

This Threshold Value is intended to address the contingency of a very rapid progression of events due to an airborne hostile attack such as that experienced on September 11, 2001. This Threshold Value is not premised solely on the potential for a radiological release. Rather the issue includes the need for assistance due to the possibility for significant and indeterminate damage from such an attack.

Although vulnerability analyses show NPPs to be robust, it is appropriate for Offsite Response Organizations to be notified and encouraged to activate to be better prepared should it be necessary to consider further actions. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant. The status and size of the plane is provided by NORAD through the NRC.

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Emergency Classification and Initial Actions L I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 106 Of 114 Initiating Condition INILMl Notification of HOSTILE ACTION within the OCA Operating Mode Applicability: All Threshold Value:

1. A notification from the site security force that an armed attack, explosive attack, airliner impact or other HOSTILE ACTION is occurring or has occurred within the OCA.

Basis:

HOSTILE ACTION: An act toward an NPP or its personnel that includes the use of violent force to destroy equipment, take hostages, and lor intimidate the licensee to achieve an end. This includes attack by air, land, or water using weapons, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities, (e.g., violent acts between individuals in the owner controlled area.)

HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

This Threshold Value is intended to address the potential for a very rapid progression of events due to a hostile attack including:

  • air attack (airliner impacting the OCA)
  • land-based attack (HOSTILE FORCE progressing across licensee property or directing projectiles at the site)
  • waterborne attack (HOSTILE FORCE on water attempting forced entry, or directing projectiles at the site)
  • BOMBs This Threshold Value is not intended to address incidents that are accidental or acts of civil disobedience, such as hunters or physical disputes between employees within the OCA or PA. That initiating condition is adequately addressed by other Threshold Values.

This Threshold Value is not premised solely on adverse health effects caused by a radiological release.

Rather the issue is the immediate need for assistance due to the nature of the event and the potential for significant and indeterminate damage. Although NPP security officers are well trained and prepared to protect against HOSTILE ACTION, it is appropriate for Offsite Response Organizations to be notified and encouraged to begin activation (if they do not normally) to be better prepared should it be necessary to consider further actions.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 1070f 114 Basis: (cont.) IMILMl This Threshold Value is intended to address the contingency for a very rapid progression of events due to an airborne hostile attack such as that experienced on September 11, 2001 and the possibility for additional attacking aircraft. It is not intended to address accidental aircraft impact as that initiating condition is adequately addressed by other Threshold Values. This Threshold Value is not premised solely on the potential for a radiological release. Rather the issue includes the need for assistance due to the possibility for significant and indeterminate damage from additional attack elements. Although vulnerability analyses show NPPs to be robust, it is appropriate for Offsite Response Organizations to be notified and to activate in order to be better prepared to respond should protective actions become necessary. If not previously notified by NRC that the aircraft impact was intentional, then it would be expected, although not certain, that notification by an appropriate Federal agency would follow. In this case, appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. However, the declaration should not be unduly delayed awaiting Federal notification. Airliner is meant to be a large aircraft with the potential for causing significant damage to the plant. The status and size of the plane is provided by NORAD through the NRC.

This IClThreshold Value addresses the immediacy of an expected threat arrival or impact on the site within a relatively short time. The fact that the site is an identified attack candidate with minimal time available for further preparation requires a heightened state of readiness and implementation of protective measures that can be effective (onsite evacuation, dispersal or sheltering) before arrival or impact.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Conditions, Threshold Values and Basis 108 Of 114 Initiating Condition 1 Natural and Destructive Phenomena Affecting the PROTECTED AREA.

Operating Mode Applicability: All Threshold Value:

1. Confirmed "Seismic Instrumentation Triggered" (Unit 1) alarm indicating horizontal acceleration greater than 0.005 g or report by plant personnel that an earthquake was "felt".

OR

2. Report by plant personnel of a tornado or high winds greater than 100 mph striking within the PROTECTED AREA.

OR

3. Crash of vehicle, large enough to cause significant damage, into plant structures containing functions or systems required for safe shutdown within the PROTECTED AREA boundary.

OR

4. Report by plant personnel of an unanticipated EXPLOSION within the PROTECTED AREA boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.

OR

5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

OR Applicable Areas are:

NOTE:

Unit 1/2 Southeast Diagonal Rooms (RHR Diagonals)

Unit 1/2 Northeast Diagonal Rooms (RHR Diagonals)

Unit 1/2 HPCI Rooms Unit 1 Southwest Diagonal Room (RCIC Diagonal)

Unit 2 Northwest Diagonal Room (RCIC Diagonal)

The plant's IPE Internal Floods Analysis determined that the RHR systems in the Unit 1/2 Southeast and Northeast Diagonal Rooms are required for all modes of operation. The high pressure ECCS systems located in the HPCI Rooms adjacent to U1 Northeast and U2 Southeast diagonals and Unit 1 Southwest Diagonal and Unit 2 Northwest Diagonal Room (RCIC Diagonals) are only included in the list because they are required by Technical Specifications for modes 1 and 2 (Le., these rooms are only applicable for this threshold in modes 1 and 2).

6. Exceeding Max Normal Operating Values specified in EOP 31 EO-EOP-014-1 (2) SC - Secondary Containment Control/RR - Radioactivity Release Control Table 5 Secondary Containment Operating Water Levels.

OR

7. Sustained hurricane force winds greater than 74 mph forecast to be at the plant site in the next four hours in accordance with 34AB-Y22-002-0.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - InitiatinQ Conditions, Threshold Values and Basis 109 Of 114 Basis: 1 PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

EXPLOSION: is a rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

VISIBLE DAMAGE: is damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes:

deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

These ICs are categorized on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Areas identified in the Threshold Values define the location of the event based on the potential for damage to equipment contained therein.

Threshold Value #1 - As defined in the EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:

An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. For most plants with seismic instrumentation, the seismic switches are set at an acceleration of about 0.01 g.

Threshold Value #2 is based on the assumption that a tornado striking or high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. The high wind 100 mph value is based on FSAR design basis (105 mph design) and the highest meter reading available (100 mph). If such damage is confirmed visually or by other in-plant indications, the event may be escalated to Alert.

Threshold Value #3 addresses crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant.

For Threshold Value #4 only those EXPLOSIONs of sufficient force to damage permanent structures or equipment within the PROTECTED AREA should be considered. No attempt is made in this Threshold Value to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The Emergency Director also needs to consider any security aspects of the EXPLOSION, if applicable.

Threshold Value #5 addresses main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids and gases to the plant environs.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions II 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating

-lnitiatinQ Conditions, Threshold Values and Basis 110 Of 114 Basis: (cont.) 1 EAL #6 addresses the effect of flooding caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant that are not designed to be wetted or submerged. Escalation of the emergency classification is based on the damage caused or by access restrictions that prevent necessary plant operations or systems monitoring. The plant's IPE Flooding Analysis (Hatch U1 Internal Floods Analysis H96-0 and Hatch U2 Internal Floods Analysis H97-0) were used to determine the areas (Unit 1/2 Southeast or Northeast Diagonal Rooms) for the condition when developing this EAL. The high pressure ECCS systems (HPCI & RCIC) required by Technical Specifications for modes 1 and 2 (Le., these rooms are only applicable for this threshold in modes 1 and 2). HPCI Room, Unit 1 Southwest Diagonal Room, and Unit 2 Northwest Diagonal Room (RCIC Diagonals) were included for completeness, even though not a part of the IPE Flooding Analysis.

Threshold Value #7 covers site-specific phenomena of the hurricane based on the severe weather mitigation procedure.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL - Initiating Conditions, Threshold Values and Basis 1110f 114 Initiating Condition H FIRE Within PROTECTED AREA Boundary NOT Extinguished Within 15 Minutes of Detection.

Operating Mode Applicability: All Threshold Value:

1. FIRE in buildings or areas contiguous to any of the following areas NOT extinguished within 15 minutes of control room notification or control room alarm unless disproved by personnel observation within 15 minutes of the alarm:

Primary Containment Reactor Building Diesel Generator Building Control Building Intake Structure Basis:

FIRE: is combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIREs. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

PROTECTED AREA: the area which normally encompasses all controlled areas within the security protected area fence.

The purpose of this IC is to address the magnitude and extent of FIREs that may be potentially significant precursors to damage to safety systems. As used here, Detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a FIRE is occurring, or indication of a VALID fire detection system alarm. Verification of a fire detection system alarm includes actions that can be taken with the control room to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by personnel dispatched to the scene.

The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREs that are readily extinguished. The list is limited and applies to buildings and areas contiguous to plant VITAL AREAs or other significant buildings or areas.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 112 Of 114 Initiating Condition Release of Toxic, Asphyxiant, or Flammable Gases Deemed Detrimental to Normal Operation of the Plant.

Operating Mode Applicability: All Threshold Values: (1 OR 2)

1. Report or detection of toxic, asphyxiant, or flammable gas that has or could enter the Owner Controlled Area in amounts greater than life threatening or flammable concentrations that can affect NORMAL PLANT OPERATIONS.

OR

2. Report by Local, County, or State Officials for evacuation or sheltering of site personnel based on an offsite toxic, asphyxiant, or flammable gas event.

Basis:

NORMAL PLANT OPERATIONS: activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures. Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure from NORMAL PLANT OPERATIONs.

This IC is based on the existence of uncontrolled releases of toxic, asphyxiant, or flammable gas that may enter the Owner controlled Area and affect normal plant operations. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. Offsite events are included through a warning by local officials as the resultant affect on NORMAL PLANT OPERATIONS would be the same. This would preclude small or incidental releases, or releases that do not impact structures needed for plant operation. The Threshold Values are not intended to require significant assessment or quantification.

The IC assumes an uncontrolled process that has the potential to affect plant operations, or personnel safety.

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 113 Of 114 Initiating Condition Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant.

Operating Mode Applicability: All Threshold Values: (1 OR 2 OR 3)

1. Security events with potential degradation in the level of safety of the plant as determined from the Security Plan and reported by Security Shift Captain or designee.

OR

2. A CREDIBLE site specific security THREAT notification.

OR

3. A validated notification from NRC providing information of an aircraft threat.

Basis:

CREDIBLE THREAT: A threat is considered credible through use of 82SS-SEC-051-0, Threat Assessment CIVIL DISTURBANCE: is a group of two or more persons violently protesting station operations or activities at the site.

STRIKE ACTION: is a work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONs.

Reference is made to security shift supervision because these individuals are the designated personnel on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant security plan.

Threshold Value #1 is based on the security plan. Consideration is given to CIVIL DISTURBANCE, and STRIKE ACTION when evaluating an event against the criteria of the security plan.

The intent of Threshold Value 2 is to ensure that appropriate notifications for the security threat are made in a timely manner. This includes information of a CREDIBLE THREAT.

The intent of Threshold Value 3 is to ensure that notifications for the security threat are made in a timely manner and that Offsite Response Organizations and plant personnel are at a state of heightened awareness regarding the credible threat. This Threshold Value is met when a plant receives information regarding an aircraft threat from the NRC. Should the threat involve an airliner, the status and size of the plane is provided by NORAD through the NRC.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT_4_ Attachment Page TITLE: HNP EAL -Initiating Conditions, Threshold Values and Basis 114 Of 114 Initiating Condition Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NUE.

Operating Mode Applicability: All Threshold Value:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

Basis:

This Threshold Value is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the NUE emergency class.

From a broad perspective, one area that may warrant Emergency Director judgment is related to likely or actual breakdown of site-specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.

MGR-0009 Rev. 5.0

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Emergency Classification and Initial Actions I 73EP-EIP-001-0 17.0 ATTACHMENT _5_ Attachment Page TITLE: EmerQency Classification Worksheet 1 Of 1 NOTE: Events should be classified based on meeting the Ie and TV for an EAL considering each unit independently. IF both units are in concurrent events, then the highest classification must be made and used for the offsite notifications with the other unit events noted on the ENN form.

1. Select the affected unit's operating mode:

D Mode 1, 2 or 3 (go to step # 2)

D Mode 4, 5 or DEFUELED (go to step # 4)

2. Evaluate the status of the fission product barriers using Attachment 1 of 73EP-EIP-001-0.
a. Select the condition for each fission product barriers:

Fuel Cladding Integrity D LOSS D POTENTIAL LOSS DINTACT Reactor Coolant System D LOSS D POTENTIAL LOSS DINTACT Containment Integrity D LOSS D POTENTIAL LOSS DINTACT

b. Determine the highest applicable fission product barrier Initiating Condition (IC):

(select one) DFG1 DFS1 DFU1 o None NOTE: Determination of the initiating condition and associated threshold value is based on events which are in progress, past events, and their impact on the current plant conditions.

3. Using Attachment 2 of procedure 73EP-EIP-001-0, evaluate and determine the highest applicable "HOT" IC and associated Threshold Value (TV). NEXT, go to step 5.

IC#

or D None

4. Using Attachment 3 of procedure 73EP-EIP-001-0, evaluate and determine the highest applicable "COLD" IC and associated TV. NEXT, go to step 5.

IC# _ _ _ _ _ _ _ _ _ or D None

5. Check the highest emergency classification level identified from EITHER step 2 or 3 OR step 4:

D None Classification based on IC#

D Notification of Unusual Event D Alert Emergency D Site-Area Emergency D General Emergency

6. IC

Description:

7. Emergency Declaration At:

DATE TIME Emergency Director Signature: ______________________

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DOCUMENT TITLE: DOCUMENT NUMBER IDOCUMENT Version No:

Emergency Classification and Initial Actions I I 73EP-EIP-001-0 17.0 ATTACHMENT 6 Attachment Page TITLE: Unit 1 and Unit 2 Pretreatment Monitor (Flow vs. mR/hr) Graphs 1 Of 2 MGR-0009 Rev. 5.0

SNC PLANT E.I. HATCH I PQ. 127 of 127 Po.

DOCUMENT TITLE: IDOCUMENT NUMBER Version No:

Emergency Classification and Initial Actions EmerQency I 73EP-EIP-001-0 17.0 ATTACHMENT 6 Attachment Page TITLE: Unit 1 and Unit 2 Pretreatment Monitor (Flow vs. mRlhr) Graphs 2 Of 2 MGR-0009 Rev. 5.0

FG1 FS1 1iI&'il FU1 Loss of ANY Two Barriers AMJ;!. Loss or Potential Loss of ANY ANY Loss or Potential Loss of ANY Loss or Potential Loss of Loss or Potential Loss ofThird Two Barriers .!illl:!.liB Fuel Clad.QB RCS Containment Barrier Barrier Barrier Fuel Clad Barrier Loss I~~----~----~--~----------------~

Primary Coolant ActiYity IJeye) (Pg.37)

Coolant Activity greater than 300 uCilgm DEIl31 .QR uCilgm total RCS activity 4 Other Indjcations (Pg.38)

Offgas pre- and post-treatment monitors Offscale High.QB Drywell Post LOCA Monitor Offscale High

......l&IllU:£II£UlJL[S£I!!L..tW1i:111l:w (Pg. 38)

Judgment by the ED that the Fuel Clad Barrier is lost Consider conditions not addressed and inability to detennine the status of the Fuel Clad Barrier 3 RCS Leak Rate (Pg.39)

~:;:~~~~

I

.... -""'....."""....,."" (Pg.

(Pg. 39) 1

39) RCS leakage GREATER THAN 50 gpm inside the drywell Main Steamline break as indicated by the failure of both MSIVs onc one line to close .QR Unisolable primary system leakage outside drywell as indicated by Secondary A. High MSL Flow Containment operating temperatures or radiation levels above Max. Nonnal Operating

.QR Values (SC - Secondary Containment Control Flowchart ~ Table 4 & Table 6)

B. High Steam Tunnel Temperature annunciators

.QR C. Turbine Building BuUding MSL leak annunciator

.QR D. Direct report of steam release 4 DryweJl 1:1o....w= ....Radiation Monitoring (Pg.

"""......!W...I!!l!J.I.WL.....,. (Pg. 39) 39)

DWRRM greater than 138 Rlhr S Other Indications (Pg. 39)

Drywell Post LOCA Monitor 4.7 I E+04 cpm Emergency R Director Judgment (Pg.40) (Pg. 40)

Judgment by the ED that the ReS RCS Barrier is lost. Consider conditions not detennine the status of the ReS addressed and inability to dctcnnine RCS Barrier Loss I

1. DmyelJ Pressure (Pg. 40) 1..... DryweJl Pressure (Pg.40) 1 ..w:mJ:JU:Jl:.<Ul!.IJ:

Rapid unexplained decrease following initial increase 56 PSIG AM!. rising

.QR .QR Drywcll pressure response not consistent with LOCA conditions Greater than or equal to 6% Hz A.IS.Il greater than or equal to 5% O2

2. Reactor yessel vessel 'Vater Level (Pg.40)

Primary containment flooding required by EOPs as indicatcd indicated by entry into procedure 31EO-EOP-112-1/2 Primary Containment Flooding.

II ~-f!!Illll!.J!m~l,mIlUi.I!DJEl!iI~.l!L.~WI.~

~.l:l!IilalJlIl1£D1..I!ilIIlI!til!Il..El!i.~..l!LIill¥iW. (Pg. 41) ar:m.~2W11stream downstream breach breach outside outside Unisolable primary system leakage outside drywell dtyWell as indicated by Secondary Containment operating temperatures or radiation levels above Max. Safe Operating Values (SC - Secondary Containment Control Flowchart - Table 4 & Table 6) 1.

4. Significant Radioactive Inventory in Containment (Pg. 41)

DWRRM greater than 110,000 I 10,000 Rlbr Rlhr Il!.o.--'1dI!lllI.I~:DlJ.[£!::!l!L8!lI!l:!nW

6. Emergency Djrector Judgment (Pg. 41) 6. Emergency Director Judgment (Pg.41)

Judgment by the ED that the Containment Barrier is lost. Consider conditions Judgment by the ED that the Containment Barrier is potentially lost. Consider conditions not not addressed and inability

. . to determine the status of the Containment Barrier addressed and inability . to determine detennine the status of the Containment Barrier o

l TVs based on TVs based on TVs based on Secondary the DWRRM thcRWL Containment Max. Safe Operating Values

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Edwin I. Hatch Nuclear Plant Enclosure 8 DOEJ-HR2100455301-M001 LOSP Evaluation of Fire Areas 2408/2409

Southern Nuclear Operating Company DOCUMENTATION OF ENGINEERING JUDGMENT DOEJ-HR2100455301-M001 LOSP Evaluation of Fire Areas 2408/2409 Version Record Version ReviewerIDate No. Si nature 1.0 John Lattner 03/10/2010 NMP-ES-039-002. Version 2.0 Documentation of Engineering Judgment

DOEJ-HR2100455301-MOOI Southern Nuclear Operating Company

Purpose:

In support of RER 2100455301, this DOEJ documents an evaluation of a specific fire in the 2F Switchgear Room (Fire Area 2408) and its effects on specific targets. The evaluation is to demonstrate a single switchgear fire or high-energy arching fault (HEAF) in Switchgear Rooms 2F and 2G will not result in a loss of offsite power (LOSP).

Design Inputs (Reference NMP-ES-042):

1. H-12618, Sheet 1, Version 4.0
2. H-23122, Sheet 1, Version 2.0
3. S-13595, Sheet 6, Version 1.0
4. S-13668, Sheet 1, Version 1.0
5. Okoguard-Okolon Type MV-90 Cable Product Data Sheet

References:

1. Inspection Report 05000321/2009005 and 05000366/2009005, February 12, 2010
2. NUREG 1805, December 2004
3. NUREG 1824, May 2007
4. NUREG 6850, September 2005
5. SFPE Handbook of Fire Protection Engineering, 3 rd Edition
6. RER 2100455301 Assumptions:
1. A 4 kV switchgear provides the worst-case scenario fire source to effect a LOSP.
2. Failure criteria is met if one of the two following conditions is met:

=

a. Target temperature 330°C (625°F)
b. Target heat flux = 11 kW/m2 (1.0 BTU/ft2)
3. NUREG-1805 Spreadsheets provide a conservative qualitative analysis of the fire scenario.
a. Steady state fire
b. Liquid pool fires are bounding for the fire scenario
4. No credit will be taken for thermal and radiant shielding due to the cable tray construction, or any other obstruction; this creates a more conservative calculation.
5. In Switchgear Fire #2, the heat release rate (HRR) is postulated by the correlation provided by the SFPE Handbook, 3rd Edition (Page 3-16).
6. In Switchgear Fire #2, the fire area is limited to the width of the bus duct, squared.

The remainder of the lower bus duct is assumed to be "damaged but not ignited" (Appendix M or NUREG 6850). This assumption also complies with the limitations of Chapter 9 of NUREG 1805 calculation which requires an approximately circular fire area.

7. All measurements are from Fire Area 2408; however due to similarities in room construction and equipment placement, the results from the calculations are assumed to be acceptable for Fire Area 2409.

1 NMp*ES*039-002, Version 2.0 Documentation of Engineering Judgment,

DOEJ-HR2100455301-MOOI Southern Nuclear Operating Company Evaluation:

General Information NRC Inspection Report 05000321/2009005 and 05000366/2009005, dated February 12, 2010, evaluated an inspection finding for the 2AEDG LOSP timer card failure as a PRELIMINARY GREATER THAN GREEN FINDING. The reasons for the PRELIMINARY GREATER THAN GREEN FINDING include the long exposure time for the timer card failure and for sequences which result in loss of offsite power (LOSP).

Fires which may induce a LOSP increase the risk.

SNC conducted specific fire modeling to provide a realistic assessment for the possibility of a switchgear fire in fire areas 2408 and 2409 to determine if a loss of both sources of offsite power due to a single fire is credible.

The targets for each fire scenario are those cables which affect LOSP and are closest to the presumed fire source. In both fire scenarios, the cables are within one of two bus ducts 2R21-S003 and 2R21-S004.

Switchgear Fire #1 Switchgear Fire #1 is defined as a "normal" fire initiated inside the switchgear. CFAST was utilized as the fire modeling program as it is acceptable tool to calculate target temperature and heat flux. CFAST requires the following inputs:

  • Compartment geometry and materials
  • Fire location
  • Heat release rate (HRR)
  • Fire duration
  • Target location and materials
  • Ventilation size and rating The compartment (Switchgear Room 2F, Fire Area 2408) is defined as a concrete enclosed volume with measurements (W x L x H) of 7.3152 m (24 ft) x 9.144 m (30 ft) x 5.7912 m (19 ft). Wall thickness was set at 0.9144 m (3 ft).

The fire, which is characterized as a point source by CFAST, is located at 0.3048 m (1 ft) below the top of frame 1. This particular switchgear frame was utilized since it was closest to the target location and guaranteed the loss of one electrical bus. The HRR is specified as 200kW per Appendix G of NUREG 6850 (98th percentile). The fire duration is set for 60 minutes (50% longer than recommended HRR profile per Appendix G of NUREG 6850).

2 NMP-ES-039'OO2, Version 2.0 Documentation of Engineering Judgment,

DOEJ-HR2100455301-MOOI Southern Nuclear Operating Company The target, which is also defined as a point in space by CFAST, was selected as the closest point on the bus duct holding the target cables. The target material was defined as steel.

Mechanical ventilation is present through two (2) redundant roof ventilator fans. Both fans are 28 %" square with a rating of 1.88 m3/s (4,000 cfm). The location of the fans cannot be modeled by CFAST; CFAST assumes the roof vent to be centered in the room.

This alternation should not significantly affect the results.

The results of the CFAST calculation was a maximum target temperature of 69.8 °C (157.6 OF) and a maximum heat flux of 0.8717 kW/m2 (0.08 BTU/ft2 ), which are well below the failure criteria. Relevant model data has been plotted in Figure 1 and Figure 2 below. Therefore, a "normal" switchgear fire does not present a significant risk to cause a LOSP event in the 2F Switchgear Room.

Switchgear Fire #1 Switchgear Fire #1 Target Tern perature Target Heat Flux 100 l00~  :

iJ i10

~

8.

E dU 50 40 20 o :

i:t

~

f 20:) ~

a ;

Failu"e t'eat ~Iu. is 1.100 W/m')

a 500 lOGO 15:'0 20')0 2500 3000 3~00 ~JOO : o 503 1000 1500 2000 2500 3,,00 3500 100:'

Figure 1 l1me lsI FIj;"re 2 Time (s)

Switchgear Fire #2 Switchgear Fire #2 is defined as a fire in an affected bus duct as a result of a high-energy arching fault (HEAF) condition. The target is a second bus duct located directly above the fire. Appendix M of NUREG 6850 provides the methodology to characterize the damage associated with a HEAF. From this method. FDTs spreadsheet inputs where developed and used to determine a target temperature. The target temperature will be conservatively approximated by the centerline plume temperature calculation (Chapter 9 of NUREG 1805) since the target is directly above the fire source. A heat flux calculation is not required since the radiant energy is absorbed by the smoke particles in the plume.

The centerline plume temperature calculation requires the following information:

  • Heat release rate
  • Distance from the top of the fuel to target
  • Fuel area
  • Ambient temperature 3

NMP-ES-039-002, Version 2.0 Documentation of Engineering Judgment.

DOEJ-HR2100455301-MOOl Southern Nuclear Operating Company Per the assumptions above, the HRR is determined by a correlation published in the SFPE Handbook:

,where ql' is the full-scale heat release rate [kW]

if:', is the bench-scale heat release rate measurement [kW/m2]

A is the exposed tray area actively pyrolyzing [m 2]

The value of 4:', is provided in a table for selected cables. The oft-site power feeders in Fire Areas 2408 and 2409 are located in the cable bus ducts. The cabling has a chlorosulfonated polyethylene jacket and ethylene-propylene rubber (EPR) insulation.

EPR insulation is not represented in the referenced table, however crosslinked polyethylene, another thermoset cable, is referenced and is assumed an acceptable substitute. Using this cable composition, 4:', = 204 kW/m2 The fire area is considered to be the square of the width of the cable tray or (15 in)2 = 1.56 fe =0.145 m2. Therefore, using the above equation, the HRR would be 13.33 kW.

The distance from the top of the fuel source (lower bus duct) to the target (upper bus duct) is estimated to be 3'-3" (.991 m).

The ambient temperature is assumed to be the temperature of the room after the HEAF, which is heated due to the large energy output. Appendix M of NUREG 6850 provides a conservative calculation for the General Room Heat Up. The use of this ambient temperature disregards any heat dissipation associated with the room construction (concrete walls, ceiling and floor act as a heat sink) and ventilation (approximately 17 air changes an hour). The calculation for General Room Heat Up is provided below:

,where I.1T is the change in temperature of the room [K]

rna is the product of air mass density in the room and the volume of the room [kglm 3 ]

Cp is the specific heat capacity of air [J/kg-K]

Q is the potential energy release to the room from the explosive failure [W] and is:

Q=V*A*'l' ,where V is the voltage of the switchgear [V]

A is the maximum available fault current [amps]

'l' is the duration of the energetic event [s]

4 NMP-ES-039-002. Version 2.0 Documentation of Engineering Judgment.

DOEJ*HR2100455301*MOOl Southern Nuclear Operating Company Therefore, b.T = V* A-r (Volume* PaJ*Cp (4160V). (42,OOOamps)* (%Os) b.T--r----~--~~--~~

b.T--r----~----~--~~

3

- (387.4m .1. 225 *YmJ998}-1g.K b.T = 14.56MJ

.4736 MYK b.T =30.74°K = 30.74°C =87.34°F Troom = T_ + b.T = 75°F +87.34° F = 162.34°F With all the factors assembled and inputted into the FDTs spreadsheet, the estimated centerline plume temperature of the cable fire at 3'-3" above the fire is 169.1 °C (336.3°F).

This estimated temperature is well below the failure criteria. See Attachment 1 for spreadsheet printout.

==

Conclusion:==

The conclusions of this DOEJ find there is no significant risk of an LOSP due to a single fire in one of the switchgear rooms; neither an internal switchgear fire or a high-energy arching fault results in a loss of both sources of offsite power.

List of Attachments:

1. FDTs Chapter 9 Spreadsheet Calculation for Switchgear Fire #2 5

NMP-ES-039-002, Version 2.0 Documentation of Engineering Judgment,

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Edwin I. Hatch Nuclear Plant Enclosure 9 Fairbanks Morse Report 5.08-6.06-0260 Genset Operation without Service Water

BF30lOB PAGE NO. 1 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVED4 TITLE: BY ~~/--

GENERAL:

This report covers calculations of the time a unit can run with the service water off from the time of start up of the unit from keep warm temperatures until such time as the system alarms are actuated for the Lube Oil and/or Jacket Water Systems. The conditions for the calculations are outlined as follows:

Ambient Temperature - 95°F (also Service Water Temperature)

Lube Oil System - Keep Warm Temperature - 1550P

- High Temperature Alarm/limit - 220°F

- a temperature difference of 65°F Jacket Water System - Keep Warm Temperature - 110°F

- High Temperature Alarm/limit - 195°F

- a temperature difference of 85°F The approach to the calculation was to compute the heat required to raise the temperature of the various fluids from the initial conditions to the final conditions, including the lube oil and jacket water contained in the engine and its associated equipment and piping, and the service water on the service water side of the heat exchangers. There was no attempt to determine the heat required to heat the metal parts of the engine or of the parts of the various systems. To do this would require the preparation of a very sophisticated model with many elements and inter-relationships which would require more time than has been allotted for preparation of this report.

The units are a 12 cylinder, mode 38TD8-1I8 x 10 Fairbanks Morse Opposed Piston (OP) engine rated at 2850 KW at 900 rpm. The engine is mounted on a skid with a generator and all system equipment and piping on the skid unit. Where possible, the drawings and data for the actual contract were used in the computations. Where assumptions have been made, they are identified in the computations or in the discussion below.

COMPUTATION BASED ON FLUID TMPERATURE RISE TIME:

The approach to the computation was to determine the amount of each fluid contained in each system.

This involved estimating the length of piping runs on each system, the volume of the various pieces of equipment (such as the lube oil filter and strainer, the lube oil sump/crankcase, the volume of the tubes on the heat exchangers, the volume of the shell side of the heat exchangers, and so forth). The volume of each element was converted to the gallons of the fluid, and that in turn was converted to the pounds of that particular fluid. Standard heat transfer formulae were then used to determine the heat required to change the temperature of that amount of fluid by the temperature difference noted above.

It was assumed that on the service water side of the heat exchanger, there is water present and the service

FAIRBANKS BF3010B PAGE NO. 20F10

~~~3'!~~0

~~~3'\~~0 an EnPro Industries company ENGINEERING REPORT FILE NO. 5.08-6.06-0260 DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSMI2, 71038, 71040, 75025, 75026 BY V. T .Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVE~

TITLE: BY~-~

water temperature would approach the system temperature (alarm point temperature noted above) by not less than 10°F (otherwise, there would be no heat transfer to the service water contained in that side of the heat exchanger, and there would be no accounting for the heat to raise that water temperature). This would result in a Service Water temperature rise of 115Gp (210-95) for the Lube Oil System and 90°F (185-95) for the Jacket Water system.

The results of the analysis and computation are given in the following table.

TABLE I.

Engine output - % 0 25 50 75 100

-KW 0 712.5 1425 2137.5 2850

-BHP (96% Gen Eff) 0 995 1990 2985 3980 LUBE OlL HR Rate - BTU/br-hp 655 Heat Rejection - BTU/hr 260,661 847,148 1.433M 2.020M 2.606M Time for Delta Temp Rise - min. 38.94 11.98 7.08 5.02 3.89 Rate of Temperature Rise - o/min 1.67 5.43 9.18 12.94 16.69 JACKET WATER HR-BTU/br-hp 640 Heat Rejection - BTU/hr 254,692 827,748 1.401M 1. 974M 2.546M Time for Delta Temp Rise - min 26.40 8.12 4.80 3.41 2.64 Rate of Temperature Rise - °/min 3.22 10.46 17.71 24.95 32.19 It should be understood that the engine does put out heat at no-load, in proportion to the horsepower required to overcome the friction and losses that exist at no load. There are estimated to be 10% of the rated horsepower. Therefore, the heat rejection at no load is taken as 10% of the heat rejection at full load. Those no-load losses are included in the full load heat rejection automatically. This 10% fraction is proportioned into the other load points (25, 50 and 75% loads) also to derive the expected heat rejections at those points based on the full load heat rejection (which is the only point that FME is usually concerned about).

Spread sheets for each of the computations are included as Appendix A and B. The data from the spread sheet, as summarized on the table above, is included in the Graph attached as Figure 1.

TEST DATA ANALYSIS:

Because the time for heating up the Jacket water in the analysis above seems very short, and less than our experience would indicate, it was decided to also make an analysis using data from a test run in 1991 on a similar unit. That test was run on a spare unit provided for Philadelphia Electric (PECO) for their Peach Bottom plant. The test consisted of running the unit at is rated output (2850 KW) until equilibrium temperatures were achieved. At that time, the service water supply (motor driven pump) was shut off.

The temperature in the Jacket Water, Lube Oil and Intercooler Water systems were monitored and when the temperature limits were obtained, the service water was turned back on. The duration of the test,

BF3010B PAGE NO. 3 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSMI2, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVED4 TITLE: BY~~-

between the equilibrium condition and the limiting conditions was 8.8 minutes. That report is included as Appendix D. A spread sheet showing that data normalized to the Georgia Power Hatch plant conditions is included as Appendix C. In that analysis, the test data was normalized to a factor of the Rate per KW

(<lp/min-kw). That rate was then multiplied by the Hatch load conditions of 0, 25, 50, 75 and 100 percent load at the equivalent KW loads computed based on a 10% friction KW load at the no-load point. This analysis indicates that the Jacket Water system is controlling, but that the time to achieve alarm point conditions from the Keep Warm conditions are much longer than computed with the calculated data method used above.

The points from this analysis for the Jacket Water and Lube Oil systems have also been included in the graph shown on Figure 1. The following table (TABLE II) summarizes this data.

TABLE II Load - % 0 25 50 75 100

- KW 0 712.5 1425 2137.5 2850

- Equiv KW 285 926.25 1567.5 2208.75 2850 LOSYSTEM-Rate of Change - deg/min 0.24 .78 1.32 1.88 2.41 Time - minutes 270 83.0 49.1 34.8 27.0 JWSYSTEM-Rate of Change - deg/min 0.58 1.87 3.16 4.46 5.75 Time - minutes 148 45.5 26.9 19.1 14.8 ICWSYSTEM-Rate of Change - deg/min 0.71 2.32 3.93 5.53 7.14 Time-min 175 53.9 31.8 22.6

17.5 CONCLUSION

AND COMMENTS:

The Jacket Water System controls in both analysis. This is true for several reasons. The lube oil system has more total fluid to heat up, 3,664Ibs. versus 1,035Ibs. in the Jacket Water system. The heat transfer coefficient (Cp) for water is 1 whereas that for oil is 0.5. Therefore, the water heats at twice the rate of the lube oil.

The analysis based on the heating of the fluid volumes from keep warm to alarm point temperatures does not take into account the heat required to simultaneously raise metal temperatures, which may be about the same as that of raising the fluid temperatures. In comparison of the test derived times with the calculated times, the conservative approach to assure the engine is not damaged by being operated without service water flow is to use the calculated times per TABLE I. If an engine is run without service water flow, the operating conditions should be monitored and the unit unloaded or shut down when the operating parameters (high jacket water temperature - alarm, or high lube oil temperature - alarm) are exceeded and alarms activated.

BF30lOB PAGE NO. 4 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVE~

TITLE: BY~~~

BASED ON UNIT AT KEEP WARM TEMPERATURE (FRO~ START UP - UNIT NOT RUNNING)

JACKErWATER FRO~ 11 oro 196 DEG F LUBI: OIL-FROM 1.55 TO 22.0DEG F

- - - - CALCULATEDDATA FROM TEST DATA INTmCOOLER WATER SYSTEM

~ UJB~ OI~sYSTEM WBI' OI"SYSTEN

~~-----

JACKErWATmSYSTEM JACKErWATffiSYSTEM

~ =:::::::

LUBE OIL- SYSTEM

+0 50 60 PERC~NT LOAD

~

o z:

~

FIGURE 1 - LOSS OF SERVICE WATER

BF30lOB PAGE NO. 5 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DA TE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVE~

TITLE: BY~-~

APPENDIX A:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT - SERVICE WATER STUDY Study of Time for Engine Operation without Service Water - Start Up and at various loads LUBE OIL SYSTEM ANALYSIS: Density 7.5 Ibs/gal Volume 231 cu.in/gal Cp for Oil 0.5 BTU/hr-Ib-OF Keep Warm Temp 155 OF Alarm point Temp 220 OF Temp Difference 65 OF Weight of Fluids in the System: LUBE OIL Equipment or system piping Volume Weight cu.in. gallons Ibs.

Lube Oil Sump at average level - 12.38" 65,114 281.88 2114.1 (30.24 wide x 174 long x 12.375 deep avg)

Lube Oil Filter 26,743 115.77 868.28 (23" ID x 68.5" high X 90% filled)

Lube Oil Strainer 6,401 27.71 207.84 (15" ID x 40.25" high x 90% filled)

Heat Exchanger - Shell (oil) side 7952 34.42 258.19 (15" ID x 96" long less tube area 16965 306 tubes 5/8" OD x 96" long) (9012)

Piping Systems: Diameter Length Volume (cu.in.)

Sump to Pump 4 24 301.59 Pump to Filter 4 84 1055.58 Filter to Amot (A) 4 37 464.96 Amot (C) to Heat Exchanger 4 30 376.99 Amot (B) to bypass junction 5 48 942.48 Junction to Strainer 5 165 3,239.77

BF30IOB PAGE NO. 6 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVE~

TITLE: BY~~~

Strainer to Engine 4 20 251.33 Total Volume 6,632.69 Total Gallons 28.71 Totallbs. (piping) 215.35 Appendix A - continued Total Pounds of Oil in System 3,663.75 Ibs Amount of Heat (BTU's) required to raise oil by Temp Difference 119,072 BTU HEAT EXCHANGER-SERVICE WATER SIDE Density 8.34 Ibs/ga I Volume Weight cu.in. Gallons Ibs Heat Exchanger - Tube (water) side Tubes {5/8" aD x .049"w x 96" x 306 6408 27.74 231.34 Volume in Ends {2 x is ID X 16" long 5655 24.48 204.16 Total Pounds of Service Water in HX Totallbs 435.50 Temperature at Start (ambient) 95 of Temperature at End Point (lOOF less than La end) 210 of Temperature Difference from Start to End Point 115 of Amount of Heat (BTU's) required to raise Service Water Temp. 50,083 BTU Total Amount of Heat to Raise Oil and Service Water Temps 169,155 BTU Engine Heat Rejection from the LUBE OIL System at various Loads Percent Load (2850 KW Rated) No Load 25 50 75 100 KW Loading 0 712.5 1425 2137.5 2,850 BHP Loading (96% Gen Efficiency) 0 995 1990 2985 3980 Heat Rejection at Rated Load - BTU/hr-bhp 655 Heat Rejection Rate - BTU/hr 260,661 847,148 1,433,636 2,020,123 2,606,610 Time to Raise Temperature by Difference 38.94 11.98 7.08 5.02 3.89 Minutes Rate of Temperature Rise - OF per minute 1.67 5.43 9.18 12.94 16.69 NOTES:

1- Sump volume may vary from 207 gallons to 369 gallons (3" above suction screen [9.25 depth] to 2" below lowest point of crankshaft [15.5 depth]).

The average value was used (282 gallons at mid-point -12.38/1 depth)

BF30lOB PAGE NO. 7 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVE~/

TITLE: BY~c~

APPENDIX B:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT - SERVICE WATER STUDY Study of Time for Engine Operation without Service Water - Start Up and at various loads JACKET WATER SYSTEM Density 8.34 Ibs/gal Volume 231 cu.in/gal Cp-H20 1 BTU/hr-lb-oF Keep Warm Temp 110 of Alarm point Temp 195 of Temp Difference 85 of Weight of Fluids in the System: JACKET WATER Equipment or system piping Volume Weight cu.in. gallons Ibs.

Volume in Cylinder Liner Jackets & Belts 50 417 Heat Exchanger - Shell (JW) side 5497 23.80 198.46 (12" 10 x 96" long less tube area 10857 182 tubes 5/8" 00 x 96" long) (5360)

Piping Systems: Diameter Length Volume Engine Inlet Piping - both sides 4 126 3166.73 Engine Cross Pipe - both sides 4 60 753.98 Engine Outlet Header 4 132 1658.76 Engine Header to Amot (A) 5 75 1472.62 Amot (C) to Heat Exchanger 5 59 1158.46 Amot (B) to bypass junction 5 92 1806.42 Junction to Pump Inlet 5 81 1,590.43 Total Volume 11,607.40 cu.in.

Total Gallons 50.25

BF30lOB PAGE NO. 8 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DA TE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSMI2, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVED4 TITLE: BY~/~

Totallbs. 419.07 Total Pounds of Water in System Amount of Heat (BTU's) required to raise oil by Temp Difference Appendix B - continued SERVICE WATER SIDE OF HEAT EXCHANGER Density 8.34 Ibs/gal Volume Weight cu.in. gallons Ibs Heat Exchanger - Tube (service water) side Tubes (5/8" 00 x .049"w x 96" x 182) 3811 16.50 137.60 Volume in Ends (2 x 12" 10 x 16" long) 3619 15.67 130.66 Total Pounds of Service Water in HX Totallbs 268.26 Temperature at Start (ambient) 95 of Temperature at End Point (10°F less than JW end) 185 OF Temperature Difference from Start to End Point 90 OF Amount of Heat (BTU's) required to raise Service Water Temp. 24,143 BTU Total Amount of Heat to Jacket Water and Service Water Temps 112,079 BTU Engine Heat Rejection from the LUBE OIL System at various Loads Percent Load (2850 KW Rated) No Load 25 50 75 100 KW Loading 0 712.5 1425 2137.5 2850 BHP Loading (96% Gen Efficiency) 0 995 1990 2985 3980 Heat Rejection at Rated Load - BTU/hr-bhp 640 Heat Rejection Rate - BTU/hr 254,692 827,748 1,400,804 1,973,861 2,546,917 Time to Raise Temperature by Difference 26.40 8.12 4.80 3.41 2.64 Minutes Rate of Temperature Rise - OF per minute 3.22 10.46 17.71 24.95 32.19

BF30lOB PAGE NO. 9 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DA TE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSM12, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER TITLE:

APPENDIX C:

SCN - Georgia Power - Hatch Nuclear Power Plant Study of Time for Engine Operation without Service Water - Start up and at various loads Based on Testing done for PECO - Peach Bottom spare unit - 1991 The following is a compilation of the Data from that testing.

Based on a unit rating of 2850 KW, the average rate of temperature rise during the testing was:

TEST UNIT RATING 2850 KW 8.8 minutes Tem peratures Begin End Delta Rate Rate per KW of of of of/min °F/min-kw LO SYSTEM 167.7 188.9 21.2 2.409 0.000845 JW SYSTEM 144.1 194.7 50.6 5.750 0.002018 ICW SYSTEM 102.3 165.1 62.8 7.136 0.002504 USING GEORGIA POWER - HATCH II RATING

% Load 0 25 50 75 100 KW Load 0 712.5 1425 2137.5 2850 Equivalent KW Load 285 926.25 1767.5 2208.75 2850 LO SYSTEM Keep Warm Temp 155 Alarm Point Temp 220

BF30lOB PAGE NO. 10 OF 10 FILE NO. 5.08-6.06-0260 an EnPro Industries company ENGINEERING REPORT DATE March 4, 2010

SUBJECT:

GEORGIA POWER - HATCH NUCLEAR POWER PLANT PREPARED RE: ENGINE SERIAL NOs - 38D871037TDSMI2, 71038, 71040, 75025, 75026 BY V.T.Stonehocker REPORT GENSET OPERATION WITHOUT SERVICE WATER APPROVED4 TITLE: BY ~(/~

Delta Temp 65 Rate of Change 0.241 0.783 1.325 1.867 2.409 Time-Minutes 269.8 83.0 49.1 34.8 27.0 Appendix C - continued JACKET WATER SYSTEM Keep Warm Temp 110 Alarm Point Temp 195 Delta Temp 85 Rate of Change 0.575 1.869 3.163 4.456 5.750 Time-Minutes 147.8 45.5 26.9 19.1 14.8 INTERCOOLER WATER SYSTEM Ambient Temp 95 Maximum Allowable 220 Delta Temp 125 Rate of Change 0.714 2.319 3.925 5.531 7.136 Time-Minutes 175.2 53.9 31.8 22.6 17.5

Edwin I. Hatch Nuclear Plant Enclosure 10 RCIC Calculation 1 Unit 2 Battery Life

SOUTHERN COMPANY A

E"'rx110 S<r ... y"", W..,."'"

Design Calculations - Nuclear Southern Company Services, Inc.

IE 0118 (Note 1)

Calculation Number Project Discipline E. I. Hatch Nuclear Plant - Unit(s): 01 12Il2 Electrical Objective Job Number To Evaluate Station Service Battery 2A Capacity for 5 Hour RCIC Operation REA HT 02-637 SubjectITitle Station Service Battery 2A Capacity for 5 Hour Duration, To Support Probabilistic Safety Assessment (PSA)

Design Engineer's Signature Vivek Srinivasan !ltAJpjVt,. A~~

I06/15/02 J Date Last Page Number 14 Contents Topic Page Attachments Number (CorT1JUler Printouts. Technical Papers, Sketclles, Correspondence) of Pages Purpose of Calculation/Summary of Conclusions 1 At. 5 Hour Battery 2A Load Profile/Sequence 8 Criteria 2 A2. Battery 2A, 5 Hour Load Profile 1 Major Equation Sources/Derivation Methods 2-4 B. Battery Sizing Worksheets 12 Assumptions 5 C. Battery Voltage Profile Worksheets 4 Listed References 5 D. Battery 2A Discharge Characteristics 14 7- E. SNC Letter Log No. HL-6252, Dated.

Body of Calculations 2 14 06/18/02 Safety Related I&J Yes 0 No Nonsafety-related That Could Impact Safety-Related 0 Yes 0 No Record of Revisions Rev. No. Descrl ption OrlQinator I Dale Reviewer I Date Supervisor I Date 0 Issued Per REA HT 02-637 I/M rob!! 0 2- Sl£.~ I b \-2-11 DO; 6/'),,101-I I I Notes:

1. This calculation is a study type calculation developed for Probabilistic Safety Assessment (PSA) and is not required to be maintained for plant modification unless required by PSA.
2. This calculation is based on Station Service Battery 2A Sizing and Voltage Profile calculation SENH 93-024, Rev.4 and must be worked along with calculation SENH 93-024 for all future use.

Nuclear -Calculation.dot 1 Rev. 1/06*30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 [8]2 E 0118 SubJecVTltle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 1 of 14 To Support Probability Risk Analysis Program (PSA) 1.0 Purpose/Scope:

This calculation perfonns an evaluation to determine if the station service battery 2A (2R42-SOOlA) has adequate capacity to meet a 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> duty cycle developed to support the Probability Risk Analysis (PRA) program. The 5-hour duty cycle is based on the LOCAILOSP profile used for calculation SENH 93-024, Rev.4 (Station Service Battery 2A Sizing and Voltage Profile) and Attachment E. Additionally, the battery is evaluated for its ability to meet battery voltage requirements with one/two cells removed from each 601120-cell bank. This calculation also verifies the capability of the nonnal and standby chargers to recharge the battery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a duty cycle discharge while simultaneously supplying the connected loads.

2.0 Summary of

Conclusions:

The calculation (Attachments Bl through B4) shows that station service battery 2A has adequate capacity with sufficient design margins as shown in the table below along with temperature correction factor and aging factor.

The station service battery 2A voltage profile is calculated using a 5-hour duty cycle, developed to support the PRA program and the discharge characteristics of the C&D LCR-25 cell, to verify the requirements of criteria 3.2 for a minimum battery terminal voltage of 105/210 volts.

However, as discussed in calculation SENH 94-021, additional battery connection resistance should be included when determining minimum battery end voltage. Since it could become necessary to temporarily remove/replace individual cells without taking the battery out of service, battery capacity was evaluated for the removal of one cell (59/118) in each 60 cell bank.

Considering the above and battery sizing and voltage profile worksheets, the following table presents the battery sizing factors/margins for comparison.

Without Additional With Additional Connection Resistance Connection Resistance 60/120 Oneffwo 60/120 Oneffwo Cells Factors/Margins Normal Cells Normal* Removed Removed Temp. Correction 8% 8% 8% 8%

Factor Aging Correction 25% 25% 25% 25%

Factor Design Margin 46.38% 39.85% 44.00% 34.90%

Ref: Att. Bl Att. B3 Att. C2 Att. C4 NlIClear *Calculation.do! I Rev. 1 10&30*99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unites): 01 1R12 E 0118 Sub/ecVTIlle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 2 of 14 To Support Probability Risk Analysis Program (PSA)

3.0 Criteria

3.1 Per IEEE Standard 485-1983, the battery will be sized to have a minimum 8% temperature correction factor, 25% aging correction factor, and 7% design margin.

3.2 The minimum battery terminal voltage will be 105/210 volts.

3.3 The maximum DC system voltage is 1401280 volts. This is based on the maximum voltage limit of devices fed from the DC system. Site procedures address and control adherence to this limit.

3.4 Battery chargers will be sized to recharge the battery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a load profile discharge and simultaneously support the connected loads.

4.0 Equation Sources/Derivation Methods:

4.1 Battery Load Profile/Sequence (Attachment A)

The battery load profile/sequence identifies the sequencing (on/off of the individual loads connected to the battery including the load amperes in one minute increments for the time period or duration (300 minutes) required for the battery.

All the loads included in the battery load profile were extracted from Attachment B of Calculation 93-24, Rev.4. The first minute of the load profile is broken down into seconds to satisfy the initial one minute requirement of IEEE 485-1983. The changes made to the LOCA profile to develop the 5-hour load profile are based on the assumptions discussed in section 5.0.

4.2 Battery Sizing Worksheet Attachments: Bl and B3 The battery sizing worksheet determines the battery cell size required to satisfy the load and design parameters. The selected cell size as well as other data is identified in the Battery Input Data box. The data in these two boxes are entered manually based on the battery manufacturer's data sheet, battery discharge curves, plant identification of the battery and other design requirements. The load profile data in the Load Duty Cycle box is manually entered from load profile (Attachments A). The Output Results box shows the battery and battery charger sizing results. The battery sizing worksheet is based on the example shown in IEEE 485-1983 using amperes/positive plate data.

Attachments: B2 and B4 Data in the "Period" column is manually entered. Data in the "Load Amps" and "Period Duration" columns of each "Section" is automatically entered from the "Load Duty cycle" box from Attachment BIIB3. Within each "Section", data for the "Load Change" column is automatically calculated by subtracting the load amps in the previous period (step) from Nuclear -Calculation. dot I Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unites): 01 1&12 E 0118 SubjectrTitle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 3 of 14 To Support Probability Risk Analysis Program (PSA) the present period. For each period in a section, data for the "Minutes To End of Section" column is automatically determined by the algebraic sum of the "Period Duration" from the present period to the last period in the section. Data in the "Amp/Pos Capacity (Adjusted)" column is determined by multiplying the "Amp/Pos Capacity (Unadjusted)"

from Attachment D by the "Cell Sizing Correction Factor". Data in the "Required #

Positive Plates" column is determined by dividing the "Load Change" by "Amp/Pos Capacity (Adjusted)", these two operations are also performed automatically.

The section requiring the greatest number of plates is the section that determines the size of the battery. The number of plates from this section is automatically entered along with other data from the "Input Data" and "Battery Input Data" boxes into a small table at the end of the battery sizing worksheet identified as "Calculated Cell Size. A "0" for Random Size is manually entered, indicating that this factor is not applicable.

"Uncorrected Size" is determined by adding "Maximum Section Size" and "Random Size" together. "Required # Positive Plates" is determined by multiplying "Uncorrected Size",

"Temperature Correction Factor", "Aging Factor", and "Design Margin" together.

Calculation of the "Continuous Amps" ("Load Duty Cycle" box) Attachment Bl provided by the battery and the "Calculated Charger Amps" ("Output Results" box) are described in the body of this calculation.

Design margin on the battery is calculated directly on the battery sizing worksheet using the relation Design margin = No. Of Positive Plates / (Max section size*Temperature correction factor* Aging factor)

Attachments: Dl and D2 Data for the "Amp/Pos Capacity column (Unadjusted)" column in Attachments Dl and D2 are determined by assuming the "Minutes To End of Section" (discharge time) to be ampere-hours/positive plate (i.e., 95 minutes wo~ld be 95 ampere-hours/positive). Locate the point of intersection for the ampere-hours/positive plate and 60 amperes/positive.

Alternately, you could. locate the point of intersection for twice the ampere-hours/positive plate and 120 amperes/positive. Draw a line between this point of intersection and the 0,0 coordinate. Read the "Amp/Pos Capacity (Unadjusted)" (amperes/positive) at the intersection of this line with the end-voltage curve identified as "End of Discharge Volts/Cell" in the "Input Data" box (i.e., 1.75 or 1.78). The end-voltage curve is determined by dividing the required end voltage ("Min System Volts") by the number of cells in the battery ("No. of Cells"), i.e.,

105 volts/59 cells = 1.78 volts/cell, Nuclear -Calculation.dot I Rev. 1 106-3()'99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 1:&12 E 0118 SubJectiTitle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 4 of 14 To Support Probability Risk Analysis Program (PSA)

If a "Cell Sizing Correction Factor" is applicable to the type cell ("Manufacturer's Type")

and cell characteristic discharge curve drawing ("Discharge Curves") identified in the "Battery Input Data" box, it will be shown in a table of values on the cell characteristic discharge curve drawing. If the discharge period you require is not shown in the table, linearly interpolate the "Cell Sizing Correction Factor" from the discharge periods shown.

The amperes/positive plate for the end of battery discharge voltages 1.75V/aell and 1.78V/cell along with their correction factors, for the interval 1-300 minutes are read from battery discharge curve (Ref.6.6 of calculation SENH 93-024, RevA) and shown in Attachments Dl, and D2 respectively.

4.3 Battery Charger Sizing Worksheet (Attachments B2 and B4)

The Battery charger sizing is done by determining the required charger output current based on the total Ampere-hours removed from the battery as per the load profile and the continuous load.

4.4 Battery Voltage Profile Worksheet (Attachments C1 through C4)

The battery voltage profile worksheet shows the end-of-discharge battery voltage for the battery load profile data and the data shown in the "Input Data" box. All data in the "Input Data" box except for "No. of Cells (NOC)", "Aging Allowance (%)", "Temp Correction Allowance (%), "Design Margin (%)", and "Combined Correction Factor (CCF)" are automatically entered from the battery sizing worksheet (Attachments Bl and B3). Data for "No. of Cells (NOC)", "Aging Allowance (%)", "Temp Correction Allowance (%), and "Design Margin (%)" are manually entered. The "Combined Correction Factor (CCF)" is determined by converting "Aging Allowance (%)", "Temp Correction Allowance (%), and "Design Margin (%)" to decimal values, adding 1 to each, and multiplying the resulting values together.

. . The data in column [1] is entered manually. The data in column [2] and [3] is automatically entered from the Load Duty cycle table of Battery sizing Worksheet (Attachments Bl and B3). The data in column [5], [6], [7], [8], [9], and [11] are calculated as indicated in the column headers. The data in column [4] is entered manually using the same method described for the battery sizing worksheet.

The data in column [10] is determined by locating the intersection of the column [7] and

[9] data on the cell discharge characteristic curve (Ref. 6.6 of calculation SENH 93-024, RevA). The location of this intersection point with respect to the cell end-of-discharge voltage curves determines the cell voltage. If the intersection point does not occur on an existing end-of-discharge voltage curve, a linear interpolation horizontally between the voltage curve to the left and right of the point determines the cell voltage.

Nuclear -Calculation.dot I Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 1R12 E 0118 SubJectITitie Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 5 of 14 To Support Probability Risk Analysis Program (PSA)

5.0 Assumptions

5.1 The RCIC startup occurs four times instead of three times as in the LOCA profile. The RCIC Startup takes place at 1st minute, 75th minute, 151st minute and the 227th minute.

Each time the RCIC Startup take place, it runs for approximately 60 minutes before it is tripped. Each time RCIC is tripped, RCIC re-initiation takes place after 15 minutes. The RCIC started for the final time runs for 72 minutes before it is tripped at the 299th minute.

5.2 The DC EMERGENCY BRG OIL PUMP 2N34-C004 and TURB L.O. EMERGENCY PUMP 2N34-C009 started during the 1st minute are tripped at the end of the 30th minute.

5.3 RCIC suction switchover will not occur any time during the 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> battery duty cycle.

5.4 Manual tripping of 4kV breakers requires about 5 amps momentarily. This is an intermittent load that may occur anytime during the two-hour sequence. Since this load will have a negligible effect on battery capacity, it is not included in the 125V panel loading.

5.5 Operating time for MOVs is taken from the Torque Switch Setting Guide.

5.6 Motor starting current is assumed to reach FLA within five (5) seconds for pump motors and one (1) second for MOVs.

5.7 Where actual data is not available, reduced voltage start pump motors have been considered. These motors are assumed to have a starting current of 400% of full load current (Ref Hatch Electrical Design Criteria).

5.8 It is conservatively assumed that the RCIC Test valve 2E51-F022 is open prior to the accident event and this valve closes on receipt of the RCIC start signal. This valve is not operated again during the 300 minute battery loading cycle.

6.0

References:

6.1 As listed in Section 6.0 of calculation SENH 93-024, Rev.4 Nuclear *Calculation.dotl Rev. 1/06-3()'99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 [8]2 E 0118 SubJectITllle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 6 of 14 To Support Probability Risk Analysis Program (PSA)

Attachments:

A. AI: -Station Service Battery 2A-Load Profile/Sequence. (8 Sheets)

A2: -Station Service Battery 2A-Load Profile graph. (iSheet)

B. Station Service Battery 2A Sizing Worksheets.

BI-B2: - Sizing calculation for 60/120 cells. (6 Sheets)-Normal B3-B4: - Sizing calculation for 59/118 cells. (6 Sheets)- 112 cells removed C. Station Service Battery 2A Voltage Profile Worksheets.

CI - Voltage Profile using 120 cells and design margin calculated from Attachment DI (1 Sheet)

C2 - Voltage Profile using 120 cells with reduced design margin to account for additional connection resistances (1 Sheet).

C3 - Voltage Profile using 118 cells and design margin calculated from Attachment D3 (1 Sheet)

C4 - Voltage Profile using 118 cells with reduced design margin to account for additional connection resistances. (1 Sheet)

D. Dl: -Discharge characteristics for C&D battery, LCR-25 at 1.75V/cell (7 Sheets).

D2: -Discharge characteristics for C&D battery, LCR-25 at 1.78Vlcell (7 Sheets).

E. SNC letter Log No. HL-6252 dated 6/18/02 (2 Sheets).

Nuclear -Calculation. dot I Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Project Calculation Calculation Number Number E. I. Hatch Nuclear Plant - Unit(s): 01 [R)2 E 0118 SubjectlTltle SubjectlTltle Sheet Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 7 of 14 Support Probability Risk Analysis Program JPSA)

To SUPQort (PSA) 7.0 Body of Calculation:

Battery sizing is determined per IEEE 485-1983 (Reference 6.1 of calculation SENH 93-024, Rev.4). Battery terminal voltages during the battery discharge period are calculated based on the battery load profile (Attachment A) and the battery discharge characteristic curves (Reference 6.6) supplied by the battery manufacturer. Battery terminal voltages are used to calculate the minimum available voltage to insure that the end devices will not be prevented from performing their intended safety-related function.

Station Service Battery 2A (2R42-S001A) is a C&D LCR-25 lead-calcium type, 120 cells, 12 positive plates per cell with the center tapped for a 125/250 V DC system. The existing Cyberex battery chargers (2R42-S026, 2R42-S027 and 2R42-S028) are model number 130/400R3-S. They have an input voltage of 550 V AC and an output voltage of 130 V DC. Battery chargers are full-wave silicon-controlled rectifier type, rated at 400A DC with an output voltage regulation of

+/- 0.75% from no-load to 2% and +/-O.5% from 2% load to full-load.

7.1 Load Summaries Refer calculation SENH 93-024, Rev. 4 section 7.1 7.2 Summary of 125 V Panel loads on Buses P-PN and PN-N Refer calculation SENH 93-024, Rev. 4 section 7.2 7.3 Load Data Profile and Step Reduction Attachment Al is a 5-hour load profile developed to support the Probability Risk Analysis (PRA) program.

Attachment A2 is the Station Service Battery Load Profile after step reduction in Graphic form. The step reduction in the Station Service Battery 2A load profile is explained below.

Per Attachment AI, the resulting battery load profile for all load sequence data is as below:

Nuclear-Calculation.dotl Rev. 1/06-30*99 Nuclear-Calculation.dotl Rev. 1/06-30*99

Design Calculations - Nuclear Southem Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 [8]2 E 0118 SubJectITltle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 8 of 14 To Support Probability ProbabiH!}I Risk Analysis Program (PSA)

TABLE-1 Step Time Duration As-Built Profile (minutes) (Amps) 1 0-1 607.99 2 1-2 435.89 3 2-3 285.59 4 3-30 248.09 5 30-60 129.39 6 60-61 262.39 7 61-62 129.39 8 62-75 116.89 9 75-76 332.89 10 76-136 129.39 11 136-137 262.39 12 137 -138 129.39 13 138 -151 116.89 14 151-152 332.89 15 152-212 129.39 16 212-213 262.39 17 213 -214 129.39 18 214-227 116.89 19 227 -228 332.89 20 228-399 129.39 21 299-300 262.39 The above 21-step load profile is reduced to 14 step load profile to accommodate battery test procedure and maintain consistency in the battery load profiles.

In the Table-2 shown below the number of steps in the load duty cycle are reduced to 14.

The Step reduction is explained below:

Step 3 in the above Table-1 is removed by calculating the change in load amps between

=

step 3 and step 4. The change in load amps is 37.5 (285.59 - 248.09 37.5). This load change is evenly spread over the time interval between the start of step 3 and end of step

4. The time interval for the step 3 in Table-2 below is 28 minutes (1 + 27). By adding 1.34 amps (37.5/28 = 1.34) to step 4 in the above Table-1 we get step 3 of the Table-2, below as 249.43 amps (248.09 + 1.34 = 249.43).

Similarly steps 7, 9, 12, 14, 17 and 19 are removed by combining these steps with 8, 10, 13, 15, 18 and 20 respectively.

Nuclear -Calculation.dot! Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 [8]2 E 0118 SubjectITitie Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 9 of 14 To Support Probability Risk Analysis Program (PSA)

TABLE-2 Step Time Duration As-Built Profile (minutes) (Amps) 1 0-1 607.60 2 1-2 435.89 3 2-30 249.43 4 30-60 129.39 5 60-61 262.39 6 61-75 117.78 7 75 -136 132.73 8 136-137 262.39 9 137 -151 117.78 10 151- 212 132.73 11 212-213 262.39 12 213 -227 117.78 13 227 -299 132.22 14 299-300 262.39 For all further battery sizing and voltage profile calculations (Attachment B 1 through B4 and Attachments Cl through C4), the above load profile with 14 steps has been used.

7.4 Battery Charger Sizing Calculation (2R42-S026, 2R42-S027, & 2R42-S028)

The required battery charger output current is determined according to the following formula.

I = L + (1.1 x AH) / T Where: =

I charger current

=

L constant DC load amps AH = Amp-Hours discharged from the battery

=

T Recharge time in hours The constant current DC load amps is determined by the step in the load profile with the lowest discharge current because non-accident (house load) current requirements are considerably less than this step. The constant current DC load amps is 117.78A (steps 6, 9 and 12).

The amp-hours (AH) discharged from the battery is calculated according to the following formula.

Nuclear-Calculation.dotl Rev. 1/06-3()'99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. J. Hatch Nuclear Plant - Unit(s): 01 [g]2 E 0118 SubjectITltie Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 10 of 14 To Support Probability Risk Analysis Program (PSA)

AH=

Where: In = Discharge current in amps for section n tn = Discharge time in minutes for section n n = Discharge section number The AH discharged from the battery is:

607.99 xl + 435.89 xl + 249.43 x 28 + 129.39 x 30 + 262.39 xl +

60 60 60 60 60 117.78 x 14+ 132.73 x 61 + 262.39 xl + 117.78x14 + 132.73 x 61 +

60 60 60 60 60 262.39 xl + 117.78 x 14 + 132.22 x 72 + 262.39 xl + =726.98AH 60 60 60 60 I = 117.78 + (1.1 X 726.98)124 = 151.10 A Design Margin = (1-151.10 /400) X 100 = 62.23%

The existing 400 Amp Battery Chargers are adequate to handle the load and fully charge the battery in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7.5 Battery Sizing Calculations The sizing worksheets shown in Attachments B 1 through B4 determine the size of the battery and the design margin available in the battery per the sizing calculation method described in IEEE-485 (Reference 6.1 of calculation SENH 93-024, Rev.4). Calculations for all the sections on the battery sizing worksheets have been performed to facilitate automatic computations on the worksheet. Attachment B1 and B2 assume all 60/120 cells are included in the battery. Since it may become necessary to temporarily remove/replace cells for maintenance, battery sizing worksheets in Attachments B3 and B4 were prepared with 591118 cells, to determine if adequate margins would be maintained. However, these sizing worksheets do not include the affect of additional connection resistance (Reference 6.11).

This additional connection resistance must be included to provide assurance that the end of discharge voltage will meet the required minimum battery terminal voltage of 1051210 volts (Criteria 3.2). The affect of additional connection resistance is addressed in the voltage profile worksheets (Attachments C2 and C4).

7.6 Battery Connection Resistance Considerations For each of the Class IE batteries, calculation SENH 94-021 (Reference 6.11) presents a tabulation of resistances for the various types of battery connections, the total voltage drop Nuclear *Calculation.dot I Rev. 1/06-30-99

Design Calculations

  • Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 [8]2 E 0118 SubjecVTitle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 11 of 14 To Support Probability Risk Analysis Program (PSA) due to these resistances, and the average volts-per-cell adjustment required to compensate for the additional voltage drop. For Station Service Battery 2A, SENH 94-021 shows that the actual minimum average volts-per-cell should be 1.756 volts with 120 cells in the battery. This voltage is determined as follows:

_ Is X RA VA - Equation 2 Nc Where: VA =additional voltage drop per cell Is = current for step that sizes the battery

=

RA additional connection resistance Nc = number of cells VN = Vo + VA Equation 3 Where: VN =new average cell voltage Vo = old average cell voltage from VP#l VA = additional voltage drop per cell From Equation 2

= 262.39 amps x 2706 j£l =.0 0059 voIts/ce11 "".0 006 voIts/ce11 120 cells From Equation 3 VN = 1.750 + 0.006 = 1.756 volts/cell This same methodology can be used to determine the actual minimum volts-per-cell with one cell removed from each battery bank. However, an additional step is required to account for having only 118 rather than 120 cells in each case respectively.

Case 1: when one cell is removed from each battery bank (59/118)

Using the same step and resistance, the additional voltage drop per cell is:

From Equation 2 x 2706 flO - 0 006 1 / 11 VA = 262.39 amps 118 cells

-. vo ts ce Nuclear -Calculation.dot I Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): Unites): 01 [&]2 E 0118 SubJect/Tille Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 12 of 14 Sup~ort Probability Risk Analysis Program (PSA)

To Support For a 120-cell battery, the minimum average cell voltage would be 1.75 volts. For a 118-cell battery, the minimum average volts-per-cell would be 1.78. This is a difference of 0.030 volts/cell. Based on the calculated minimum average cell voltage of 1.75 for a 120-cell battery in VP#1 (Attachments Cl), the minimum average cell voltage for a l18-cell battery would be 1.75 plus an adjustment for the minimum average volts-per-cell difference (0.030) and the voltage drop due to additional resistance (0.0060).

The new required minimum cell voltage for a lIS-cell battery is:

1.750+ 0.030 + 0.006 =1.786 volts/cell 7.7 Voltage Profile Worksheets Generally, voltage profiles are prepared by calculating the discharged amperes per positive plate during each step and the cumulative discharged ampere-hours per positive plate at the end of each step of the load profile based on the indicated battery sizing parameters and the load profile used in sizing the battery (Attachments Al and A2). The amperes per positive plate and cumulative ampere-hours per positive plate for the load profile step that determines the battery size is plotted on the battery discharge characteristic curve for the C&D type LCR-25 cell (Reference 6.6 of calculation SENH 93-024, RevA) to determine the minimum average per-cell voltage. The minimum battery voltage is determined by the minimum average per-cell voltage and the number of cells in the battery.

Note: Only the battery and minimum average per-cell voltages for profile step 14 will be Calculated in the voltage profiles. This is the profile step that determines the size of the battery The design margin calculated in Attachment Bl is used in voltage profile #1 (Attachment Cl) to determine the minimum average per-cell voltage that satisfies the minimum battery voltage requirements (Criteria 3.2) excluding the affect of additional connection resistance. The battery sizing design parameters (aging factor of 25%, temperature correction factor of 8%, and the design margin) are used to calculate the battery terminal voltage. The minimum required and calculated terminal voltage for station service battery 2A is 105/210 V.

Voltage profile #1 (Attachment Cl) establishes the base minimum average per-cell voltage used to calculate the per-cell voltage in voltage profile #2 (Attachment C2) that includes the voltage drop due to additional connection resistance. Calculation SENH 94-021 identifies the per-cell voltage that is required to compensate for the voltage drop due to additional connection resistance. The difference between this voltage and the minimum average designed per-cell voltage of 1.75 volts in sizing worksheet #1 (Attachment Bl) is added to the per-cell voltage in step 14 of voltage profile #1 (Attachment Cl). The Nuclear -Calculation.dot I Rev. 1 106-30-99

Design Calculations - Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unites}: 01 [8]2 E 0118 SubjectfTltle Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 13 of 14 To Support Probability Risk Analysis Program (PSA) resulting voltage of 1.756 volts is the minimum per-cell voltage limit for voltage profile

  1. 2 (Attachment C2) that determines the minimum battery voltage due to the affect of additional connection resistance. By trial and error, a design margin less than that shown in sizing worksheet #1 (Attachment Bl) is selected and the resulting discharged amperes per positive plate and cumulative ampere-hours per positive plate are plotted to determine a per-cell voltage. This trial and error process is repeated until the per-cell voltage is equal to or slightly greater than the minimum per-cell voltage limit of 1.756 volts. As shown in voltage profile #2 (Attachment C2), the design margin will be reduced.

Additionally, voltage profiles #3 and #4 (Attachments C3 and C4) show that battery is capable of meeting required load and voltage requirements with one (591118) cell removed from each 601120 cell bank.

These voltage profiles show that including the additional battery connection resistance; the battery has adequate capacity with an aging factor of 25%, a temperature correction factor of 8% and reduced design margin in the case of one cell removed. The minimum required terminal voltage for station service battery 2A is 105/210 V. However, if Attachment Cl calculates the end voltage to be less than 210.0 volts. The possible reason for this difference is described in the next section.

7.8 Calculation Anomalies A cell type and size is selected after a sizing calculation is performed using the application design requirements. For various reasons, the final cell selection usually exceeds the application design requirements and has some additional capacity. If this additional capacity is factored into the sizing calculation, all of the available plates in the cell are required. This will produce the worst case (designed minimum) end-of-discharge cell voltage for the load profile step that determines the cell size. This voltage will be equal to the end-of-discharge volts-per-cell used in the sizing calculation, e.g., 1.75 volts-per-cell for a 120 cell battery with an end-of-discharge voltage of 210 volts.

Therefore, a voltage profile calculation for the above case should verify the designed end-of-discharge volts-per-cell at the end of the load profile step that determines the cell size.

The design parameters in the battery sizing worksheet #1 (Attachment Bl) should produce a worst case end-of-discharge volts-per-cell of 1.75 volts as shown in voltage profile #1 (Attachment Cl). Similarly design parameters in worksheet #2 (Attachment B3) with one cell removed (59/118) should produce a worst case end of discharge volts-per-cell of 1.78 volts as shown in voltage profiles #3 (Attachment C3). However, occasionally these voltages do not match the expected 1.75 and 1.78 volts-per-cell in each case respectively.

These potential voltage discrepancies are assumed to be associated with the method in which the battery discharge characteristic curves are created. The basis for this assumption is as follows. Known data points for discharge times of one minute and 2 Nuclear *Calculation.dot I Rev. 1 106-30-99

Design Calculations

  • Nuclear Southern Company Services, Inc.

Project Calculation Number E. I. Hatch Nuclear Plant - Unit(s): 01 002 E 0118 SubJectITitie Sheet Station Service Battery 2A Capacity for 5 Hour Duration, 14 of 14 To Support Probability Risk Analysis Program (PSA) hours or greater are used to generate the discharge characteristic curves. Additionally, battery perfonnance tests verify the discharge characteristic curve data for the 2-hour and greater regions of the curves. However, the calculated data points from the voltage profile calculation fall into the region of the characteristic curves, which are not commonly tested values. Curve-smoothing techniques used for this region may account for the discrepancy.

It is important to note that the accuracy of the sizing methodology used is not in question and the voltage profile calculations yield conservative results. However, the battery design parameters used in the sizing calculations will be adjusted in the voltage profile calculations such that the results will be not produce voltages that are less than the required end-of discharge voltage limit.

Nuclear -Calculation.dotl Rev. 1 106-30-99

5 HOUR BATTERY 2A LOAD PROFILE/SEQUENCE AnACHMENT A1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 1 of 8 TIME 2831* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2E51* 2N34- 2N34- 2R25* 2R25* 2R25* 2R25- TOTAL Profile C005A C002*1 C002*2 FOO8 F010 F013 F019 F022 F029 F031 F045 F046 F524 C004 C009 S001 S003 S077 S129 AMPS Step Ot01see 53.09 34.96 28.57 16.57 133.19 1 to 2 see 53.09 34.96 28.57 16.57 133.19 2 to 3 see 53.09 34.96 28.57 16.57 133.19 3t04see 364.40 110.40 53.09 34.96 28.57 16.57 607.99 STP#1 4t05see 364.40 110.40 53.09 34.96 28.57 16.57 607.99 5 106 see 364.40 110.40 37.09 18.96 28.57 16.57 575.99 6 to 7 see 364.40 110.40 37.09 18.96 28.57 16.57 575.99 7108see 364.40 110.40 37.09 18.96 28.57 16.57 575.99

  • 8t09see 91.10 27.60 37.09 18.96 28.57 16.57 219.89 9 to 10 see 14.10 91.10 27.60 37.09 18.96 28.57 16.57 233.99 10t011see 14.10 91.10 27.60 37.09 18.96 28.57 16.57 233.99 11 to 12 see 14.10 91.10 27.60 37.09 18.96 28.57 16.57 233.99 12 to 13 see 14.10 91.10 27.60 37.09 18.96 28.57 16.57 233.99 13 to 14 see 14.10 91.10 27.60 37.09 18.96 28.57 16.57 233.99 14 to 15 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 15 to 16 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 16 to 17 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 17 to 18 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 18 to 19 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 19 to 20 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 20 to 21 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 21 to 22 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 22 to 23 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 23 to 24 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 24 to 25 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 25 to 26 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 261027 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 27 to 28 see 4.70 91.10 27.60 37.09 1S.96 28.57 16.57 224.59 28 to 29 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 29 to 30 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 30 t031 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 31 to 32 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 32 to 33 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 33 to 34 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 34 to 35 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 35 to 36 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 36 to 37 see 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 37 to 38 see 4.70 ---_ ..... _-

91.10 27.60 37.09 18.96 28.57 16.57 224.59

5 HOUR BATTERY 2A LOAD PROFILE/SEQUENCE ATTACHMENT A1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 2 of 8 TIME 2831* 2E51* 2E51- 2E51* 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2N34- 2N34- 2R25- 2R25* 2R25- 2R25* TOTAL Profile COO5A COO2-1 COO2-2 F008 F010 F013 F019 F022 F029 F03l F045 F046 F524 COO4 Coo9 SOOl SOO3 S077 S129 AMPS Step 38 to 39 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 39 to 40 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 40 to 41 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 41 to 42 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 42 to 43 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 43 to 44 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 44 to 45 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 45 to 46 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 46 to 47 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 47 to 48 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 48 to 49 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 49 to 50 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 50 to 51 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 51 to 52 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 52 to 53 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 53 to 54 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 54 to 55 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 55 to 56 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 56 to 57 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 57 to 58 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 58 to 59 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 59 to 60 sec 4.70 91.10 27.60 37.09 18.96 28.57 16.57 224.59 1 to 2 min 4.70 44.00 29.40 42.50 46.50 29.40 19.50 91.10 27.60 37.09 18.96 28.57 16.57 435.89 STP#2 2t03min 4.70 50.00 11.00 91.10 27.60 37.09 18.96 28.57 16.57 285.59 STP#3 3t04 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 STP#4 4 t05 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 5 t06 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 6t07min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 7t08min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 8t09min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 9 to 10 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 10 to 11 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 11 to 12 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 12 to 13 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 13 to 14 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 14 to 15 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 15 to 16 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09

5 HOUR BAITERY 2A LOAD PROFILE/SEQUENCE ATTACHMENT A1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 3 of 8 TIME 2831* 2E51* 2E51- 2E51- 2E51- 2E51- 2E51* 2E51* 2E51- 2E51- 2E51- 2E51* 2E51- 2N34- 2N34- 2R25- 2R25- 2R25- 2R25- TOTAL Profile Coo5A C002-1 C002-2 FOOS F010 F013 F019 F022 F029 F031 F045 F046 F524 C004 C009 S001 S003 son S129 AMPS Step SteD 1610 17 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 1710 18 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 1810 19 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 191020 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 201021 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 211022 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 221023 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 231024 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 "24 1025 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 25 to 26 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 26 1027 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 27 to 28 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 28 to 29 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 29 1030 min 4.70 12.50 11.00 91.10 27.60 37.09 18.96 28.57 16.57 248.09 30t031 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#5 31 to 32 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 32 1033 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 33 1034 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 34 to 35 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 35 to 36 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 36 to 37 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 371038min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 38 to 39 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 39 to 40 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 401041min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 41 10 42 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 421043 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 431044 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 441045 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 451046 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 46 to 47 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 471048 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 48 to 49 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 49 1050 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 501051 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 51 to 52 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 521053 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 531054 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39

5 HOUR BATIERY 2A LOAD PROFILE/SEQUENCE ATTACHMENT A1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 4 of 8 TIME 2831- 2ES1- 2ES1- 2E51- 261- 2E51- 2E51- 261- 2E51- 261- 2E51- 2E51- 261- 2N34- 2N34- 2R25- 2R25- 2R25- 2R25- TOTAL Profile C005A C002-1 C002-2 F008 F010 F013 F019 F022 F029 F031 F045 F046 F524 C004 C009 SOOl S003 son S129 AMPS Ste~

Step 54 to 55 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 55 to 56 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 56 to 57 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 57 to 58 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 58 to 59 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 59 to 60 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 60 to 61 min 4.70 12.50 11.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 262.39 STP#6 I 61 to 62 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#7 62 to 63 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 STP#8 63 to 64 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 64 to 65 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 65 to 66 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 66 to 67 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 67 to 68 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 68 to 69 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 69 to 70 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 70t071 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 71 to 72 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 72 to 73 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 73 to 74 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 74 to 75 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 75 to 76 min 4.70 50.00 44.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 332.89 STP#9 76 to 77 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#10 77 1078 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 78 to 79 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 79 to 80 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 801081 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 81 to 82 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 82 to 83 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 83 to 84 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 84 to 85 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 85 to 86 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 86t087 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 87 to 88 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 88 to 89 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 89 to 90 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 90 t091 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 91 to 92 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57. 129.39

5 HOUR BATTERY 2A LOAD PROFILE/SEQUENCE AITACHMENT A1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 5 of 8 TIME 2831- 2E51- 2E51- 2E51* 2E51- 2E51- 2E51* 2E51* 2E51* 2E51- 2E51- 2E51- 2E51- 2N34- 2N34- 2R25- 2R25* 2R25- 2R25- TOTAL Profile COO5A COO2-1 COO2*2 FOO8 F010 F013 F019 F022 F029 F031 F045 F046 F524 C004 COO9 SOO1 SOO3 S077 S129 AMPS StE!~

Step 92 1093 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 93 to 94 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 94 to 95 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 95 to 96 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 96 to 97 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 97 to 98min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 98 to 99 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 99 to 100 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 ioo to 101min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 101 to 102 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 102 to 103 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 103 to 104 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 104 to 105 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 105 to 106 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 106 to 107 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 107 to 108 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 108 to 109 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 109 to 110 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 110 to 111 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 111 to 112 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 112 to 113 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 113 to 114 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 114 to 115 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 115to 116 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 116 to 117 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 117 to 118 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 118 to 119 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 119to120min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 120 to 121 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 121 to 122 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 122 to 123 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 12310124 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 124 to 125 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 125 to 126 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 126 to 127 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 127 to 128 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 128 to 129 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 129 to 130 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39

5 HOUR BATIERY 2A LOAD PROFILE/SEQUENCE AITACHMENT A1 TO CALC. E 0118, REV. 0 TO SUPPORT PROBABILITY RISK ANALYSIS Page 6 of 8 TIME 2831- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2N34- 2N34- 2R25- 2R25- 2R25- 2R25- TOTAL Profile i COO5A COO2-1 COO2-2 FOO8 F010 F013 F019 F022 F029 F031 F045 F046 F524 COO4 C009 S001 S003 son S129 AMPS Step 130 to 131 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 131 to 132 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 132 to 133 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 133 to 134 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 13410135 min 4.70 12.50 11.00 3'1.09 18.96 28.57 16.57 129.39 135 t0136 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 13610137 min 4.70 12.50 11.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 262.39 STP#11 137 t0138 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#12 138 t0139 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 STP#13 13910140 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14010141 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89  !

141 t0142 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14210143 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 143 t0144 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 144 toMS min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14510146 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14610147 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14710148 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14810149 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 14910150 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 15010151 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 15110152 min 4.70 50.00 44.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 332.89 STP#14 15210153 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#15 15310154 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 15410155 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 15510156 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 15610157 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 15710158 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 158 t0159 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 15910160 min 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16010161 min 4.70 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 161 10162 min 162 t0163 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16310164 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16410165 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16510166 min 16610167 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 37.09 18.96 28.57 16.57 129.39 167 10168 min 4.70 12.50 11.00

5 HOUR BATTERY 2A LOAD PROFILE/SEQUENCE ATTACHMENT A 1 TO SUPPORT PROBABILITY RISK ANALYSIS TO CALC. E 0118, REV. 0 Page 7 of 8 TIME 2831- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2N34- 2N34- 2R25- 2R25- 2R25- 2R25- TOTAL Profile C005A C002-1 C002*2 F008 F010 F013 F019 F022 F029 F031 F045 F046 F524 C004 C009 8001 S003 S077 S129 AMPS Steo 16010161 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16810169 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 16910170 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17010171 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17110172 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17210173 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17310174 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17410175 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17510176 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 176 to177 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17710178 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 17810179 min 4.70 12.50 11.00

  • 37.09 18.96 28.57 16.57 129.39 179 t0180 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18010181 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 181 10182 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18210183 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18310184 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 184 to185 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18510186 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18610187 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 18710188 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 188 to189 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 I 18910190 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 I 19010191 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 I 191 10192 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 192 to193 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 I 19310194 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 I 19410195 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 195 to196 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 196 10197 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 19710198 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 19810199 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 199 to 200 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 20010201 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 201 to 202 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 202 10 203 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 203 to 204 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 204 to 205 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39

5 HOUR BATTERY 2A LOAD PROFILE/SEQUENCE ATTACHMENT A1 TO CALC. E 0118, REV. 0 TO SUPPORT PROBABILITY RISK ANALYSIS Page 8 of 8 TIME 2831- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2E51- 2N34- 2N34- 2R25- 2R25* 2R25- 2R25- TOTAL Profile COO5A C002-1 C002-2 FOOS F010 F013 F019 F022 F029 F031 F045 F046 F524 COO4 C009 S001 S003 son S129 AMPS Step 205 to 206 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39

. 206 to 207 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 207 to 208 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 208 to 209 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 209 to 210 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 210 to 211 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 211 to 212 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 212 to 213 min 4.70 12.50 11.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 262.39 STP#16 213t0214 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#17 214 t0215 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 STP#18 215 t0216 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 216 to 217 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 217 to 218 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 218 to 219 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 219 to 220 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 220 to 221 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 221 to 222 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 222 to 223 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 223 to 224 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 224 to 225 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 225 to 226 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 226 to 227 min 4.70 11.00 37.09 18.96 28.57 16.57 116.89 227 to 228 min 4.70 50.00 44.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 332.89 STP#19 228 to 299 min 4.70 12.50 11.00 37.09 18.96 28.57 16.57 129.39 STP#20 299 to 300 min 4.70 12.50 11.00 29.40 42.50 29.40 19.50 12.20 37.09 18.96 28.57 16.57 262.39 STP#21

5-HOUR LOAD PROFILE FOR BATTERY 2A Attachment A2 To Calc. E0118, Rev. 0 CALCULATION EOl18, REV. 0 Page 1 of 1 1200 1100 I 1000 900 800 700 (f) 0... 607. ~9 607.99 0... 600

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ATTACHMENT B1 TO CALC. E 0118, REV. 0 Page 1 of 1 BATTERY SIZING WORKSHEET INPUT DATA Plant/Substation  : E. I. Hatch Nuclear Plant Unit No. :2 Battery Name :Station Service Battery 2A Equipment No. :2R42-S001 A Nom. System Volts  : 125/250 V End of Discharge V/Cell  : 1.75 V Min. System Volts  : 105/210 V Min Design Temp.  : 65 Deg.F Max. System Volts  : 140/280 V Design Margin 46.38  %

No. of Cells  : 60/120 Reqd. Charge Time :24 Hours Battery Charger Rating  : 400 A LOAD DUTY CYCLE BATTERY INPUT DATA Step Load Duration Manufacturers Name :C&D Amps Minutes Manufacturers Type  : LCR-25 1 607.99 1 Battery Type  : Calcium 2 435.89 1 Discharge Curves  : D-841 S-53155 Rev. 1 3 249.43 28 Aging Allowance :25%

4 129.39 30 No. of Plates/Cell  : 25 5 262.39 1 No. of Pas. Plates Available :12 6 117.78 14 7 132.73 61 8 262.39 1 OUTPUT RESULTS 9 117.78 14 10 132.73 61 No. of Pas. Plates Available: 12 11 262.39 1 No. of Pas. Plates Required: 12 12 117.78 14 Calculated Amp-Hr 13 132.22 72 Discharged 726.98 14 262.39 1 Temperature Allowance 8.00  %

300 Aging Allowance 25.00  %

Design Margin 46.38  %

Continuous Amps 117.78 A Additional Margin 0  %

AH Discharged from 726.98 AH Calculated Charger Amps 151.10 A the battery Charger Design Margin 62.23  %

COMMENTS:

1.DetelTT'line design margin that requires all of the plates available ( no additional margin).

2.Step 14 detelTT'lines the size of the battery.

3.Additional margin for added battery connection resistance is not included.

AITACHMENT 82 TO CALC. E 0118, REV. 0 Page 1 of 5 BATTERY SIZING WORKSHEET PlanVSubstation  : E. I. Hatch Nuclear Plant Manufacturers Name :C&D Unit No. :2 Manufacturers Type  : LCR-25 Battery Name :Station Service Battery 2A No. of Cells  : 60/120 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 1 FIRST PERIOD ONLY IF SECTION 2 > SECTION 1, GO TO SECTION 2 1 607.990 607.990 1 140.000 1.071 149.940 4.055 Section 1 Total = 4.055 SECTION 2 PERIODs 1 - 2 ONLY IF SECTION 3 > SECTION 2, GO TO SECTION 3 1 607.990 607.990 1 2 138.200 1.067 147.459 4.123 2 435.890 -172.100 1 1 140.000 1.071 149.940 -1.148 Section 2 Total = 2.975 SECTION 3 PERIODs 1 - 3 ONLY IF SECTION 4 > SECTION 3, GO TO SECTION 4 1 607.990 607.990 1 30 96.500 1.052 101.518 5.989 2 435.890 -172.100 1 29 97.600 1.052 102.695 -1.676 3 249.430 -186.460 28 28 98.700 1.052 103.872 -1.795 Section 3 Total = 2.518 SECTION 4 PERIODs 1 - 4 ONLY IF SECTION 5 > SECTION 4, GO TO SECTION 5 1 607.990 607.990 1 60 73.300 1.040 76.232 7.976 2 435.890 -172.100 1 59 73.940 1.040 76.927 -2.237 3 249.430 -186.460 28 58 74.580 1.041 77.623 -2.402 4 129.390 -120.040 30 30 96.500 1.052 101.518 -1.182 Section 4 Total = 2.154

ATTACHMENT 82 TO CALC. E 0118, REV. 0 Page 2 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 5 PERIODs 1 - 5 ONLY IF SECTION 6> SECTION 5, GO TO SECTION 6 1 607.990 607.990 1 61 72.840 1.040 75.760 8.025 2 435.890 -172.100 1 60 73.300 1.040 76.232 -2.258 3 249.430 -186.460 28 59 73.940 1.040 76.927 -2.424 4 129.390 -120.040 30 31 95.500 1.052 100.428 -1.195 5 262.390 133.000 1 1 140.000 1.071 149.940 0.887 Section 5 Total = 3.036 SECTION 6 PERIODs 1 - 6 ONLY IF SECTION 7 > SECTION 6, GO TO SECTION 7 1 607.990 607.990 1 75 66.500 1.041 69.243 8.781 2 435.890 -172.100 1 74 66.900 1.041 69.654 -2.471 3 249.430 -186.460 28 73 67.300 1.041 70.065 -2.661 4 129.390 -120.040 30 45 83.000 1.046 86.818 -1.383 5 262.390 133.000 1 15 115.000 1.055 121.325 1.096 6 117.780 -144.610 14 14 116.800 1.055 123.224 -1.174 Section 6 Total = 2.188 SECTION 7 PERIODs 1 - 7 ONLY IF SECTION 8> SECTION 7, GO TO SECTION 8 1 607.990 607.990 1 136 46.867 1.046 49.038 12.398 2 435.890 -172.100 1 135 47.050 1.046 49.226 -3.496 3 249.430 -186.460 28 134 47.233 1.046 49.414 -3.773 4 129.390 -120.040 30 106 54.200 1.044 56.576 -2.122 5 262.390 133.000 1 76 66.000 1.041 68.728 1.935 6 117.780 -144.610 14 75 66.500 1.041 69.243 -2.088 7 132.730' 14.950 61 61 72.840 1.040 75.760 0.197 Section 7 Total = 3.051 SECTION 8 PERIODs 1 - 80NLY IF SECTION 9 > SECTION 8, GO TO SECTION 9 1 607.990 607.990 1 137 46.683 1.046 48.850 12.446 2 435.890 -172.100 1 136 46.867 1.046 49.038 -3.510 3 249.430 -186.460 28 135 47.050 1.046 49.226 -3.788 4 129.390 -120.040 30 107 53.900 1.044 56.267 -2.133 5 262.390 133.000 1 77 65.500 1.041 68.213 1.950 6 117.780 -144.610 14 76 66.000 1.041 68.728 -2.104 7 132.730 14.950 61 62 72.380 1.040 75.287 0.199 8 262.390 129.660 1 1 140.000 1.071 149.940 0.865 Section 8 Total = 3.924

ATTACHMENT B2 TO CALC. E 0118, REV. 0 Page 3 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 9 PERIODS 1

  • 90NLY IF SECTION 10> SECTION 9, GO TO SECTION 10 1 607.990 607.990 151 43.800 1.048 45.884 13.251 2 435.890 -172.100 1 150 44.000 1.048 46.090 -3.734 3 249.430 -186.460 28 149 44.200 1.047 46.296 *4.028 4 129.390 -120.040 30 121 49.833 1.045 52.080 -2.305 5 262.390 133.000 1 91 58.760 1.043 61.262 2.171 6 117.780 -144.610 14 90 59.000 1.043 61.508 -2.351 7 132.730 14.950 61 76 66.000 1.041 68.728 0.218 8 262.390 129.660 1 15 115.000 1.055 121.325 1.069 9 117.780 -144.610 14 14 116.800 1.055 123.224 -1.174 Section 9 Total = 3.117 SECTION 10 PERIODS 1 *10 ONLY IF SECTION 11 > SECTION 10, GO TO SECTION 11 1 607.990 607.990 1 212 34.760 1.048 36.442 16.684 2 435.890 -172.100 1 211 34.880 1.048 36.570 -4.706 3 249.430 -186.460 28 210 35.000 1.049 36.698 -5.081 4 129.390 -120.040 30 182 38.000 1.050 39.896 -3.009 5 262.390 133.000 1 152 43.600 1.048 45.678 2.912 6 117.780 -144.610 14 151 43.800 1.048 45.884 -3.152 7 132.730 14.950 61 137 46.683 1.046 48.850 0.306 8 262.390 129.660 1 76 66.000 1.041 68.728 1.887 9 117.780 -144.610 14 75 66.500 1.041 69.243 -2.088 10 132.730 14.950 61 61 72.840 1.040 75.760 0.197 Section 10 Total = 3.949 SECTION 11 PERIODS 1 *11 ONLY IF SECTION 12 > SECTION 11, GO TO SECTION 12 1 607.990 607.990 1 213 34.640 1.048 36.315 16.742 2 435.890 -172.100 1 212 34.760 1.048 36.442 -4.723 3 249.430 -186.460 28 211 34.880 1.048 36.570 -5.099 4 129.390 -120.040 30 183 37.900 1.050 39.789 -3.017 5 262.390 133.000 1 153 43.400 1.048 45.472 2.925 6 117.780 -144.610 14 152 43.600 1.048 45.678 -3.166 7 132.730 14.950 61 138 46.500 1.047 48.662 0.307 8 262.390 129.660 1 77 65.500 1.041 68.213 1.901 9 117.780 -144.610 14 76 66.000 1.041 68.728 -2.104 10 132.730 14.950 61 62 72.380 1.040 75.287 0.199 11 262.390 129.660 1 1 140.000 1.071 149.940 0.865 Section 11 Total = 4.830

ATIACHMENT B2 TO CALC. E 0118, REV. 0 Page 4 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 12 PERIODS 1 -12 ONLY IF SECTION 13> SECTION 12, GO TO SECTION 13 1 607.990 607.990 1 227 32.890 1.048 34.457 17.645 2 435.890 -172.100 1 226 33.020 1.048 34.595 -4.975 3 249.430 -186.460 28 225 33.150 1.048 34.733 -5.368 4 129.390 -120.040 30 197 36.500 1.049 38.294 -3.135 5 262.390 133.000 1 167 40.667 1.049 42.656 3.118 6 117.780 -144.610 14 166 40.833 1.049 42.827 -3.377 7 132.730 14.950 61 152 43.600 1.048 45.678 0.327 8 262.390 129.660 1 91 58.760 1.043 61.262 2.116 9 117.780 -144.610 14 90 59.000 1.043 61.508 -2.351 10 132.730 14.950 61 76 66.000 1.041 68.728 0.218 11 262.390 129.660 1 15 36.800 1.049 38.614 3.358 12 117.780 -144.610 14 14 38.100 1.050 40.003 -3.615 Section 12 Total = 3.961 SECTION 13 PERIODS 1 -13 ONLY IF SECTION 14> SECTION 13, GO TO SECTION 14 1 607.990 607.990 1 299 26.187 1.0441 27.340 22.238 2 435.890 -172.100 1 298 26.273 1.0441 27.432 -6.274 3 249.430 -186.460 28 297 26.360 1.0442 27.524 -6.775 4 129.390 -120.040 30 269 28.193 1.0456 29.478 -4.072 5 262.390 133.000 1 239 31.330 1.0471 32.804 4.054 6 117.780 -144.610 14 238 31.460 1.0471 32.942 -4.390 7 132.730 14.950 61 224 33.280 1.0478 34.871 0.429 8 262.390 129.660 1 163 41.333 1.0486 43.341 2.992 9 117.780 -144.610 14 162 41.500 1.0485 43.513 -3.323 10 132.730 14.950 61 148 44.400 1.0473 46.502 0.321 11 262.390 129.660 1 87 60.500 1.0423 63.056 2.056 12 117.780 -144.610 14 86 61.000 1.0422 63.572 -2.275 13 132.220 14.440 72 72 67.700 1.0410 70.476 0.205 Section 13 Total = 5.187

ATTACHMENT B2 TO CALC. E 0118, REV. 0 Page 5 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 14 PERIODS 1 -14 ONLY IF SECTION 15> SECTION 14, GO TO SECTION 15 1 607.990 607.990 1 300 26.100 1.0440 27.248 22.313 2 435.890 -172.100 1 299 26.187 1.0441 27.340 -6.295 3 249.430 -186.460 28 298 26.273 1.0441 27.432 -6.797 4 129.390 -120.040 30 270 28.100 1.0455 29.379 -4.086 5 262.390 133.000 1 240 31.200 1.0470 32.666 4.071 6 117.780 -144.610 14 239 31.330 1.0471 32.804 -4.408 7 132.730 14.950 61 225 33.150 1.0478 34.733 0.430 8 262.390 129.660 1 164 41.167 1.0487 43.170 3.003 9 117.780 -144.610 14 163 41.333 1.0486 43.341 -3.337 10 132.730 14.950 61 149 44.200 1.0474 46.296 0.323 11 262.390 129.660 1 88 60.000 1.0423 62.540 2.073 12 117.780 -144.610 14 87 60.500 1.0423 63.056 -2.293 13 132.220 14.440 72 73 67.300 1.0411 70.065 0.206 14 262.390 130.170 1 1 140.000 1.0710 149.940 0.868 Section 14 Total = 6.073 Calculated Cell Size Maximum Temperature Required #

Section Random Uncorrected Correction Aging Design Positive Size Size Size Factor Factor Margin Plates 6.073 0.000 6.073 1.080 1.250 1.464 12 Required Battery Charger DC Output Amps Continuous Amps + (AH Discharged

  • 1.1) / Battery Recharge Time Charger Output DC Amps = 151.10 A

AITACHMENT 83 TO CALC. E0118, REV. 0 Page 1 of 1 BATIERY SIZING WORKSHEET INPUT DATA Plant/Substation  : E. I. Hatch Nuclear Plant Unit No. :2 Battery Name :Station Service Battery 2A Equipment No. :2R42-S001 A Nom. System Volts  : 125/250 V End of Discharge VlCell  : 1.78 V Min. System Volts  : 105/210 V Min Design Temp  : 65 Deg.F Max. System Volts  : 140/280 V Design Margin 39.85  %

No. of Cells  : 59/118 Reqd. Charge Time  : 24 Hours Battery Charger Rating  : 400 A LOAD DUTY CYCLE BATTERY INPUT DATA Step Load Duration Manufacturers Name :C&D Amps Minutes Manufacturers Type  : LCR-25 1 607.99 1 Battery Type  : Calcium 2 435.89 1 Discharge Curves  : D-841 S-53155 Rev. 1 3 249.43 28 Aging Allowance :25%

4 129.39 30 No. of Plates/Cell  : 25 5 262.39 1 No. of Pos. Plates Available  : 12 6 117.78 14 7 132.73 61 8 262.39 1 OUTPUT RESULTS 9 117.78 14 10 132.73 61 No. of Pos. Plates Available: 12 11 262.39 1 No. of Pos. Plates Required: 12 12 117.78 14 Calculated Amp-Hr 13 132.22 72 Discharged 726.98 14 262.39 1 Temperature Allowance 8.00  %

300 Aging Allowance 25.00  %

Design Margin 39.85  %

Continuous Amps 117.78 A Additional Margin a  %

AH Discharged from 726.98 AH Calculated Charger Amps 151.10 A the battery Charger Design Margin 62.23  %

COMMENTS:

1.Determine design margin that requires all of the plates available ( no additional margin).

2.Step 14 determines the size of the battery.

3.Additlonal margin for added battery connection resistance is not included.

ATTACHMENT B4 TO CALC. E0118, REV. 0 Page 1 of 5 BATTERY SIZING WORKSHEET PlanVSubstation  : E. I. Hatch Nuclear Plant Manufacturers Name :C&D Unit No. :2 Manufacturers Type  : LCR-25 Battery Name :Station Service Battery 2A No. of Cells  : 59/118 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 1 FIRST PERIOD ONLY IF SECTION 2 > SECTION 1, GO TO SECTION 2 1 607.990 607.990 1 1 121.000 1.071 129.591 4.692 Section 1 Total = 4.692 SECTION 2 PERIODs 1 - 2 ONLY IF SECTION 3 > SECTION 2, GO TO SECTION 3 1 607.990 607.990 1 2 120.000 1.067 128.040 4.748 2 435.890 -172.100 1 1 121.000 1.071 129.591 -1.328 Section 2 Total = 3.420 SECTION 3 PERIODs 1 *3 ONLY IF SECTION 4 > SECTION 3, GO TO SECTION 4 1 607.990 607.990 1 30 88.000 1.052 92.576 6.567 2 435.890 -172.100 1 29 89.000 1.052 93.646 -1.838 3 249.430 -186.460 28 28 90.000 1.052 94.716 -1.969 Section 3 Total = 2.761 SECTION 4 PERIODs 1 *4 ONLY IF SECTION 5 > SECTION 4, GO TO SECTION 5 1 607.990 607.990 1 60 68.200 1.040 70.928 8.572 2 435.890 -172.100 1 59 68.800 1.040 71.580 -2.404 3 249.430 -186.460 28 58 69.400 1.041 72.232 -2.581 4 129.390 -120.040 30 30 88.000 1.052 92.576 -1.297 Section 4 Total = 2.290

ATTACHMENT B4 TO CALC. E0118, REV. 0 Page 2 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTIONS PERIODs 1

  • 5 ONLY IF SECTION 6 > SECTION 5, GO TO SECTION 6 1 607.990 607.990 1 61 67.800 1.040 70.518 8.622 2 435.890 -172.100 1 60 68.200 1.040 70.928 -2.426 3 249.430 -186.460 28 59 68.800 1.040 71.580 -2.605 4 129.390 -120.040 30 31 87.240 1.052 91.742 -1.308 5 262.390 133.000 1 1 121.000 1.071 129.591 1.026 Section 5 Total = 3.308 SECTION 6 PERIODs 1 - 6 ONLY IF SECTION 7 > SECTION 6, GO TO SECTION 7 1 607.990 607.990 1 75 62.200 1.041 64.766 9.388 2 435.890 -172.100 1 74 62.620 1.041 65.198 -2.640 3 249.430 -186.460 28 73 63.040 1.041 65.630 -2.841 4 129.390 -120.040 30 45 76.900 1.046 80.437 -1.492 5 262.390 133.000 1 15 104.000 1.055 109.720 1.212 6 117.780 -144.610 14 14 105.200 1.055 110.986 -1.303 Section 6 Total = 2.324 SECTION 7 PERIODs 1 -7 ONLY IF SECTION 8 > SECTION 7, GO TO SECTION 8 1 607.990 607.990 1 136 43.800 1.046 45.829 13.266 2 435.890 -172.100 1 135 44.000 1.046 46.035 -3.738 3 249.430 -186.460 28 134 44.233 1.046 46.275 -4.029 4 129.390 -120.040 30 106 51.440 1.044 53.695 -2.236 5 262.390 133.000 1 76 61.760 1.041 64.313 2.068 6 117.780 -144.610 14 75 62.200 1.041 64.766 -2.233 7 132.730 14.950 61 61 67.800 1.040 70.518 0.212 Section 7 Total = 3.310 SECTION 8 PERIODs 1 - 8 ONLY IF SECTION 9 > SECTION 8, GO TO SECTION 9 1 607.990 607.990 1 137 43.600 1.046 45.624 13.326 2 435.890 -172.100 1 136 43.800 1.046 45.829 -3.755 3 249.430 -186.460 28 135 44.000 1.046 46.035 -4.050 4 129.390 -120.040 30 107 51.080 1.044 53.323 -2.251 5 262.390 133.000 1 77 61.320 1.041 63.860 2.083 6 117.780 -144.610 14 76 61.760 1.041 64.313 -2.249 7 132.730 14.950 61 62 67.400 1.040 70.107 0.213 8 262.390 129.660 1 1 121.000 1.071 129.591 1.001 Section 8 Total = 4.317

A ITACHMENT B4 TO CALC. E0118, REV. 0 Page 3 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 9 PERIODS 1 - 9 ONLY IF SECTION 10> SECTION 9, GO TO SECTION 10 1 607.990 607.990 1 151 40.833 1.048 42.776 14.213 2 435.890 -172.100 1 150 41.000 1.048 42.948 -4.007 3 249.430 -186.460 28 149 41.200 1.047 43.154 -4.321 4 129.390 -120.040 30 121 47.267 1.045 49.398 -2.430 5 262.390 133.000 1 91 55.540 1.043 57.905 2.297 6 117.780 -144.610 14 90 55.800 1.043 58.172 -2.486 7 132.730 14.950 61 76 61.760 1.041 64.313 0.232 8 262.390 129.660 1 15 104.000 1.055 109.720 1.182 9 117.780 -144.610 14 14 105.200 1.055 110.986 -1.303 Section 9 Total = 3.377 SECTION 10 PERIODS 1 -10 ONLY IF SECTION 11 > SECTION 10, GO TO SECTION 11 1 607.990 607.990 212 32.773 1.048 34.360 17.695 2 435.890 -172.100 1 211 32.887 1.048 34.480 -4.991 3 249.430 -186.460 28 210 33.000 1.049 34.601 -5.389 4 129.390 -120.040 30 182 36.667 1.050 38.496 -3.118 5 262.390 133.000 1 152 40.667 1.048 42.605 3.122 6 117.780 -144.610 14 151 40.833 1.048 42.776 -3.381 7 132.730 14.950 61 137 43.600 1.046 45.624 0.328 8 262.390 129.660 1 76 61.760 1.041 64.313 2.016 9 117.780 -144.610 14 75 62.200 1.041 64.766 -2.233 10 132.730 14.950 61 61 67.800 1.040 70.518 0.212 Section 10 Total = 4.261 SECTION 11 PERIODS 1 -11 ONLY IF SECTION 12> SECTION 11, GO TO SECTION 12 1 607.990 607.990 1 213 32.660 1.048 34.239 17.757 2 435.890 -172.100 1 212 32.773 1.048 34.360 -5.009 3 249.430 -186.460 28 211 32.887 1.048 34.480 -5.408 4 129.390 -120.040 30 183 36.500 1.050 38.320 -3.133 5 262.390 133.000 1 153 40.500 1.048 42.434 3.134 6 117.780 -144.610 14 152 40.667 1.048 42.605 -3.394 7 132.730 14.950 61 138 43.400 1.047 45.418 0.329 8 262.390 129.660 1 77 61.320 1.041 63.860 2.030 9 117.780 -144.610 14 76 61.760 1.041 64.313 -2.249 10 132.730 14.950 61 62 67.400 1.040 70.107 0.213 11 262.390 129.660 1 1 121.000 1.071 129.591 1.001 Section 11 Total = 5.273

ATTACHMENT B4 TO CALC. E0118, REV. 0 Page 4 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 12 PERIODS 1-120NLY IF SECTION 13> SECTION 12, GO TO SECTION 13 1 607.990 607.990 227 31.140 1.048 32.624 18.636 2 435.890 -172.100 226 31.220 1.048 32.709 -5.262 3 249.430 -186.460 28 225 31.300 1.048 32.795 -5.686 4 129.390 -120.040 30 197 34.300 1.049 35.986 -3.336 5 262.390 133.000 1 167 38.300 1.049 40.174 3.311 6 117.780 -144.610 14 166 38.400 1.049 40.275 -3.591 7 132.730 14.950 61 152 40.667 1.048 42.605 0.351 8 262.390 129.660 1 91 55.540 1.043 57.905 2.239 9 117.780 -144.610 14 90 55.800 1.043 58.172 -2.486 10 132.730 14.950 61 76 61.760 1.041 64.313 0.232 11 262.390 129.660 1 15 104.000 1.055 109.720 1.182 12 117.780 -144.610 14 14 105.200 1.055 110.986 -1.303 Section 12 Total = 4.289 SECTION 13 PERIODS 1 -13 ONLY IF SECTION 14> SECTION 13, GO TO SECTION 14 1 607.990 607.990 1 299 25.560 1.044 26.686 22.783 2 435.890 -172.100 1 298 25.620 1.044 26.750 -6.434 3 249.430 -186.460 28 297 25.680 1.044 26.814 -6.954 4 129.390 -120.040 30 269 27.780 1.046 29.045 -4.133 5 262.390 133.000 1 239 30.180 1.047 31.600 4.209 6 117.780 -144.610 14 238 30.260 1.047 31.685 -4.564 7 132.730 14.950 61 224 31.413 1.048 32.915 0.454 8 262.390 129.660 1 163 38.833 1.049 40.720 3.184 9 117.780 -144.610 14 162 39.000 1.049 40.892 -3.536 10 132.730 14.950 61 148 41.400 1.047 43.360 0.345 11 262.390 129.660 1 87 57.120 1.042 59.533 2.178 12 117.780 -144.610 14 86 57.560 1.042 59.987 -2.411 13 132.220 14.440 72 72 63.460 1.041 66.062 0.219 Section 13 Total = 5.340

A ITACHMENT B4 TO CALC. E0118, REV. 0 Page 5 of 5 Cell Sizing Period Minutes Amp/Pos Correction Amp/Pos Required #

Load Load Duration To End Capacity Factor Capacity Positive Period Amps Change (Minutes) of Section (Unadjusted) (CSCF) (Adjusted) Plates SECTION 14 PERIODS 1 -14 ONLY IF SECTION 15> SECTION 14, GO TO SECTION 15 1 607.990 607.990 1 300 25.500 1.044 26.622 22.838 2 435.890 -172.100 1 299 25.560 1.044 26.686 -6.449 3 249.430 -186.460 28 298 25.620 1.044 26.750 -6.971 4 129.390 -120.040 30 270 27.700 1.046 28.960 -4.145 5 262.390 133.000 1 240 30.100 1.047 31.515 4.220 6 117.780 -144.610 14 239 30.180 1.047 31.600 -4.576 7 132.730 14.950 61 225 31.300 1.048 32.795 0.456 8 262.390 129.660 1 164 38.667 1.049 40.548 3.198 9 117.780 -144.610 14 163 38.833 1.049 40.720 -3.551 10 132.730 14.950 61 149 41.200 1.047 43.154 0.346 11 262.390 129.660 1 88 56.680 1.042 59.079 2.195 12 117.780 -144.610 14 87 57.120 1.042 59.533 -2.429 13 132.220 14.440 72 73 63.040 1.041 65.630 0.220 14 262.390 130.170 1 1 121.000 1.071 129.591 1.004 Section 14 Total = 6.356 Calculated Cell Size Maximum Temperature Required #

Section Random Uncorrected Correction Aging Design Positive Size Size Size Factor Factor Margin Plates 6.356 0.000 6.356 1.080 1.250 1.398 12 Required Battery Charqer DC Output Amps Continuous Amps + (AH Discharged

  • 1.1) / Battery Recharge Time Charger Output DC Amps = 151.10 A

ATTACHMENT C1 TO CALC. E 0118, REV. 0 Page 1 of 1 Battery Voltage Profile Worksheet INPUT DATA PlanVSubstatlon  : E. I. Hatch Nuclear Plant Unit No: 2 Battery Name :Station Service Battery 2A Equipment No: 2R42-S001A No. of Cells (NOC) :120 Aging Allowance: 25  % (a)

Battery Manufacturers :C&D Temperature Correction Factor: 8  % (b)

Manufacturers Type  : LCR-25 Design Margin : 46.38  % (c)

Positive Plates/Cell  : 12 Combined Correction Factor (CCF) : 1.97611 (a* b* c)

Discharge Curves  : D-841 S53155 Rev. 1 Voltage Profile

[1] [2] [31 [4] [5] [6] [7] [8] [91 [10] [11]

Cell Sizing Cell Sizing CCF VPC Correction Adjusted Adjusted Amps Amp -Hr Cumlatlve [7J x [91 Battery Time Amps Factor Amps Amps Per PP perPP Amp-Hr From Volts Step (Min) Discharged CSCF [3J/[4] [5]xCCF {6} / pp [7Jx[2]/60 PerPP Curve [10] x NOC 1 1 607.99 1.0710 567.68 1121.81 93.48 1.56 0 *

  • 1.56 2 1 435.89 1.0710 406.99 804.26 67.02 1.12 1.56 *
  • 2.68 3 28 249.43 1.0524 237.01 468.36 39.03 18.21 2.68 *
  • 20.89 4 30 129.39 1.0520 122.99 243.05 20.25 10.13 20.89 *
  • 31.02 5 1 262.39 1.0710 245.00 484.14 40.34 0.67 31.02 *
  • 31.69 6 14 117.78 1.0550 111.64 220.61 18.38 4.29 31.69 *
  • 35.98 7 61 132.73 1.0401 127.61 252.18 21.02 21.37 35.98 *
  • 57.34 8 1 262.39 1.0710 245.00 484.14 40.34 0.67 57.34 *
  • 58.02 9 14 117.78 1.0550 111.64 220.61 18.38 4.29 58.02 *
  • 62.31 10 6i 132.73 1.0401 127.61 252.18 ' 21.02 21.37 62.31 *
  • 83.67 11 1 262.39 1.0710 245.00 484.14 40.34 0.67 83.67 *
  • 84.34 12 14 117.78 1.0550 111.64 220.61 18.38 4.29 84.34 *
  • 88.63 13 72 132.22 1.0410 127.01 250.99 20.92 25.10 88.63 *
  • 113.73 14 1 262.39 1.0710 245.00 484.14 40.34 0.67 113.73 114.40 1.750 210.0 15 * * * * * * * * *
  • COMMENTS:
1. Design Margin was selected to require all of the plates available (no additional margin)
2. Since step 14 detennines the size of the battery. it is the only step calculated.
3. Additional margin for added battery connection resistance is not included.

ATTACHMENT C2 TO CALC. E 0118, REV. 0 Page 1 of 1 Battery Voltage Profile Worksheet INPUT DATA Plant/Substation  : E. I. Hatch Nuclear Plant Unit No: 2 Battery Name :Station Service Battery 2A Equipment No : 2R42-S001A No. of Cells (NOC)  : 120 Aging Allowance: 25% (a)

Battery Manufacturers :C&D Temperature Correction Factor: 8% (b)

Manufacturers Type  : LCR-25 Design Margin: 44.00 % (c)

Positive Plates/Cell  : 12 Combined Correction Factor (CCF) : 1.944 (a

  • b
  • c)

Discharge Curves  : 0.841 S53155 Rev. 1 Voltage Profile

[1J [2J [3J [4] [5] [6J [7] [8] [9J [1 OJ [11]

Cell Sizing Cell Sizing CCF VPC Correction Adjusted Adjusted Amps Amp-Hr Cumlalive [7] x [9J Battery Time Amps Factor Amps Amps PerPP perPP Amp-Hr From Volts Step (Min) Discharged CSCF [3J / [4J [5]xCCF (6}/pp [7]x[2]/60 PerPP Curve [10J xNOC 1 1 607.99 1.0710 567.68 1103.58 91.96 1.53 0 *

  • 1.53 2 1 435.89 1.0710 406.99 791.20 65.93 1.10 1.53 *
  • 2.63 3 28 249.43 1.0524 237.01 460.75 38.40 17.92 2.63 *
  • 20.55 4 30 129.39 1.0520 122.99 239.10 19.93 9.96 20.55 *
  • 30.51 5 1 262.39 1.0710 245.00 476.27 39.69 0.66 30.51 *
  • 31.17 6 14 117.78 1.0550 111.64 217.03 18.09 4.22 31.17 *
  • 35.39 7 61 132.73 1.0401 127.61 248.08 20.67 21.02 35.39 *
  • 56.41 8 1 262.39 1.0710 245.00 476.27 39.69 0.66 56.41 *
  • 57.07 9 14 117.78 1.0550 111.64 217.03 18.09 4.22 57.07 *
  • 61.29 10 61 132.73 1.0401 127.61 248.08 20.67 21.02 61.29 *
  • 82.31 11 1 262.39 1.0710 245.00 476.27 39.69 0.66 82.31 *
  • 82.97 12 14 117.78 1.0550 111.64 217.03 18.09 4.22 82.97 *
  • 87.19 13 72 132.22 1.0410 127.01 246.91 20.58 24.69 87.19 *
  • 111.88 14 1 262.39 1.0710 245.00 476.27 39.69 0.66 111.88 112.55 1.756 210.7 15 * * * * * * * * *
  • COMMENTS:
1. Design Margin was selected considering battery connection resistances to require all of the plates available.
2. Since step 14 determines the size of the battery. it is the only step calculated.

ATTACHMENT C3 TO CALC. E 0118, REV. 0 Page 1 of 1 Battery Voltage Profile Worksheet INPUT DATA PlanVSubstation  : E. I. Hatch Nuclear Plant Unit No: 2 Battery Name :Station Service Battery 2A Equipment No : 2R42-S001A No. of Cells (NOC)  : 118 Aging Allowance: 25% (a)

Battery Manufacturers :C&D Temperature Correction Factor: 8% (b)

Manufacturers Type  : LCR-25 Design Margin: 39.85 % (c)

Positive Plates/Cell :12 Combined Correction Factor (CCF) : 1.887975 (a* b* c)

Discharge Curves  : 0-841 S53155 Rev. 1 Voltage Profile

[1] [2] [3] [4] [5] [6] [7] [8] [9] [10] [11]

Cell Sizing Cell Sizing CCF VPC Correction Adjusted Adjusted Amps Amp-Hr Cumlallve [7] x [9] Battery Time Amps Factor Amps Amps PerPP perPP Amp-Hr From Volts Step (Min) Discharged CSCF [3]1 [4] [5] xCCF {6} 1 pp [7]x{2]160 PerPP Curve [10]xNOC 1 1 607.99 1.0710 567.68 1071.77 89.31 1.49 0 *

  • 1.49 2 1 435.89 1.0710 406.99 768.39 64.03 1.07 1.49 *
  • 2.56 3 28 249.43 1.0524 237.01 447.47 37.29 17.40 2.56 *
  • 19.96 4 30 129.39 1.0520 122.99 232.21 19.35 9.68 19.96 *
  • 29.63 5 1 262.39 1.0710 245.00 462.55 38.55 0.64 29.63 *
  • 30.28 6 14 117.78 1.0550 111.64 210.77 17.56 4.10 30.28 *
  • 34.37 7 61 132.73 1.0401 127.61 240.93 20.08 20.41 34.37 *
  • 54.79 8 1 262.39 1.0710 245.00 462.55 38.55 0.64 54.79 *
  • 55.43 9 14 117.78 1.0550 111.64 210.77 17.56 4.10 55.43 *
  • 59.53 10 61 132.73 1.0401 127.61 240.93 20.08 20.41 59.53 *
  • 79.94 11 1 262.39 1.0710 245.00 462.55 38.55 0.64 79.94 *
  • 80.58 12 14 117.78 1.0550 111.64 210.77 17.56 4.10 80.58 *
  • 84.68 13 72 132.22 1.0410 127.01 239.80 19.98 23.98 84.68 *
  • 108.66 14 1 262.39 1.0710 245.00 462.55 38.55 0.64 108.66 109.30 1.780 210.0 15 * * * * * * * * *
  • COMMENTS:
1. Design Margin was selected to be same as determined in the battery sizing worksheet, with one cell removed from each bank.
2. Since step 14 determines the size of the battery, it is the only step calculated.
3. Additional margin for added battery connection resistance is not included.

ATTACHMENT C4 TO CALC. E0118, REV. 0 Page 1 of 1 Battery Voltage Profile Worksheet INPUT DATA Plant/Substation  ; E. I. Hatch Nuclear Plant Unit No; 2 Battery Name :Station Service Battery 2A Equipment No ; 2R42-S001A No. of Cells (NOC)  ; 118 Aging Allowance: 25 % (a)

Battery Manufacturers :C&D Temperature Correction Factor: 8% (b)

Manufacturers Type  : LCR-25 Design Margin: 34.90 % (c)

Positive Plates/Cell :12 Combined Correction Factor (CCF) ; 1.82115 = (a

  • b
  • c)

Discharge Curves  : 0-841 S53155 Rev. 1 Voltage Profile

[1] [2] [3] [4] [5} [6] [7] [8} [9} [10] [11]

Cell Sizing Cell Sizing CCF VPC Correction Adjusted Adjusted Amps Amp-Hr Cumlative [7] x [9} Battery Time Amps Factor Amps Amps PerPP perPP Amp-Hr From Volts Step (Min) Discharged CSCF (3)/[4) [5] xCCF {6} / pp [7]x[2]160 PerPP Curve [10] x NOC 1 1 607.99 1.0710 567.68 1033.84 86.15 1.44 0 *

  • 1.44 2 1 435.89 1.0710 406.99 741.20 61.77 1.03 1.44 *
  • 2.47 3 28 249.43 1.0524 237.01 431.63 35.97 16.79 2.47 *
  • 19.25 4 30 129.39 1.0520 122.99 223.99 18.67 9.33 19.25 *
  • 28.58 5 1 262.39 1.0710 245.00 446.17 37.18 0.62 28.58 *
  • 29.20 6 14 117.78 1.0550 111.64 203.31 16.94 3.95 29.20 *
  • 33.16 7 61 132.73 1.0401 127.61 232.41 19.37 19.69 33.16 *
  • 52.85 8 262.39 1.0710 245.00 446.17 37.18 0.62 52.85 *
  • 53.47 9 14 117.78 1.0550 111.64 203.31 16.94 3.95 53.47 *
  • 57.42 10 61 132.73 1.0401 127.61 232.41 19.37 19.69 57.42 *
  • 77.11 11 1 262.39 1.0710 245.00 446.17 37.18 0.62 77.11 *
  • 77.73 12 14 117.78 1.0550 111.64 203.31 16.94 3.95 77.73 *
  • 81.68 13 72 132.22 1.0410 127.01 231.31 19.28 23.13 81.68 *
  • 104.81 14 1 262.39 1.0710 245.00 446.17 37.18 0.62 104.81 105.43 1.786 210.7 15 * * * * * * * * *
  • COMMENTS:
1. A reduced design margin was selected to account for the battery connection resistance with one cell removed from each bank.
2. Since step 14 determines the size of the battery, it is the only step calculated.

AnACHMENT D1 TO CALC. E0118, REV.O Page 1 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75v/CELL 60/120 CELLS (C&D CURVE S-53155 0-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 300 26.10 1.0440 27.248 299 26.19 1.0441 27.340 298 26.27 1.0441 27.432 297 26.36 1.0442 27.524 296 26.45 1.0442 27.616 295 26.53 1.0443 27.707 294 26.62 1.0443 27.799 293 26.71 1.0444 27.891 292 26.79 1.0444 27.983 291 26.88 1.0445 28.075 290 26.97 1.0445 28.167 289 27.05 1.0446 28.259 288 27.14 1.0446 28.350 287 27.23 1.0447 28.442 286 27.31 1.0447 28.534 285 27.40 1.0448 28.626 284 27.45 1.0448 28.676 283 27.49 1.0449 28.726 282 27.54 1.0449 28.777 281 27.59 1.0450 28.827 280 27.63 1.0450 28.877 279 27.68 1.0451 28.927 278 27.73 1.0451 28.977 277 27.77 1.0452 29.027 276 27.82 1.0452 29.077 275 27.87 1.0453 29.128 274 27.91 1.0453 29.178 273 27.96 1.0454 29.228 272 28.01 1.0454 29.278 271 28.05 1.0455 29.328 270 28.10 1.0455 29.379 269 28.19 1.0456 29.478 268 28.29 1.0456 29.577 267 28.38 1.0457 29.676 266 28.47 1.0457 29.775 265 28.57 1.0458 29.874 264 28.66 1.0458 29.973 263 28.75 1.0459 30.072 262 28.85 1.0459 30.171 261 28.94 1.0460 30.270 260 29.03 1.0460 30.369 259 29.13 1.0461 30.468

ATTACHMENT 01 TO CALC. E0118, REV.O Page 2 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75VJCELL 601120 CELLS (C&D CURVE SM53155 0-841 TirneTo Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 258 29.22 1.0461 30.567 257 29.31 1.0462 30.666 256 29.41 1.0462 30.765 255 29.50 1.0463 30.864 254 29.61 1.0463 30.984 253 29.73 1.0464 31.104 252 29.84 1.0464 31.225 251 29.95 1.0465 31.345 250 30.07 1.0465 31.465 249 30.18 1.0466 31.585 248 30.29 1.0466 31.705 247 30.41 1.0467 31.825 246 30.52 1.0467 31.945 245 30.63 1.0468 32.065 244 30.75 1.0468 32.186 243 30.86 1.0469 32.306 242 30.97 1.0469 32.426 241 31.09 1.0470 32.546 240 31.20 1.0470 32.666 239 31.33 1.0471 32.804 238 31.46 1.0471 32.942 237 31.59 1.0472 33.079 236 31.72 1.0472 33.217 235 31.85 1.0473 33.355 234 31.98 1.0473 33.493 233 32.11 1.0474 33.630 232 32.24 1.0474 33.768 231 32.37 1.0475 33.906 230 32.50 1.0475 34.044 229 32.63 1.0476 34.182 228 32.76 1.0476 34.319 227 32.89 1.0477 34.457 226 33.02 1.0477 34.595 225 33.15 1.0478 34.733 224 33.28 1.0478 34.871 223 33.41 1.0479 35.009 222 33.54 1.0479 35.147 221 33.67 1.0480 35.284 220 33.80 1.0480 35.422 219 33.92 1.0481 35.550 218 34.04 1.0481 35.677

  • 217 34.16 1.0482 35.805 216 34.28 1.0482 35.932

ATTACHMENT 01 TO CALC. E0118, REV.O Page 3 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75V/CELL 60/120 CELLS (C&D CURVE S-53155 0-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 215 34.40 1.0483 36.060 214 34.52 1.0483 36.187 213 34.64 1.0484 36.315 212 34.76 1.0484 36.442 211 34.88 1.0485 36.570 210 35.00 1.0485 36.698 209 35.12 1.0486 36.825 208 35.24 1.0486 36.953 207 35.36 1.0487 37.080 206 35.48 1.0487 37.208 205 35.60 1.0488 37.336 204 35.72 1.0488 37.463 203 35.84 1.0489 37.591 202 35.96 1.0489 37.718 201 36.08 1.0490 37.846 200 36.20 1.0490 37.974 199 36.30 1.0491 38.081 198 36.40 1.0491 38.187 197 36.50 1.0492 38.294 196 36.60 1.0492 38.401 195 36.70 1.0493 38.507 194 36.80 1.0493 38.614 193 36.90 1.0494 38.721 192 37.00 1.0494 38.828 191 37.10 1.0495 38.935 190 37.20 1.0495 39.041 189 37.30 1.0496 39.148 188 37.40 1.0496 39.255 187 37.50 1.0497 39.362 186 37.60 1.0497 39.469 185 37.70 1.0498 39.576 184 37.80 1.0498 39.682 183 37.90 1.0499 39.789 182 38.00 1.0499 39.896 181 38.10 1.0500 40.003 180 38.20 1.0500 40.110 179 38.37 1.0499 40.282 178 38.53 1.0498 40.454 177 38.70 1.0498 40.625 176 38.87 1.0497 40.797 175 39.03 1.0496 40.969 174 39.20 1.0495 41.140 173 39.42 1.0494 41.365

ATTACHMENT D1 TO CALC. E0118, REV.O Page 4 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75v/CELL 60/120 CELLS (C&D CURVE S-53155 0-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 172 39.63 1.0493 41.589 171 39.85 1.0493 41.813 170 40.07 1.0492 42.037 169 40.28 1.0491 42.261 168 40.50 1.0490 42.485 167 40.67 1.0489 42.656 166 40.83 1.0488 42.827 165 41.00 1.0488 42.999 164 41.17 1.0487 43.170 163 41.33 1.0486 43.341 162 41.50 1.0485 43.513 161 41.72 .1.0484 43.736 160 41.93 1.0483 43.960 159 42.15 1.0483 44.184 158 42.37 1.0482 44.407 157 42.58 1.0481 44.631 156 42.80 1.0480 44.854 155 43.00 1.0479 45.060 154 43.20 1.0478 45.266 153 43.40 1.0478 45.472 152 43.60 1.0477 45.678 151 43.80 1.0476 45.884 150 44.00 1.0475 46.090 149 44.20 1.0474 46.296 148 44.40 1.0473 46.502 147 44.60 1.0473 46.707 146 44.80 1.0472 46.913 145 45.00 1.0471 47.119 144 45.20 1.0470 47.324 143 45.42 1.0469 47.547 142 45.63 1.0468 47.770 141 45.85 1.0468 47.993 140 46.07 1.0467 48.216 139 46.28 1.0466 48.439 138 46.50 1.0465 48.662 137 46.68 1.0464 48.850 136 46.87 1.0463 49.038 135 47.05 1.0463 49.226 134 47.23 1.0462 49.414 133 47.42 1.0461 49.602 132 47.60 1.0460 49.790

. 131 47.83 1.0459 50.030 130 48.07 1.0458 50.270

ATTACHMENT D1 TO CALC. E0118, REV.O Page 5 of 7 STATION SERVICE SATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75V1CELL 60/120 CELLS (C&D CURVE S-53155 0-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section { Not Correction Capacity (Mins) Adjusted} Factor (Adjusted) 129 48.30 1.0458 50.510 128 48.53 1.0457 50.750 127 48.77 1.0456 50.990 126 49.00 1.0455 51.230 125 49.17 1.0454 51.400 124 49.33 1.0453 51.570 123 49.50 1.0453 51.740 122 49.67 1.0452 51.910 121 49.83 1.0451 52.080 120 50.00 1.0450 52.250 119 50.36 1.0449 52.622 118 50.72 1.0448 52.994 117 51.08 1.0448 53.366 116 51.44 1.0447 53.738 115 51.80 1.0446 54.109 114 52.04 1.0445 54.356 113 52.28 1.0444 54.602 112 52.52 1.0443 54.848 111 52.76 1.0443 55.095 110 53.00 1.0442 55.341 109 53.30 1.0441 55.650 108 53.60 1.0440 55.958 107 53.90 1.0439 56.267 106 54.20 1.0438 56.576 105 54.50 1.0438 56.884 104 54.80 1.0437 57.193 103 55.10 1.0436 57.501 102 55.40 1.0435 57.810 101 55.70 1.0434 58.118 100 56.00 1.0433 58.427 99 56.36 1.0433 58.798 98 56.72 1.0432 59.168 97 57.08 1.0431 59.539 96 57.44 1.0430 59.910 95 57.80 1.0429 60.281 94 58.04 1.0428 60.526 93 58.28 1.0428 60.771 92 58.52 1.0427 61.017 91 58.76 1.0426 61.262 90 59.00 1.0425 61.508 89 59.50 1.0424 62.024 88 60.00 1.0423 62.540 87 60.50 1.0423 63.056

ATTACHMENT D1 TO CALC. E0118, REV.O Page 6 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75v/CELL 60/120 CELLS (C&D CURVE S-53155 D-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 86 61.00 1.0422 63.572 85 61.50 1.0421 64.088 84 62.00 1.0420 64.604 83 62.50 1.0419 65.120 82 63.00 1.0418 65.636 81 63.50 1.0418 66.151 80 64.00 1.0417 66.667 79 64.50 1.0416 67.182 78 65.00 1.0415 67.698 77 65.50 1.0414 68.213 76 66.00 1.0413 68.728 75 66.50 1.0413 69.243 74 66.90 1.0412 69.654 73 67.30 1.0411 70.065 72 67.70 1.0410 70.476 71 68.10 1.0409 70.886 70 68.50 1.0408 71.297 69 69.00 1.0408 71.812 68 69.50 1.0407 72.326 67 70.00 1.0406 72.841 66 70.50 1.0405 73.355 65 71.00 1.0404 73.870 64 71.46 1.0403 74.342 63 71.92 1.0403 74.815 62 72.38 1.0402 75.287 61 72.84 1.0401 75.760 60 73.30 1.0400 76.232 59 73.94 1.0404 76.927 58 74.58 1.0408 77.623 57 75.22 1.0412 78.319 56 75.86 1.0416 79.016 55 76.50 1.0420 79.713 54 77.20 1.0424 80.473 53 77.90 1.0428 81.234 52 78.60 1.0432 81.996 51 79.30 1.0436 82.757 50 80.00 1.0440 83.520 49 80.60 1.0444 84.179 48 81.20 1.0448 84.838 47 81.80 1.0452 85.497 46 82.40 1.0456 86.157 45 83.00 1.0460 86.818 44 83.80 1.0464 87.688

ATTACHMENT D1 TO CALC. E0118, REV.O Page 7 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS, C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.75V/CELL 601120 CELLS (C&D CURVE S*53155 0-841 Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 43 84.60 1.0468 88.559 42 85.40 1.0472 89.431 41 86.20 1.0476 90.303 40 87.00 1.0480 91.176 39 87.90 1.0484 92.154 38 88.80 1.0488 93.133 37 89.70 1.0492 94.113 36 90.60 1.0496 95.094 35 91.50 1.0500 96.075 34 92.50 1.0504 97.162 33 93.50 1.0508 98.250 32 94.50 1.0512 99.338 31 95.50 1.0516 100.428 30 96.50 1.0520 101.518 29 97.60 1.0522 102.695 28 98.70 1.0524 103.872 27 99.80 1.0526 105.049 26 100.90 1.0528 106.228 25 102.00 1.0530 107.406 24 103.16 1.0532 108.648 23 104.32 1.0534 109.891 22 105.48 1.0536 111.134 21 106.64 1.0538 112.377 20 107.80 1.0540 113.621 19 109.40 1.0542 115.329 18 110.80 1.0544 116.828 17 112.20 1.0546 118.326 16 113.60 1.0548 119.825 15 115.00 1.0550 121.325 14 116.80 1.0550 123.224 13 118.60 1.0550 125.123 12 120.40 1.0550 124.285 11 122.20 1.0550 128.921 10 124.00 1.0550 130.820 9 126.00 1.0550 132.930 8 128.00 1.0550 135.040 7 129.80 1.0550 136.939 6 131.50 1.0550 138.733 5 133.00 1.0550 140.315 4 134.80 1.0590 142.753 3 136.80 1.0630 145.418 2 138.20 1.0670 147.459 1 140.00 1.0710 149.940

ATTACHMENT D2 TO CALC. E0118, REV.O Page 1 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V1CELL 59/118 CELLS (C&D CURVE S-53155 0-841)

Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 300 25.50 1.0440 26.62 299 25.56 1.0441 26.69 298 25.62 1.0441 26.75 297 25.68 1.0442 26.81 296 25.74 1.0442 26.88 295 25.80 1.0443 26.94 294 25.86 1.0443 27.01 293 25.92 1.0444 27.07 292 25.98 1.0444 27.13 291 26.04 1.0445 27.20 290 26.10 1.0445 27.26 289 26.16 1.0446 27.33 288 26.22 1.0446 27.39 287 26.28 1.0447 27.45 286 26.34 1.0447 27.52 285 26.40 1.0448 27.58 284 26.49 1.0448 27.67 283 26.57 1.0449 27.77 282 26.66 1.0449 27.86 281 26.75 1.0450 27.95 280 26.83 1.0450 28.04 279 26.92 1.0451 28.13 278 27.01 1.0451 28.22 277 27.09 1.0452 28.32 276 27.18 1.0452 28.41 275 27.27 1.0453 28.50 274 27.35 1.0453 28.59 273 27.44 1.0454 28.68 272 27.53 1:0454 28.78 271 27.61 1.0455 28.87 270 27.70 1.0455 28.96 269 27.78 1.0456 29.05 268 27.86 1.0456 29.13 267 27.94 1.0457 29.22 266 28.02 1.0457 29.30 265 28.10 1.0458 29.39 264 28.18 1.0458 29.47 263 28.26 1.0459 29.56 262 28.34 1.0459 29.64 261 28.42 1.0460 29.73 260 28.50 1.0460 29.81 259 28.58 1.0461 29.90

AITACHMENT 02 TO CALC. E0118, REV.O Page 2 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V/CELL 59/118 CELLS (C&D CURVE S-53155 0-841 0-841)

Time To AmpslPos End Of Capacity Cell Sizing AmpslPos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 258 28.66 1.0461 29.98 257 28.74 1.0462 30.07 256 28.82 1.0462 30.15 255 28.90 1.0463 30.24 254 28.98 1.0463 30.32 253 29.06 1.0464 30.41 252 29.14 1.0464 30.49 251 29.22 1.0465 30.58 250 29.30 1.0465 30.66 249 29.38 1.0466 30.75 248 29.46 1.0466 30.83 247 29.54 1.0467 30.92 246 29.62 1.0467 31.00 245 29.70 1.0468 31.09 244 29.78 1.0468 31.17 243 29.86 1.0469 31.26 242 29.94 1.0469 31.34 241 30.02 1.0470 31.43 240 30.10 1.0470 31.51 239 30.18 1.0471 31.60 238 30.26 1.0471 31.69 237 30.34 1.0472 31.77 236 30.42 1.0472 31.86 235 30.50 1.0473 31.94 234 30.58 1.0473 32.03 233 30.66 1.0474 32.11 232 30.74 1.0474 32.20 231 30.82 1.0475 32.28 230 30.90 1.0475 32.37 229 30.98 1.0476 32.45 228 31.06 1.0476 32.54 227 31.14 1.0477 32.62 226 31.22 1.0477 32.71 225 31.30 1.0478 32.79 224 31.41 1.0478 32.91 223 31.53 1.0479 33.04 222 31.64 1.0479 33.16 221 31.75 1.0480 33.28 220 31.87 1.0480 33.40 219 31.98 1.0481 33.52 218 32.09 1.0481 33.64 217 32.21 1.0482 33.76 216 32.32 1.0482 33.88

ATTACHMENT 02 TO CALC. E0118, REV.O Page 3 of 7 STATION SERVICE SATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V1CELL 59/118 CELLS (C&D CURVE S-53155 0-841) 0-841)

Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 215 32.43 1.0483 34.00 214 32.55 1.0483 34.12 213 32.66 1.0484 34.24 212 32.77 1.0484 34.36 211 32.89 1.0485 34.48 210 33.00 1.0485 34.60 209 33.10 1.0486 34.71 208 33.20 1.0486 34.81 207 33.30 1.0487 34.92 206 33.40 1.0487 35.03 205 33.50 1.0488 35.13 204 33.60 1.0488 35.24 203 33.70 1.0489 35.35 202 33.80 1.0489 35.45 201 33.90 1.0490 35.56 200 34.00 1.0490 35.67 199 34.10 1.0491 35.77 198 34.20 1.0491 35.88 197 34.30 1.0492 35.99.

196 34.40 1.0492 36.09 195 34.50 1.0493 36.20 194 34.67 1.0493 36.38 193 34.83 1.0494 36.55 192 35.00 1.0494 36.73 191 35.17 1.0495 36.91 190 35.33 1.0495 37.08 189 35.50 1.0496 37.26 188 35.67 1.0496 37.44 187 35.83 1.0497 37.61 186 36.00 1.0497 37.79 185 36.17 1.0498 37.97 184 36.33 1.0498 38.14 183 36.50 1.0499 38.32 182 36.67 1.0499 38.50 181 36.83 1.0500 38.67 180 37.00 1.0500 38.85 179 37.10 1.0499 38.95 178 37.20 1.0498 39.05 177 37.30 1.0498 39.16 176 37.40 1.0497 39.26 175 37.50 1.0496 39.36 174 37.60 1.0495 39.46 173 37.70 1.0494 39.56

ATTACHMENT 02 TO CALC. E0118, REV.O Page 4 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR*25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78VJCELL 59/118 CELLS (C&O CURVE S"53155 0"841)

Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 172 37.80 1.0493 39.66 171 37.90 1.0493 39.77 170 38.00 1.0492 39.87 169 38.10 1.0491 39.97 168 38.20 1.0490 40.07 167 38.30 1.0489 40.17 166 38.40 1.0488 40.28 165 38.50 1.0488 40.38 164 38.67 1.0487 40.55 163 38.83 1.0486 40.72 162 39.00 1.0485 40.89 161 39.17 1.0484 41.06 160 39.33 1.0483 41.23 159 39.50 1.0483 41.41 158 39.67 1.0482 41.58 157 39.83 1.0481 41.75 156 40.00 1.0480 41.92 155 40.17 1.0479 42.09 154 40.33 1.0478 42.26 153 40.50 1.0478 42.43 152 40.67 1.0477 42.61 151 40.83 1.0476 42.78 150 41.00 1.0475 42.95 149 41.20 1.0474 43.15 148 41.40 1.0473 43.36 147 41.60 1.0473 43.57 146 41.80 1.0472 43.77 145 42.00 1.0471 43.98 144 42.20 1.0470 44.18 143 42.40 1.0469 44.39 142 42.60 1.0468 44.60 141 42.80 1.0468 44.80 140 43.00 1.0467 45.01 139 43.20 1.0466 45.21 138 43.40 1.0465 45.42 137 43.60 1.0464 45.62 136 43.80 1.0463 45.83 135 44.00 1.0463 46.04 134 44.23 1.0462 46.28 133 44.47 1.0461 46.52 132 44.70 1.0460 46.76 131 44.93 1.0459 47.00 130 45.17 1.0458 47.24

ATTACHMENT 02 TO CALC. E0118, REV.O Page 5 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V1CELL 59/118 CELLS (C&D CURVE S-53155 D-841)

Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 129 45.40 1.0458 47.48 128 45.63 1.0457 47.72 127 45.87 1.0456 47.96 126 46.10 1.0455 48.20 125 46.33 1.0454 48.44 124 46.57 1.0453 48.68 123 46.80 1.0453 48.92 122 47.03 1.0452 49.16 121 47.27 1.0451 49.40 120 47.500 1.0450 49.64 119 47.760 1.0449 49.91 118 48.020 1.0448 50.17 117 48.280 1.0448 50.44 116 48.540 1.0447 50.71 115 48.800 1.0446 50.98 114 49.040 1.0445 51.22 113 49.280 1.0444 51.47 112 49.520 1.0443 51.72 111 49.760 1.0443 51.96 110 50.000 1.0442 52.21 109 50.360 1.0441 52.58 108 50.720 1.0440 52.95 107 51.080 1.0439 53.32 106 51.440 1.0438 53.69 105 51.800 1.0438 54.07 104 52.040 1.0437 54.31 103 52.280 1.0436 54.56 102 52.520 1.0435 54.80 101 52.760 1.0434 55.05 100 53.000 1.0433 55.30 99 53.300 1.0433 55.61 98 53.600 1.0432 55.91 97 53.900 1.0431 56.22 96 54.200 1.0430 56.53 95 54.500 1.0429 56.84 94 54.760 1.0428 57.11 93 55.020 1.0428 57.37 92 55.280 1.0427 57.64 91 55.540 1.0426 57.91 90 55.800 1.0425 58.17 89 56.240 1.0424 58.63 88 56.680 1.0423 59.08 87 57.120 1.0423 59.53

ATTACHMENT 02 TO CALC. E0118, REV.O Page 6 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V1CELL 591118 CELLS (C&D CURVE S-53155 0-841)

Time To AmpslPos End Of Capacity Cell Sizing AmpslPos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted)

(Adjusted) 86 57.560 1.0422 59.99 85 58.000 1.0421 60.44 84 58.400 1.0420 60.85 83 58.800 1.0419 61.26 82 59.200 1.0418 61.68 81 59.600 1.0418 62.09 80 60.000 1.0417 62.50 79 60.440 1.0416 62.95 78 60.880 1.0415 63.41 77 61.320 1.0414 63.86 76 61.760 1.0413 64.31 75 62.200 1.0413 64.77 74 62.620 1.0412 65.20 73 63.040 1.0411 65.63 72 63.460 1.0410 66.06 71 63.880 1.0409 66.49 70 64.300 1.0408 66.93 69 64.680 1.0408 67.32 68 65.060 1.0407 67.71 67 65.440 1.0406 68.10 66 65.820 1.0405 68.49 65 66.200 1.0404 68.88 64 66.600 1.0403 69.29 63 67.000 1.0403 69.70 62 67.400 1.0402 70.11 61 67.800 1.0401 70.52 60 68.200 1.0400 70.93 59 68.800 1.0404 71.58 58 69.400 1.0408 72.23 57 70.000 1.0412 72.88 56 70.600 1.0416 73.54 55 71.200 1.0420 74.19 54 71.660 1.0424 74.70 53 72.320 1.0428 75.42 52 72.889 1.0432 76.04 51 73.440 1.0436 76.64 50 74.000 1.0440 77.26 49 74.580 1.0444 77.89 48 75.160 1.0448 78.53 47 75.740 1.0452 79.16 46 76.320 1.0456 79.80 45 76.900 1.0460 80.44 44 77.620 1.0464 81.22

ATTACHMENT 02 TO CALC. E0118, REV.O Page 7 of 7 STATION SERVICE BATTERY 2A DISCHARGE CHARECTERISTICS C&D BATTERY LCR-25 AMPS/POS. PLATE, AT DISCHARGE CURVE OF 1.78V1CELL 59/118 CELLS (C&D CURVE S-53155 0-841)

Time To Amps/Pos End Of Capacity Cell Sizing Amps/Pos Section ( Not Correction Capacity (Mins) Adjusted) Factor (Adjusted) 43 78.340 1.0468 82.01 42 79.060 1.0472 82.79 41 79.780 1.0476 83.58 40 80.500 1.0480 84.36 39 81.240 1.0484 85.17 38 81.980 1.0488 85.98 37 82.720 1.0492 86.79 36 83.460 1.0496 87.60 35 84.200 1.0500 88.41 34 84.960 1.0504 89.24 33 85.720 1.0508 90.07 32 86.480 1.0512 90.91 31 87.240 1.0516 91.74 30 88.000 1.0520 92.58 29 89.000 1.0522 93.65 28 90.000 1.0524 94.72 27 91.000 1.0526 95.79 26 92.000 1.0528 96.86 25 93.000 1.0530 97.93 24 94.000 1.0532 99.00 23 95.000 1.0534 100.07 22 96.000 1.0536 101.15 21 97.000 1.0538 102.22 20 98.000 1.0540 103.29 19 99.200 1.0542 104.58 18 100.400 1.0544 105.86 17 101.600 1.0546 107.15 16 102.800 1.0548 108.43 15 104.000 1.0550 109.72 14 105.200 1.0550 110.99 13 106.400 1.0550 112.25 12 107.600 1.0550 113.52 11 108.800 1.0550 114.78 10 110.000 1.0550 116.05 9 111.200 1.0550 117.32 8 112.000 1.0550 118.16 7 113.800 1.0550 120.06 6 115.000 1.0550 121.33 5 116.000 1.0550 122.38 4 117.200 1.0590 124.11 3 118.600 1.0630 126.07 2 120.000 1.0670 128.04 1 121.000 1.0710 129.59

JUN.19.2002 7:59AM ENG/LIC/P&P Southern Nuclear Attachment E Opsrating Company, 1110.

p. O. Box 1295 To Calc. E 0118, Rev.O Birmingham. Alabama 35201.1295 Page lof2 Tel 205.992.6000 Mr. Larry L. Rowe Bechtel Power Corporation 5325 Spectrum Drive Frederlck, MD 21703-8388 Hatch Project SuppoJ:to. Licensing B#ended Battery Life LogNo. HL-62S2

Dear Sir:

REAHT02-637 was writtell"to prepate calcu1atiOllS for extended battery life for Hatch Unit ~ and Unit 2 A Station Service Batteries. The calculations are perfonned to evaluate 'a. speo.ifio ~

for the Hatch Probabilfstic Safety Assessment. The scenario is a long tenn Station Blackout and is based on the following assumpti01lll. .

1. The RCIC startup oocurs fout times bread of 3 times as in tho LOCA profile. The RCI~

startup tabs place at 1at, 7st!', IS 11\ 227111 minutes, Eaoh time the RCIe startup takes p~ it runs for approximately 60 minutes before it is tripped. Each time ReIC is trlpp~ RCIe 1'&0 initiation takes place after 15 minutes. The RCIC started for the final time runs for 72

. minutes before it is tripped at the 299th minute. ,

I

2. The.RFPT lA Emergency Oil Pump IN34-COO7 and Ernergenoy Bea.rblg oil pump IN34-C012 for Unit 1 and RFPT 2A. Tu.rbine Lube on Emergency ~ 2N34-CO09 and DC Btnergenoy Bearing Oil Pump 2N34-C004 started during the let minute mIl be tripped a.t the end ofthe 301h minu.tc.
3. RcrC suction switcbover will not occur at any t:itne during the 5..hour duty ~o.

i The ca1cula1:ions are EO 113 for Unit 1 and EO 1I 8 fOl' Unit 2 with both eve1uating station s~

battety capacities for a S-hour duration. .~

The PSA scenario for SBO g6es beyond the FSAR Coping Analysis and into possibilities of not getting AC power back. for some time. Rele, in this easo, becomes the high-pressure source for water injection to the reactor vessel. The A Station Servi~ Battery System powers RCIe thus it becomes the source under investigation. The above USl.llDptiOllB allow for socuring selected pc ,

powered pumps in order to co~ battery power. These pumps are the DC coastdown unit, for the RFPTs and the Main TuriJine. During the SBO, main condenser vacuum will be lost1b.er0by stopping the turbines most likely within 30 minutes. Securing the oil coastdown pumps becomes of little consequenoe for turbine protection after thi.s time frame, but of major consequence :in batt~ power OOllBervatiOn. Tn any case, battery conservation is of the utmost importance in 1;h.e SBO scenario. The RCle suction source swapover will also not be allowed in order to limit batterY duty due to cycling the DC powere4 valves. The cycling times for RCre are se1eotcd based on conservatism and .in reality are probably going to be less, not more.

  • JUN.19.2002 8:00AM ENG/LIC/P&P Attachment E To Calc. EO! 1'8, Rev.O Page 2 of2 Mr. Larry L. Rowe June 18, 2002 Pagel Certain items from the above discussion will be inoluded in plant abnormal procedures, 34AB-R2U03-1S and.2S: STATION BLACKOUT. . f I

If you have any quomom, please call Ed I:ngram at 2OS~992..6701, DonM.Crowe Licensing Service Manager OCV/eb oc: Sashi K. Ahujll SteveF.Cortis Ed Ingram JituRathod Bill Snider DOO1.Ul1eJlt Management

Edwin I. Hatch Nuclear Plant Enclosure 11 RCIC Calculation 2 Core Cooling from AC-Independent Systems Until Battery Depletion

SCENARIO CLASS DEFINITION Long Term Station Blackout (Loss of All RPV Injection at Battery Depletion)

This class of scenarios involves loss of all on-site AC power (Le., normal and emergency) such that core cooling is provided by AC-independent systems until battery depletion.

  • Initial Condition: Operation at 100% reactor power
  • Station Blackout at t=O
  • Batteries Depleted at t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
  • Loss of all RPV injection at battery depletion SCENARIO TIMELINE Event Time Event Initiation T = 0 sec Battery Depletion and Feedwater Loss T= 5.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Core Uncovery T1 = 6.68 hours7.87037e-4 days <br />0.0189 hours <br />1.124339e-4 weeks <br />2.5874e-5 months <br /> Core Damage T2 = 7.59 hours6.828704e-4 days <br />0.0164 hours <br />9.755291e-5 weeks <br />2.24495e-5 months <br /> Vessel Failure T3 = 11.32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> Containment Failure T4 = 11.33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> Basis:

Comments:

This case did not scram at time=O, Rx scram occurred @ 3.5 seconds.

See attached figures.

"b x

1.5 .jilillhllfliiihhii,ih,iiilliliillhli",illllliilliiiiihiiiiiiliiiiilliiiiiiil"iI& 150 QiiJiiiililiiiiliiiiiiiliiilillii Iliiiilliiii i lilliill 300

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> 0 2 3 4 5 6 7 8 9 0 2 3 4 5 6 7 8 9 0 2 3 4 5 6 7 8 9

?

.... x 10' x 10' x 10' TIME. SECONDS TIME. SECONDS TIME . SECONDS z

0

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N 100

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TIME. SECONDS x 10' TIME. SECONDS x 10' TIME. SECONDS x 10' e>

~ SOURCE GCl FILE IS C"Hatah,SNC023'CASES.GCl

Edwin I. Hatch Nuclear Plant Enclosure 12 United Controls International Repair Report 3697 Failed LOCA CARD Test Report

United Controls International Repair Report 36971 Revision 0 5139 S. Royal Atlanta Drive Tucker, Georgia 30084 Tel: (770) 496-1406 Fax: (770) 496-1422 I.

i I

I

'. 'Failed LOCA.CARDTestReport I

I II UCI Job Number: 3697 I* Georgia Power P/ON6".:* "60826!?9 .'.

Customer SIN: CIS 1601 Seq. A Date: 12/18/2009*

\

\ Purchase Order Line* Item Number: 001 .

R~ported Failure / As Found Condition: .Timing sequence reset at approximately 14 seconds during the second of four scheduled tests of the 2C Diesel: . The

. se1quence A card has no normal changes in output state at 14 seconds.

su1pseqpent, multiple tests in the CS1600 Test Set revealed no problems; I

Te~t Summary: No mechanical or electrical problems could be found with the card\ in bench testing and no external stresses could induce a timing failure .

.\

\

~~~~~ .

'>1)\ Physical Examination - The card was reviewed for bad solder joints, IC~s

\not seated in their sockets properly, broken component leads, defective t~work, and cracked traces. Visual examination did not show any defects

'.....*.. CI*I~hough workmanship is poor as on all of these cards .

... 2)lJ!>jpowered Tests - All diodes and resistors were checked with a meter to ..

  • '/-.**:~h~ure the diodes were functioning and resistors were the proper value.
  • . AII'6,hecked okay.

9)PoW~red . , Tests with 125 VDC and 24VDCinputs:.

8; '\+5 Volt Power Supply - measured 5.1 VDC with a peak to peak

'.' ,.,:r;ipple

," ,", .; .\

of 142 mV which is within the 5% tolerance specifications.

<.>J:'b:.t:R¢set Circuit - Tested pushbutton' reset switch for sensitive

'... , '. "op'E?ration with vibration and minor manipulation of the actuator. 'No fals~ reset signal was produced .. Switch had to be pushed in until.

the actuator bottomed out before a reset signal was generated .

.Monit~red reset signal during multiple timing sequences' and

'8,Xterr'la/:'stresses with no false glitches.

Page 1 of2 United Controls International ~ 5139 S.RoyaJ Atlanta Drive ~ Tucker, GA 30084 www.unitedcontrols.com

Repair Report 3697, Revision 0

4) Powered Tests with 125 VDC and 24VDC inputs: (cont) .: .,r ** : "'., '.: .: "'~I:

.. . .....,. :c', Vibratiol)' and H,~Clt. Stress - T~e car.d was subjected to mechanical, ::",'" "':'

':", " ::shock duringtirning ~eque'nces with no reset occu~ring. App'iying<::::.. ,,;, ,",

pressure to and heating of components with hot air did not induce.

~ 'any timing failures or resets., " ':,

d. EMI Sellsitivity - While the card was timing at elevated:: , ' .. '

temperatu're, a brushed ac motor was brought into close proximity, ' ", ','"

to the card to induce noise. The card performed normally.

"e. Watchdog Timer Circuit:- The processor was contin,uously resetting' "

the watchdog ,timer every 5 milli-seconds. That is twice as fast as needed and'no marginal timing was noted.

f. Output Disable Circuit - This circuit shuts off all output drivers if a

.watchdqg, tim$r, or, 1/0' error 099urs; Output maintained proper'.

saturation, voltag,e levels during timing.

g. External Reset -: Measured the voltage levels on the output of. the opto-coupler of the external 125 VDC Reset signal to ensure the' phototrarisistor was switching between +5 VDC and Ground.

Excessive leakage of this transistor could cause false reset signals:

The levels were well within specifications of the CMOS 4049 receiver.

h. Low Voltage Operation -: The 125 VDC and 24 VDC power supplies were lowered to 1DOVDC and 20VDC respectively during timing sequences to see if power supply droop would cause a reset. The card functioned properly through every timer sequence.

As Left Condition: Since the failure could not be replicated, no repairs were performed on this board. The as left condition ,is the same as the as found condition.

Conclusion:

There is no apparent reason for the card to have failed the timing sequence,' and' the failure could not be duplicated even when stressed to help reveal a problem with the card. Since the external 125 VDC Reset signal has precedence over operation ~nd will immediately reset the outputs in the middle of, a timing sequence, it is a possibility that a 125 VDC'Reset was somehow'generated.,

Another possibility could be a' severe, short duration drop in the 24 VDC and/or on-card +5 VDC power supply voltage that caused a legitimate processor reset. '

, ~

Prepared By: frtlr1#-vS~Yf¥\A f2../~3Jo&tReviewed By: ~ ,a~lo")

UCI Engineering UCI Engineering Approved By: (\A:? l~ I?/2j>/o:J UCI Quality Assurance Page 2 of2

, United Controls International"" 5139 S.Royal Atlanta Drive,." Tucker, GA 30084

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