Text

Revision 29 to Plant Operation Manual, EM-225, Duties of the Technical Support Center Accident Assessment Team.
	ML14036A206
Person / Time
Site:	Crystal River
Issue date:	01/27/2014
From:	; Duke Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	FOIA/PA-2015-0005 EM-225, Rev 29
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PASSPORT DOCUMENT To NRC NRC Facility CR3 Department : TRANSMITTAL Address CR3-01242 / MAIL CODE: N/A DC DESK-ATTN: DENNIS HAGAN SAFEGUARDS TO B.GAPP-NA2S Page: 1 From Address :

CR3DOCSVCS DUKE ENERGY FLORIDA CRYSTAL RIVER COMPLEX Attention:

15760 WEST POWERLINE STREET DOCSVCS NAlE / PLNTSUPT NR2 1l1111 11111 Ilill I111 1 City : CRYSTAL RIVER State: FL Postal Code: 34428-6708 Country : UNITED STATES Email Contact Date/Time 01/27/2014 15:20 Transmittal Group Id: 0000082334 Trans No. 000568872

Title:

Total Items: 00001 Item Facility Type Sub Document Number Sheet Doc Status Revision Doc Date Copy # Media Cpys 0001 CR3 POM EMG EM0225 ACTIVE 029 H 01 If a document was not received or is no longer required check the response below and return to sender.

I Documents noted above not received (identify those not received).

I no longer require distribution of these documents (identify those no longer required).

Date: Signature:

1045 A11-4

I DUKE C*ol)ENERGY. Information Use CRYSTAL RIVER UNIT 3 PLANT OPERATING MANUAL EM-225 DUTIES OF THE TECHNICAL SUPPORT CENTER ACCIDENT ASSESSMENT TEAM REVISION 29

TABLE OF CONTENTS

. SECTION PAGE 1 .0 P UR P O S E ............................................................................................................................. 3

2.0 REFERENCES

...................................................................................................................... 3 2.1 Developmental References ................................................................................... 3 2.2 Im plementing References ...................................................................................... 5 3.0 DEFINITIONS ....................................................................................................................... 6 4.0 RESPONSIBILITIES ..................................................................................................... 7 5.0 PREREQUISITES ............................................................................................................... 11 6.0 PRECAUTIONS, LIM ITATIONS AND NOTES ................................................................ 11 7.0 SPECIAL TOOLS AND EQUIPM ENT ............................................................................. 11 8.0 ACCEPTANCE CRITERIA ............................................................................................. 11 9.0 INSTRUCTIONS ................................................................................................................. 11 9.1 Accident Assessm ent Initiation ............................................................................. 11 10.0 RECORDS .......................................................................................................................... 11 ENCLOSURES (All Enclosures are "Optional Records Non-Quality")

1 TSC Guidance for EOPs ..................................................................................... 12 2 EDG Scenarios to Establish Adequate Loading ................................................... 20 ATTACHMENTS 1 AAT Coordinator Checklist ................................................................................. 21 2 TSC Briefing Guideline ........................................................................................ 23 3 TSC Ringdown Com m unicator Checklist ............................................................ 24 4 AAT Operations Support Checklist ....................................................................... 25 5 AAT Engineers Checklist ...................................................................................... 28 6 Control Room Ringdown Com m unicator Checklist ............................................... 29 7 NRC Com m unicator Checklist ............................................................................. 30 8 Critical Safety Function Checklist ......................................................................... 31 9 Dose Assessment Team Notification .................................................................. 37 10 Core Damage Assessm ent .................................................................................. 40 11 OSC Request Form ............................................................................................ 45 12 SPDS or RECALL Display Setup for TSC Projection Screens ............................ 46 Sum mary of Changes .................................................................................................................... 47 EM-225 Rev. 29 Page 2 of 47

1.0 PURPOSE

1. This procedure provides guidance for the establishment and operation of the Technical Support Center Accident Assessment Team (AAT), for the determination of core and fission product barrier status, and for the interface with the Radiation Controls Coordinator. Information from these assessments will be used in conjunction with other guidance for development of accident mitigation strategies. This procedure also provides guidance to the AAT to perform actions described in the EOPs [NOCS 062718].

2. This procedure is an emergency plan implementing procedure. Any revisions to this procedure must be carefully considered for emergency plan impact.

2.0 REFERENCES

2.1 Developmental References

1. Response Technical Manual (RTM-96); USNRC; Volume 1, Rev. 3

2. Radiological Emergency Response Plan

3. Emergency Operating Procedures (EOPs)

4. NUREG-1228, Source Term Estimation during Incident Response to Severe Nuclear Power Plant Accidents

5. B&W Technical Bases Document

6. FPC IOC CR97-0122, Dated 12/23/97

7. NEI 91-04, Revision 1, Severe Accident Issue Closure Guidelines

8. FPC IOC SE99-0184, Dated 9/14/99

9. EEM-99-018, Rev. 0 Operating Limits for SWP-1A/SWP-1B under Minimum Flow Conditions

10. EM-202, Duties of the Emergency Coordinator

11. EM-102, Operation of Technical Support Center

12. EM-1 03, Operation and Staffing of the CR-3 Control Room During Emergency Classification

13. CP-151, External Reporting Requirements

14. Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors

15. EC 58982, RB Sump Strainer Modification

16. EC 59476, RB Sump Level Instrumentation Modifications

17. EC 55315, Alternate AC Diesel Generator

18. EC 66671, Installation of CR3 Intrusion Detection System (Firewall) and Refinement of the CR3 Cyber Security Defensive Model

19. INPO IER 11-2, Fukushima Daiichi Nuclear Station Sent Fuel Pool Loss of Cooling and Makeup

20. 10CFR50.54(x) or Section 24 of the Physical Security Plan A

21. 1 OCFR50.72(a)(1)(i) and 10CFR50.72(b)(2)(i) a V

EM-225 Rev. 29 Page 3 of 47

Subsection 2.1, Developmental References (Cont'd)

22. EEM-01-021

23. FSAR Table 4-10

24. IOC SE-99-0184

25. NOCS 62718, 62764, 62767, 96042, 100056, 100408, 100441, 100483, and 100568

26. IER 11-46 Extended Emergency Power Operations Following A Loss Of Off-Site Power

27. EC 84553, Minimum Expected EGDG Electrical Loads For R16 Extended Outage

28. Calc M89-0063, Waste Gas Decay Tank Rupture Environmental Condition

29. EC 92486, CR3 Response to NRC Request for Information Regarding Rescinding Fukushima Order

30. EC 76363, Radiation Monitors RM-A1/RM-A2 Replacement EM-225 Rev. 29 Page 4 of 47

2.2 Implementing References

1. CR-3 Severe Accident Guideline

2. Emergency Response Personnel Roster

3. AP-770, Emergency Diesel Generator Actuation

4. AP-990, Shutdown from Outside the Control Room

5. CH-632, Post Accident Sampling and Analysis of the Reactor Coolant, Decay Heat, and Reactor Building Sump

6. CP-151, External Reporting Requirements

7. EMG-NGGC-0002, Off-Site Dose Assessment

8. EM-103, Operation and staffing of CR-3 Control Room during Emergency

9. EM-202, Duties of the Emergency Coordinator

10. EM-206, Emergency Response Organization Notification

11. EM-225A, Post Accident RB Hydrogen Control

12. EM-225B, Post-Accident Boron Concentration Management

13. EM-225C, Post Accident Monitoring for Reactor Building Temperature

14. EM-225D, Guidance for Dry OTSG Tube to Shell Delta T Monitoring and Control

15. EM-225E, Guidelines for Long Term Cooling

16. EM-225F, Long Term Emergency Feedwater Management

17. EOP-3, Inadequate Subcooling Margin

18. EOP-5, Excessive Heat Transfer

19. EOP-6, Steam Generator Tube Rupture

20. EOP-7, Inadequate Core Cooling

21. EOP-8A, Loca Cooldown

22. EOP-8B, HPI Cooldown

23. EOP-9, Natural Circulation Cooldown

24. EOP-12, Station Blackout

25. EOP-14, Emergency Operating Procedure Enclosures

26. MP-575, Hydrogen Recombiner Installation

27. MP-815, Installation of Post Accident H2 Purge Flow Instruments

28. OP-1 03C, Cycle 17 Reactivity Worth Curves

29. OP-41 7B, Operation of the Post Accident Hydrogen Recombiner

30. SP-306, Routine Surveillance Log EM-225 Rev. 29 ae o

3.0 DEFINITIONS

1. Accident Assessment Team (AAT): Consists of Coordinator, TSC Ringdown Communicator, Control Room Ringdown Communicator, Engineer, Operations Support, and NRC Communicator.

2. Candidate High Level Actions (CHLA): Actions described in the CR-3 Severe Accident Guideline which could be taken to mitigate a Severe Accident and are deemed appropriate based on Plant Damage Conditions.

3. Critical Safety Functions (CSFs): Those functions needed to ensure adequate core cooling and to preserve the integrity of the fission product barriers thereby protecting the health and safety of the general public and plant personnel. They include:

reactivity control, coolant inventory control, decay heat removal capability, fission product barrier status, electrical power availability and control complex status.

4. Inadequate Core Cooling: Accident conditions that result in a loss of core cooling that requires entering EOP-7, Inadequate Core Cooling.

5. Emergency Action Levels (EALs): Conditions or indications that may be used as thresholds for initiating specific emergency measures (see EM-202, Duties of the Emergency Coordinator, Enclosure 1).

6. Plant Damage Conditions (PDC): Damage conditions used in the CR-3 Severe Accident Guideline to describe the status of the reactor coolant system, reactor core, and the containment during the progression of a Severe Accident.

7. Protective Action Recommendations (PARs): Emergency measures recommended for purposes of preventing or minimizing radiological exposures to the Energy Complex personnel or members of the general public.

8. Severe Accident: An accident (beyond that assumed in the CR-3 design and licensing basis) that results in catastrophic fuel rod failure, core degradation and fission product release into the Rx vessel, Reactor Building or the environment.

9. Full HPI: The conditions necessary to ensure > the minimum required HPI flow assumed in the plant design basis. These conditions include: at least 1 MUP running with HPI flow through all 4 HPI nozzles (all 4 HPI valves open, or HPI crossties open with 1 train of HPI valves open) with one of the following:

HPI recirc to sump, MUP recirc, MU flowpath to the RCS, and RCP seal injection flowpaths isolated.

Total HPI flow is in the "Acceptable Region" of the "Minimum Required HPI Flow" figure.

10. Less Than Full HPI (Inadequate HPI flow, maximum cooldown in progress per EOP-3, Inadequate Subcooling Margin): Not all portions of the HPI flow path satisfy the independence criteria discussed in the CR3 ITS. Specifically, the HPI flow path downstream of the HPI/Makeup pumps is not separable into two distinct trains, and is therefore, not independent. As such, in the event of a postulated break in the HPI injection piping, injection flow is required through one of the following alignments:

A minimum of three (3) intact injection legs, assuming one pump operation

A minimum of two (2) intact injection legs, assuming two HPI pump operation.

a w

EM-225 Rev. 29 Page 6 of 47

4.0 RESPONSIBILITIES

1. Control Room Ringdown Communicator:

Reports to the Control Room and establishes communication with the TSC Ringdown Communicator on the Accident Assessment Ringdown phone. Brief TSC Ringdown Communicator on operator actions that are in progress.

Relays status of overall plant conditions, operator activities and questions to the TSC AAT.

Relays instructions to Control Room Operators for mitigating actions as directed by the Emergency Coordinator (EC).

If a Severe Accident is occurring, directs Control Room personnel regarding actions to take to mitigate the Severe Accident, based on actions approved by the TSC EC.

Relay request for support from the Control Room to OSC teams, via TSC Ringdown Communicator.

Once TSC is operational, request extra plant operators (if available) be sent to OSC for in plant support.

Inform TSC of in plant operator actions that are being performed.

EM-225 Rev. 29 Page 7 of 47

Section 4.0, RESPONSIBILITIES (Cont'd)

2. AAT Coordinator:

Informs the EC of any developments in plant status that may impact EALs and PARs.

Ensures appropriate AAT personnel have staffed the TSC.

Ensures additional AAT members are notified as needed.

Identifies plant parameters to be tracked.

Coordinates AAT activities and ensures that team members remain focused on objectives.

Keeps the EC informed of AAT activities.

If a Severe Accident is occurring, reviews recommended Candidate High Level Actions and mitigation plans prior to submitting to the Emergency Coordinator.

[NOCS 100056]

If a Severe Accident is occurring, coordinates efforts of the Accident Assessment team to ensure the development of mitigation strategies using the CR-3 Severe Accident Guideline.

If additional resources are needed, coordinates with the EOF Technical Support Team to provide required support.

Establishes communications with the Emergency Operating Facility (EOF)

Technical Support Team, if the EOF is staffed.

Approve Attachment 11, OSC Request Form to request operator actions outside CCHE or maintenance repair activities that have been initiated by the Control Room or AAT. This request should be processed through the OSC Manager to the OSC.

Notifies Shift Manager in requal training for additional Operations support.

Ensures TSC display screen computers are logged in. If computer room door is locked, contact Security for access.

3. TSC Ringdown Communicator:

Establishes communications with the Control Room Ringdown Communicator on the Accident Assessment Ringdown phone.

Relays information on changing radiological conditions and maintenance activities to the Control Room.

Relays plant conditions from the Control Room to the TSC AAT.

Maintains the Accident Assessment Team Log.

Relays information and directions to the Control Room of actions required to mitigate a Severe Accident based on approved Candidate High Level Actions.

Monitors progression through EOPs and APs.

Initiate Attachment 11, OSC Request Form to request operator actions outside CCHE or maintenance repair activities for the OSC that is requested by the Control Room or AAT.

EM-225 Rev. 29 Page 8 of 47

Section 4.0, RESPONSIBILITIES (Cont'd)

4. AAT Engineers:

Assesses plant conditions and provides engineering support for developing accident mitigation strategies as needed.

0 Aids in determining additional Engineering resources.

Monitors plant parameters for indications of core damage and status of fission product barriers.

During Severe Accident conditions, evaluates plant parameters, determines Plant Damage Conditions, and develops Candidate High Level Action recommendations using appropriate calculational aids from the CR-3 Severe Accident Guideline.

5. AAT Operations Support:

Monitors overall plant status during an emergency with emphasis on Critical Safety Functions.

Functions as a technical resource for Operations in assessing plant conditions and in development of accident mitigation strategies that are outside the scope of Emergency Operating Procedures (EOPs).

0 Maintains the CSF Status Board at the TSC.

0 During Severe Accident Conditions, provides support to the AAT Engineers in determining Plant Damage Conditions and developing mitigation strategies using the CR-3 Severe Accident Guideline.

Coordinates/processes requests for operator actions or maintenance support activities through the OSC Manager using Attachment 11, OSC Request Form.

Determine emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements.

6. NRC Communicator: [NOCS 96042]

Maintains an open, continuous communication line on the Emergency Notification System with the NRC Operations Center upon request by the Headquarters Operations Officer.

Log times NRC is notified of Emergency Classification changes and Protective Action Recommendations.

Make emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements.

EM-225 Rev. 29 Page 9 of 47

Section 4.0, RESPONSIBILITIES (Cont'd)

7. EOF Technical Support Team:

Functions as a technical resource for the EOF Director in development of PARs by monitoring plant conditions (particularly the CSFs).

Assists the TSC AAT team as needed in development of mitigation strategies and in research of solutions to plant problems.

Responsible for the development of long-term recovery plans.

8. Emergency Coordinator (EC) or designee:

Controls all activities at CR-3 during activation of the Radiological Emergency Response Plan.

Implements EM-202, Duties of the Emergency Coordinator.

0 Determines EAL and PAR changes based on information obtained from the Accident Assessment Team and Radiation Controls Coordinator.

Functions as the decision maker during a Severe Accident. The EC will approve all recommended Severe Accident mitigation strategies prior to implementation.

Is authorized to declare 10CFR50.54(x and y) to implement emergency actions deemed necessary to protect the health and safety of the public. A separate notification is required to the NRC for each occasion. Once a Severe Accident is declared, only one notification to the NRC is required.

9. Radiation Controls Coordinator:

Supports the Accident Assessment team with on-site radiological data and with chemical and radiological analysis of samples as needed to assess the accident.

Provides Plant Radiation Monitor readings and assessments.

Provides projected radiological data (on-site and off-site doses, dose rates, and deposition) (> 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to obtain).

Provides capability to obtain RCS samples for boron concentration.

Provides capability to obtain grab samples for RB Atmosphere and RB/AB Vent.

Provides in-plant radiological data.

Provides chemical and radiological analysis of OTSGs and secondary samples.

Provides Reactor Building sump boron concentration (> 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to obtain).

.EM-225 Rev. 29 Page 10 of 47

5.0 PREREQUISITES None 6.0 PRECAUTIONS, LIMITATIONS AND NOTES Under Severe Accident Conditions, plant instrumentation may provide false or highly inaccurate readings due to harsh environments beyond their qualifications.

Several instruments should be monitored along with trends to assess plant conditions.

7.0 SPECIAL TOOLS AND EQUIPMENT None 8.0 ACCEPTANCE CRITERIA None 9.0 INSTRUCTIONS NOTE: EM-225 attachments have been pre-marked with not applicable areas or known conditions based on a defueled reactor and Waste Gas Decay Tanks purged.

9.1 Accident Assessment Initiation

1. [AAT Coordinator or designee] PERFORM the duties of Attachment 1, AAT Coordinator Checklist.

2. [TSC Ringdown Communicator] PERFORM the duties of Attachment 3, TSC Ringdown Communicator Checklist.

3. [AAT Operations Support member] PERFORM the duties of Attachment 4, AAT Operations Support Checklist.

4. [AAT Engineers] PERFORM the duties of Attachment 5, AAT Engineers Checklist.

5. [Control Room Ringdown Communicator] REPORT to the Control Room AND PERFORM the duties of Attachment 6, Control Room Ringdown Communicator Checklist.

6. [NRC Communicator] PERFORM the duties of Attachment 7, NRC Communicator Checklist.

10.0 RECORDS All attachments are quality records EM-225 Rev. 29 Page 11 of 47

ENCLOSURE 1 Page 1 of 8 TSC GUIDANCE FOR EOPS

[NOCS 62718, 62764, 62767]

This enclosure provides the relationship with the EOPs and TSC guidance during emergency events. It is management's expectation that the guidance steps will be implemented, based on the emergency condition of the plant, by either invoking 10 CFR 50.54 (x), (y),

formal 10 CFR 50.59 reviews and approvals, or by existing approved procedures.

PARAMETER EOP TBD REF. TS'C GUIDANCE RB Hydrogen EOP-3, EOP-6, HPIC, 5.4 1. Align hydrogen monitoring equipment using EOP-14, Enclosure 2, PPO Control EOP-7, EOP-8A, Post Event Actions.

Ill.F, 6.2,10.0, 12.6b, EOP-8B 13.6b 2. Monitor hydrogen concentrations using EOP-14, Enclosure 21, RB LBLO 4.4, 6.3 Hydrogen Monitor Log.

SBLO 12.4, 20.3, 9.3 3. Purge RB when authorized per EM-225A. [NOCS 62767]

Interfacing references are:

EM-206 for telephone number for procurement representative to obtain recombiners

" MP-575 for installation of recombiners

OP-417B for operation of recombiners

MP-815 for installina H2 Durae flow indicators Building EOP-3, None If RB sump strainer blockage occurs consider alternate criteria for BSP Spray EOP-8A, shutdown (See EM-225E, Section 9.6)

Termination EOP-8B Verify all of the following before terminating Building Spray:

Criteria EOP-14,Enc 19 1. BS has been on for > or equal to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

2. RB pressure is < 10 psig.

3. RB pressure is stable or lowering.

4. RB atmosphere is < 13 lpci/cc 1-131.

5. RB temperature is stable or lowering (also refer to EM-225C).

6. Concurrence is obtained from EC and Dose Assessment to terminate BS.

SFP Level EOP-06, VI-IIIE Perform EOP-14 Enc 24, Monitoring Spent Fuel Parameters and EOP-8A, VI-IIIA Interfacing references are:

Temperature EOP-8B, VI-IVA 1. AP-406, Loss of SFP Cooling Trending EOP-10 2. AP-1080, Refueling Canal, SFP level Lowering EOP-12 3. AAG-05, Contingencies for Loss of SFP Level EOP-14 Enc 24 EM-225 Rev. 29 Page 12 of 471

ENCLOSURE 1 Page 2 of 8 TSC GUIDANCE FOR EOPs Continue EOP-6, FF, 11.5 Verify all of the following:

Cooldown With EOP-8A, NC, 11.4 1. Begin establishing a Post Accident Recovery Plan (this can be done during plant DHR System EOP-8B cooldown).

2. The reactor is being cooled by DHR.

3. DHR cooling is consistent with maintaining adequate SCM.

4. The RCS is subcooled (use DH cooler outlet temperature for cooldown rates).

5. The RCS is depressurized.

6. Prohibit establishing any flow path that was isolated by the ES system unless the potential for radioactive releases is evaluated and the release path, doses, and methods have been approved by the EC.

7. Control of containment penetrations has been established.

8. Monitor and maintain RCS boron concentration for required shutdown margin.

Steaming an EOP-6 III.E Steaming an affected OTSG may be desirable for the following reasons:

isolated OSTG 0 Increase cooldown rate for TRACC 0 Prevent challenging tube to shell dT limits a Prevent idle loop voiding when in natural circulation.

All of the following conditions should be evaluated to determine if steaming an affected OTSG is appropriate:

1) BWST > 35 ft (1)

AND

2) Affected OTSG Level < 90%(2)

AND

3) Any of the following conditions exists:

" Steaming is required to avoid core damage

1. Estimated OTSG leakage times RCS DE 1-131 concentrations is < 0.4 OTSG Leakage (gpm) X Initial RCS DE 1-131 (gci/gm) < 0.4

" Wind is blowing off-shore (Off-shore winds originate from NNE to SE sectors 011.2* to 146.30)

Note (1) - If BWST level is < 35 ft, then determine if adequate BWST level is available for long term cooldown (Ref calc M89-1089) prior to steaming the OTSG.

(2) - If OTSG level is > 90%, then determine if OTSG level is low enough to prevent water carry-over. As long as water level can be ensured to be below the bottom of the main steam outlet nozzles there should not be any carry-over concern.

EM-225 Rev. 29 Page 13 of 471

ENCLOSURE 1 Page 3 of 8 TSC GUIDANCE FOR EOPs BWST Makeup EOP-6 lII.E Monitor BWST level trend and evaluate depletion rate. Ensure adequate BWST inventory is available to support RCS cooldown to DHR. Evaluation should include the following:

" Primary to secondary leak rate

" BWST available inventory

BWST depletion rate

" Current RCS temperature

BWST volume required to support cooldown (refer to OP-304)

Potential for leak rate increase (leak before break)

IF ECCS water supplies are insufficient to support cooldown to DHR, THEN, make preparations to initiate BWST makeup from spent fuel pools.

a Refer to EM-225E, Enclosure 11, BWST Refill from Spent Fuel Pool RCS Leakage No EOP-8A, None 1. The RCS is capable of being cooled by DHR.

Longer Exists EOP-8B 2. Prohibit establishing any flow path that was isolated by the ES system unless the potential for radioactive releases is evaluated and the release path, doses, and methods have been approved by the EC.

3. Begin DHR.

Break size > 1 HPI EOP-8A, None 1. Establish a Post Accident Recovery Plan. This plan is dependent on the Pump Capability or EOP-8B scope of the applicable Emergency Event.

Unable to transition 2. The Post Accident Recovery Plan is approved by the PNSC, and applicable to DHR regulatory agencies as determined by FPC Management.

3. Prohibit establishing any flow path that was isolated by the ES system unless the potential for radioactive releases is evaluated and the release path, doses, and methods have been approved by the EC.

4. The availability of borated water sources for required shutdown margin is maintained until the actions of the Post Accident Recovery Plan are completed or to the extent that plant and public safety is ensured.

5. Post and label protected train boundaries for the borated water sources and components that are available.

EM-225 Rev. 29 Page 14 of 47]

0 ENCLOSURE 1 Page 4 of 8 TSC GUIDANCE FOR EOPs EOP-8A I. I ransluon to uLr'* cooiuuwn.

EOP-8B 2. Establish a Post Accident Recovery Plan. This plan is dependent on the scope of the applicable Emergency Event.

Pump Capability and 3. The Post Accident Recovery Plan is approved by the PNSC, and applicable able to transition to DHR regulatory agencies as determined by FPC Management.

4. Prohibit establishing any flow path that was isolated by the ES system unless the potential for radioactive releases is evaluated and the release path, doses, and methods have been approved by the EC.

5. The availability of borated water sources for required shutdown margin is maintained until the actions of the Post Accident Recovery Plan are completed or to the extent that plant and public safety is ensured.

6. Post and label protected train boundaries for the borated water sources and components that are available.

Establishing Primary to SS-2 1. Refer to the entry conditions and recommendations of the Emergency Operating Secondary Heat Transfer Procedures Technical Basis Document (TBD), Section SS-2 for guidance related to One or Both OTSGs to establishing primary to secondary heat transfer to one or both OTSGs.

2. Accident Assessment personnel in the TSC will provide recommended guidance to the EC for when and how to establish heat transfer using one or both OTSGs.

3. The EC will approve any actions recommended.

Termination of HPI and EOP-8A LBLO, 2.2, 3.0 1. Recommended guidance is to stop HPI pumps and trip running RCPs when LPI Shutdown of RCPs EOP-8B flow has been in excess of 1400 gpm in each injection line for at least 20 minutes.

Accident Assessment personnel will evaluate plant conditions and provide recommendations to the EC.

2. The EC will approve any actions recommended.

Control of Radioactive EOP-8A SBLO 12.0 1. Prohibit establishing any flow path that was isolated by the ES system unless the Release Paths from EOP-8B potential for radioactive releases is evaluated and the release path, doses and Containment Penetration methods have been approved by the EC.

Valves Monitoring of RB Sump EOP-8A None NOTE: With the installation of the TSP baskets, pH data is not required but still Level, RB Sump Boron EOP-8B desired iffeasible.

Concentration, RB Other: IOC CR 1. Accident Assessment personnel to monitor and trend RB sump level, boron Sump pH and RB Sump 97-0122 concentration, pH and RB Sump strainer AP at intervals recommended by the EC.

strainer AP 2. Data for sump pH and boron concentration to be obtained using CH-632 or other PNSC approved alternate methods dependent on the Emergency Event.

EM-225 Rev. 29 Page 15 of 47

0 ENCLOSURE 1 Page 5 of 8 TSC GUIDANCE FOR EOPs Venting of Non- EOP-8A None 1. Once subcooling margin is regained, all of the noncondensable gas Condensable Gases EOP-8B production will have ceased. However, as the RCS is depressurized these gases will come out of solution and should be vented. If natural circulation is lost to an available OTSG, Accident Assessment personnel will recommend to the EC when to vent noncondensable gases.

2. The EC will approve any actions recommended.

Reactor is Being EOP-8A SBLO, 17.7 1. Verify TBVs/ADVs are closed.

Adequately Cooled EOP-8B 2. Fill available OTSGs to 90%.

Using HPI or LPI and 3. Close EFW/AFW/MFW Valves.

OTSG Cooling is No 4. Stop all EFW/AFW Pumps.

Longer Desired 5. Stop MFWPs and MFWBPs.

Boron Concentration EOP-8A None Refer to EM-225B Management When EOP-8B Adequate Sub NOTE: If a failure of ES MCC 3AB has occurred, ensure repair efforts are Cooling Margin Does EOP-14, initiated to repower auxiliary pressurizer spray valve RCV-53 prior to the Not Exist (Boron Enc. 20 onset of boron precipitation.

Precipitation)

RB Temperature Refer to EM-225C Monitoring (To Preserve EQ Standards)

Feeding a Dry OTSG EOP-5, Ill.D, 12.0 Refer to EM-225D (Tube to Shell EOP-9, Ill.E, 17.7 Delta T Monitoring EOP-14, NC, 5.2, 5.3, 6.4 and Control) Enc. 3 Long-Term Core EOP-8A LBLO, 6.4a, 6.4b, 6.6, Refer to EM-225E Cooling Using the RB EOP-8B 6.7 Sump EFW or AFW is EOP-14, Refer to EM-225F Operating Enc. 7 Enc. 22 rEM-225 Rev. 29 Page 16 of 47

0 ENCLOSURE 1 Page 6 of 8 TSC GUIDANCE FOR EOPs TBP-3 is Running. EOP-14, TBP-3 will drain non-l E battery during LOOP. Stopping TBP-3 before 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months may TBP-2 is Not Running. Enc. 14 result in Turbine bearing damage.

Generator Purge Complete Refer to IOC SE-99-0184 Concentrated BA EOP-14, None If concentrated BA is allowed to remain in the boron injection path piping addition made and flush Enc 18 (letdown/DH purification piping) the BA will eventually cool down and solidify. Timely water not available. action is required to preclude this condition.

1. Direct the control room to reestablish a continuous BA injection at a flow rate of 2 - 3 GPM (Batch controller is the preferred method).

2. Monitor RCS boron concentration. DO NOT allow RCS boron concentration to exceed the values listed in FSAR Table 4-10.

3. Evaluate the following options.

" If plant conditions permit, expedite restoration of RCS letdown (or DH purification).

" If plant condition permit expedite restoration of power to at least one source of flush water (DWP-1A, DWP-1 B, WDP-5A, WDP-5B, or WDP-5C).

" If BA flow rate and AB temperature conditions permit, evaluate securing continuous BA addition and performing periodic batch additions to prevent boron solidification.

4. IF letdown or DH purification flow is established, THEN direct the control room to STOP concentrated BA additions.

5. IF any flush water source becomes available, THEN direct the control room to STOP concentrated BA additions and perform a line flush using EOP-14, Enclosure 18.

Refer to EEM-01-021, FSAR Table 4-10 IEM-225 I Rev. 29 Page 17 of 47

ENCLOSURE 1 Page 7 of 8 TSC GUIDANCE FOR EOPs Indications of RB EOP-14, None ECCS pumps have been aligned to the RB sump and are now showing signs of sump sump strainer Enc. 19 strainer blockage (flow oscillations, pump amp swings, high RB strainer AP). Per EOP-14, blockage have Enclosure 19, ECCS Suction Transfer, LPI flows should have been reduced to 1400 gpm occurred. per pump. At least one BSP should be secured. If both trains of LPI are in service HPI should be secured. If only one train of LPI is in service, one train of HPI must be aligned to the operable LPI pump in piggy back mode. [NOCS 100483 and 100408]

1. Verify proper ECCS pump configuration

2. Closely monitor ECCS pump parameters, Incore temperatures and RB sump strainer AP (Ref. Recall Point 79).

3. Expedite BWST refill operation from spent fuel pool using EM-225E Enclosures 11 and 12.

4. Expedite mixing of boric acid for BAST makeup per OP-403B, Section 4.2, Boric Acid Production.

5. Refer to EM-225E, Section 9.6, Contingency Actions for RB Sump Strainer Blockage, for specific guidance.

Non-Vital Battery EOP-12 None During operation of the Alternate AC Diesel the potential exists for hydrogen accumulation Hydrogen AP-770 in the Non-vital battery room. A portable fan must be set up within the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of continuous operation of the Alternate AC Diesel to promote ventilation and the dilution of any hydrogen gas. A 120 VAC duplex receptacle exists adjacent to ACDP-1 76.

CFT isolation not EOP-3 EOP-3, EOP-8A and EOP-12 have requirements to maintain RCS pressure above 140 psig closed, RCS EOP-8A if LPI is not available and CFT isolation valves are not closed to prevent the CFT from depressurizing EOP-1 2 further being depleted. The continued depletion of the CFT below an RCS pressure of 140 psig can allow nitrogen gas used to pressurize the CFTs to enter the RCS. This nitrogen intrusion may result in voiding and interfere with primary to secondary heat transfer (Natural Circulation). Maintain RCS pressure > 140psig, re-establish LPI or close the CFT isolation valves.

EM-225 Rev. 29 Page 18 of 47]

ENCLOSURE 1 Page 8 of 8 TSC GUIDANCE FOR EOPs PARAMETER I EOP I TBD REF I TSC GUIDANCE Station Blackout EOP-12 None The AAC Diesel has more limitations then the ES Diesels based on additional loads on Guidance with only 4160 Rx Aux Bus 3 (i.e. Non-IE battery chargers). The AAC Diesel does not have the AAC Diesel capability to load sequence and therefore ES is blocked from actuating when the AAC supplying power to Diesel is powering the ES 4160V Bus. All actions for mitigating the event must be ES 4160V Bus. (no done manually. Actions outside of EOP-12 or procedures not referenced by EOP-12 ES Diesel or offsite may not work for all situations. Based on the situations which are beyond the scope of power available)

EOP-12 (i.e. LOCA, inadequate heat transfer, excessive heat transfer, SGTR, etc) the MAT needs to evaluate the specific guidance within other EOPs to develop mitigation strategy for the specific conditions. Based on the duration of the loss of offsite power or ES Diesels the following parameters need to be monitored and additional guidance given:

" Monitor EFT depletion rate and heatup rate. Based on the depletion of EFT-2 give guidance to establish alternate EFW sources from the CST, FST or hotwell using cross-connect lines between the tanks if necessary (Ref EOP-14 Enclosures 22, Secondary Inventory Management). Refer to EM-225F.

Monitor BWST depletion rate. Based on BWST depletion rate establish guidance for transferring the ECCS or MUP suction to the RB sump if necessary (Ref EOP-14 Enclosure 19, ECCS Suction Transfer). Refer to EM-225E for guidance for filling BWST or allowable BWST level to support only a makeup pump.

Monitor Containment Temperatures. Ensure adequate RB cooling is being maintained. Refer to EM-225C.

Inadequate HPI EOP-3 If multiple component/equipment failures or plant configuration results in inadequate HPI flow, maximum flow, then an alternate evaluation of less than FULL HPI flow can be considered based on cooldown in ITS 3.5.2 bases, which states that the following injection flow path options can provide progress. adequate HPI flow without meeting the configuration requirements of FULL HPI, as described in the definition section of this procedure:

0 A minimum of three (3) intact injection legs, assuming one pump operation

A minimum of two (2) intact injection legs, assuming two HPI pump operation.

EM-225 Rev. 29 Page 19 of 47

ENCLOSURE 2 Page 1 of 1 EDG SCENARIOS TO ESTABLISH ADEQUATE LOADING The following pumps and fans combinations will place an electrical loading on the EDGs greater than 600 kW (Reference EC 84553).

Case 1 Case 2 EGDG-1A Operatingq Loads EGDG-1A Operatingq Loads Potential EDG Loading total - 686.0 KW Potential EDG Loading total - 689.4 KW

272.9 KW - SWP-1A running 9 63.4 KW - DCP-1A running

383.3 KW - RWP-2A running (Note 1) 0 160.1 KW - RWP-3A running

29.8 KW - SFP-1A running

215.4 KW - DHP-1A in recirculation mode to BWST at 3000 gpm (Note 2)

155.4 KW - BSP-1A in recirculation mode to BWST at 1500 gpm (Note 2)

95.1 KW- AHF-1A orAHF-1C running in high speed EGDG-1B Operatingq Loads EGDG-1B Operating Loads Potential EDG Loading total - 689.4 KW Potential EDG Loading total - 686.0 KW e 63.4 KW - DCP-1B running 0 272.9 KW - SWP-1B running

160.1 KW- RWP-3B running 9 383.3 KW- RWP-2B running (Note 1) e 215.4 KW - DHP-1B in recirculation mode 0 29.8 KW - SFP-lB running to BWST at 3000 gpm (Note 2)

155.4 KW - BSP-1B in recirculation mode to BWST at 1500 gpm (Note 2)

95.1 KW-AHF-1B orAHF-1C running in high speed Note: (1) Operating both redundant pumps together can affect EDG loading in some cases due to load sharing between pumps (example: RWP-2A and RWP-2B).

(2) Monitor BWST water temperature to maintain below 92.5 deg F. Use of DHHE to cool the BWST may be required.

Other identified loads that should be available for operation to further increase the load if required or replace loads (example: BSP-1A/1B secured due to BWST temperature). Reference EC 84553 evaluation.

Load Nameplate 85% Nameplate Comments KW/HP KW CHHE-1A/1B 194 kW 164.9 Electrical load is dependent on the CC heat load.

AHF-17A/17B 60 HP 38.0 AHF-18A/18B AHF-19A/19B 20 HP 12.7 AHF-24A/24B 15 HP 9.5 AHF-29A/29B 10 HP 6.3 EM-225 Rev. 29 Page 20 of 47

ATTACHMENT 1 Page 1 of 2 AAT COORDINATOR CHECKLIST NOTE: Attachment steps can be completed in any order .................................... E.

1. BADGE IN at TSC card reader AND PLACE name on TSC Staffing Board ............. El

2. NOTIFY the EC that the Accident Assessment Team is operational when ALL of the following are accomplished:

INITIATE Critical Safety Functions evaluation (Attachment 2, T S C B riefing G uideline) ................................................................................. El

ESTABLISHED Communication via phone link with the control room or ability to monitor plant via computer (e.g. SPDS ................................

3. EVALUATE plant conditions AND ASSIST the EC in making timely and proper Emergency Classifications and Protective Action R ecom m endatio ns .................................................................................................... El

4. ENSURE Attachment 2, TSC Briefing Guideline is complete.

(norm ally by AAT O perations support) .....................................................................

5. ENSURE Radiological Emergency Conditions Status Board is u p d a te d .................................................................................................................... ED

6. ENSURE phone link between Control Room and TSC Ringdown C o m m unicato rs ........................................................................................................ FD

7. ENSURE each AAT position is staffed. REQUEST Security to contact additional AAT members as needed. (Refer to "Emergency Response Personnel Roster".)

Operations Support:

TSC Ringdown Communicator:

Control Room Ringdown Communicator:

Two Engineers:

NRC Communicator:

8. ENSURE all AAT members have badged in at TSC Card Reader ...........................

9. DETERMINE parameters or parameter groups (SPDS and RECALL) to monitor AND ENSURE the desired parameters are displayed (R eference Attachm ent 12) ......................................................................................

10. ENSURE times and results of significant actions are documented throughout the em ergency ........................................................................................

I EM-225 Rev. 29 Page 21 of 47

ATTACHMENT 1 Page 2 of 2 AAT COORDINATOR CHECKLIST

11. ENSURE AAT performs applicable attachments in EM-225 .................................... El

12. ENSURE OSC repair priorities are appropriate for plant conditions ......................... El

13. ENSURE the EC is informed of significant AAT activities and changes in plant status ............................................................................................ El

14. IF the EOF is staffed, THEN ESTABLISH communication with the EOF Technical Support Team via Accident Assessment Ringdown line, or extensions 6720/6205 ..................................................................... N/A El El

15. NOTIFY Off-Duty Shift Manager (requal crew or OSS crew) for additional O perations support .................................................................................. El

16. APPROVE Attachment 11, OSC Request Form: (This request should go through the OSC Manager to the OSC)

Requests for operator actions outside CCHE ................................................ F1 OR

Maintenance repair activities that have been initiated by the C ontrol R oom or AA T .................................................................................... 0

17. REVIEW Attachment 9, Dose Assessment Team Notification AND ENSURE updates are provided as plant conditions change .................................... El NOTES: 1 AAG-005, Contingencies for Loss of SFP Level, contains guidance for establishing SF Pool makeup due to an extended loss of SF Pool cooling ....... E]

2 Additional off-site equipment maybe needed to support guidance in AA G -0 0 5 . ........................................................................................................ .

3 Duke Energy Policy Statement PY-AD-ALL-0002 establishes an agreement with Duke nuclear fleet to provide an alternate PPIP.

1-1/2 inch fire hose may be obtained from on-site fire hose stations, local fire departments, fire equipment vendors, or other Duke sites, based on a v a ila b ility . .......................................................................................................... L

18. IF additional equipment to control SF Pool level OR Spent Fuel Cooling is required, THEN request equipment from off-site Duke fleet resources with Emergency Coordinator approval ................................ N/A LI El EM-225 I Rev. 29 Page 22 of 471

ATTACHMENT 2 Page 1 of 1 TSC BRIEFING GUIDELINE NOTE: REFER TO Attachment 8, Critical Safety Function Checklist and Attachment i 10, Core Damage Assessment to aid in this evaluation .......................................

1.1 REACTOR SHUTDOW N ................................................................................. Yes M No --

1.2 CORE ADEQUATELY COOLED (1)................................................................ Yes E No [L 1.3 FISSION PRODUCT BARRIER ASSESSMENT ................................ Z N/A (Defueled) EL (Use Attachment 8,Critical Safety Function Checklist, Table 3)

1. Fuel Clad: Intact El Potential Loss Ri Loss Fi

2. RCS: Intact EL Potential Loss Ri Loss Fi

3. Containment: Intact El Potential Loss Ri Loss EL 1.4 SPENT FUEL POOL STATUS:

1. SF Pool Cooling Available? .................................... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Yes Li No El

2. S F C lad Intact? ............................................. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Y es Ri No F-]

3. SF Pool Level Stable (Indication On scale)? ......................................... Yes EL No E]

4. SF Pool Temperature Stable? .............................................................. Yes E] No Rl 1.5 EMERGENCY ELECTRICAL POWER STATUS

1. Off-Site Power Available? ..................................................................... Yes El No Li

2. ES Bus Energized? ............................................................................... Yes Li No LI

3. Emergency Diesel Generator Available? .............................................. Yes Li No Li

4. Alternate AC Diesel Generator Available? ............................................ Yes Li No Li

5. DC Power Available? ............................................................................ Yes Li No Li 1.6 CONTROL COMPLEX STATUS

1. Ventilation / Cooling Available? ............................................................. Yes EL No Li

2. Necessary Instrumentation Available? (2) ............................................. Yes [i: No Li 1.7 OTHER CONDITIONS / CHALLENGES Note (1) - Inadequate Core Cooling is accident conditions that result in a loss of core cooling that requires entering EOP-7, Inadequate Core Cooling (Ref step3.o.4 )

(2) - Necessary refers to specific instruments and annunciators that are needed to identify, diagnose, and track the problems that are causing the emergency.

EM-225 Rev. 29 Page 23 of 47

ATTACHMENT 3 Page 1 of 1 TSC RINGDOWN COMMUNICATOR CHECKLIST

NOTE: Attachment steps can be completed in any order ...............................................

1. ESTABLISH contact with the Control Room Communicator via the Accident Assessment Ringdown phone (receiver needs to be off the hook to use the headset) .................................................................................... FD

2. ENSURE the Control Room is informed of:

changing radiological conditions .................................................................... ED 0 ongoing TSC maintenance and repair activities ............................................ ED 0 accident m itigation priorities .......................................................................... El 0 operator actions outside the CCHE ............................................................... 0

3. IF the EOF is staffed, THEN ESTABLISH communication with the EOF Technical Support Team via Accident Assessment Ringdown line, or extensions 6720/6205 ..................................................................... N/A [] I

4. MAINTAIN the Accident Assessment Team log book with all significant events, changes in plant status, and requests to and from the C ontrol R oom .............................................................................................

5. RELAY information and directions to the Control Room as appropriate ...................

6. MONITOR progression through EOPs and APs,

Anticipate problems created by unavailable equipment or other unusual plant conditions ....................................................................... EL

MARK place keeping aids as appropriate to allow other AAT members to determine status of procedure usage ........................................

0 PROVIDE periodic status to AAT Operations Support member .................... El

7. INITIATE Attachment 11, OSC Request Form:

Requests for operator actions outside CCHE ................................................

OR

Maintenance repair activities for the OSC that is requested by the Control Room or AAT .........................................................................

EM-225 Rev. 29 Page 24 of 47 1

ATTACHMENT 4 Page 1 of 3 AAT OPERATIONS SUPPORT CHECKLIST

[NOCS 62764]

[NOTE: Attachment steps can be completed in any order .......................................... j.... L.

1. BEGIN assessment of Critical Safety Functions to ensure adequate core cooling and fission product barrier preservation, USING Attachment 8, Critical Safety Function Checklist as a p p lica b le .................................................................................................................

2. COMPLETE Attachment 2, TSC Briefing Guideline AND PROVIDE the results to the AAT Coordinator. Attachment 2, TSC Briefing Guideline should be completed periodically or as conditions c h a n g e ..................................................................................................................... D

3. MAINTAIN the CSF Status Board at the TSC .......................................................... El

4. COMPLETE Attachment 9, Dose Assessment Team Notification AND PROVIDE the results to the TSC Radiation Controls Coordinator and the EOF Dose Assessment Team Leader. If conditions change, Attachment 9, Dose Assessment Team Notification should be reassessed and submitted to the Radiation C o ntrols C oo rd inato r ................................................................................................ El

5. Coordinates/processes requests for operator actions or maintenance support through the Repairs Coordinator using Attachment 11, OSC Request Form. REFER TO SP-306 for a list of EO B and EO L locations and contents .................................................................. El

6. IF RCS LOCA conditions exist, THEN COORDINATE performance of EM-225A, Post Accident RB Hydrogen Control [NOCS 62767] .............. N/A Z Li

7. IF RCS LOCA conditions exist, THEN COORDINATE performance of EM-225E, Guidelines For Long Term Cooling ......................................... N/A Z EL

8. IF SGTR exists, THEN MONITOR BWST depletion rate AND INITIATE BWST MU early in the event if necessary (see Enclosure 1, TSC Guidance for EOPs page 2 of this procedure) ................ N/A E Li

9. IF EFW or AFW is operating, THEN COORDINATE performance of EM-225F, Long Term Emergency Feedwater Management ................... N/AZ Li

10. IF a Severe Accident is in progress, THEN ASSIST engineering in developing appropriate mitigation strategies using the Candidate High Level Actions in the CR-3 Severe Accident G uideline. [NO CS 100056] .......................................................................... N/A 0Li

11. PROVIDE appropriate input to the Communication/Report Coordinator to update Florida Nuclear Plant Emergency Notification Form Supplemental Data Sheet ....................................................... l EM-225 Rev. 29 Page 25 of 47 1

ATTACHMENT 4 Page 2 of 3 AAT OPERATIONS SUPPORT CHECKLIST

[NOCS 62764]

12. IF any diesel operated equipment is running, THEN EVALUATE the following parameters (OSC support and local observation might be required to obtain information on support systems and operating param eters) ........................................................................... N /A ]D1:

Diesel support systems (i.e., ventilation, fuel transfer, cooling, e tc .) ............................................................................................................... 0 Establish periodic monitoring of diesel operating parameters to ensure proper operation. Monitor every 4 hrs and adjust monitoring intervals as required based on trending results.

For EDGs refer to OP-707, Operation Of The ES Emergency Diesel Generators and SP-354C, Functional Test Of The Alternate AC Diesel Generator EGDG-1 C for operating param eters .................................................................................... El

Operating EDG load limitation (loaded and unloaded) .................................. EL

Fuel and lube oil supplies .............................................................................. EL NOTE: Low load/no load EDG Operation will accumulate unburned oil in exhaust system due to low exhaust temperatures, and may result in a challenge to EDG availability due to excessive exhaust back-pressure.

A key symptom of excessive exhaust back-pressure is erratic RPM/Frequency control followed by EDG Stall. When load is raised to burn off oil, ignited oil will be present in exhaust system. Dark smoke or possibly flame may be visible from the exhaust stack until excess oil is consumed, or removed ............. El

13. IF any ES diesel loading is < 600 KW, THEN perform the following.

1) Ensure EDG is being monitored and Increase monitoring frequency to hourly interval and continue trending EDG parameters to determine if more frequent monitoring will be re q u ire d . ......................................................................................................... El NOTE: EDG Operational issues are expected after 2 hrs if no load or approx 6 hrs if loaded between 400 KW to 600 KW.

2) Determine equipment that can be added to maintain the EDG > 600 KW. Refer to Enclosure 2, EDG Scenarios To Establish A dequate Loading .........................................................................

3) Develop strategy for adding EDG loads ........................................................ L]

4) Obtain EC approval on developed strategies ................................................ L]

5) Coordinate with the MCR on implementing strategies ................................... EI EM-225 I Rev. 29 Page 26 of 47

ATTACHMENT 4 Page 3 of 3 AAT OPERATIONS SUPPORT CHECKLIST

[NOCS 62764]

14. IF DHV-3 is required to be manually opened due to a Control Room fire resulting in AP-990, Shutdown from Outside the Control Room, entry, THEN COORDINATE RB entry activities ............................... N/A Z EI

15. DETERMINE emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements ................................

EM-225 Rev. 29 Page 27 of 47

ATTACHMENT 5 Page 1 of 1 AAT ENGINEERS CHECKLIST

[NOCS 62764]

NOTE: Attachment steps can be completed in any order ............................................... E]

1. PERFORM Attachment 10, Core Damage Assessment. PERFORM an initial and periodic assessment of core damage and fission product barriers, AND PROVIDE the results to the AAT Operations Support Member and the Radiation Controls Coordinator ....................................... ED

2. IF RCS LOCA conditions exist, THEN COORDINATE performance of EM-225B, Post-Accident Boron Concentration Management .................. N/A E El

3. IF RCS LOCA conditions exist, THEN OBTAIN RB atmosphere 1131 concentration AND TRANSMIT value to control room (for BS pump shutdow n decision m aking) ......................................................................... N/A E D

4. MAINTAIN the Plant Parameters Status Board (if required). Based on plant conditions, PLACE key parameters on status board for tre n d ing .................................................................................................................... Li

5. MONITOR for conditions listed in Enclosure 1, TSC Guidance for EOPs. PROVIDE the AAT Operations Support member with recom m ended actions .............................................................................................. EI

6. IF RB temperatures are elevated, THEN COORDINATE the performance of EM-225C, Post Accident Monitoring Of Reactor Building Tem perature .................................................................................. N/A Z LI

7. IF any OTSG level is < 12.5 inches (indicating a dry OTSG), THEN COORDINATE the performance of EM-225D, Guidance For Dry OTSG Tube To Shell Delta T Monitoring And Control ................................. N/A [ LI

8. EVALUATE the effects of proposed maintenance repair activities and operational manipulations on plant equipment .................................................. El

9. DEVELOP contingency plans AND SUPPORT emergency repair effo rts as applicable ................................................................................................. E]

10. IF a Severe Accident is in progress, THEN DEVELOP mitigation strategies using the Candidate High Level Actions in the CR-3 Severe Accident G uideline .......................................................................... N/A []

F

11. Within 7 days, ENSURE SW minimum flow requirements are maintained. IF ES or RBIC has actuated and either SWV-353 or 354 has failed closed, THEN ESTABLISH flow to the RB coolers OR ENSURE only 1 SW pum p is running ................................................................ EL

12. IF additional computers are required THEN obtain, as needed, from nuclear administrative building (i.e., engineering laptop computers), that can be used to access documentation on the netw o rk ........................................................................................................ N/A RI L]

13. IDENTIFY AAT priorities using the AAT priority board in the AAT ro o m .........................................................................................................................

EM-225 Rev. 29 Page 28 of 47

ATTACHMENT 6 Page 1 of 1 CONTROL ROOM RINGDOWN COMMUNICATOR CHECKLIST NOTE: Attachment steps can be completed in any order ...............................................

1. ESTABLISH communication with the TSC Ringdown Communicator on the Accident Assessment Ringdown phone in the Control Room. BRIEF TSC Ringdown Communicator on operator actions that are in progress ........................................................................ El

2. RELAY status of overall plant conditions, operator activities and questions to the T S C AA T ........................................................................................ El

3. RELAY instructions to Control Room Operators for mitigating actions as directed by the EC ................................................................................... FI

4. INFORM Control Room Operators of the following:

0 Changes in Emergency Classifications ......................................................... El 0 T S C repa ir effo rts ..........................................................................................

Operators activities dispatched from the TSC/OSC ......................................

0 Changing radiological conditions ................................................................... EL

Mitigatio n prio rities ......................................................................................... EL

5. MONITOR EOPs or APs in use by Control Room ....................................................

6. IF a Severe Accident is in progress, THEN DIRECT Control Room personnel regarding mitigation strategies, based on actions approved by the TSC Emergency Coordinator ............................................ N/A LI 0

7. RELAY requests for support from the Control Room to OSC teams, via the TSC Ringdown Communicator .......................................................... El

8. Once TSC is operational, REQUEST extra plant operators (if available) be sent to OSC for in plant support. (Ref. EM-1 03, Enclosure 1, Dispatching of Resources During Emergency Plan E n try ) .......................................................................................................................

9. INFORM TSC of operator actions being performed ................................................. El EM-225 Rev. 29 1 Page 29 of 47

ATTACHMENT 7 Page 1 of 1 NRC COMMUNICATOR CHECKLIST NOTE: Attachment steps can be completed in any order ............................................... ....[E

1. CONTACT the Communication/Report Coordinator to determine if continuous communication with the NRC is required ...............................................

2. OBTAIN copies of any previously submitted NRC reports ................................... E]

3. IF the NRC has requested continuous communication, THEN ESTABLISH communication with the NRC on the Emergency Notification System (ENS) ...................................................................... N/A El E

4. MAINTAIN a log book of significant communications between the NRC and CR-3, including a summary of responses to NRC questions and transmittal of information .............................................................. El

5. MAINTAIN an open line on the ENS until the NRC agrees to term inate com m unications .......................................................................................

6. LOG time(s) when TSC notifies NRC of Emergency Classification c h a n g e s .................................................................................................................... El

7. LOG time(s) when TSC notifies NRC of Protective Action R eco m m e nd atio ns ...................................................................................................

8. WHEN communication with the NRC is not required, THEN PROVIDE support to other AAT members as needed ...................................

9. MAKE emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements. Examples include, but are not limited to:

Suspension of Safeguards (invoked under 10 CFR50.54(x),

or Section 24 of the Physical Security Plan) ............................................. El

The declaration of any of the Emergency Classes specified in EM-202, Duties of the Emergency Coordinator

{1 OC F R 50 .72 (a )(1 )(i)} ....................................................................................

0 The condition of CR-3, including its principal safety barriers, being seriously degraded ..............................................................................

CR-3 being in an unanalyzed condition that significantly deg rades pla nt safety ....................................................................................

One Hour Security Notifications for sabotage (10 CFR 73.71, A p pe nd ix G I(a )(2 )) ........................................................................................

EM-225 Rev. 29 Page 30 of 47

1 ATTACHMENT 8 Page 1 of 6 CRITICAL SAFETY FUNCTION CHECKLIST NOTE: The parameter tables below are for reference only. It is not intended that the tables be completed during each evaluation. Plant computer point numbers or SPDS/RECALL point numbers are listed, if available. Using pre-established RECALL Groups based on accident type in progress is recommended.

1. MONITOR the parameters associated with the Critical Safety Functions.

Qualified OSI PI EP folder has established Accident groups.("Start-Program-Business Apps-PI System-CR3 QPIM") ...................................................................

2. NOTIFY the AAT Coordinator immediately if any of the CSFs cannot be verified ..................................................................................................... El TABLE 1: Reactor Shutdown Status Reactivity Control .............................................. N/A E Li All Rods at in-limits Y/N P057 RECL-375 Intermediate Range P212 RECL-150 detector NI-3 amps Intermediate Range P213 RECL-1 51 detector NI-4 amps Source Range NI-1 cps P202 RECL-152 Source Range P203 RECL-1 53 NI-2 cps Adequate Shutdown OP-1 03C Margin Curve 18&19 FEM-225 Rev. 29 Page 31 of 47

ATTACHMENT 8 Page 2 of 6 CRITICAL SAFETY FUNCTION CHECKLIST TABLE 2: Core Cooling Status ECCS/Support Status ....................................... N/A [ LI bubcooiing Margin I _I I I _III A HPI Pump operating RECL-209 B HPI Pump operating RECL-210 C HPI Pump operating RECL-211 MUV-23 flow W704 RECL-52 MUV-24 flow W706 RECL-54 MUV-25 flow W703 RECL-51 MUV 26 flow W705 RECL-53 DHPs operating A/B X063 RECL-207 (run/stop) X064 RECL-208 DHP-1A flow W409 RECL-55 DHP-1B flow W410 RECL-56 CFT A level P200 CFT B level P201 CFT A press CFT B press BWST level (ft) X335 RECL-57 RWP-1 operating X060 RWP-2A operating X061 RECL-222 RWP-2B operating X062 RECL-223 RWP-3A operating RECL-217 RWP-3B operating RECL-218 DCP-1A operating RECL-220 (yes/no)

DCP-1 B operating RECL-221 (yes/no)

SWP-1A operating RECL-219 SWP-1 B operating SWP-IC operating RB Sump Strainer AP RECL-79 EM-225 Rev. 29 Page 32 of 47

ATTACHMENT 8 Page 3 of 6 CRITICAL SAFETY FUNCTION CHECKLIST TABLE 2: Core Cooling Status (Cont'd)

Secondary System Status ............................... N/A E LI LI-UL; U I Z5 A press VV44V KIL;L-ZZ I EFIC OTSG B press W452 RECL-255 OTSG A level S285 RECL-92 OTSG B level S286 RECL-93 MFW flow A S301 RECL-100 MFW flow B S302 RECL-101 EFPs operating 1/2/3/7 EFP-1/3 Flow to A OTSG RECL-246 EFP-1/3 Flow to B OTSG RECL-245 EFP-2 Flow to A OTSG RECL-248 EFP-2 Flow to B OTSG RECL-247 Total EFW Flow to A OTSG S300 RECL-408 Total EFW Flow to B OTSG S312 RECL-409 EFW Tank Level RECL-236 I EM-225 I Rev. 29 Page 33 of 47

ATTACHMENT 8 Page 4 of 6 CRITICAL SAFETY FUNCTION CHECKLIST TABLE 3: Fission Product Barrier Assessment ............. N/A IZ E]

[] INTACT [] POTENTIAL LOSS [] LOSS

Does NOT *RCS condition warrant entry *RCS conditions in (or previously in) meet the into EOP-07, Inadequate Core Region 3 or Severe Accident Region criteria for Cooling *RCS activity >300pCi/gr 1131 dose "Potential Core Exit Thermocouples > equivalent. Additional indication is 100 Loss" or 700 degrees F mR/hr measured on RM-G3 or at one "Loss" foot from sample lines in Nuclear Sample Room

RM-G29/30 > 100 R/hr for ? 15 minutes

Attachment 10, Core Damage Assessment indicates failed fuel El INTACT Fl POTENTIAL LOSS L] LOSS

Does NOT 0 RCS leak or OTSG tube leak 0 RCS leak resulting in loss of adequate meet the requiring one or more injection subcooling margin criteria for valves to maintain adequate 0 OTSG Tube Rupture resulting in loss of "Potential subcooling margin adequate subcooling margin Loss" or 0 RCS pressure /Tincore

RM-G29/30 >1 OR/hr for -215 minutes "Loss" relationship violates NDT limits

RCS leak or OTSG tube leak results in ES actuation on low RCS pressure.

a HPI/PORV or HPI/Code Safety valve coolingiinpors LI INTACT LI POTENTIAL LOSS LI LOSS

Does NOT

RB pressure > 54 psig

Containment isolation is incomplete and meet the

RB hydrogen concentration > release path to environment exists.

criteria for 4% Confirmation may be from elevated "Potential

RB pressure > 30 psig with radiation readings in areas adjacent to Loss" or NO building spray available the RB.

"Loss"

RMG-29 or 30 reading > 0 OTSG Tube Rupture > 10 gpm exists 25,000 R/hr and prolonged steaming to atmosphere

Core conditions in severe or an unisolable steam leak outside RB accident region of ICC curves from affected OTSG.

for >15 mrin Containment pressure or sump level response NOT consistent with LOCA conditions 0 Rapid unexplained RB pressure decrease following an initial increase I EM-225 Rev. 29 Page 34 of 47 1

ATTACHMENT 8 Page 5 of 6 CRITICAL SAFETY FUNCTION CHECKLIST TABLE 4: Spent Fuel Pool Status PARAM ETER YES NO SF Pool Cooling

SF Pump, SWP, RW running Available OR

DHP aligned for SF Pool Cooling, DCP, RWP Running SF Clad Intact ; RM-G14/15 rising dose rate indication STABLE UNSTABLE ISF Pool Level Indication On scale (Available MCB only)

(Refer to EOP-14 Enclosure 24)

SF Pool-Temperature Ref to EOP-14 Enclosure 241 TABLE 5: Emergency Electrical Power Status Off-Site Power 500 KV SWITCHYARD 230 KV SWITCHYARD OFF-SITE POWER XFRM BEST ES Buses A-ES 4160V BUS B-ES 4160V BUS A- ES 480V BUS (1)

B-ES 480V BUS (1)

Emergency Diesel Generator A-EDG RECL-133,171 B-EDG RECL-134,172 Alternate AC Diesel N/A DC Electrical A-BATTERY B-BATTERY C-BATTERY Note i*) - Batery Tawure will occur ITassociated oattery cnargers are oe-energizea.

EM-225 I Rev. 29 1 Page 35 of 471

ATTACHMENT 8 Page 6 of 6 CRITICAL SAFETY FUNCTION CHECKLIST TABLE 6: Control Complex Status Control Complex Ventilation Status A-TRAIN EMERGENCY RECIRC 4 4 4 B-TRAIN EMERGENCY RECIRC 4 4

A-CHILLER 4 4

B-CHILLER

________________ _____________ _____________ L ______________

Control Room Instrumentation Status NNI-X 4 I NNI-Y 4 1 ICS 4 1 EFIC I I RPS IESASI I COMMENTS:

Performed By: Date: / / Time:

EM-225 Rev. 29 Page 36 of 47

ATTACHMENT 9 Page 1 of 3 DOSE ASSESSMENT TEAM NOTIFICATION NOTES: 1. The TSC/AAT supplies the TSC Radiation Control Coordinator and EOF DAT Leader with Attachment 9 to assist with the projection of off-site doses ............ Li 2 MARK items N/A or unknown based on information available at the time .......... .I--

3 PROVIDE readily available information promptly AND FOLLOW UP with additional forms as more information becomes available .................................... El 4 Attachment 9 can be completed in any order ..................................................... El LOSS-OF-COOLANT ACCIDENT: Rx Trip Date/Time: / ...... N/A N FI

a. Rx Fuel Cladding status: (from Attachment 10, Core Damage Assessment)

El Normal Activity (Spike Factor _ ) (See General Information - Pg 3)

E] Clad Damage El Fuel melt

b. Rx Core Uncovered? (Rx Core covered is based on RCS NOT superheated)

F-1 NO (Not in EOP-07, Inadequate Core Cooling)

Li YES Uncovered - Date/Time /

Recovered - Date/Time /

c. Start of release to containment (start of the LOCA)

L1 Unknown n Date/Time

d. Release to atmosphere?

DNO DYES Date/Time / Estimated duration Release path (from where to where)

Release path flow rate (unmonitored releases):

Li Estimated hole size _ _ Diameter (in) or El Area (in2 )

Containment pressure (RECL-82/83) PSIG OR Li % RB Volume /day (RB Volume is 2 X10 6 cu ft) OR F1 gpm __ OR Li Design Basis Leakage

e. Reactor Building Spray Actuated? (RECL-212/213)

FD NO Fi YES Dates/Times

f. Rx Bldg Vent flow rate (AH-1 003-TIR Channel 4)

Charcoal banks in service Li YES / Li NO

g. Auxiliary Building ventilation (W351): flow rate Charcoal banks in service [] YES / Li NO

h. Loose Parts Monitor indications?

i- Unavailable / Li NO L-i YES Location EM-225 Rev. 29 Page 37 of 47

ATTACHMENT 9 Page 2 of 3 DOSE ASSESSMENT TEAM NOTIFICATION

WASTE GAS DECAY TANK RUPTURE: (See General Information-Pg 3) ......... N/A Z El

a. Release pathway:

Fl Tank rupture OR El Valve leakage OR El Other

b. Tank volume (Each WGDT volume = 1753 Wf 3)

Tank pressure (RW203/204/205)

c. Release rate El Unknown OR E _ _ CFM

d. Start of release El Unknown OR E] Date/Time Estimated duration

e. Auxiliary Building Ventilation (W351) Flow rate Charcoal banks in service El YES / El NO STEAM GENERATOR TUBE RUPTURE: Rx Trip Date/Time: / N/A N El

a. Primary-to-secondary leak rate:

El Unknown OR El gpm OR El Number of tubes

b. Rx Fuel Cladding status: (From Attachment 10, Core Damage Assessment)

El Normal Activity (Spike Factor _ ) (See General Information - Pg 3)

El Clad Damage El Fuel melt

c. Rx Core Uncovered? (Rx Core cover is based on RCS NOT superheated)

El NO (Not in EOP-07, Inadequate Core Cooling)

El YES Uncovered - Date/Time /

Recovered - Date/Time /

d. Leaking OTSG isolated? El NO / El YES Date/Time

e. Release Point:

El MSSV/ADV (intermittent/continuous) OR El Condenser

f. Start of leak El Unknown OR El Date/Time

g. OTSG Water Mass (See General Information - Pg 3)

El OTSG Water Mass OR El RASCAL default value of 93000 Ibm

h. OTSG Steaming Rate (See General Information - Pg 3)

El OTSG Steaming rate OR El RASCAL default 75000 Ibm/hr

i. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in service El YES / El NO EM-225 Rev. 29 Page 38 of 47

ATTACHMENT 9 Page 3 of 3 DOSE ASSESSMENT TEAM NOTIFICATION SPENT FUEL ACC IDENT: ................................................................................. N/A [] FI

a. Spent Fuel uncovered?

[] NO II YES Date/Time Recovered Date/Time /

b. Spent Fuel Pool Empty?

] NO LI YES Date/Time Recovered Date/Time /

c. Fuel assembly damaged by dropped component/handling?

(See General Information below)

F] NO LI YES Number of assemblies damaged __ Damage Date/Time Last irradiation Date (use last refueling outage)

Damaged Fuel assembly underwater [i YES / FI NO / LI UNKNOWN

d. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in service nI YES / F] NO

e. Dry Cask lost cooling?

M N/A FI NO LI YES <24 hrs >24 hr,, ____ Cask on fire ;

Number assemblies in cask _ Type of dry cask GENERAL INFORMATION:

1. Initially use a spiking factor of 100. Adjusted spiking factor based on RCS sampling and analyses when accident conditions allow sampling to be performed.

The Spiking Factor is used in RASCAL to distinguish the change in concentration of the fission products in the RCS due to a rapid drop in RCS pressure. The sudden RCS pressure drop increases the rate at which the radioactive fission products in the fuel rod cladding gap escape to the RCS.

2. Waste Gas Decay Tank Rupture information does not need to be completed if RM-A2 is in service monitoring release.

3. Use RASCAL default value if data is not known at the time of completing Attachment 9.

4. Fuel assembly damage is associated with damage from a dropped component and not clad failures due to overheating or flaws.

COMMENTS:

Performed By: Date: Time:

Reviewed By Accident Assessment Team Coordinator:

EM-225 I Rev. 29 Page 39 of 47

ATTACHMENT 10 Page 1 of 5 CORE DAMAGE ASSESSMENT

1. This attachment does not apply if the core is defueled ............................... N/A Li Z

2. DETERMINE if core damage has occurred using one or more of the following methods AND ESTIMATE the extent of the damage and status of the fission product barriers ........................................................... N/A M 0I

DETERMINE/ESTIMATE core damage based on R M-G 29/30 radiation levels ........................................................................... D NOTES: 1. Use of RM-G29/30 for determining core status requires a failure of the RCS (i.e., LOCA or PORV open).

2. Low monitor reading does not necessarily indicate lack of core damage.

The release from the core may bypass the Containment, may be retained in the RCS, may be over a long period of time, or may not be uniformly mixed.

3. Inconsistent readings may be due to the uneven mixing in the Containment (e.g., steam rising to the top). It may take several hours for uniform mixing.

ASSUMPTIONS:

The below table assumes a short release. A long-term release cannot be characterized using these tables:

TIME _ ____

RM-G29 R/HR RPHR RIHR R/HR R/HR RM-G30 R/HR R/HR R/HR R/HR R/HR No co re dam age .................................................................................. ... .

< 100 R/HR Possible clad failure and gas gap release .....................................................

100 - 25,000 R/HR with RB spray

100 - 75,000 R/HR without RB spray P ossible core m elting ....................................................................................

> 25,000 RPHR with RB spray

> 75,000 R/HR without RB spray EM-225 I Rev. 29 1 Page 40 of 47 1

ATTACHMENT 10 Page 2 of 5 CORE DAMAGE ASSESSMENT

DETERMINE/ESTIMATE core damage based on iodine ratios ..................... 0 NOTE: C.ore damage assessment based on Iodine Ratios will be evaluated by the F{adiation Controls Coordinator using EMG-NGGC-0002, Off-Site Dose A\ssessment. Contact Dose Assessment Team to coordinate the activity of e stimating core damage using this method. This method can take several hours b ased on the requirements to perform a gamma isotopic of a grab sample.

No co re da m a ge ............................................................................................ F-

1-131/Total Iodine < 0.05 Possible clad failure and gas gap release / possible core melting(There is no way to distinguish between a gap release and a core melt release using iodine ratios) ..................................... F-

1-131/Total Iodine >-0.05 I EM-225 Rev. 29 Page 41 of 471

ATTACHMENT 10 Page 3 of 5 CORE DAMAGE ASSESSMENT CORE DAMAGE ASSESSMENT BASED ON ICC CURVE DETERMINE/ESTIMATE core damage by plotting RCS pressure/Tincore temperature on the ICC curve below ............................. D NOTES: 1. Regions 1 and 2 indicate no fuel damage (normal RCS activity).

2. Severe Accident Region indicates possible core melt.

3. Region 3 indicates possible gas gap failure.

2500 2250 2000 1750 1500 IL1250 1000 750 500 250 0 * -

'----4,~---- i ~i 200 300 400 500 600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 1800 Tincore (*F)

EM-225 Rev. 29 Page 42 of 47

ATTACHMENT 10 Page 4 of 5 CORE DAMAGE ASSESSMENT CORE DAMAGE PROGRESSION ONCE UNCOVERED

3. IF inadequate subcooling margin exists, THEN DETERMINE if the core is uncovered ........................................................................................ N/A Z FI NOTES: 1. Reactor Coolant Inventory Tracking System (RCITS) provides a continuous indication of reactor vessel head and hot leg coolant inventory trend with the reactor coolant pumps in operation or tripped. RCITS consists of an RCS Hot Leg Level Subsystem, Reactor Vessel Level Subsystem and RC Void Trending Subsystem.

2. The RCS Hot Leg Level Subsystem (RC-163A/B-LR1) can monitor the top of the hot leg to the bottom of the hot leg with zero flow conditions. The Reactor Vessel Level Subsystem (RC-164A/B-LR1) can monitor the top of the reactor vessel to the bottom of the hot leg with zero flow conditions.

The bottom of the hot leg is approximately two feet above the top of the fuel. An off-scale low reading would indicate a high probability of loss of level below core level. Any flow (including natural circulation) in the RCS will result in a lower than actual reading. Thus, any indicated level will provide assurance that coolant level is above the core.

3. The Reactor Void Trend Subsystem (RC-1 69-XR) monitors void trends in the RCS when RCPs are running. RCP motor power and TcoId are used to infer average density of fluid passing through the pump (liquid or two-phase). A 0% reading infers no voiding, while 100% reading infers complete voiding.

4. Recorders are on the PSA panel in the Control Room and display on RECALL (points 62, 63, 64, 65, 70, 71).

A...HOT LEG. BA~.~G A~ESI -ESL VO1ID TREND~T RC-163A-LR1 RC-163B-LR1 RC-1 64A-LR1 RC-164B-LR1 RC-1 69-XR RECALL PT 63 RECALL PT 70 RECALL PT 62 RECALL PT 65 RECALL PT 64,71 C ore rem ains covered ................................................................................... 0

Tincore indicates saturated conditions

RCITS indicates any level U ncovered for 15 to 45 m inutes .................................................................... 0

Core temperature 1800-2400'F

Fuel cladding failure (occurred in 34 minutes at Three Mile Island)

Rapid hydrogen generation

Release of fission products out of fuel pin gap (gas gap failure)

Local fuel melt EM-225 Rev. 29 Page 43 of 47

ATTACHMENT 10 Page 5 of 5 CORE DAMAGE ASSESSMENT CORE DAMAGE PROGRESSION ONCE UNCOVERED (Cont'd)

Uncovered for 30 to 90 m inutes .................................................................... 0

Core temperature 2400-4200OF

Possible un-coolable core

Possible slump of molten core

Rapid release of volatile fission products (grain boundary release)

Uncovered for 1 to 3+ hours .......................................................................... El

Core temperature > 4200°F

Maximum core melt and hydrogen generation

Maximum in-vessel fission product release

Possible melt-through of vessel

4. Report the results of the evaluation to the AAT operations support member and the Radiation Controls Coordinator .....................................................

5. Continue to re-assess core and fission product barrier status as co nd itio ns cha n g e ....................................................................................................

EM-225 Rev. 29 Page 44 of 47 1

ATTACHMENT 11 Page 1 of 1 (SAMPLE)

OSC REQUEST FORM INSTRUCTIONS:

1. Use this form for each requested action from the Control Room, or Accident Assessment Team (multiple steps of EOPs/APs may be covered by one request)

2. Obtain approval from the AAT Coordinator

3. Obtain acknowledgement from OSC Manager

4. Make copy and give original to OSC Manager

5. Give copy to TSC Ringdown Communicator

6. Feedback to the Control Room on status of request.

REQUEST NUMBER: INITIATED BY:(AAT MEMBER) TIME DATE (UNIQUE NUMBER)

REQUESTED ACTION(S):

CONSEQUENCES IF NOT PERFORMED:

TIME RE FRAME TAG NO: TRAIN: LOCATION:

REQ'D APPROVAL TIME:

(AAT COORDINATOR)

RECEIVED BY: TIME:

(TSC REPAIR COORDINATOR)

FEEDBACK PROVIDED TO MAIN CONTROL ROOM TIME (TSC RINGDOWN COMMUNICATOR)

EM-225 Rev. 29 Page 45 of 47

ATTACHMENT 12 Page 1 of 1 SPDS OR RECALL DISPLAY SETUP FOR TSC PROJECTION SCREENS NOTE: TSC plant computers are labeled EMCO-81 and EMCO-50. Rebooting any of the TSC plant computers will result in an alarm in the main control room. Call the control room before rebooting any TSC plant computer. Rebooting an NGG standard desktop computer will not result in an alarm.

1. In the projector room, CONNECT computer monitor to the desired computer interface box along the side of computer rack. IF the desired computer is not connected to an interface box, THEN CONNECT the video display cable from an interface box to the desired co m puter ..................................................................................................... Dl

2. IF the computer is an NGGC standard desktop computer, THEN USE the desired computer mouse to select the specific display desired : ........................................................................................... N/A EZ- I

a. LOG into the computer with a corporate ID ................................................... El

b. SELECT start - programs - engineering - CR3 - CR3 PICS .......................... El

c. LOG into the pics access control client using system "CR3 PPCS" and "TSC" as the username and password .......................................

d. For a specific recall display, SELECT "recall display program", SELECT one of the pre-established displays from the "workspaces" drop down menu, AND CLICK "open".............................. EL

e. For a specific SPDS display, SELECT "SPDS display" AND CLICK on the desired SPDS display buttons OR USE the keyboard (refer to the laminated card for commands).

CNTRL H displays or hides the button bar. The button bar mimics the function panel on the MCB (Ref OP-509) .................................... El

f. For the subcooling margin monitor display, SELECT "t sat"......................... El

3. IF the computer is a TSC plant computer, THEN USE the desired computer mouse to select the specific display desired: .............................. N/A El El For a specific recall display, SELECT one of the pre-established displays from the "workspaces"drop down m enu, A ND C LIC K "open"................................... .........................................

For a specific SPDS display, just CLICK on the desired SPDS display buttons OR USE the keyboard (refer to the lam inated card for com m ands) ......................................................................

4. GO TO the touch screen (located in the main TSC room) which controls the projection screens. SELECT the desired projector room computer from the associated screen location (left, ce nter, o r right scree n) .....................................................................................

EM-225 I Rev. 29 1 Page 46 of 47 1

SUMMARY

OF CHANGES PRR 625518 NOTES: 1. Writers and Reviewers: Ensure that any changes to this procedure that affect information contained in Emergency Response Facility posters, enclosures, briefing cards, guidelines etc. are made to those items as well.

2. Writers and Reviewers: Changes to certain parts of this procedure may impact other Emergency Plan Implementing Procedures. Ensure appropriate PRRs are initiated as needed.

. .... . .. . .. . ..... * . . .. . . , =. . * *. ..... .*.

SECTION . ... CHANGE . ..

Throughout Change Revision to '29' Section 2.1, Item Add NOCS 100568 as a Developmental Reference (EDITORIAL) 25 Throughout Change references to 'Repairs Coordinator' to 'OSC Manager' (EDITORIAL) , Item Added new bullet for 10 CFR 73.71 - One Hour Security Notifications 9 for sabotage.

Deleted 33 bullet for plant shutdowns required by ITS, as the plant is permanently shut down. (PRR# 650001)

EM-225 I Rev. 29 1 Page 47 of 47

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