ML021050243
| ML021050243 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/01/2002 |
| From: | Florida Power Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML021050243 (33) | |
Text
TlPCUO PASSPORT DOCUMENT To NRC NRC Facility CR3 Department : TRANSMITTAL Address CR3-01242 / MAIL CODE: N/A From Address DC DESK CR3DOCSVCS Attention: DOCUMENT SERVICES FLORIDA POWER CORPORATION 111111 Page:
iiiIH CRYSTAL RIVER COMPLEX 15760 WEST POWERLINE STREET City CRYSTAL RIVER State:FL Postal Code: 34428-6708 Country UNITED STATES Email Contact Date/Time 04/01/02 11:50 Transmittal Group Id:0000001855 Trans No. 000059903
Title:
Total Items: 00001 Item Facility Type Sub Document Number Sheet Doc Status Revision Doc Date Copy # Media Copies 0001 CR3 POM EMG EM0225 ACTIVE 011 H 01 If a document was not received or is no longer required check the response below and return to sender.
Documents noted above not received (identify those not received).
I no longer require distribution of these documents (identify those no longer required).
Date: Signature:
('j
- .NNC-1 Information A s Use FLORIDA POWER CRYSTAL RIVER UNIT 3 PLANT OPERATING MANUAL EM-225 DUTIES OF THE TECHNICAL SUPPORT CENTER ACCIDENT ASSESSMENT TEAM EM-225 Rev. 11 Page 1 of 32
TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE ................................................................................................................................ 3
2.0 REFERENCES
......................................................................................................................... 3 2.1 Developm ental References ............................................................................... 3 3.0 PERSONNEL INDOCTRINATIO N....................................................................................... 4 3.1 Definitions ................................................................................................................ 4 3.2 Responsibilities ................................................................................................... 5 3.3 Lim its And Precautions ...................................................................................... 8 4.0 INSTRUCTIONS ...................................................................................................................... 9 4.1 Accident Assessm ent Initiation ......................................................................... 9 ENCLOSURES (All Enclosures are "Optional Records Non-Quality")
1 AAT Coordinator Checklist ....................................................................................................... 10 2 TSC Briefing Guideline .................................................................................................................. 11 3 TSC Ringdown Com m unicator Checklist .................................................................................. 12 4 AAT Operations Support Checklist ............................................................................................ 13 5 AAT Engineers Checklist ............................................................................................................... 14 6 Control Room Ringdown Com m unicator Checklist ................................................................... 15 7 NRC Com m unicator Checklist .................................................................................................. 16 8 Critical Safety Function Checklist .............................................................................................. 17 9 Dose Assessm ent Team Notification ....................................................................................... 22 10 Core Damage Assessment ................................................ 23 11 TSC Guidance For EOPs ............................................................................................................. 26 12 TSC Acccident Assessm ent Team OSC Request Form ......................................................... 31 EM-225 Rev. 11 Page 2 of 32
1.0 PURPOSE 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 Dose Assessment Team. 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 0627181
2.0 REFERENCES
2.1 Developmental References 2.1.1 Response Technical Manual (RTM-96); USNRC; Volume 1, Rev. 3 2.1.2 Radiological Emergency Response Plan 2.1.3 Emergency Operating Procedures (EOPs) 2.1.4 NUREG-1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents 2.1.5 B&W Technical Bases Document 2.1.6 FPC IOC CR97-0122, Dated 12/23/97 2.1.7 NEI 91-04, Revision 1, Severe Accident Issue Closure Guidelines 2.1.8 FPC IOC SE99-0184, Dated 9/14/99 2.1.9 EEM-99-018, Rev. 0 Operating Limits for SWP-1A/SWP-1B under Minimum Flow Conditions.
2.1.10 EM-202, Duties of the Emergency Coordinator 2.1.11 EM-102, Operation of Technical Support Center 2.1.12 EM-103, Operation and Staffing of the CR-3 Control Room During Emergency Classification 2.1.13 CP-151, External Reporting Requirements EM-225 Rev. 11 Page 3 of 32
3.0 PERSONNEL INDOCTRINATION 3.1 Definitions 3.1.1 Accident Assessment Team (AAT) - Consists of Coordinator, TSC Ringdown Communicator, Control Room Ringdown Communicator, Engineer, Operations Support, and NRC Communicator.
3.1.2 Candidate Hicih 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.
3.1.3 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 complex status.
3.1.4 Emergiency Action Levels (EALs) - Conditions or indications that may be used as thresholds for initiating specific emergency measures (see EM-202, Enclosure 1).
3.1.5 Plant Damagie 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.
3.1.6 Protective Action Recommendations (PARs) - Emergency measures recommended for purposes of preventing or minimizing radiological exposures to the Generating Complex personnel or members of the general public.
3.1.7 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 Rx vessel, Reactor Building or the environment.
EM-225 Rev. 11 Page 4 of 32
3.2 Responsibilities 3.2.1 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.
0 Relays instructions to Control Room Operators for mitigating actions as directed by the Emergency Coordinator (EC).
0 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.
0 Relay request for support from the Control Room to OSC teams, via TSC Ringdown Communicator.
a Once TSC is operational, request extra plant operators (if available) be sent to OSC for in plant support.
0 Inform TSC of in plant operator actions that are being performed.
EM-225 Rev. 11 Page 5 of 32
3.2.2 AAT Coordinator
- 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 Enclosure 12 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 TSC Repairs Coordinator to the OSC.
3.2.3 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.
- Initiate Enclosure 12 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. 11 Page 6 of 32
3.2.4 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.
0 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.
3.2.5 AAT Operations Support
- Monitors overall plant status during an emergency with emphasis on Critical Safety Functions.
0 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). [NOCS 13010]
- 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.
0 Coordinates/processes requests for operator actions or maintenance support activities through the TSC Repairs Coordinator using Enclosure 12.
- Determine emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements.
3.2.6 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.
3.2.7 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.
EM-225 Rev. 11 Page 7 of 32
3.2.8 Emergency Coordinator (EC) or designee
- Controls all activities at CR-3 during activation of the Radiological Emergency Response Plan.
- Implements EM-202.
- Determines EAL and PAR changes based on information obtained from the Accident Assessment Team and Dose Assessment Team.
- 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.
3.2.9 Dose Assessment Team
- 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain).
- Provides RCS PASS data (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain) on Radionuclide composition, Chloride concentration, Dissolved Hydrogen concentration, and Boron concentration.
- 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).
- 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to obtain).
- Performs CH-632A, Post Accident Sampling and Analysis of the Reactor Coolant System Enclosure 2 (if applicable) to determine core damage.
3.3 Limits And Precautions 3.3.1 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.
EM-225 Rev. 11 Page 8 of 32
4.0 INSTRUCTIONS 4.1 Accident Assessment Initiation 4.1.1 AAT Coordinator or designee: PERFORM the duties of Enclosure 1, AAT Coordinator Checklist.
4.1.2 TSC Ringdown Communicator: PERFORM the duties of Enclosure 3, TSC Ringdown Communicator Checklist.
4.1.3 AAT Operations Support member: PERFORM the duties of Enclosure 4, AAT Operations Support Checklist.
4.1.4 AAT Engineers: PERFORM the duties of Enclosure 5, AAT Engineers Checklist.
4.1.5 Control Room Ringdown Communicator: REPORT to the Control Room and PERFORM the duties of Enclosure 6, Control Room Ringdown Communicator Checklist.
4.1.6 NRC Communicator: PERFORM the duties of Enclosure 7, NRC Communicator Checklist.
EM-225 Rev. 11 Page 9 of 32
ENCLOSURE 1 AAT COORDINATOR CHECKLIST El Badge IN at TSC card reader and place name on TSC Staffing Board.
El Notify the EC that the Accident Assessment Team is operational when ALL of the following are accomplished:
El Determine Critical Safety Functions (Enclosure 2)
EJ Ability to brief EC on plant status to include impact of EALs or PARs through use of either SPDS or phone link established with Control Room.
0l Determine current plant status and conditions.
O Ensure Enclosure 2 is complete. (normally by AAT Operations support)
O Ensure Critical Safety Functions Status Board is updated.
O Ensure phone link between Control Room and TSC Ringdown Communicators.
El Evaluate plant conditions and assist the EC in making timely and proper Emergency Classifications and Protective Action Recommendations.
El Ensure each AAT position is staffed. Request Security to contact additional AAT members as needed. (Refer to "Emergency Response Personnel Roster".)
O Operations Support:
El TSC Ringdown Communicator:
El Control Room Ringdown Communicator:
El 2 Engineers:
El NRC Communicator:
El Ensure all AAT members have badged in at TSC Card Reader.
0l Determine parameters or parameter groups (SPDS and RECALL) to monitor and ensure the desired parameters are displayed.
El Ensure times and results of significant actions are documented throughout the emergency.
El Ensure AAT performs applicable enclosures in EM-225.
El Ensure OSC repair priorities are appropriate for plant conditions.
El Ensure the EC is informed of significant AAT activities and changes in plant status.
El If the EOF is staffed, establish communication with the EOF Technical Support Team using plant extensions (6720, 6205)
El During TSC briefing ensure Critical Safety Functions are addressed.
El Approve Enclosure 12 requests for operator actions outside CCHE or maintenance repair activities that have been initiated by the Control Room or AAT. This request should go through TSC Repairs Coordinator to the OSC.
EM-225 Rev. 11 Page 10 of 32
ENCLOSURE 2 TSC BRIEFING GUIDELINE Refer to Enclosures 8 and 10 to aid in this evaluation.
I REACTOR SHUTDOWN Yesil] No F II. CORE ADEQUATELY COOLED Yes F NoDF Ill. FISSION PRODUCT BARRIERS ASSESSMENT Fuel clad Li Intact SChallenged INLost E]Regained RCS i Intact -- Challenged F-1 Fl Lost F-I Fl Regained Containment I] Intact E- Challenged Lost Regained IV. EMERGENCY ELECTRICAL POWER STATUS Off-Site Power Available? Yes n No El ES Buses Energized? Yes -] No i-1 Emergency Diesel Generator's Available? Yes [ No I]
DC Power Available? Yes --] No V. CONTROL COMPLEX STATUS Ventilation/Cooling Available? Yes E No[l Necessary instrumentation Available?* Yes E- NoD 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.
EM-225 Rev. 11 Page 11 of 32
ENCLOSURE 3 TSC RINGDOWN COMMUNICATOR CHECKLIST Dl Establish contact with the Control Room Communicator via the Accident Assessment Ringdown phone.
El Ensure the Control Room is informed of changing radiological conditions, ongoing TSC maintenance and repair activities, accident mitigation priorities and operator actions outside the CCHE.
NOTE The EOF Technical Support Team can monitor the Accident Assessment Ringdown phone, but cannot be heard.
El Maintain the Accident Assessment Team log book with all significant events, changes in plant status, and requests to and from the Control Room.
El Relay information and directions to the Control Room as appropriate.
El Monitor progression through EOPs and APs (using a copy of the applicable procedures), anticipating problems created by unavailable equipment or other unusual plant conditions. Mark place keeping aids as appropriate to allow other AAT members to determine status of procedure usage. Provide periodic status to AAT Operations Support member.
El Initiate Enclosure 12 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. 11 Page 12 of 32
ENCLOSURE 4 AAT OPERATIONS SUPPORT CHECKLIST
[NOCS 62764]
"- Begin assessment of Critical Safety Functions to ensure adequate core cooling and fission product barrier preservation, using Enclosure 8 as applicable.
El Complete Enclosure 2 and provide the results to the AAT Coordinator.
Enclosure 2 should be completed periodically or as conditions change.
O Maintain the CSF Status Board at the TSC.
0 Complete Enclosure 9 and provide the results to the Dose Assessment Team Leader. If conditions change, Enclosure 9 should be reassessed and submitted to the Dose Assessment Team.
"El Coordinates/processes requests for operator actions or maintenance support through the Repairs Coordinator using Enclosure 12. Refer to SP-306 for a list of EOB and EOL locations and contents.
"El If RCS LOCA conditions exist, coordinate performance of EM-225A, "Post Accident RB Hydrogen Control." [NOCS 62767]
"El If RCS LOCA conditions exist, coordinate performance of EM-225E, "Guidelines For Long Term Cooling."
"El If EFW or AFW is operating, coordinate performance of EM-225F, "Long Term Emergency Feedwater Management."
"El If a Severe Accident is in progress, assist engineering in developing appropriate mitigation strategies using the Candidate High Level Actions in the CR-3 Severe Accident Guideline. [NOCS 100056]
"El Supply appropriate input to the Communication/Report Coordinator to update Florida Nuclear Plant Emergency Notification Form Supplemental Data Sheet for plant conditions information.
"El If any diesel operated equipment is running, evaluate the following parameters (OSC support and local observation might be required to obtain information on support systems and operating parameters):
"* Diesel support systems (i.e., ventilation, fuel transfer, cooling, etc.)
"* Diesel operating parameters
"* Operating EDG load limitation (loaded and unloaded)
"* Fuel and lube oil supplies "El Determine emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements.
EM-225 Rev. 11 Page 13 of 32
ENCLOSURE 5 AAT ENGINEERS CHECKLIST
[NOCS 62764]
EJ Perform Enclosure 10. 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 Dose Assessment Team Leader.
El If RCS LOCA conditions exist, coordinate performance of EM-225B, "Post-Accident Boron Concentration Management."
El Maintain the Plant Parameters Status Board (if required). Based on plant conditions, place key parameters on status board for trending.
El Monitor for conditions listed in Enclosure 11. Provide the AAT Operations Support member with recommended actions.
El If RB temperatures are elevated, coordinate the performance of EM-225C, "Post Accident Monitoring Of Reactor Building Temperature."
O3 If any OTSG level is < 12.5 inches (indicating a dry OTSG), coordinate the performance of EM-225D, "Guidance For Dry OTSG Tube To Shell Delta T Monitoring And Control."
El Evaluate the effects of proposed maintenance repair activities and operational manipulations on plant equipment.
El Develop contingency plans and support emergency repair efforts as applicable.
El If a Severe Accident is in progress, develop mitigation strategies using the Candidate High Level Actions in the CR-3 Severe Accident Guideline.
El 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, establish flow to the RB coolers or ensure only 1 SW pump is running.
El Additional computers may be obtained, as needed, from nuclear administrative building (i.e., engineering laptop computers), that can be used to access documentation on the network.
EM-225 Rev. 11 Page 14 of 32
ENCLOSURE 6 CONTROL ROOM RINGDOWN COMMUNICATOR CHECKLIST El 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 Relay status of overall plant conditions, operator activities and questions to the TSC AAT.
El Relay instructions to Control Room Operators for mitigating actions as directed by the EC.
El Inform Control Room Operators of the following:
Changes in Emergency Classifications TSC repair efforts Operators activities dispatched from the TSC/OSC Changing radiological conditions Mitigation priorities El EOPs or APs in use by Control Room.
El If a Severe Accident is in progress, direct Control Room personnel regarding mitigation strategies, based on actions approved by the TSC Emergency Coordinator.
El Relay requests for support from the Control Room to OSC teams, via the TSC Ringdown Communicator.
El Once TSC is operational, request extra plant operators (if available) be sent to OSC for in plant support. (Ref. EM-103, Enclosure 1, Dispatching of Resources During Emergency Plan Entry)
El Inform TSC of operator actions being performed.
EM-225 Rev. 11 Page 15 of 32
ENCLOSURE 7 NRC COMMUNICATOR CHECKLIST Dl Contact the TSC Report Preparation to determine if continuous communication with the NRC is required.
El Obtain copies of any previously submitted NRC reports.
El If the NRC has requested continuous communication, establish communication with the NRC Communicator on the Emergency Notification System (ENS).
[NOCS 3054, 9405]
El 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 Maintain an open line on the ENS until the NRC agrees to terminate communications. [NOCS 10528]
El Log time(s) when NRC is notified of Emergency Classification changes.
El Log time(s) when NRC is notified of Protective Action Recommendations.
El When communication with the NRC is not required, provide support to other AAT members as needed.
El Make emergency and non-emergency notifications to the NRC as defined in CP-151, External Reporting Requirements.
EM-225 Rev. 11 Page 16 of 32
ENCLOSURE 8 Page 1 of 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 AAT Coordinator immediately if any of the CSFs cannot be verified.
I. REACTOR SHUTDOWN STATUS REACTIVITY CONTROL PARAMETER COMPUTER RECALL POINT POINT All Rods at in-limits P057 RECL-375 YIN 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-1 53 NI-2 cps Low Range RECL NI-14/15 102,103 Adequate Shutdown OP-1 03C Margin Curve 18&19 EM-225 Rev. 11 Page 17 of 32
ENCLOSURE 8 Page 2 of 5 II. CORE COOLING STATUS:
ECCS/SUPPORT STATUS PARAMETER COMPUTER RECALL POINT POINT 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 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 RWPs operating 1/2A/2B/3A/3B DCPs operating A/B (yes/no)
SWPs operating A/B/C SECONDARY SYSTEM STATUS PARAMETER COMPUTER RECALL POINT 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 EM-225 Rev. 11 Page 18 of 32
ENCLOSURE 8 Page 3 of 5 Ill. FISSION PRODUCT BARRIER ASSESSMENT:
-FUEL CLADDING BARRIER:
0 INTACT 0 CHALLENGED 0 LOST 0 REGAINED
- No indication
- RCS condition warrant entry into
- RCS conditions in (or
- Cooling restored, of cladding EOP-07 previously in) Region 3 or no further damage 0 Core Exit Thermocouples > 700F Severe Accident Region degradation
- PASS indicates increased expected.
RCS activity >300pCi/gr 1131 (refer to CH-632A) 0 RM-G29/30 > 100 R/hr for
> 15 minutes
- Enclosure 10 indicates failed fuel
--REACTOR COOLANT SYSTEM BARRIER O INTACT 0 CHALLENGED 0 LOST E0 REGAINED
- Leakage is
- 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 0 Subcooling capacity
- RCS pressure /Tincore subcooling margin Margin restored relationship violates NDT limits 0 RM-G29/30 >1 OR/hr for and leak isolated
& HPI/PORV or HPI/Code Safety valve cooling is in progress
... .... ... CNITAINMVENT BARRIER O INTACT 0 CHALLENGED El LOST 0 REGAINED
"* No evidence a RB pressure > 54 psig
- Containment isolation is
- Repair efforts of
- RB hydrogen concentration > 4% incomplete and release path have isolated containment & RB pressure > 30 psig with no to environment exists. leak leakage building spray available Confirmation may be from
- Containment
"* Tube rupture & RMG-29 or 30 reading elevated radiation readings in pressure has release is > 25,000 R/hr areas adjacent to the RB. reduced to stop only through
- Core conditions in severe
>15 min prolonged steaming to atmosphere or an unisolable steam leak outside RB from affected OTSG.
"* Containment pressure or sump level response not consistent with LOCA conditions
"* Rapid unexplained RB pressure decrease following an initial increase Performed By: Date: Time:
E=M-225 Rev. 11 Page 19 of 32
ENCLOSURE 8 Page 4 of 5 IV. EMERGENCY ELECTRICAL POWER STATUS:
OFF-SITE POWER PARAMETER AVAILABLE UNAVAILABLE 500 KV SWITCHYARD 230 KV SWITCHYARD OFF-SITE POWER XFRM BEST ES BUSES PARAMETER AVAILABLE UNAVAILABLE A-ES 4160V BUS B-ES 4160V BUS A- ES 480V BUS (Note 1)
B-ES 480V BUS (Note 1)
EMERGENCY DIESEL GENERATOR PARAMETER RECALL PT. LOADED AVAILABLE UNAVAILABLE A-EDG RECL-133,171 B-EDG RECL-134,172 DC ELECTRICAL PARAMETER Note (1) AVAILABLE UNAVAILABLE A-BATT'ERY B-BA-FTERY C-BATTERY Note (1) Battery failure will occur if associated battery chargers are de-energized.
EM-225 Rev. 11 Page 20 of 32
ENCLOSURE 8 Page 5 of 5 V. CONTROL COMPLEX STATUS:
CONTROL COMPLEX VENTILATION STATUS PARAMETER AVAILABLE OPERATING UNAVAILABLE A-TRAIN EMERGENCY RECIRC B-TRAIN EMERGENCY RECIRC A-CHILLER B-CHILLER CONTROL ROOM INSTRUMENTATION STATUS PARAMETER AVAILABLE UNAVAILABLE NNI-X NNI-Y ICS EFIC RPS ESAS COMMENTS:
EM-225 Rev. 11 Page 21 of 32
ENCLOSURE 9 DOSE ASSESSMENT TEAM NOTIFICATION
- 1. The Accident Assessment 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 __ W G Decay Tank Rupture __ OTSG Tube Leak
___ Fuel Handling __ Other:
"o LOSS-OF-COOLANT ACCIDENT TIME OF RX TRIP:
- a. Normal Activity Clad damage Fuel melt (from Enclosure 10)
- 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. Aux. Bldg. Ventilation: Flow rate Charcoal banks in service A:B:C:D
- g. Containment pressure 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. Aux. Bldg. Ventilation: Flow rate Charcoal banks in service A:B:C:D "ElSTEAM 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. Aux. Bldg. Ventilation: Flow rate Charcoal banks in service A:B:C:D "0 FUEL HANDLING ACCIDENT
- a. Location of damaged fuel: Pool A Pool B__ Number of Elements
- b. Damage caused by: Mechanical impact Overheating__ Unknown__
- c. Aux. Bldg. Ventilation: Flow rate Charcoal banks in service A:B:C:D
- d. Release pathway: _ Unknown__
- e. Estimated duration Unknown Status as of Date: _ Completed By:
EM-225 Rev. 11 Page 22 of 32
ENCLOSURE 10 Page 1 of 3 CORE DAMAGE ASSESSMENT Determine if core 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 AAT operations support member and the dose assessment team leader. Continue to re-assess core and fission product barrier status as conditions change.
"El ESTIMATE CORE DAMAGE BASED ON RCS SAMPLES.
Core damage assessment based on Reactor Coolant samples will be evaluated by the Dose Assessment Team using CH-632A Enclosure 2. The results will be submitted to the AAT. (May take >2 hours to obtain results)
"C ESTIMATE CORE DAMAGE BASED ON RM-G29/30 RADIATION LEVELS [NOCS 002153]
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 below table assumes a short release. A long-term release cannot be characterized using these tables.
TIME : : : : :
RM-G29 R/HR R/HR R/HR R/HR R/HR RM-G30 R/HR R/HR R/HR R/HR R/HR 11 NO CORE DAMAGE
< 100 R/HR "O POSSIBLE CLAD FAILURE AND GAS GAP RELEASE 100 - 25,000 R/HR WITH RB SPRAY 100 - 75,000 R/HR WITHOUT RB SPRAY "O POSSIBLE CORE MELTING
> 25,000 R/HR WITH RB SPRAY
> 75,000 R/HR WITHOUT RB SPRAY EM-225 Rev. 11 Page 23 of 32
ENCLOSURE 10 Page 2 of 3 CORE DAMAGE PROGRESSION ONCE UNCOVERED El IF inadequate subcooling margin exists, THEN determine if the 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-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.
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 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).
A-HOT LEG B-HOT LEG A-VESSEL B-VESSEL VOID TREND 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 "El CORE REMAINS COVERED TINCORE indicates saturated conditions RCITS indicates any level "El UNCOVERED FOR 15 TO 45 MINUTES Core temperature 1800-2400OF 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 "El 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)
"El 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 EM-225 Rev. 11 Page 24 of 32
ENCLOSURE 10 Page 3 of 3 CORE DAMAGE ASSESSMENT BASED ON ICC CURVE El ASSESS CORE DAMAGE BY PLOTTING RCS PRESSUREINCORE TEMPERATURE ON THE ICC CURVE BELOW.
"El Regions 1 and 2 indicate no fuel damage (normal RCS activity).
"C Region 3 indicates possible gas gap failure.
"El Severe Accident Region indicates possible core melt.
2600 4!
2400 Saturation I 2200 2000 21600
"* ~Region 1 l LU Region 2 IZO° 1200 I-/
1000 IRegion 3 I I IZTcladT.1400°FIi 800/ f'/
600 Severe Accident Region1
/ / / Tcnad >-"1800 °F 400
- 200; / ,
200 300 400 500 600 700 800 900 1000 1100 1200 1300 Tincore (OF)
EM-225 Rev. 11 Page 25 of 32
EN7 'SURE 11
. age 1 of 5 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. TSC GUIDANCE RB Hydrogen EOP-3, 6, HPIC, 5.4 1. Align hydrogen monitoring equipment using EOP-14, Enclosure 2, PPO Control 7, 8 5ll.F, 6.2,10.0, 12.6b, 13.6b Post Event Actions.
LBLO 4.4, 6.3 2. Monitor hydrogen concentrations using EOP-14, Enclosure 21, SBLO 12.4, 20.3, 9.3 RB Hydrogen Monitor Log.
- 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 installing H2 purge flow indicators Building Spray EOP-3, 8 None Verify all of the following before terminating Building Spray:
Termination 1. BS has been on for > or equal to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Criteria 2. RB pressure is < 10 psig.
- 3. RB pressure is stable or lowering.
- 4. RB atmosphere is < 13 pci/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.
Continue EOP-6, 8 FF, 11.5 Verify all of the following:
Cooldown With NC, 11.4 1. Begin establishing a Post Accident Recovery Plan (this can be done during DHR System plant 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.
EM-225 Rev. 11 Page 26 of 32
.. '"SURE11 age 2 of,5 TSC GUIDANCE FOR EOPs PARAMETER EOP TBD REF. TSC GUIDANCE Steaming an EOP-6 IIL.E Steaming an affected OTSG may be desirable for the following reasons:
isolated OSTG for 0 Increase cooldown rate TRACC o Prevent challenging tube to shell dT limits o 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*
AND
- 2) Affected OTSG Level < 90%**
AND
- 3) Any of the following conditions exists:
- Steaming is required to avoid core damage
- Estimated OTSG leakage times RCS DE 1-131 concentrations is < 0.4 OTSG Leakacqe (.qpm) X Initial RCS DE 1-131 (iLci/qm) < 0.4
- Wind is blowing off-shore (Off-shore winds originate from NNE to SE sectors 011.20 to 146.30)
- 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.
- 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.
RCS Leakage No EOP-8 None 1. The RCS is capable of being cooled by DHR.
Longer Exists 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-8 None 1. Establish a Post Accident Recovery Plan. This plan is dependent on the Pump Capability or 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 I and components that are available.
EM-225 Rev. 11 Page 27 of 32
EN(7 'SURE 11 age 3 of 5 TSC GUIDANCE FOR EOPs PARAMETER EOP TBD REF. TSC GUIDANCE Break size < 1 HPI EOP-8 None 1. Transition to DHR cooldown.
Pump Capability 2. Establish a Post Accident Recovery Plan. This plan is dependent on the and able to scope of the applicable Emergency Event.
transition to 3. The Post Accident Recovery Plan is approved by the PNSC, and applicable 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 SS-2 1. Refer to the entry conditions and recommendations of the Emergency Primary to Operating Procedures Technical Basis Document (TBD), Section SS-2 for Secondary Heat guidance related to establishing primary to secondary heat transfer to one Transfer to One or or both OTSGs.
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 EOP-8 LBLO, 2.2, 3.0 1. Recommended guidance is to stop HPI pumps and trip running RCPs and Shutdown of when LPI flow has been in excess of 1400 gpm in each injection line for at RCPs 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 EOP-8 SBLO 12.0 1. Prohibit establishing any flow path that was isolated by the ES system Radioactive Release unless the potential for radioactive releases is evaluated and the release Paths from path, doses and methods have been approved by the EC.
Containment Penetration Valves Monitoring of RB EOP-8 None NOTE: With the installation of the TSP baskets, pH data is not required Sump Level, RB but still desired if feasible.
Sump Boron Other: IOC CR 97-0122 1. Accident Assessment personnel to monitor and trend RB sump level, boron Concentration, and concentration, and pH at intervals recommended by the EC.
RB Sump pH 2. Data for sump pH and boron concentration to be obtained using CH-632D or other PNSC approved alternate methods dependent on the Emergency Event.
EM-225 Rev. 11 Page 28 of 32
ENCLOSURE 11 TSC GUIDANCE FOR EOPs Page 4 of 5 PARAMETER EOP TBD REF. TSC GUIDANCE Venting of Non- EOP-8 None 1. Once subcooling margin is regained, all of the noncondensable gas Condensable Gases 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-8 SBLO, 17.7 1. Verify TBVs/ADVs are closed.
Adequately Cooled 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 EOP-8 None Refer to EM-225B Concentration EOP-14, Management When Enc. 20 NOTE: If a failure of ES MCC 3AB has occurred, ensure repair efforts are Adequate Sub initiated to repower auxiliary pressurizer spray valve RCV-53 prior Cooling Margin to the onset of boron precipitation.
Does Not Exist (Boron Precipitation)
RB Temperature Refer to EM-225C Monitoring (To Preserve EQ Standards)
Feeding a Dry OTSG EOP-5, 9 Ill.D, 12.0 Refer to EM-225D (Tube to Shell EOP-14, Ill.E, 17.7 Delta T Monitoring Enc. 3 NC, 5.2, 5.3, 6.4 and Control)
Long-Term Core EOP-8 LBLO, 6.4a, 6.4b, 6.6, 6.7 Refer to EM-225E Cooling Using the RB Sump EFW or AFW is EOP-14, Refer to EM-225F Operating Enc. 7 Enc. 22 TBP-3 is Running. EOP-14, TBP-3 will drain non-1 E battery during LOOP. Stopping TBP-3 before 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> TBP-2 is Not Enc. 14 may result in Turbine bearing damage.
Running. Generator Purge Complete Refer to IOC SE-99-0184 EM-225 Rev. 11 Page 29 of 32
ENCLOSURE 11 TSC GUIDANCE FOR EOPs Page 5 of 5 PARAMETER EOP TBD REF. TSC GUIDANCE Concentrated BA EOP-14, None Ifconcentrated BA is allowed to remain in the boron injection path piping addition made and Enc 18 (letdown/DH purification piping) the BA will eventually cool down and solidify.
flush water not Timely action is required to preclude this condition.
available.
- 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.
- Ifplant conditions permit, expedite restoration of RCS letdown (or DH purification).
- Ifplant condition permit expedite restoration of power to at least one source of flush water (DWP-IA, 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 solidificaiton.
- 4. IF letdown or DH purification flow is established, THEN direct the control room to STOP concentrated BA additions.
- 5. IFany 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 EM-225 Rev. 11 Page 30 of 32
ENCLOSURE 12 (SAMPLE)
TSC ACCIDENT ASSESSMENT TEAM OSC Request Form INSTRUCTIONS:
- 1. Use this form for each requested action from the Control Room, or Accident Assessment Team
- 2. Obtain approval from the AAT Lead Coordinator
- 3. Obtain acknowledgement from TSC Repairs Coordinator
- 4. Make copy and give original to TSC Repairs Coordinator
- 5. Give copy to Ringdown Communicator
- 6. Feedback to the Control Room on status of request.
EM-225 Rev. 11 Page 31 of 32
Revision Summary SECTION I CHANGE Inclosure 11, page 5 If concentrated BA is allowed to remain in the boron injection path piping (letdown/DH purification piping) the BA will eventually cool down and solidify.
EOP-14, Enclosure 18 provides guidance for flushing BA from the injection line.
If a flush water source is not available Enclosure 18 directs the control room to notify the TSC.
This revision to enclosure 11 adds TSC guidance for mitigating this problem.
This change is listed as "change of intent" solely because it involves a "procedure related to Emergency Preparedness."
References:
EEM-01 -021, Boration Flush Volume Flow Necessary for Boron Solubility NCR 42781 FSAR , Section 4.2 and Table 4-10 EM-225 Rev. 11 Page 32 of 32