ML022240115
| ML022240115 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/29/2002 |
| From: | Florida Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| REP-12, Rev 3 | |
| Download: ML022240115 (26) | |
Text
PASSPORT DOCUMENT To : NRC NRC TRANSMITTAL Facility CR3 Department :
Address CR3-01242 / MAIL CODE: N/A DC DESK Page: 1 From Address
- CR3DOCSVCS Attention: DOCUMENT SERVICES - SA2A FLORIDA POWER CORPORATION CRYSTAL RIVER COMPLEX EhhhhhMmENl111 15760 WEST POWERLINE STREET City CRYSTAL RIVER State:FL Postal Code: 34428-6708 Country UNITED STATES Email Contact Date/Time 08/01/02 13:35 Transmittal Group Id:0000004188 Trans No. 000074204
Title:
'rta1 Items : 00001 1 . . . Ii Sheet Doc Status Revision Doc Date Copy # Media Copies Item Facility Type Sub Document Number ACTIVE 003 H 02 0001 CR3 MGEN MAN REP12 below and return to sender.
If a document was not received or is no longer required check the response Documents noted above not received (identify those not received).
no longer required).
I no longer require distribution of these documents (identify those Date: Signature:
-9
Rrda Pmwe EMERGENCY PREPAREDNESS REP-12 Revision 3 APM.f.usEwWw UNIT PROCEDURE Date 07/29/02 Page 1 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM REQUIRED APPROVALS cAiAZ Tupervidor, Emergency Preapdness t*ate I
FRorida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 AProsEuraCnpeny UNIT PROCEDURE Date 07/29/02 Page 2 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM 1.0 PURPOSE This procedure provides guidance for establishment and operation of the Emergency Operations Facility (EOF) Technical Support Team (TST).
2.0 SCOPE 2.1 The EOF Technical Support Team primarily functions as a technical resource for the EOF Technical Support Director and the EOF Director in development of public PARs. Duties also include assisting the TSC Accident Assessment Team as needed in development of mitigation strategies and in research of solutions to plant problems (however, the EOF TST personnel are NOT considered Severe Accident Management evaluators), interfacing with the Dose Assessment Team, and assisting in the development of recovery plans.
3.0 APPLICABLE REFERENCE DOCUMENTS 3.1.1 REP-03A, "Setup of the Emergency Operations Facility" 3.1.2 EM-225, "Duties of the Technical Support Center Accident Assessment Team" 3.1.3 EM-225A, "Post-Accident RB Hydrogen Control" 3.1.4 EM-225B, "Post-Accident Boron Concentration Management" 3.1.5 EM-225C, "Post-Accident Monitoring of Reactor Building Temperature" 3.1.6 EM-225D, "Guidance for Dry OTSG Tube to Shell Delta T Monitoring and Control" 3.1.7 EM-225E, "Guidelines for Long-Term Cooling" 3.1.8 EM-225F, "Long Term Emergency Feedwater Management" 3.1.9 Emergency Operating Procedures (EOPs) 3.1.10 MP-575, "Hydrogen Recombiner Installation" 3.1.11 OP-417B, "Operation of the Post-Accident Hydrogen Recombiners" 3.1.12 B&W Technical Bases Document 3.1.13 CR-3 Severe Accident Guideline
Florida Power EMERGENCY PREPAREDNESS REP-1 2 Revision 3 APreg Ea Copny UNIT PROCEDURE Date 07/29/02 Page 3 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM 3.2.14 Response Technical Manual (RTM-96), USNRC, Vol. 1, Rev. 3 3.2.15 Radiological Emergency Response Plan 3.2.16 NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents" 3.2.17 FPC IOC CR97-0122, dated 12/23/97 3.2.18 CH-632A, "Post-Accident Sampling and Analysis of the Reactor Coolant System" 4.0 DEFINITIONS AND ABBREVIATIONS 4.1 DEFINITIONS 4.1.1 Technical Support Team (TST)
Consists of EOF Technical Support Coordinator, EOF Technical Support Engineer, and EOF Technical Support Operations Representative.
4.1.2 Candidate High Level Actions (CHLAs)
Actions described in the CR-3 Severe Accident Guideline that could be taken to mitigate a severe accident and are deemed appropriate based on PDCs.
4.1.3 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 Room status.
4.1.4 Emergency Action Levels (EALs)
Conditions or indications that may be used as thresholds for initiating specific emergency measures. (See EM-202, Enclosure 1) 4.1.5 Protective Action Recommendations (PARs)
Emergency measures recommended for purposes of preventing or minimizing radiological exposures to Energy Complex personnel or members of the public.
k Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 A.Progmu EnergyCompany UNIT PROCEDURE Date 07/29/02 Page 4 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM 4.1.6 Plant Damage Conditions (PDCs)
Damage conditions used in the CR-3 Severe Accident Guideline to describe the status of the RCS, reactor core, and the containment during the progression of a severe accident.
4.1.7 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 reactor vessel, RB, or environment.
4.2 ABBREVIATIONS AAT - Accident Assessment Team ADV - Atmospheric Dump Valve AP - Abnormal Procedure BSP - Building Spray Pump BWST - Borated Water Storage Tank CFM - Cubic Feet per Minute CFT - Core Flood Tank CHLA - Candidate High Level Action CRNP- Crystal River Nuclear Plant CSF - Critical Safety Function DCP - Decay Heat Closed Cycle Cooling System Pump DHP - Decay Heat Pump EAL - Emergency Action Level ECCS - Emergency Core Cooling System EDG or EGDG - Emergency Diesel Generator EFIC - Emergency Feedwater Initiation and Control EFP - Emergency Feedwater Pump EFW - Emergency Feedwater EOF - Emergency Operations Facility EOP - Emergency Operating Procedure EP - Emergency Preparedness FP - Florida Power HPI - High Pressure Injection ICC - Inadequate Core Cooling IOC - Inter-Office Communication LOCA - Loss of Coolant Accident MFW - Main Feedwater MSSV - Main Steam Safety Valve OTSG - Once-Through Steam Generator PAR - Protective Action Recommendation
PREPAREDNESS REP-12 Revision 3 FRorida A Power EMERGENCY Pgr=EnrgC=May UNIT PROCEDURE Date 07/29/02 Page 5 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM PASS - Post-Accident Sampling System PDC - Plant Damage Condition PORV - Pilot-Operated Relief Valve PNSC - Plant Nuclear Safety Committee PSIG - Pounds per Square Inch (Gauge)
QC - Quality Control RB - Reactor Building RCITS - Reactor Coolant Inventory Tracking System RCP - Reactor Coolant Pump RCS - Reactor Coolant System RPS - Reactor Protection System RWP - Raw Water Pump Rx - Reactor SPDS - Special Parameter Display System SWP - Nuclear Services Closed Cycle Cooling System Pump TSC - Technical Support Center TST - Technical Support Team USNRC - U.S. Nuclear Regulatory Commission XFRM - Transformer 5.0 RESPONSIBILIES AND ACTIONS 5.1 RESPONSIBILITIES 5.1.1 EOF Technical Support Director Notifies the EOF Technical Support Coordinator and the Radiation Controls Manager that the EOF is activated and provides direction and control to the EOF functions associated with accident mitigation and assessment.
5.1.2 EOF Technical Support Coordinator
- a. Keeps the EOF Technical Support Director informed of TST activities and developments in plant status, especially those that may impact EALs and PARs.
- b. Notifies the EOF Technical Support Operations Representative and EOF Technical Support Engineer that the EOF has been activated and provides coordination and support for the TST.
- c. Ensures communication is established with the TSC on the TSC/EOF Ringdown until EOF Communicator arrives.
- d. Assists in the setup of the EOF Technical Support Work Area.
REP-12 Revision 3 Florida Power EMERGENCY PREPAREDNESS A PSEmn Cofpay UNIT PROCEDURE Date 07/29/02 Page 6 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- e. Performs "plant conditions" portion of the EOF briefings using the briefing guidelines provided in Attachment 4.
- f. Monitors the Accident Assessment Ringdown (communication between the Control Room and TSC AAT).
- g. Provides support to the TSC AAT in determining the causes and consequences of the emergency.
- h. Ensures interface is established with the EOF Dose Assessment Team using Section 5.1.5 and Attachment 7 as guidance.
- i. Refers to attachments for additional accident assessment guidance and information.
- j. Notifies Simulator support personnel when necessary (e.g., for testing mitigation strategies).
- k. Monitors CSFs and provides status to EOF Technical Support Director and EOF personnel during briefings, as needed.
- 1. Supplies input for completion of the Florida Nuclear Plant Emergency Notification Form Supplemental Data Sheet for plant conditions information. (Refer to Attachment 5, Part III, Fission Product Barrier Assessment).
5.1.3 EOF Technical Support Operations Representative
- a. Verifies that the SPDS computer is properly set up and operational.
- b. Operates the SPDS computer.
- c. Monitors plant parameters and provides status updates to the EOF Technical Support Coordinator.
- d. Monitors communications between the Control Room and the TSC AAT via speaker in the EOF Technical Support Room.
- e. Assists in the setup of the EOF Technical Support Work Area.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 APrras Emarg Company UNIT PROCEDURE Date 07/29/02 Page 7 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM 5.1.4 EOF Technical Support Engineer
- a. Assesses plant conditions and provides Engineering support for developing accident mitigation strategies as needed.
- b. Provides Engineering support to the TSC Technical Support Team.
- c. Notifies additional Engineering resources when necessary.
- d. Assists in the setup of the EOF Technical Support Work Area.
5.1.5 Dose Assessment Team
- a. Supports the TSC Accident Assessment Team with on-site radiological data and with chemical and radiological analysis of samples as needed to assess the accident.
- b. Provides plant radiation monitor readings and assessments.
- c. Provides projected radiological data (on-site and off-site doses, dose rates, and deposition) (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain).
- d. Provides RCS PASS data (> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to obtain results) on radionuclide composition, Chloride concentration, dissolved Hydrogen concentration, and Boron concentration.
- e. Provides Reactor Building and/or Auxiliary Building atmosphere radionuclide composition (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain).
- f. Provides in-plant radiological data.
- g. Provides chemical and radiological analysis of OTSGs and secondary samples.
- h. Provides Reactor Building sump boron concentration (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain).
- i. Performs CH-632A, Post Accident Sampling and Analysis of the RCS, Enclosure 2 (ifapplicable) to determine core damage.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 APogrmEnrgyComnjiy UNIT PROCEDURE Date 07/29/02 Page 8 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM 5.2 INSTRUCTIONS 5.2.1 The EOF Technical Support Coordinator or designee performs the duties of Attachment 1.
5.2.2 The EOF Technical Support Operations Representative performs the duties of Attachment 2.
5.2.3 The EOF Technical Support Engineer performs the duties of Attachment 3.
5.2.4 The completion of the Fission Product Barrier Status portion of the Supplemental Data Sheet of the Florida Nuclear Plant Emergency Notification Form requires the use of the information contained in Attachment 5, Part Il1.
6.0 INTERPRETATION CONTACT Supervisor Emergency Preparedness 7.0 REVISION HISTORY Rev. 0 Date: 01/26/99 Initial Issue Rev. 1 Date: 10/09/00 General revision to update attachments to match EM-225.
Rev. 2 Date: 07/18/01 NUPOST comment to make FPB Assessment table consistent with EM-202 &
EM-225. Changed name of team to Technical Support Team. Other changes include organizational titles, deletion of cancelled reference and consistency-type changes.
Rev. 3 Date: 07/29/02 Change Emergency Preparedness Department to Emergency Preparedness Unit and Corporate Health Physicist to Radiation Controls Manager. Added remaining DAT responsibilities. Deleted 10CFR50.59 exemption statement. Added responsibility to Technical Support Coordinator to supply FPB status information for Florida Nuclear Plant Emergency Notification Form using Attachment 5, Part II1. Corrected incorrect reference to attachment on Attachment 7.
Corrected room numbers in Attachment 1.
Replaced placekeeping lines in Attachments 1-3 with check boxes. Added reference to EM-225F.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 APrmEwCony UNIT PROCEDURE Date 07/29/02 Page 9 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 1 EOF TECHNICAL SUPPORT COORDINATOR CHECKLIST Check LI 1. Perform telephone notifications (Refer to Emergency Response Personnel Roster):
El EOF Technical Support Operations Representative El EOF Technical Support Engineer (3)
Ll 2. Check-in with EOF Director upon arrival and ensure name is posted on staffing board.
L[ 3. Ensure Accident Assessment Ringdown speaker is functional. (Refer to Attachment 9) (This is for monitoring only. Notify the TSC AAT by plant extension as necessary.)
L3 4. Ensure CSFs are being assessed. (Refer to Attachment 4) For the initial assessment, this information may be readily available from the TSC AAT. For subsequent evaluations, discuss with the EOF Technical Support Director whether TSC information is sufficient or if independent evaluations are desirable.
Li 5. Ensure PARs are appropriate per EM-202, Enclosure 8.
Qi 6. Provide input for completion of the Florida Nuclear Plant Emergency Notification Form Supplemental Data Sheet for plant conditions information section (Refer to Attachment 5, Part III, Fission Product Barrier Assessment).
L3 7. Ensure Room 124 SPDS computer is functional. (Refer to Attachment 9)
Li 8. Ensure communication is established with TSC on TSC/EOF Ringdown, until EOF Communicator arrives. (Refer to Attachment 9)
Li 9. Ensure EOP/AP procedure book is retrieved from Operations Briefing Room (217) or Simulator Control Booth (220).
Li 10. Brief the EOF Technical Support Director on CSFs and plant status.
Li 11. Provide CSFs and status reports to the EOF Staff during briefings. Use guidelines on Attachment 4 as necessary.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 AfPg*,M Ecainy UNIT PROCEDURE Date 07/29/02 Page 10 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 1 (Continued)
EOF TECHNICAL SUPPORT COORDINATOR CHECKLIST Check
- 12. Establish interface with EOF Dose Assessment Team. Attachment 7 lists information needed by the Dose Assessment Team; Section 5.1.5 lists information available from the Dose Assessment Team.
L1 13. Ensure the Status Board Coordinator is provided current CSFs information in the Main Conference Room.
L1 14. Notify Simulator personnel for support, if needed (e.g., for testing mitigation strategies).
- 15. Provide assistance to EOF Director on recovery plan. (Refer to Attachment 8)
T'*RFlorida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 A rogr=.E..WGuwry UNIT PROCEDURE Date 07/29/02 Page 11 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 2 EOF TECHNICAL SUPPORT OPERATIONS REPRESENTATIVE CHECKLIST Check S1. Confirm SPDS computer is operational. (Refer to Attachment 9)
U 2. Ensure work area is set up and functional. (Refer to REP-03A, Attachment 1)
L1 3. Provide plant parameters/status to EOF Technical Support Coordinator.
[- 4. Ensure Accident Assessment Ringdown speaker is functional. (Refer to Attachment 9) (This is for monitoring only. Notify the TSC MAT by plant extension as necessary.)
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 AP*g.e. Esrg*Company UNIT PROCEDURE Date 07/29/02 Page 12 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 3 EOF TECHNICAL SUPPORT ENGINEER CHECKLIST Check
[l 1. Ensure work area is set up and functional. (Refer to REP-03A, Attachment 1)
Ll 2. Establish communication with TSC AAT by plant phone as necessary Li 3. Obtain applicable drawings/documents, as needed.
Ll 4. Establish additional Engineering resources, if necessary.
11 5. Provide assistance for recovery plan development. (Refer to Attachment 8)
Ll 6. Obtain reactor core status. (Refer to Attachment 6)
Lj 7. Provide initial core assessment to EOF Technical Support Coordinator.
[l 8. Provide core status updates to EOF Technical Support Coordinator.
[3 9. Complete Attachment 7 for the Dose Assessment Team and provide to the EOF Technical Support Coordinator.
17 10. Notify additional resources (engineers, etc.) for core damage assessment, if needed.
Ll 11. If requested by the TSC to perform an assessment of core damage based on reactor coolant sample, use Enclosure 2 from CH-632A.
RS Florida ,,wW A ftgrm Power Cwny EMERGENCY PREPAREDNESS UNIT PROCEDURE Date 07/29/02Revision REP-12 3 1_ Page 13 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 4 BRIEFING GUIDELINE Refer to Attachments 5 and 6 to aid in this evaluation.
I REACTOR SHUTDOWN Yes F- No F-I1. CORE ADEQUATELY COOLED Yes E- NoF]
III. FISSION PRODUCT BARRIERS ASSESSMENT (Use Attachment 5, Part III)
Fuel Clad Fl Intact E] Challenged El Lost El Regained RCS El Intact [-] Challenged E] Lost E] Regained Containment E-] Intact E] Challenged El] Lost El] Regained IV. EMERGENCY ELECTRICAL POWER STATUS Off-Site Power Available? Yes E No I-l ES Buses Energized? Yesr No Emergency Diesel Generator's Available? Yes El No E-l DC Power Available? Yes l No E-l V. CONTROL COMPLEX STATUS Ventilation/Cooling Available? Yesrl Norl Necessary Instrumentation Available?* Yes - No l VI. OTHER CONDITIONS/CHALLENGES
- Necessary refers to specific instruments and annunciators that are needed to identify, diagnose, and track the problems that are causing the emergency.
REP-12 Revision 3 Florida Power EMERGENCY PREPAREDNESS AProgress EnergyCopany UNIT PROCEDURE Date 07/29/02 Page 14 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 5 CRITICAL SAFETY FUNCTION CHECKLIST Monitor the parameters associated with the Critical Safety Functions. 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.
Notify the Technical Support Coordinator immediately if any of the CSFs CANNOT be verified.
I. REACTOR SHUTDOWN STATUS:
REACTIVITY CONTROL PARAMTE4 C&OMUTER, :RECALL I
__________ <POINT$ POINT~>_______
All Rods at in-limits P057 RECL-375 Y/N Intermediate Range P212 RECL-150 detector NI-3 amps Intermediate Range P213 RECL-151 detector NI-4 amps Source Range P202 RECL-152 NI-1 cps Source Range P203 RECL-153 NI-2 cps Low Range RECL NI-14/15 - 102.103 Adequate Shutdown Margin J OP-10oC Curve 18&19
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3
- APrDgmssEnergy xnpary UNIT PROCEDURE Date 07/29/02 Page 15 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT5 (Continued)
II. CORE COOLING STATUS:
ECCS/SUPPORT STATUS SPAR...A..ETER CM..T.E. RECALLPOINT Subcooling Margin M114 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 (run/stop) X063 RECL-207 X064 RECL-208 DHP-1A flow W409 RECL-55 DHP-1 B flow W410 RECL-56 CFT A level P200 CFT B level P201 CFT A press OFT B press BWST level (ft) X335 RECL-57 RWPs operating 1/2A/2B/3A/3B DCPs operating A/B (yes/no)
SWPs operating A/B/C SECONDARY SYSTEM STATUS 4
~ P;RAMETER~ COMPUTER~ ktdEALL POINT EFIC OTSG A press W449 RECL-252 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 EFW flow to A OTSG S300 RECL-245 EFW flow to B OTSG S312 RECL-247
Revision 3 Florida Power EMERGENCY PREPAREDNESS REP-12 AProgres Energy CoTnpany UNIT PROCEDURE Date 07/29/02 Page 16 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT5 (Continued) -
Ill. FISSION PRODUCT BARRIER ASSESSMENT:
___________FUELýCLADDING`BARRIER V O INTACT 0 CHALLENGED 0 LOST 0 REGAINED
- NO indication
- Cooling restored, of cladding EOP-07 previously in) Region 3 or NO further damage
- Core Exit Thermocouples > 700 Severe Accident Region degradation degrees F a PASS indicates increased expected.
RCS activity >300gtCi/gr I3 (refer to CH-632A) a RM-G29/30 > 100 R/hr for>
15 minutes
- Attachment 6 indicates failed fuel
__________REACTOR COOLANT-SYSTEM BARRIER v o INTACT 0 CHALLENGED 0 LOST C REGAINED
- RCS leak resulting in loss of
- HPI/PORV or HPI within normal requiring one or more injection adequate subcooling margin Code Safety makeup valves to maintain adequate
- OTSG Tube Rupture resulting Cooling stopped pump subcooling margin in loss of adequate
- Subcooling capacity
- RM-G29/30 >1 OR/hr for> 15 and leak isolated
- HPI/PORV or HPI/Code Safety valve cooling is in progress 2 - ~~~CONTAINMENT ,BARRIER 2 O INTACT 0 CHALLENGED [I LOST 0 REGAINED
"* NO evidence e RB pressure > 54 psig
- Containment isolation is 0 Repair efforts of
- RB pressure > 30 psig with NO to environment exists. leak leakage building spray available Confirmation may be from
- Containment
"* Tube rupture e RMG-29 or 30 reading > 25,000 elevated radiation readings in pressure has release is R/hr areas adjacent to the RB. reduced to stop only through
- Core conditions in severe - OTSG Tube Rupture > 10 leakage condenser accident region of ICC curves for gpm exists and prolonged
>15 min steaming to atmosphere or an unisolable steam leak outside RB from affected OTSG.
- Rapid unexplained RB pressure decrease following an initial increase Performed By: Date: Time:
Revision 3 Florida Power EMERGENCY PREPAREDNESS REP-12 AProgres Energy Company UNIT PROCEDURE Date 07/29/02
____Page 17 of 25 K>EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 5 (Continued)
IV. EMERGENCY ELECTRICAL POWER STATUS:
OFF-SITE POWER 500 SWIA.MTC R U VA-IAbLE 500 KV SWITCHYARD 230 KV SWITCHYARD OFF-SITE POWER XFRM BEST ES BUSES 46VAILA P ftAk&& EU LE A-ES 4160V BUS B-ES 4160V BUS A-ES 480V BUS (Note 1)
B-ES 480V BUS (Note 1)
~.PARAMVETbft RKdALLPT VA L UNAVIAL A-EDG RECL-133,171 B-EDG RECL-134,172 DC ELECTRICAL A-BATTERY B-BATTERY C-BATTERY Note (1) Battery failure will occur if associated battery chargers are de-energized.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 AProgres Ener Company UNIT PROCEDURE Date 07/29/02 Page 18 of 25 "EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 5 (Continued)
V. CONTROL COMPLEX STATUS:
CONTROL COMPLEX VENTILATION STATUS PARAMETER ~ AVAILABLE IOPR'ATIRG : UhAVAILABLE A-TRAIN EMERGENCY RECIRC B-TRAIN EMERGENCY RECIRC A-CHILLER B-CHILLER CONTROL ROOM INSTRUMENTATION STATUS NNI-X A UNAVAILAB NNI-X NNL-Y ICS EFIC RPS ESAS COMMENTS:
Revision 3 Florida Power EMERGENCY PREPAREDNESS REP-12 AProgmssEnergyComparry UNIT PROCEDURE Date 07/29/02 Page 19 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 6 CORE DAMAGE ASSESSMENT Determine ifcore damage has occurred using one or more of the following methods. Estimate the extent of the damage. Evaluate the status of the fission product barriers. Report the results of the evaluation to the EOF Technical Support Operations Representative and the Dose Assessment Team Leader. Continue to reassess core and Fission Product Barrier status as conditions change.
EO - ESTIMATE CORE DAMAGE BASED ON RCS SAMPLES.
Core damage assessment based on Reactor Coolant samples is evaluated by the Dose Assessment Team using CH-632A Enclosure 2. The results are submitted to the TST. (May take >2 hours to obtain results) o ESTIMATE CORE DAMAGE BASED ON RM-G29/30 RADIATION LEVELS.
NOTE: (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 table below assumes a short release. A long-term release CANNOT be characterized using these tables.
T;JIME: ____
RM-G29 '* R/HR R/HR R/HR R/HR R/HR RM-G3n'0 R/HR R/HR R/HR R/HR R/HR O] NO CORE DAMAGE
< 100 R/HR
[0 POSSIBLE CLAD FAILURE AND GAS GAP RELEASE 100 - 25,000 R/HR WITH RB SPRAY 100 - 75,000 R/HR WITHOUT RB SPRAY
[I POSSIBLE CORE MELTING
> 25,000 R/HR WITH RB SPRAY
> 75,000 R/HR WITHOUT RB SPRAY
. Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 APrgressEnergyCUnparry UNIT PROCEDURE Date 07/29/02 Page 20 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 6 (Continued)
CORE DAMAGE PROGRESSION ONCE UNCOVERED 0 IFinadequate subcooling margin exists, THEN determine ifthe core is uncovered.
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.
The RCS Hot Leg Level Subsystem (RC-1 63A/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-1 64A/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.
The Reactor Void Trend Subsystem (RC-169-XR) monitors void trends in the RCS when RCPs are running.
RCP motor power and Tcold are used to infer average density of fluid passing through the pump (liquid or two-phase). A 0% reading infers NO voiding, while a 100% reading infers complete voiding.
- . Recorders are on the PSA panel in the Control Room and display on RECALL (points 62, 63, 64, 65, 70, 71).
~ AIHOTýLEG V B-'H0T LEG W A-VESSEL> B-'VESSEL VOID TREND RC-1 63A-LR1 RC-163B-LR1 RC-1 64A-LR1 RCo1 64B-LR1 FICA169-XR RECALL PT 63 RECALL PT 70 RECALL PT 62 RECALL PT 65 RECALL PT 64,71 o3 CORE REMAINS COVERED TINCORE indicates saturated conditions RCITS indicates any level O UNCOVERED FOR 15 TO 45 MINUTES 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 -1.
o UNCOVERED FOR 30 TO 90 MINUTES Core temperature 2400-4200'F Possible uncoolable core Possible slump of molten core Rapid release of volatile fission products (grain boundary release)
O UNCOVERED FOR 1 TO 3+ HOURS Core temperature > 4200°F Maximum core melt and hydrogen generation Maximum in-vessel fission product release Possible melt-through of vessel
kFlorida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 AProgress Energy Company UNIT PROCEDURE Date 07/29/02
__Page 21 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 6 (Continued)
CORE DAMAGE ASSESSMENT BASED ON ICC CURVE ASSESS CORE DAMAGE BY PLOTTING RCS PRESSURE/INCORE TEMPERATURE ON THE ICC CURVE BELOW.
"o Regions 1 and 2 indicate NO fuel damage (normal RCS activity).
"o Region 3 indicates possible gas gap failure.
"o Severe Accident Region indicates possible core melt.
2600 2400 SaturationI II 2200 2000 1800 1600 In 0.
1400 0:
ILl C',
(n Cr.
nw Of 03 1200 1000 800 600
= Accident Region zTclad
- - -Severe
- - - -Z x100 'F 400 200
-0 200 300 400 500 600 700 800 900 1000 1100 1200 1300 Tincore (OF)
REP-12 Revision 3 Florida Power EMERGENCY PREPAREDNESS AProgress EnergyCanpary UNIT PROCEDURE Date 07/29/02 Page 22 of 25 K>EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 7 DOSE ASSESSMENT TEAM NOTIFICATION
- 1. The Technical Support Team is responsible for supplying the Dose Assessment Team with an evaluation of the accident type, the release pathway, and the release flow rate. The accident type affects the radionuclide distribution (i.e., percentage of each isotope) used by Dose Assessment to predict off-site doses.
- 2. The accident type is determined by physical parameters and instrument readings throughout the plant.
- 3. Complete the checklist below to the extent possible and give to the Dose Assessment Team Leader.
ACCIDENT TYPE
__ LOCA __ Waste Gas Decay Tank Rupture __ OTSG Tube Leak
___ Fuel Handling __ Other:
[I LOSS-OF-COOLANT ACCIDENT TIME OF RX TRIP:
- a. Normal Activity__ Clad Damage __ Fuel melt __ (from Attachment 6)
- b. Release pathway information (leak from where to where)
- c. Release path flow rate (estimated for unmonitored releases)
- d. Estimated duration Unknown.
- e. Reactor Building spray on/off times
- f. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in service A: B: C: D
- g. Containment pressure (X147) _ PSIG
- h. Loose Parts Monitor indications No -__Yes___ Location:
El WASTE GAS DECAY TANK RUPTURE
- a. Release pathway: Tank rupture __ Valve leakage __ Other_
- b. Tank volume pressure
- c. Release rate: Unknown__ Estimate CFM
- d. Estimated duration: Unknown Time __
- e. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in service A: B: C: D
[3 STEAM GENERATOR TUBE RUPTURE TIME OF RX TRIP:
- a. Primary-to-secondary leak rate: - gpm
- b. Core status: Cladding damage Fuel melt __ Normal__
C. Leaking OTSG isolated: Yes __ No __
- d. MSSV Open: Yes_ No_ ADV Open: Yes __ No __
- e. Condenser vacuum: Yes __ No __ RM-A2 In Service?: Yes __ No __
- f. Potential for change in status of leak: Yes __ No __
- g. Estimated duration of leak:
- h. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in servilce A: B: C: D El FUEL HANDLING ACCIDENT
- a. Location of damaged fuel: Pool A__ Pool B Number of Elements __
- b. Damage caused by: Mechanical impact Overheating__ Unknown
- c. Auxiliary Building Ventilation (W351): Flow rate Charcoal banks in service A: B: C: D
- d. - Release pathway: Unknown
- e. Estimated duration Unknown Status as of Date: Completed By:
Revision 3
,j Florida Power EMERGENCY PREPAREDNESS REP-12 AProgress Energy Compary UNIT PROCEDURE Date 07/29/02 Page 23 of 25 K>EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- A1TACHMENT 8 SHORT-TERM RECOVERY PLAN GENERIC OUTLINE PHASE I - INCIDENT STABILITY
- 1. Verify Security System integrity.
- 2. Assess integrity of systems required for Iongterm cooling by system walkdown:
- Decay Heat
- Spent Fuel
- Ventilation
- 3. Continue cooldown using an appropriate heat removal method.
- 4. Verify termination of release.
PHASE II- DATA GATHERING Auxiliary Building Filter Changeout and Analysis
- 2. Plant and Off-Site Radiation Surveys and Dose Assessments
- 3. Primary System and RB Atmosphere Sampling
- 4. Debrief key personnel.
- 5. Equipment inspection/develop damage report:
- Emergency Feedwater System (including electrical)
- Makeup System (HPI Valve)
- PORV and Block Valves
- Fuel Handling Area
- Diesel Generator
- 6. Community Reaction Survey
- 7. Develop detailed incident report.
- 8. Establish whole body counting capability for emergency workers.
EMERGENCY PREPAREDNESS REP-12 Revision 3 ArFogriedaEPwf UNIT PROCEDURE Date 07/29/02 Page 24 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT 8 (Continued)
SHORT-TERM RECOVERY PLAN GENERIC OUTLINE PHASE III - RESTORATION Based on results of Phase II assessment:
- 1. Prepare procedures as required.
- 2. Begin repair efforts.
- 3. Establish team for system cleanup and waste disposal activities.
- 4. Establish community educational and public relations activities.
- 5. Establish Recovery Team organization and off-site support liaison.
- 6. Re-establish normal site operations.
- 7. Establish claim office.
- 8. Assure regulatory communication.
- 9. Establish technical assessment team (FPC, Framatome Technologies, other Architect/Engineer, etc.).
- 10. Develop long-term organizational recovery responsibilities and plant status objectives.
NOTE: The completed recovery plan and implementing procedures shall be submitted to the PNSC for approval before implementation.
Florida Power EMERGENCY PREPAREDNESS REP-12 Revision 3 Company APmgress Enerfg UNIT PROCEDURE Date 07/29/02 Page 25 of 25 EMERGENCY OPERATIONS FACILITY TECHNICAL SUPPORT TEAM
- ATTACHMENT9 EQUIPMENT INSTRUCTIONS Accident Assessment Rinqdown Speaker
- 1. Ensure the speaker is connected to phone jack 124-D5.
- 2. Turn on power using the knob on the front of the speaker.
- 3. Activate the speaker circuit by removing receiver from the Accident Assessment Ringdown phone in the Simulator Control Room. This phone must be off the hook for the Room 124 speaker to function.
- 1. Ensure both computers and monitors designated for SPDS in northwest comer of Room 124 are turned on.
- 2. In the "Access Control Client" dialog box, ensure "CR3 SPDS" is selected in the drop-down box (or "SIM SPDS" for drills).
- 3. Click "LOGON."
TSC/EOF Rincqdown Phone The TSC/EOF Ringdown phone is located in southwest comer of Room 124. This is an automatic circuit that connects when either the TSC phone or EOF phone is taken off the hook.