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| Virginia Electric and Power Company I
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| North Anna Unit 1 l July 15,1987 Steam Generator Tube Rupture Event Report I t
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| l July 29,1987 Revision 0
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| VIRGINIA ELECTRIC AND POWER COMPANY i
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| NORTH ANNA UNIT I I JULY 15,1987 STEAM GENER ATOR TUBE RUPTURE EVENT REPORT !
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| JULY 29,1987 l
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| REV.0
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| NORTH ANNA UNIT 1 JULY 15, 1987 STEAM CENERATOR TUBE RUPTURE EVENT REPORT l
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| TABLE OF CONTENTS
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| CHAPTER PAGE l I. EXECUTIVE
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| ==SUMMARY==
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| A. Purpose 6 B. Brief Description of North Anna Power Station 6 C. Overview of Event 7 D. Virginia Electric and Power Company and NRC 8
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| ===Response===
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| 1 II. DESCRIPTION OF EVENT A. Introduction 12 B. Conditions Prior to the Event 12 C. Event Description and Analysis 13 D. Operational Analysis 15 E. Procedure Utilization and Analysis 17 III. RADIOLOGICAL EFFECTS OF EVENT A. Summary 25 B. Evaluation of Radiological Release 25 IV. EMERGENCY RESPONSE A. Emergency Plan Implementation 42
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| : b. Emergency Response Computer System 45 V. SAFETY EVALUATION A. Comparison of North Anna Unit 1 Tube Rupture Event 47 to UFSAR Analysis B. Fuel Integrity Evaluation 47 C. Conclusions 48 i
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| Page 1
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| NORTH ANNA UNIT 1 JULY 15, 1987 STEAM GENERATOR TUBE RUPTURE EVENT REPORT 1
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| TABLE OF CONTENTS 1
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| CHAPTER PAGE r
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| l VI STEAM GENERATOR EVALUATION l
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| A. North Anna Steam Generator Operating Experience 52 i B. Current Outage Inspection Activities 54
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| ! VII. LESSONS LEARNED I
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| i LIST OF FIGURES FIGURE NUMBER DESCRIPTION I-1 Recovery Organization I-2 Nuclear Steam Supply System I-3 Westinghouse Series 51 Steam Generator ;
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| II-1 Pressurizer Level 11-2 Reactor Coolant Pressure (Loop A) l II-3 Reactor Coolant Temperature (Loop A) l II-4 Steam Generator "C" Level II-5 Steam Generator "C" Pressure II-6 "A", "B" and "C" Steam Cenerator Levels II-7 North Anna Unit 1 Power History Prior to I Event II-8 On-line Chemistry Monitoring System Data V-1 Reactor Core Safety Limits and State Point I Prior to Trip VI-1 Map of Plugged Tubes - Steam Generator "A"
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| VI-2 Map of Plugged Tubes - Steam Generator ugn VI-3 Map of Plugged Tubes - Steam Generator ,
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| "C" 1
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| LIST OF TABLES TABLE NUMBER TITLE II - 1 North Anna Unit 1 Operating Conditions Prior to Steam Generator Tube Rupture II - 2 Reactor Coolant System Leak Rate Prior to Tube Rupture Event II - 3 Unit 1 Air Ejector Activity Prior to Tube Rupture Event III - 1 Calcuated Release Activity From Condenser Air Ejector For Time Period From 0630 to 0756 Houre l
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| III - 2 Calculated Noble Gas Release Activity From Steam Driven '
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| Auxiliary Feedwater Pump Exhaust For Time Period From 0636 to 0648 Hours III - 3 Caluclated Release Activity (Excluding Noble Gases) From Steam Driven Auxiliary Feedwater Pump Exhaust For Time Period From 0636 to 0648 Hours III - 4 Calculated Release Activity from Steam Driven Auxiliary Feedwater Pump Exhaust For Time Period From 0648 to 0745 )
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| Hours l l
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| III - 5 Reactor Coolant Radiogas Concentrations and Corresponding Release Activities Based on Mass Balance Calculations III - 6 Environmental Samples - 15 July 1987 l V-1 North Anna Unic 1 Steam Generator Tube Rupture -
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| Comparison to UFSAR Accident Analysis V-2 North Anna Unit 1 Steam Generator Tube Rupture -
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| COBRA /WRB-1 Point Estimate of Minimum DNBR VI - 1 Steam Generator Tube Plugging By Indication Type During 1987 Refueling Outage l VI - 2 North Anna Unit 1 Tube Plugging Summary Page 4 4
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| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _________.__J
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| LIST OF ATTACHMENTS ATTACHMENT TITLE 1 Chronology of Steam Generator Tube Rupture Event 2 Significant Operational and Maintenance Activities Prior to Event 3 Sequence of Events - Health Physics l 4 Emergency Plan Chronology i
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| 5 Emergency Operating Procedure EP-0, Reactor Trip or Safety i Injection 1 6 Emergency Operating Procedure EP-3, Steam Generator Tube Rupture 7 Emergency Operating Procedure ES-3.1, Post Steam Generator Tube Rupture Cooldown Using Backfill ,
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| L__ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ . _ _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ . _ _ _ _ _ _ _ . _ . . _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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| I. EXECUTIVE
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| ==SUMMARY==
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| A. Purpose This report provides a detailed description of the July 15, 1987 steam generator tube rupture event at North Anna Unit 1, and the ensuing evaluations and actions performed and planned by the Virginia Electric and Power Company (Virginia Power).
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| {
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| The detailed sequence of events including the response of the operators and j key plant equipment is discussed in Chapter II. The radiological effects of the event are evaluated in Chapter III. An evaluation of the implementation of the Emergency Plan is discussed in Chapter IV. The safety consequences and significance of the event are discussed and evaluated in Chapter V. The cause of the steam generator tube failure and the actions taken to address the tube failure are discussed in Chapter VI. And finally the lessons learned and corrective actions taken or planned by Virginia Power as a result of the event are discussed in Chapter VII.
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| B. Brief Description of North Anna Power Station The North Anna Power Station is a two unit station owned jointly by Virginia Electric and Power Company and Old Dominion Electric Cooperative and operated by Virginia Electric and Power Company (Virginia Power). It is located on the southern shore of Lake Anna, in Louisa County, approximately 40 ]
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| miles north of Richmond, Virginia. Each unit includes a three-loop pressurized j water reactor nuclear steam supply system (Figure I-2) and turbine generator furnished by Westinghouse Electric Corporation. The balance of plant was designed and constructed by Virginia Power with the assistance of Stone and ,
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| ! Webster Engineering Corporation. Each nuclear unit is licensed to operate at a I core power of 2893 MW thermal. I i
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| Each unit has three Westinghouse Series 51 steam generators illustrated on Figure I-3. These Series 51 steam generators have 3388 tubes, 0.875 inch 0.D.
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| by 0.050 inch wall thickness. The tubing is Inconel 600 in the mill annealed condition. The tubes are hardrolled above the bottom of the tubesheet, and were explosively expanded prior to operation to eliminate the open annular ;
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| crevice that remained within the tubesheet. Each steam generator has seven l tube support plates which are carbon steel with drilled tube holes having a small clearance between the tube and support plate.
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| Unit 1 achieved commercial operation in June 1978 and has completed operation of its sixth fuel cycle. To date, Unit I has operated for approx-imately six effective full power years.
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| C. Overview'of-Event t
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| At approximately 0635 hours on July 15, 1987, North Anna Unit I was manually tripped from 100% power due to indications of a steam generator tube rupture in the "C" steam generator. Approximately twenty seconds later, the safety injection system automatically initiated. ):
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| i At 0639 hourc, a " Notification of Unusual Event" was declared after )
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| indications of the tube rupture (primary to secondary leakage exceeding one j gallon per minute) in "C" Steam Generator were evaluated by control room j personnel. Initial notifications to State and local governments were completed by 0651 hours. By 0654 hours, the event was upgraded to an " Alert" classification at the direction of the Interim Station Emergency Manager.
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| Notifications to the NRC and offsite agencies that were required as a result of the upgraded classification were completed by 0702 hours. All major plant equipment functioned properly, and the unit was stabilized in accordance with Station Emergency Operating Procedures. In addition. Emergency Plan Implementing Procedures respective to an " Alert" classification were ,
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| implemented to augment the onshift staff. By 0728 hours, an accountability of q all station personnel had been conducted. By 0757 hours, the Technical Support Center and Operational Support Center were fully manned and activated.
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| By 0915 hours, the Local Emergency Operations Facility was fully manned and activated.
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| I 1
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| The initial indications available to the operators were from the main i steam line radiation monitors which went into alarm at approximately 0630 hours on July 15, 1987. Subsequently, pressurizer pressure and level began to decrease rapidly. Pressurizer pressure decreased (after the manual reactor
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| . trip) to approximately 1730 psig which is below the. safety inj ec tion (SI) setpoint of 1765 psig. Pressurizer level decreased from its'100%' power ,
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| program level of approximately 65% to offscale low prior to recovery. The I maximum primary to secondary leak rate has been estimated to be 550 to 650 gpm. The re-establishment of pressurizer level and isolation of the ruptured j "C" steam generator was completed under Station Emergency Operating Procedures (EP-0 Reactor Trip or Safety Injection and EP-3, Steam Generator Tube Rupture Revision 1). EP-0 and EP-3 are included as Attachments 5 and 6.
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| Cooldown and depressurization of the Reactor Coolant System (RCS) was initiated at 0710 hours in accordance with Station Emergency Operating Procedure ES-3.1, Revision 1, Post Steam Generator Tube Rupture Cooldown Using Backfills (Attachment 7) and proceeded routinely. At 1108 hours, the RCS was cooled down to below 350 degrees F and Unit I was placed in Mode 4, Hot Shutdown. At 1336 hours, the RCS was cooled down to less than 200 degrees F.
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| and the unit was placed in Mode 5, Cold Shutdown. Plant conditions'were stable and the emergency was terminated at 1336 hours, at which time r termination notifications were made to the NRC and of fsite agencies. !
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| Page 7
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| During the course of the event, effluents were released from "C" steam generator prior to its. 1 solation through the condenser air ejector and auxiliary. feedwater pump turbine exhaust. The releases rates were calculated based on isotopic analysis of samples ar.d radiation monitoring system data.
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| The noble gas release rate averaged over a one-hour period was 0.33% of the Technical Specification limit.
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| D. Virginia Electric and Power Company and NRC Response Following the termination of the emergency, Virginia Electric and Power !
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| Company (Virginia Power) management immediately initiated recovery activities.
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| l An organization was established and resources identified for evaluating the j incident and recommending recovery actions. In addition, the NRC Region II, in Atlanta,' Georgia, notified Virginia Power that an 1nspection team was being dispatched to North Anna. The NRC's inspection team was on site and first met with Virginia Power's recovery organization at 1455 hours on July 15, 1987.
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| The team was upgraded to an Augmented Inspection Team (AIT) after arrival.
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| The AIT was designated to examine the Virginia Power's response to the incident and perform a separate investigation. Virginia Power's recovery organization was in place and functional immediately following termination of the emergency.
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| The Virginia Power Recovery Organization (Figure I-1) was divided into six groups reporting to the Recovery Manager. Overall coordination of the event evaluation and recovery actions was the responsibility of the Recovery Manager and the Local Emergency Operations Facility (LEOF) Command Center.
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| The Maintenance and Repair group was responsible for outage planning and steam generator maintenance activities. The Technical Evaluation group was responsible for the steam generator inspection program, tube failure analysis, steam generator integrity evaluation and tube plugging criteria and recommendations. The Nuclear Safety and Licensing group was responsible for the event evaluation including evaluation of the response of the plant and i emergency response organizations, as well as interfaces with the NRC. The l Industry Interface group was responsible for interfacing with the media as well as INPO, State and local authorities, and Old Dominion Electric Cooperative. The Health Physics group was responsible for evaluating the radiological effects of the event.
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| Page 8
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| l 1
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| I J
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| I II. DESCRIPTION OF EVENT A. Introduction i The following narrative describes the significant operating events for the North Anna Unit 1 steam generator tube rupture which occurred on July 15, 1987.
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| Attachment 1 provides a more concise and detailed chronology of events. The narrative and sequence of events were compiled from a variety of sources including 1) the Sequence of Events Recorder (SER), 2) the Control Room P-250 process computer and alarm printouts, 3) the Emergency Response Facility Computer System (ERFCS) historical file, 4) interviews with licensed Control Room Operators (CRO) and Senior Reactor Operators (SRO), 5) CR0 and SRO logs 6)
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| Emergency Response Facilities logs and 7) strip charts from Control Room recorders. The response of important reactor and secondary system parameters, including pressuritar level, RCS pressure and temperature and ruptured steam generator level and pressure are provided on Figures 11-1 through II-6 to assist the reader in following the event and associated recovery actions.
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| B. Conditions Prior.to the Event Prior to the event North Anna Unit I was operating at 100% power (which was first achieved on July 14, 1987) after having been returned to service fron a refueling outage on June 29, 1987 (See Figure II-7 for power history). Reactor and secondary system conditions were normal and are shown on Table II-1.
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| Reactor Coolant System (RCS) leak rate measurements taken the day before the event indicated less than 0.25 gpm unidentified leakage. Results of RCS leak l rate measurements taken prior to the event are shown on Table II-2. Two mair.
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| feedwater pumps, three high pressure heater drain pumps, two low pressure heater drain pumps, and three condensate pumps were in service on the secondary system.
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| Operation of charging and letdown was normal with one charging pump in 1 I
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| operation. Se rvice watet and component cooling systems were normal with one component cooling pump and one service water pump on each unit in service.
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| The air ejector radiation monitor (RM-RMS-121) had been declared inoperable at 0809 hours on July 13, 1987 when it was det' ermined to be indicating low. ;
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| After condensation was drained from the monitor, it was returned to operable ]
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| status at 0815 hours on July 14, 1987. Later, the monitor again operated i erratically and was once again declared inoperable at 2238 on July 14, 1987. No ;
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| other safety-related or other equipment important to the tube rupture event was j out of service at the time of the event, e a result of the air ejector radiation monitor being inoperable, grab i samples were being taken every eight to twelve hours in accordance with Station Health Physics procedures. The Technical Specifications requirements was to l l obtain a backup " grab" sample on a 12 hour. frequency. A sample taken at 2245 J l hours on July 14, 1987 showed slight but not significant increase in air ejector i activity from the previous sample at 0114 hours on the same day. However, the sample taken at 0620 hours on July 15, 1987, and analyzed at 0623 hours, showed a significant increase in activity. However, there was not sufficient time to report this inform < zion to the Shift Supervisor prior to the event. Results of isotopic analysi; of grab samples taken prior to the event are shown on Table II-3. 1 Page 12 '
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| r i _ _ _ - - - - - - - - . - - - - - - - - - _ - - - - - - - _ - _ _ _ a
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| 7 I
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| l Unit 1. secondary system chemistry data, including silica, chloride, sodium, i sulfate, cation conductivity, boron, pH, were reviewed beginning with the j startup from the refueling outage through July 15, 1987.
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| Although the plant experienced several chemistry transients due to multiple startup and shutdowns 3 over this period, there were no precursor indications of the tube rupture. The
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| ]
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| on-line chemistry monitoring system, which is undergoing pre-operational j checkout, was partially in service during the time of the event. The on-line j system indicated a decrease in "C" main steam pH and an increase in specific conductivity at approximately 0624 hours on July 15, 1987 (Figure II-8).
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| A chronological listing of significant operational and maintenance events ;
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| involving Unit 1 prior to the time of the event is provided in Attachment 5.
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| At .the time of the event, the operating shift was manned by three Senior Reactor Operators, including the Shift Supervisor, four licensed Reactor 1 Operators and seven unlicensed Reacter Operators. A Shift Technical Advisor j (STA) with an active SRO license was also assigned to the operating shift.
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| l i
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| l C. Event Description and Analysis At 0630 hours on 7/15/87, an alarm was received on the Unit 1 Annunciator Panel for Main Steam (High Range) Radiation Monitor. Upon investigation, it was determined that the NRC (Nuclear Research Corporation) Radiation Monitors on the l "A" and "B" main steam lines were in " ALERT" and the "C" main steam line monitor was in "HIGH" alarm. Subsequently, the Unit 1 Control Room Operator (CRO) noted pressurizer level decreasing rapidly, and informed the Unit 2 Senior Reactor Operator (SRO), since the Unit 1 SRO and the Shift Supervisor were not in the Control Room at this time. The two SR0s were recalled to the Control Room and the Superintendent of Operations was contacted at home by the Unit 2 SRO. The Shift Technical Advisor (STA) and NRC Resident Inspector also responded to the
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| ;SFO recall and reported to the Control Room. Meanwhile, the Unit 1 CR0 took
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| ' manual control of makeup and set the normal charging flow control valve (FCV-1122) to full open.
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| l Upon return to the Control Room, the Shift Supervisor was advised by the l Unit 2 SR0 that a steam generator tube rupture may have occurred. The Shift l Supervisor the took command of Unit 1 and directed that letdown from the Reactor Coolant System (RCS) to Chemical Volume Control System (CVCS) be isolated. The Unit 1 CR0 also aligned charging pump suction from the Volume
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| , Control Tank (VCT) to the Refueling Water Storage Tank (RWST) due to the rapid (inventoryloss. The Shift Supervisor also directed that a power ramp down at 2%
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| lper minute be initiated. The Unit 1 SRO returned to the Control Room and l assisted the Shift Supervisor during the event.
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| Pressurizer level and pressure. continued to decrease and at approximately 45% level (100% program level is 65%) and 2100 psig (normal operating pressure lis 2235 psig), the Shift Supervisor directed the Unit CR0 and Backboards CR0 l(who had assumed Unit 1 Balance of Plant duties) to manually trip the unit. The (manualreactorandturbinetripswereperformedat0635hoursandEP-0, Revision l1 (Reactor Trip or Safety injection) was implemented. An additional licensed ICR0 from the shift assumed the responsibility of " Procedure Reader".
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| 1 I
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| Page 13 l
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| .c e
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| Approximately 20 seconds after the manual Reactor Trip, an automatic Safety Injection (S1) was received on " low-low" pressurizer pressure (less than 1765 psig on 2 out of 3 channels). EP-0 was completed through Step 23 at which time the Shif t Supervisor directed a transition to EP-3, Revision 1 (Steam Generator Tube Rupture).
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| A " Notification of Unusual Event" (NOUE) was declared at 0639 hours. The Unit 2 SRO became the Interim Station Emergency Manager and assigned two unlicensed Control Room Operators to be communicators to the State / Local governments and to the NRC. Notifications of the NOUE were made within 15 minutes to the State and Local governments as required by the Emergency Plan.
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| Communication was established with the NRC and con:inuously maintained throughout the event.
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| The basic strategy for the response to and mitigation of a steam generator tube rupture is as follows: 1) Identify the ruptured steam generator (s), 2)
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| Isolate the flow from the ruptured steam generator (s), 3) Initiate RCS cooldown (to establish subcooling margin), 4) Depressurize the RCS (to minimize breakflow and refill the pressurizer), and 5) terminate SI (to prevent going solid in the RCS).
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| At Step 3 of EP-3, the ruptured steam generator is identified. Although there were several possible indications (i.e., steam generator level and steam line radiation) that "C" steam generator was the ruptured steam generator, the Shift Supervisor was not absolutely sure and decided to continue on without identifying the ruptured steam generator (i.e., the Response Not Obtained column of Step 3 was entered).
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| l Subsequently, after resetting S1 and the Phase A Containment Isolation signals, the Shift Supe rvisor directed that Auxiliary Feedwater flow to be isolated to "C" steam generator for verification. After isolation, it was noted that level in "C" steam generator continued to rise. Based on this
| |
| " uncontrolled" increase in level, the Shift Supervisor identified "C" steam generator as the ruptured generator and proceeded with the steps in EP-3 to i isolate steam flow from the "C" steam generator. Once isolation of "C" steam i generator was completed, a rapid cooldown of the RCS to approximately 480*F was q initiated. '
| |
| During the above evolutions, the Interim Station Emergency Manager upgraded the "NOUE" to an " ALERT" status. Within 8 minutes of this declaration, the required notifications to the State / Local governments and to the NRC were made.
| |
| ' Following the " ALERT" declaration, a callout of offsite station emergency i
| |
| response personnel was made along with the accountability of all personnel at ;
| |
| ( the station.
| |
| The Health Physics shift leader was requested to report to the Control Room to implement EPIP 4.01, Radiological Assessment Director Controlling Procedure.
| |
| The rapid cooldown of the RCS as required by EP-3 was accomplished by dumping steam from the "A" and "B" steam generators to the main condenser via two steam dump valves. After the RCS was cooled down, RCS pressure was rapidly I reduced and equalized with the ruptured steam generator's pressure.
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| Depressurization was accomplished by using both pressurizer spray valves and one pressurizer power operated relief valve (PORV). As S1 flow increased over the steam generator primary to secondary leakage flow, due to RCS depressurization, ,
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| Page 14 i
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| pressurizer level came back on scale and continuously increased. When pressure equalized between the RCS and the ruptured steam generator, the primary to secondary leakage slowed and the level in "C" steam generator stopped increasing. Subsequently, the criteria for SI termination was met, and one Charging /High Head Safety Injection pump was secured. Normal makeup to the RCS was then established, and the SI flow path through the Boron Injection Tank (BIT) was isolated. Normal letdown from the RCS to the CVCS was established i next.
| |
| An orderly cooldown and depressurization of the RCS to Cold Shutdown was !
| |
| initiated at 0718 hours (48 minutes from the initial indication) by transitioning from EP-3 to ES-3.1, Revision 1 (Post-SGTR Cooldown Using Backfill). Operating procedures were also referred to and utilized as appropriate during the cooldown. The cooldown rate was limited to approximately.
| |
| 42' F per hour in order to allow depressurization of the ruptured steam generator at the same rate as the depressurization of the RCS. The cooldown rate was maintained initially by dumping steam from the intact steam generators to the condenser via two steam dump valves and then by the RHR system. Cold Shutdown (Mode 5) was reached at 1330 hours, and the emergency was terminated at 1236 hours hours and the " Recovery" phase was initiated.
| |
| The transition from EP-3 to the post-SCTR cooldown procedure allows for three options. These options are: 1) Post-SGTR Cooldown Using backfilling (ES 3.1), 2) Post-SGTR Cooldown using Blowdown (ES 3.2), and 3) Post-SGTR Cooldown using steam dump (ES 3.3). Since RCS letdown to CVCS was available, cooldown by backfilling (ES 3.1) was the option of choice. This option minimized the amount of RCS (i.e. contamination) in the secondary system while maintaining the most rapid cooldown rate allowed by Technical Specifications. The methodology used in ES 3.1 is to backfill the ruptured steam generator secondary side inventory into the primary through the ruptured steam generator tube. This is accomplished by cooling the RCS and depressurizing the RCS until subcooling 1 dropped below 30*F on the core exit thermocouple or pressurizer level increased above 70%. Letdown was used to control pressurizer level between 30% and 70%. 3 Auxiliary feedwater and condensate feed were used to maintain the required level I in both the intact and ruptured steam generators. The above sequence was repeated until the RCS was lower than 120'F and 100 psig. After the "C" RCS loop was isolated and drained, the secondary side of the "C" steam generator (the ruptured generator) was purged with nitrogen and vented through the Gaseous Waste process vent charcoal and HEPA filter. Later, the secondary side )
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| inventory was drained to the Liquid Waste system and processed. The secondary side was also flushed to remove any remaining boron.
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| l Prior to isolating the "C" loop, the "C" steam generator was still at l satursted conditien. As a result, the "C" steam generator responded like a
| |
| " pressurizer", and as it cooled down, pressure in the steam generator decreased.
| |
| With decreasing pressure in the steam generator, RCS pressure had to be reduced in order to preclude refilling of the steam generator with RCS, Once the loop was isolated, this concern was eliminated. However, until the ruptured steam generator was cooled sufficiently to allow draining, the RCS water in the secondary side of the steam generator continued to degas. Prior to venting, the hydrogen concentration had increased to approximately 15%.
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| I Page 15
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| 1 i
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| I l
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| D. Operational Analysis The purpose of this section is to analyze the performance of the Operating and Health Physics shifts on duty during the event and the transitions of key emergency management responsibilities.
| |
| The Unit 1 Control Room Operators' (CR0s), Senior Reactor Operators' (SR0s), l and the Shift Technical Advisor's (STA) overall response to the steam generator tube rupture was excellent. The reactor was manually tripped and immediate actions (as required by EP-0, Revision 1, Reactor Trip or Safety Injection) were ;
| |
| performed in a timely manner.. The automatic initiation of Safety Injection (SI) l and the indications of a tube rupture were quickly recognized. The ruptured j steam generator was isolated within 13 minutes of' safety injecting and did not i overfill. Pressurizer level was restored and pressurizer pressure was stabilized within 34 minutes of safety injecting. Pressurizer level was restored without going solid, and the reactor trip and cooldown were .
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| accomplished without lifting any steam generator power operated relief valves I (PORVs) or safety valves. The cooldown to Cold Shutdown was conducted in a .I controlled manner and without incident. The Unit 2 CR0 performed Abnormal i
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| ' Procedure 47 (AP-47, Unit Operation During Opposite Unit Emergency) to verify l that the Unit 1 transient had no adverse impact on the continued safe operation I of Unit 2. The operational performance exhibited is attributed to the classroom )
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| and simulator training received on steam generator tube ruptures which is structured to integrate the Emergency Operating Procedures, Emergency Plan Implementing Procedures, and the Critical Safety Functions.
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| The Emergency Plan was implemented by the Interim Station Emergency Manager l (the Unit 2 SRO) in an excellent manner. The correct emergency classifications were made in a timely manner, and the predesignated communicators for the NRC and State / Local governments performed their initial notifications and updates as ;
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| required by the Emergency Plan. l During the early stages of the event, the Interim Station Emergency Manager i consulted, by telephone, with the Station Manager and the Superintendent of Operations. Upon arrival of the Station Manager, the Interin Station Emer6ency Manager began an orderly and detailed turnover. This was followed by an orderly
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| , turnover of the Emergency Plan Implementing Procedures and communication ,
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| responsibilities to the Assistant Station Manager. Upon completion of these !
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| l activities, the Station Emergency Manager (Station Manager) and Procedures
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| , Coordinator (Assistant Station Manager) relocated to the already mobilized
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| , Technical Support Center.
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| l' The Interim Station Emergency Manager interface with Health Physics was adequate, although the demands to classify the emergency and implement the Emergency Plan hampered thorough communications with Health Physics. The interim Radiological Assessment Director was not given a detailed briefing on where the radioactive release paths had occurred from. This was compounded by the very small amount of radiation being released (no radiation monitors were in ,
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| ! alarm after the ruptured steam generator had been isolated), and the fact that !
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| the air ejector radiation monitor was inoperable. It should also be noted that by the time the interim Radiological Assessment Director reported to the Control Room, the "C" steam generator had been completely isolated.
| |
| Most of the operating and health physics shifts on duty at the time of the event had been on duty since at least midnight. The health physics normal shift relief was arriving about the time the event initiated. The operations shift, Page 16
| |
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| which was not due to be relieved until 0800 hours, was quickly supplemented by the " Relief Office" crew which was already on site. The " Relief Office" crew (operations personnel assigned to tasks other than direct day to day operations ;
| |
| of the plant) consisted of two licensed CR0s and three experienced, unlicensed '
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| CRos which were utilized by the Shift Supervisor and Interim Station Emergency !
| |
| Manager as required. Upon arrival of the " day" operating shift, they reported to Operations Support Center (OSC) as required by the Emergency Plan. When the OSC had assembled more than sufficient operational personnel, the " day" shift was reassigned to the Control Room to begin a shift turnover. The turnover for licensed operators took place one at a time and was performed by having the replacement operator work with the on duty operator for 20 to 60 minutes to ensure continuity and full understanding of plant conditions by the oncoming operators. !
| |
| The STA made a major contribution to the mitigation of the event and assisted the Shif t Supervisor in keeping up with key plant responses. The STA advised the Shift Supesvisor that the Source Range Nuclear Instrumentation did not automatically energize and calculated cooldown rate information. The STA also monitored the Critical Safety Functions and the Safety Parameter Display System. Two " Yellow" paths on the Critical Safety Functions were monitored during the event. The STA also monitored the Integrated Core Cooling Monitoring .
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| System which provided RCS inventory (RVLIS) and core subcooling data. l l
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| E. Emergency Operating Procedures (EOPs) Utilization and Analysis The Revision 1 E0Ps being utilized at North Anna Power StLtion were iiaplemented April 30, 1987. The current edition of the Unit 1 E0Ps used during the Steam Generator Tube Rupture event was June 12, 1987 (i.e. Unit 1 E0Ps were {
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| updated at the completion of the refueling outage to reflect modifications !
| |
| installed during the outage). Prior to April 30, 1987, each licensed operator l and STA had received classroom training on the E0Ps (to include ECAs and FRPs) l and simulator training on most of the E0Ps. The training was conducted during i' two cycles of the Licensed Operators Requalification Program. As a result of the training, licensed operators had a good working knowledge, as well as, confidence in the accuracy and effectiveness of the E0Ps.
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| Based on interviews with the licensed operators on the shift during the event, it was concluded that the E0Ps were very effective. However, there have been several areas identified that may need further review. First is Step 3 of EP-3, which is used to identify the rupturoj steam generator. During the event and subsequently confirmed by the modeling faf the event on the North Anna Simulator, Step 3 of EP-3 was reached bef;re a " positive" identification could be made. The procedure can accommodate inability to immediately identify the ruptured steam generator by the Response No': Obtained (RNO) section of the step.
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| However, by using the RNO, the jsolation of the ruptured steam generator (and ansociated radioactivity relvase path) is delayed and the Low Head Safety Injection tumps are secured based on RCS pressure. The second is the lack of a specific e Lep in the procedure to evaluate the air ejector exhaust requirements.
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| The final concern is the delay until Step 32 for stopping and shutting down the Emergency Diesel Generators (EDG) which may result in excessive time in running with the EDGs unloaded.
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| I Page 17 m__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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| l 1
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| 1 i
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| It should be noted that none of the above concerns affected the operator's l nbility to adequately use the procedure to respond to and mitigate the tube rupture in a timely manner. However, a thorough study of these potential
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| :ence rns , as well as, the entire use of E0Ps during the Steam Generator Tube Rupture event should be conducted by the Westinghouse Owners Group (WOG)
| |
| )perations Subcommittee.
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| From the perspective of the operator, the initial indications were all three i gain steam line radiation monitors ("A" and "B" were in " Alert" alarm and "C" I sas in "High" alarm). This is symptomatic of a steam generator tube rupture
| |
| [SGTR). The next indications available were rapidly decreasing pressurizer Level and pressure. This is symptomatic of either a Loss of Coolant Accident
| |
| [LOCA) or a SGTR. The Shift Supervisor directed action to isolate normal Le tdown and start a second charging pump. (A second charging pump was )
| |
| inadvertently not started but the normal makeup flow control valve was opened 1 fully and charging pump suction was shifted from the VCT to the RWST.) The Bhift Supervisor then checked containment conditions and found them normal. The absence of radiation monitor alarms in the Auxiliary Building indicated no tormal RCS makeup or letdown line leak outside of containment. There was no abnormal indication on steam generator levels or flow rates at this time, and
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| :herefore even after enter 1ng EP-0 and safety injecting, the Shift Supervisor ras not positive of what the problem was (although a SGTR was suspected). At step 23 of EP-0, a transition was made to EP-3 based on steam generator main steam line radiation monitors being in or having been in alarm. Prior to meering EP-3, the steam dump valves had modulated closed as RCS T decreased
| |
| :o and then below 547'F due to full auxiliary feedwater flow to aiY*three steam
| |
| ;enerators. The closing of the steam dump valves stopped steam flow (except for
| |
| ;he steam flow to the steam driven auxiliary feedwater pump) and the main steam Line radiation monitor alarms cleared. When step 3 of EP-3 was reached approximately 0640 hours) the only indication available for identifying the mptured steam generator was " Unexpected Increase in Steam Generator Narrow lange Level". Figure 11-6 depicts Narrow Range level in all three steam ,
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| ;enerators during this time period. (Note: A steam generator sample had been I
| |
| >equested, but was not yet available and the steam generator blowdown radiation Bonitors were isolated by the Phase A Containment Isolation). The Shift supervisor was aware that "C" steam generator level was higher than "A" which zas unusual because "A" steam generator is supplied by the steam driven exiliary feedwater pump which refills faster than the motor pumps which supply
| |
| ;he "B" and "C" steam generators. The Shift Supervisor was also aware that "C" min steamline radiation monitor had been reading higher, but at this time was to longer in alarm. As a result, no decision to isolate a steam generator was lade because of the overriding concern of isolating the wrong steam generator. I he concern w1th isolating the wrong steam generator is the loss of that steam
| |
| ;@nerator for cooldown via the condenser. The steam generator would still be Lvailable for cooldown via the PORV to the atmosphere but this couldown path is mch less desirable.
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| From Step.3, the Shift Supervisor prc:eeded to Steps 6 through 13. At Step
| |
| ;, all steam generators are assumed to be intact and narrow range level is )
| |
| :hecked. (See Figure II-6.) Although level in any steam generator did not sceed 50% at this time, "C" steam generator level was higher, and auxiliary bedwater could have been (but was not) throttled to ensure that the level acrease was controllable. At 0644 hours, Step 13 was completed and Step 14
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| '@ turned the Shift Supervisor to Step 4 to isolate flow from the ruptured steam Page 18
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| generator. At approximately 0646 hours, auxiliary feedwater. to "C" steam generator was isolated and level was noted to still increase. (See Figure II-6.) Based on this indication, the Shift Supervisor identified "C" as the ruptured steam generator, and it was completely isolated by 0648 hours (18 minutes into the event or 13 minutes af ter safety injection.)
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| The diversion status of the air ejector exhaust was focused on between 0720 and 0730 hours. However, the final decision to divert to containment was not made until 0756 hours. A specific step in EP-3 addressing the air ejector exhaust divert status would probably have identified and isolated this release path much sooner in the event. It should be noted that once the steam dump valves went closed (less than 10 minutes into the event) the release through the 4 air ejector exhaust was minimized, and once the ruptured steam generator was isolated (approximately 18 minutes into the event or 13 minutes after safety
| |
| )
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| I injection), the release was essentially eliminated. ]
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| The EDGs could have been shutdown at Step 12 rather than Step 32. However, the decision to do this would be manpower dependent, since more shift manpewer would probably be available at Step 32 than at Step 12. l 1
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| During the initial portion of the event (i.e. approximately 5 minutes l between the first indication and the manual reactor and turbine trip), operator !
| |
| actions were driven by several Abnormal Procedures (APs) and Annunciator Responses (ARs). Specifically, ARs were in for " Main Steam Loops 1-A-B-C Auxiliary Steam Loop Hi Radiation and Rad Alert" as well as for "PZR LO LEVEL" ;
| |
| and "PZR HI-LO PRESS". The required operator actions for these ARs were carried out except for Step 2.3.5 (Carry Out AP-44, Loss of RCS Pressure) of the PZR HI-LO PRESS" AR. Health Physics was notified that a possible steam generator ttibe rupture had occurred and were requested to get steam generator samples.
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| The immediate actions of AP-24.1 " Steam Generator Tube Leak" were carried out except for starting the second charging pump. As previously stated, the Shift
| |
| ' Supervisor directed the start of this pump but inadvertently the pump was not started. Also, the immediate actions of AP-5.1 and 5.2 were carried out.
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| However, none of these APs were formerly entered prior to the manual trip.
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| lAP-24.] was completed during the event but AP-5.1 and AP-5.2 were not formally
| |
| ! entered since they were not required once the E0Ps were initiated. AP-47 " Unit jOperationDuringOppositeUnit Emergency" was completed on Unit 2 as directed by lEF-0.
| |
| The " Procedure Reader" was a licensed CR0 uhich complied with the North Anna Power Station standard that the procedure reader be a licensed SRO or CRO.
| |
| However, the standard does state that the first preference for " Procedure
| |
| , Reader" is an available SR0 who it not assigned other priority duties. In igeneral, the Procedure. Reader interacted well with the Shift Supervisor and the two CRos on the Unit I control board. Based on interviews with the licensed operators, none indicated any problems with hearing the " Procedure Reader" or being heard. However, not all steps were read, or if read, acknowledged by the CR0s on the control board. In these cases, the " Procedure Reader" directly confirmed the step. Also, the " Procedure Reader" did not check off steps in the iE0Ps as they were completed. (These methods are consistent with the way the
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| " Procedure Reader" is trained on the simulator).
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| The number of personnel in the Control Room was maintained less than 15 as required by Station Administrative Procedure ADM-20.11. " Control Room Access".
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| Page 19 l
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| During the time interval between the manual reactor and turbine trip and the closing of the steam dump valves, excessive noise'was created in the Control Room each time the Unit 1 side Control Room door was opened. This resulted in distraction to the Unit 1 CR0s and SR0s. To limit access, the Shift Supe rvisor requested Security to post an officer at the Unit 1 Control Room Door. The primary sources of noise in the Control Room were generated by telephones ringing, the conduct of the EPIPs by the Interim Station Emergency Manager, and the " Procedure Reader".
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| Three additional minor problems were identified with the EPIPs and concerned !
| |
| or affected available shift resources prior to activation of the TSC. The first problem dealt with the lack of a Backboards Operator. As previously stated, l
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| this operator had assumed Unit 1 control board responsibilities as balance-of-plant operator. As a result, the transmission of radiation monitor readings and meteorological data was slowed down. It would have been helpful if the Health Physics personnel reporting to the Control Room had been trained in-obtaining this information.
| |
| l The second problen concerned the detailed and large quantity of information l required by the NRC. The NRC communicator was an experienced unlicensed I operator with responsibility for the Safeguards Watchstation but was not licensed. (When the TSC was activated, NRC communications were performed by a licensed SRO.) Information was provided to the NRC as requested but at some impact on the limited shift resources initially available.
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| J l
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| l t
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| Page 20
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| TABLE II-1 NORTH ANNA UNIT 1 OPERATING l CONDITIONS PRIOR TO STEAM GENERATOR TUBE RUPTURE
| |
| * Reactor Power: 100%
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| RCS Pressure: 2235 psig l
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| Pressure Level: 64.5%
| |
| RCS Average Temperature: 586.5*F RCS Loop Temperatures: "A" "B" "C" T F 621 618 617 h
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| T F 556 555 553 C
| |
| Boron Concentration: 1320 ppm Hydrogen Concentration: 36.9 cc/kg l
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| Volume Control Tank:
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| l Level 42.5%
| |
| Pressure 23 psig Temperature 112*F Charging Flow: 67.5 gpm Letdown Flow: 82 gpm Reactor Coolant Pump Seal Injection:
| |
| "A" RCP: 7.6 gpm "B" RCP: 7.7 gpm "C" RCP: 7.7 gpm Steam Generators: "A" "B" "C" Pressure, psig 870 870 870 Level, % 45 45 45 6 6 6 Feed Flow, lbm/m 4.3X10 4. 10 .23X10 6 6 6 Steam Flow, Ibm /m 4.35X10 4.35X10 4.42X10 Containment Conditions:
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| Temperature: 93*F Pressure: 9.45 psia o Source: Unit 1 Reactor Operator Log 1-LOG-4 Date: 07-15-87 Time: 0400
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| l i
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| TABLE II-2 REACTOR COOLANT SYSTEM LEAK RATES j PRIOR TO TUBE RUPTURE EVENT l
| |
| IDENTIFIED UNIDENTIFIED MODE LEAK-RATE LEAK-RATE
| |
| , 07-05-87 1 1.2 GPM 0.14 GPM
| |
| ! 2107 07-08-87 1 1.9 GPM 0.16 GPM 2023 07-11-87 3 2.0 GPM 0.01 GPM 1904 07-14-87 1 1.9 GPM 0.25 GPM 2108 l
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| l l
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| Page 22
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| l l
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| l TABLE II-3 UNIT 1 AIR EJECTOR ACTIVITY (pCi/CC)
| |
| PRIOR TO TUBE RUPTURE EVENT 13 July 1987 13 July 1987 14 July 1987 14 July 1987 15 July 1987 0900 Hours- 1740 Hours 0114 Hours 2245 Hours 0620 Hours 85.5% Power 87.8% Power 99% Power 100% Power 100% Power
| |
| ,r-41 5.94E-6 7.18E-7 4.59E-5 6.02E-5 1.34E-3 I-85m r - -
| |
| 1.68E-6 3.61E-5 5.84E-5 fr-87 - -
| |
| 3.21E-6 6.62E-6 1.36E-4 lr-88 - -
| |
| 4.78E-6 7.29E-6 1.50E-4
| |
| !@-133 - -
| |
| 4.61E-6 5.90E-6 1.49E-4 (a-135m 2.78E-6 -
| |
| 2.00E-5 4.92E-5 2.22E-4 p-135 2.02E-6 9.31E-7 1.84E-5 3.43E-5 4.92E-4 fe-138 - - -
| |
| 3.08E-3 6.34E-4 l
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| l 1
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| l i
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| Page 23 i
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| FIGURE H-8 ONLINE CHEMISTRY MONITORING SYSTEM DATA Page 31
| |
| | |
| I i
| |
| l I
| |
| l III. RADIOLOGICAL RELEASES A. Summary Re' lease data during the event has been evaluated and independently calculated. This review indicates that the reported release of 2.65E-1 curies on July 15, 1987, was valid based on the sampling data available at the time of the event. Followup calculations were performed to bound the release. The maximum inventory available for release based on reactor coolant to secondary side mass balance calculations was 2.24 curies. This is the. maximum amount of activity that could have been released.
| |
| Calculations based on sampling and radiation monitor data indicated that only a total of 1.59E-1 curies were released.
| |
| I Radiological data showed that approximately 84% of the release consisted of radiogases. Calculated whole body dose rates were found to be less than background at the site boundary in the affected sector (s).
| |
| B. Evaluation of Radiological Release
| |
| : 1. Onsite l
| |
| l During the course of the event, effluents consisting primarily of radiogases were released from the "C" steam generator via the two release paths prior to its isolation. A release pathway through the condenser air ejector exhaust existed from approximately 0630 until 0646 hours when the "C" main steam trip valve was closed isolating the "C steam generator from the condenser. (It should be noted that the steam dump valves closed and stayed closed prior to 0640 hours, effectively isolating "C" steam generator from the condenser). Following isolation of the "C" steam generator, only minimal activity from "A" and "B" steam generators was being released via the air ejector, and the release path was completely isolated at 0756 hours when the air ejector exhaust was diverted to the containment. A total gaseous activity of 9.90E-2 curies was calculated to have been released via the air ejector release pathway. This calculation was based on the Kaman normal range Vent-Vent "A" stack monitor data and the air ejector grab sample taken at 0650 hours and assumed a release duration of 86 minutes from 0630 to 0756 hours (see Table III-1).
| |
| The second release pathway was from the discharge of the steam driven auxiliary feedwater (AFW) pump. This pathway conservatively existed from approximately 0636 until 0745 hours when the steam driven AFW pump was shutdown. The steam supply to the steam driven AFW pump consisted of steam from the three steam generators from 0636 until 0648 hours at which time the "C" steam generator supply was manually isolated. Following 0648 hours, steam to the steam driven AFW pump was provided by "A" and "B" steam generators only, effectively eliminating the steam driven auxiliary feedwater pump as a release pathway. A total activity of. 6.04E-2 curies was calculated to have been released via this pathway. The noble gas component (Table III-2) of this calculation was based on the' Kaman normal range monfter data by assuming that the average release rate through the Page 32
| |
| | |
| condenser air ejector immediately prior to the turbine trip was also the release rate for the- AFW pump exhaust. The tritium, particulate, and halogen component of the total activity is based on isotopic analysis of blowdown samples taken at 0943'(Table III-3). Partition factors were used to determine the halogen- and particulate components. A total release duration of 12 minutes from 0636 to 0648 hours was assumed. The total release of activity through the AFW pump exhaust from 0648 to 0745 hours was insignificant (Table III-4). Gaseous activity discharged through the air ejector and the AFW pump exhaust was 1.37E-1 curies.
| |
| The above air ejector and steam driven AFW pump releases equate to a total of 1.59E-1 curies. This compares to a radioactivity release of 2.65E-1 curies that was estimated by the dose assessment team during the tube rupture event.
| |
| A primary to secondary mass balance calculation, based on a transfer of 55,013 pounds of reactor coolant from 0630 to 0648 indicated a total of 2.24 curies of radiogas available for release. (Table III-3)
| |
| A review of the inplant radiation monitoring charts confirms that the only release paths were as discussed above. No increase was noted on the Vent-Vent "B" ventilation =*cck gas and particulate monitors or the process vent gaseous and particulate monitors. In addition, no increase was noted on the Unit I containment gas and particulate monitors or the Unit 1 manipulator crane area monitors. This was expected since no activity was fj released to the containment until the condenser air ejector exhaust was diverted to containment at 0756 hours, and the "C" steam generator had been I isolated from the condenser for more than one hour before the exhaust was I diverted.
| |
| ! Increases were observed on the Unit 1 "A", "B" and "C" main steam monitors and the steam driven AFW pump exhaust monitor. "C" main steam radiation monitor readings in the range of 7 to 9 mR/hr were indicated l during the time period of the reactor trip. Readings on the "A" and "P" l main steam line monitoro were in the 0.4 to 1 mR/hr range during this same ;
| |
| period. No response was observed, although one would have been expected, I on the Unit 1 "A", "B", and "C" steam generators blowdown monitors. The response of these monitors during the event is probably due to the fact that the rup tured- tube did not start leaking until shortly before it ruptured. The impact of the ruptured tube on the RCS resulted in safety injection and isolation of the steam generator blowdown radiation monitors.
| |
| The Unit I condenser air ejector radiation monitor was inoperable during the event. This monitor has subsequently been repaired and returned to se rvice .
| |
| : 2. Offsite Whole body and thyroid average dose rates at the site boundary were calculated based on the release rates discussed in the previous section. A whole body dose rate of 4.04E-4 mrem /hr and a thyroid dose rate of 8.30E-4 mrem /hr were calculated. For the release, a whole body dose of 5.7BE-4 mrem and thyroid dose of 1.14E-3 mrem were calculated.
| |
| Page-33
| |
| _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ - _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ a
| |
| | |
| l Given the low concentrations of activity released, a background count ,
| |
| rate of approximately 0.02 mR/hr, and the lower limits of detection for j field and laboratory analyses of air particulate and radiciodine, no I detectable increase in dose rate or radioactivity would have been expected during the event at the site boundary in the affected sector (s). This conclusion is supported by the survey and sampling results and -dose rate readings taken during the event. !
| |
| Additional environmental samples were collected after the event and forwarded to Dan environmental consultant for analysis (see Table III-6).
| |
| The air particulate and radioiodine environmental samples of sectors H and K were counted at the station prior to being forwarded to the consultant.
| |
| 'These analyses indicated, no increase in normal background levels .of radioactivity. On July 20, 1987, the consultant reported that all samples had been analyzed and that no increases in normal environmental radioactivity levels had been indicated. This supports the conclusion- !
| |
| that the release consisted primarily of radiogases at very low concentrations.
| |
| i l
| |
| 1 1
| |
| I l
| |
| I 1
| |
| l Page 34 ;
| |
| | |
| ~ . _ _ . ,
| |
| l l
| |
| l <
| |
| TABLE III - 1 CALCULATED RELEASE ACTIVITY FROM CONDENSER AIR EJECTOR FOR TIME PERIOD FROM 0630 to 0756 HOURS
| |
| , , (BASED ON AIR EJECTOR SAMPLE TAKEN AT 0650 HOURS) i i
| |
| l CALCULATED l
| |
| SAMPLE RELEASED l ACTIVITY ACTIVITY l ISOTOPES UCi/CC uC1 Ar-41 4.90E-3 4.73E+4 Kr-85m 2.08E-4 1.98E+3 Kr-87 4.48E-4 4.36E+3 Kr-88 5.30E-4 5.15E+3 l
| |
| Xe-133 5.79E-4 5.54E+3 Xe-135m 3.99E-4 3.86E+3 Xe-135 1.83E-3 1.76E+4 Xe-138 1.36E-3 1.32E-4 l
| |
| TOTAL CURIES = 9.90E-2 l
| |
| l l
| |
| Page 35 l
| |
| i L_----________--____---___--______-_---__. _ - _ _ - - - _ _ _ - - -- - - _ _ - - - - - - - - - - - - - - - - - - - - _ .- - - - - - - - - - - - - - - - - - - - -
| |
| | |
| r l
| |
| l l
| |
| 1 l
| |
| l
| |
| ! TABLE III - 2 l
| |
| CALCULATED NOBLE GAS RELEASE ACTIVITY FROM STEAM DRIVE AUXILIARY FEEDWATER PUMP EXHAUST FOR TIME PERIOD FROM 0636 TO 0648 HOURS CALCULATED RELEASED ACTIVITY ISOTOPES uCi i
| |
| Ar-41 1.82E+4
| |
| ; Kr-85m 7.39E+2 Kr-87 1.67E+3 Kr-88 1.98E+3 Xe-133 2.13E+3 Xe-135m 1.48E+3 Xe-135 6.75E+3 Xe-138 5.05E+3 l
| |
| }
| |
| l TOTAL CURIES = 3.80E-2 1 i
| |
| i 1
| |
| i l
| |
| l
| |
| \
| |
| i l
| |
| l 4
| |
| Page 36 I
| |
| L _ ._ _ _ . _ _ . . . ._ _ . _ _ _ _ _ _ _ . _ _ _ . _ _
| |
| | |
| TABLE III - 3 CALCULATED RELEASE ACTIVITY (EXCLUDING NOBLE GASES) FROM STEAM DRIVEN AUXILIARY FEEDWATER PUMP EXHAUST FOR TIME PERIOD FROM 0636 to 0648 HOURS (BASED ON STEAM GENERATORS SAMPLES TAKEN Ar 0943 HOURS)
| |
| SAMPLE ANALYSIS - 0943 CALCULATED A-S/G B-S/G C-S/G RELEASED ISOTOPES A A AC m m AC m m uC1/cc uC1/cc pC1/cc uCi I-131 2.88E-6 ND* 5.07E-4 4.50E+0 l I-l32 1.98E-5 ND 3.11E-3 2.80E+1 l' I-133 3.86E-5 7.46E-7 6.03E-3 5.40E+1 J I-134 1.34E-5 ND 2.16E-3 1.74E+1 1-135 6.28E-5 1.06E-6 9.51E-3 8.45E+1 Na-24 3.01E-5 6.02E-7 4.19E-3 3.70E+1 ]
| |
| Co-58 1.35E-6 ND 1.99E-4 1.80E+0 l Sb-122 6.llE-6 5.36E-7 7.39E-4 6.60E+0 )
| |
| Cs-134 5.02E-5 5.20E-6 6.31E-3 5.60E+1 i Cs-137 3.67E-5 5.11E-6 4.53E-3 4.00E+1 j H-3 3.22E-4 1.2E-4 2.44E-2 2.20E+4 Sr-91 ND ND 7.21E-4 6.40E+0 Rh-106 1.50E-5 ND 2.73E-3 2.40E+1 1 Cs-138 ND ND 5.44E-4 4.80E+0 l TOTAL CURIES = 2.24E-2 l l cND - None Detectable l
| |
| Page 37
| |
| | |
| I i
| |
| i 1
| |
| I !
| |
| ! l l ;
| |
| l 1 i
| |
| ! I l TABLE III - 4 I
| |
| CALCULATED RELEASE ACTIVITY FROM STEAM DRIVEN AUXILIARY FEEDWATER PUMP l EXHAUST FOR TIME PERIOD FROM 0648 to 0745 HOURS ,
| |
| (BASED ON STEAM GENERATORS SAMPLES TAKEN AT 0943 HOURS) l l
| |
| l SAMPLE ANALYSIS - 0943 CALCULATED A-S/G B-S/G RELEASED ACTIVITY ACTIVITY ACTIVITY ISOTOPES UCi/cc uCi/cc uCi I-131 2.88E-6 ND* 3.63E-1 l I-132 1.98E-5 ND 2.49E+0 I-133 3.86E-5 7.64E-7 2.48E+0 l
| |
| I-134 1.34E-5 ND 1.69E+0 I-135 6.28E-5 1.06E-6 4.02E+0 Na-24 3.01E-5 6.02E-7 1.93E+0 Co-58 1.35E-6 ND 1.70E-1 Sb-122 6.11E-6 5.36E-7 4.19E-1 Cs-134 5.02E-5 5.20E-6 3.49E+0 Cs-137 3.67E-5 5.11E-6 2.63E+0 H-3 3.22E-4 1.20E-4 2.78E+1 Sr-91 ND ND ND Rh-106 1.50E-5 ND 1.89E+0 Cs-138 ND ND ND I TOTAL CURIES = 4.93E-5 l
| |
| *ND - None Detectable 3 l
| |
| I l
| |
| q Page 38 k
| |
| | |
| 4 i
| |
| TABLE III - 5
| |
| ~
| |
| l REACTOR COOLANT RADIOGAS CONCENTRATIONS AND CORRESPONDING
| |
| ( RELEASE ACTIVITIES BASED ON MASS BALANCE CALCULATIONS l CALCULATED S/JIPLE RELEASED ACTIVITY ACTIVITY l ISOTOPES uC1/CC uCi i
| |
| Ar-41 2.5E-2 6.25E+5 Kr-85m 1.4E-3 3.50E+4 i Kr-87 3.0E-3 7.50E+4 l Kr-88 3.5E-3 8.75E+4 Xe-133 4.7E-3 1.18E+5 l Xe-135m 3.2E-2 8.00E+5 Xe-135 2.0E-2 5.00E+5 TOTAL CURIES = 2.24E+0 Page 39 I___________________.-------____.____
| |
| | |
| TABLE III - 6 15 July 1987 ENVIRONMENTAL SAMPLES i
| |
| l
| |
| ! SECTOR MILES FROM PLANT ** TYPE STATION TYPE SAMPLE F 0 12 Environmental TLD* Quarterly TLD
| |
| #11/43 F 4.75 Environmental TLD* Quarterly TLD
| |
| #12/44 G 0.75 Environmental TLD* Quarterly TLD
| |
| #13/45 G 5.88 Environmental TLD* Quarterly TLD
| |
| #14/46 H 0.93 Environmental TLD* Quarterly TLD
| |
| #15/47 H 2.33 Environmental TLD* Quarterly TLD
| |
| #16/48 H 7.30 Environmental Air Air Particulate Sampler #07 Radiciodine H 0.93 Environmental Air Air Particulate Sampler #23 Radiciodine l
| |
| l J 0.22 Environmental TLD* Quarterly TLD i
| |
| #17/49 J 1.55 Environmental TLD* Quarterly TLD
| |
| #18/50 K 0.36 Environmental TLD* Quarterly TLD
| |
| #19/51
| |
| ~
| |
| NOTE:
| |
| * TLD - Thermoluminescent Dosimetry
| |
| ** Miles from plant - distance from Unit #1 Containment Page 40
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| i I
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| l t
| |
| TABLE III - 6 (Cont.)
| |
| 15 July 1987 ENVIRONMENTAL SAMPLES l
| |
| ! 1 j
| |
| l
| |
| ; SECTOR MILES FROM PLANT ** TYPE STATION TYPE SAMPLE )
| |
| I K 5.30 Environmental TLD* Quarterly TLD ]
| |
| #20/52 l K 5.30 Environmental Air Air Particulate Sampler #02 Radiciodine i
| |
| ! L 0.30 Environmental TLD* Quarterly TLD 1 l #21/53 1
| |
| : 1 L 4 . ',6 Environmental TLD* Quarterly TLD l #22/54 l
| |
| J 0.33 Offsite Monitoring Soil ;
| |
| Team #1 Vegetation j J 1.60 Offsite Monitoring Soil l Team #2 Vegetation 1 I
| |
| NOTE:
| |
| * TLD - Thermoluminescent Dosiuetry
| |
| ** Miles from plant - distance from Unit #1 Containment i
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| Page 41
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| 'l l
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| l IV. EMERGENCY RESPONSE A. Emergency Plan Implementation l l
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| Staffing and Responsibility Specific assignments to the emergency organization were made in accordance l with the Station Emergency Plan and were adequate to respond to the event. !
| |
| Timely activation, staffing, and operation of Emergency Response Facilities was accomplished. ]
| |
| Emergency Response Support and Resources 1
| |
| Support resources (i.e. Westinghouse, OMNI, FiDCOMP, etc) were requested and effectively used during the event. State resources that participated at the LEOF were adequately accommodated. In general response agencies demonstrated their ability to augment the Station response.
| |
| Emergency Classification The emergency classification and action level scheme, as provided in the-Emergency Plan and Implementing Procedures, was adequate for the classification of the event. Emergency Plan Implementating Procedures adequately provided for I initial and continuing assessment and response actions during the emergency.
| |
| Notification Methods and Procedures 1
| |
| Procedures established .for notification of Federal, State and local I responce organizations and emergency personnel were promptly and effectively implemented. Content of initial and followup- messages was adequate and consistent with the requirements of the Emergency Plan. Furthermore, i provisions in place to alert Virginia Power's augmented emergency response organization were executed and found to be effective.
| |
| Notification of NRC was initiated at 0702 hours and continuously maintained to termination of the event.
| |
| State / local notifications were accomplished as follows:
| |
| MESSAGE # TIME 1 0651 2 0702 3 0718 4 0745 5 0805 6 0828 7 0846 8 0913 9 0935 10 1009 11 1043 Page 42 u__-_____-________-___-__-_--__-_____-___-___---__--
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| | |
| I,
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| .1 i
| |
| 12 1115
| |
| ]
| |
| 13 1148 1 14 1220 l 15 1252 16 1326 1 17 1343 Station call-out was initiated at 0655 hours.
| |
| Corporate call-out was initiated at 0710 hours.
| |
| Station accountability was initiated at 0655 hours and completed at 0728 hours.
| |
| 1 Emergency Communications l i
| |
| Conrunications between emergency facilities, principal response l organi: .tions and emergency personnel were promptly established and continuously maintained throughout the event. Ne cessa ry equipment and {
| |
| procedures were in place in order to provide adequate communications. No !
| |
| significant equipment malfunctions occurred. However, there were some problems with the audioconferencing and speaker systems at the Local Media Center. This was not prohibitive during the event, and corrective action has been initiated. !
| |
| Media and Public Information .
| |
| 1 Information concerning the emergency was made available to the media for dissemination to the public. Four press releases were issued during the event (at 0800, 0900, 1100, and 1400 hours respectively). In addition, press briefings were conducted at the Corporate Public News Center (CPNC) and Local 1 Media Center (LMC) throughout the day. The above facilities were activated at !'
| |
| 0830 and 1000 houes, respectively.
| |
| l Following termination of the event, the media was brought from the LMC to l the station for an indepth briefing and plant tour.
| |
| ! Emergency Facilities and Equipment __
| |
| Emergency facilitics were activated and promptly staffed during the event,
| |
| . and the facilities were adequately equipped to support the response effort.
| |
| ' The activation times for the Emergency Response facilities are provided below.
| |
| Technical Support Center - 0757 hours
| |
| : Operational Support Center - 0753 hours Local Emerger.cy Operations Facility - 0915 hours Corporate Emergency Response Center - 0820 hours Corporate Public News Center - 0830 hours Local Media Center - 1000 hours It should be noted that the Technical Support Center (TSC) was fully manned prior to 0757 hours. However, the Station Emergency Manager did not relocate
| |
| !from the Control Room to the TSC until 0757 hours, and therefore, the TSC was not " officially" activated until that time.
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| Page 43 L____-___-_-___---_.. _ _ _ _ _ _ - - - - _ _ _ _ _ - _ _ _ - _ - . - _ . - _ _ _ - :_ _-
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| | |
| Accident Assessment Methods, systems, and equipment for asseesing and monitoring actual onsite ccasequences of the event were adequate. Assessment and response activities included both an engineering assessment of plant status and an assessment of radiological . hazards. Inplant, onsite and offsite monitoring teams were promptly dispatched and effectively used throughout the event.
| |
| Recovery Planning Prior to termination of the event, a Recovery Organization was formulated.
| |
| This team was tasked with developing specific plans and procedures for recovery. The Recovery Organization was established under the direction of the ,
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| Vice President - Nuclear Operations Department. The organization created a
| |
| " command center" in the LEOF, established .a long-term interface arrangement with the NRC and State, and formulated the following support structure:
| |
| : 1. Maintenance and Repair Team
| |
| : 2. Technical Evaluation Team
| |
| : 3. Nuclear Safety and Licensing Team 4.
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| Industry Interface Team l
| |
| : 5. Health Physics Team The recovery organization was implemented upon termination of the event.
| |
| Summary l The July 15, 1987 SGTR event demonstrated that emergency response' personnel within the Virginia Power organization are capable of satisfactorily responding to an emergency of this nature. Furthermore, the NRC and emergency l organizations within the State of Virginia demonstrated the capability to l respond to requests for assistance during such an emergency.
| |
| In conclusion, the North Anna Power Station and those emergency response agencies that support it effectively mitigated the transient event and adequately protected the health and safety of the public.
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| Pxge 44
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| | |
| B. Emergency Response Facility Computer System Overview The North Anna Power Station Emergency Response Computer System (ERCS) operates the Data Acquisition System (DAS) for Unit 1 and Unit 2. The DAS !
| |
| covers the collection and communication of plant data required by Supplement 1 j to' NUREG-0737 (NRC Generic Letter 82-33) to be available to each unit Cont d j Room and the Emergency Response Facilities. A subset of the DAS serves as the j Safety Parameter Display System (SPDS) required by Supplement 1 to NUREG-0737, and also includes other information determined to be useful to plant and emergency response personnel. The plant SPDS and DAS were extensively used, t during and following the steam generator tube rupture event. AS evaluation of the performance and utilization of the SPDS and DAS by emergency response personnel in the Control Room and other emergency response facilities, .and personnel involved in the event analysis was conducted by interviewing selected individuals. The objectives of this evaluation were to determine: 1) whether the SPDS and DAS satisfied regulatory requirements and guidelines, and 2) whether the SPDS and DAS served as an effective aid, above and beyond the -j regulatory objectives, in the assessment. recovery, and analysis of the July 1 15, 1987 event.
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| {
| |
| Control Room Utilization This evaluation showed that overall the SPDS worked well and that displays were used to provide helpful information to the Control Room personnel during the event. Because of the nature of this particular event (ie, the cause of the transient was quickly recognized and the proper mitigative actions were i taken in a timely manner), the " accident assessment" function provided by the ;
| |
| information which was available on the SPDS (eg, trend plots) was not utilized !
| |
| to its fullest extent. The Shift Technical Advisor (STA) who was the principal user in the Control Room noted that the display of SPDS information was quite useful and convenient. Information provided by the DAS 'were also used by l Control Room personnel and found to be helpful both during the initial phase of j the event and for confirming the successful completion of accident response actions taken later in the event.
| |
| ERF Utilization The SPDS displays were also available in the Emergency Response Facilities (ERFs) and were mainly used by the personnel manning these facilities to obtain i
| |
| current plant status. ERF personnel made extensive use of the information provided by the DAS to assess and trend plant conditions as mitigative actions
| |
| , were taken. An example of this was the fact that Technical Support Center !
| |
| t personnel were able to use DAS information to assist Control Room personnel in responding to the event. Because of the nature of the July 15, 1987 event and the manner in which it was mitigated, the radiological and meteorological capabilities of the DAS were not used to their fullest extent. Significantly, ERF personnel did not interfere with Control Room operations at any time during the event. as the information provided by the DAS was complete enough for the performance of required tasks by ERF personnel.
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| Page 45
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| 'l l
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| Event Analysis -
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| 1
| |
| > The event data stored by the DAS was found to be an excellent event analysis tool, providing more extensive and more detailed data than was previously available. The resolution of the data collected and stored by the i DAS (ie, sampling frequency) will be evaluated further to determine if higher resolution data would be advantageous for possible future events.
| |
| Summary While the SPDS demonstrated that it satisfied the regtslatory requirements and guidelines, a number of shortcomings were identified that prevented the l SPDS from serving as a fully ef fective aid to Control Room personnel. These shortcomings appear to fall into two broad categories, data point treatment by i the SPDS software and human factors concerns. Specifically, SPDS alarm -l setpoints appear to be overly conservative and lack the necessary degree of flexibility to reflect plant status during a rapid transient. The amount of necessary interactions between the user and the SPDS, and the presentation of certain information in the SPDS displays do not appear to meet the -best human factors engineering criteria.
| |
| Overall, the performance of the DAS in the functioning of the ERFs met I applicable regulatory requirements and guidelines. As for the SPDS, the j deficiencies identified during the event and the event progression prevented q the DAS from being used to its full potential. Some of the problems l encountered are identical to the SPDS probleus discussed above, and concern the j manner in which data is handled by the DAS (alarm setpoints/ data flexibility) !
| |
| and human factors /information presentation concerns (more detailed piping and l Instrumentation diagrams). The need for additional training of ERF personnel l was also identified.
| |
| The problems highlighted by the personnel being interviewed will be l j addressed and resolved. However, these problems were not detrimental to the '
| |
| overall performance of the Emergency Response Computer System (ERCS) as an l effective tool for managing off normal events at Virginia Power nuclear facilities.
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| l l 1 l
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| I
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| \
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| l
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| ! Page 46 l
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| l l
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| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _____________________J
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| | |
| - i V. SAFETY EVALUATION I A. Comparison of North Anna Unit 1 Tube Rupture Event to UFSAR Analysis An analysis of a steam generator tube rupture is presented in Section ;
| |
| 15.4.3 of the North Anna UFSAR. The analysis considers a double-ended rupture of a single tube at hot full power. Initial break flow rates for this case are calculated to be approximately 710 gpm. An evaluation of the break flow rate 1 which results from reaching equilibrium conditions with the safety injection flow is presented. A total primary to secondary mass transfer is. calculated by ,
| |
| conservatively assuming this equilibrium break flow remains at a constant value j for a full half-hour. For calculating whole body and thyroid doses at the site ;
| |
| boundary, this mass is conservatively increased by 15%. The site boundary i doses are calculated based on the assumption of steam release through the main !
| |
| steam safety valves with an initial coolant activity level corresponding to
| |
| ! 1% of the fuel in the core being failed. Even with these conservative ]
| |
| ! assumptions, the calculated offsite doses were only a small fraction (less than 2%) of the 10 CFR 100 limits.
| |
| Table V-1 presents a summary comparison of the North Anna event of July 15, 1987 with the analysis results and assumptions of the FSAR. As can be seen
| |
| , by every significant measure, the actual conditions o'ecurring during the incident were bounded by the UFSAR assumptions, in some cases by orders of l magnitude. The primary to secondary mass transfer has been calculated by both a RETRAN transient simulation of the event and an independent calculation based on the measured boron concentration in the secondary side of the faulted generator following the event. These results are in good agreement.
| |
| Examination of the faulted tube showed that the break was a double-ended break, as addressed in the UFSAR. Break flow rates are highly geometry dependent. The actual break occurred high on the cold leg side of the generator, and the tubes were not completely offset from one another. The effects of the tube length upstream of the break and the likelihood of significant interaction between the flow streams from the two tube segments are therefore the probable cause for the observed break flow rates being less than 6
| |
| predicted in the UFSAR.
| |
| B. Fuel Integrity Evaluation
| |
| [ During the event, the plant was not in any condition which could have
| |
| ! compromised the integrity of the fuel. A review of the plant data shows that there was no evidence of reactor coolant system voiding, so core uncovery. was precluded. The core thermal limits were not violated at any time during the event. The closest approach to the thermal limits was during the initial
| |
| .depressurization phase, just prior to the manual reactor trip. A representation of this statepoint along with the thermal limits for the reactor condition just prior to trip is shown in Figure V-1. Note there were substantial margins to both the DNB and vessel exit boiling portions of the limit curves.
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| Page 47
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| | |
| A point estimate of the DNBR in the hottest coolant channel in the core at this point just prior to trip was performed using the COBRA code. Assumptions j are presented in Table V-2. The minimum calculated WRB-1 DNBR was 1.85 t (compared to a 1.23 limit).
| |
| The addition of soluble boron from the Safety Injection System increased the boron concentration of the reactor coolant system by over 300 ppm over the first two and one half hours of the event. This increase provided adequate shutdown reactivity margin throughout the cooldown evolution.
| |
| C. Conclusions The above evaluation has demonstrated that the consequences of the steam generator t Pe rupture event on July 15, 1987 on North Anna Unit No. I were bounded by the analysis presented in the accident analysis chapter (Chapter 15) of the UFSAR. In terms of offsite public health consequences, the event had no effect. The core was effectively and adequately cooled throughout the event.
| |
| lThe integrity of the fuel was not compromised in any way.
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| i 1
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| d Page 48
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| | |
| i TABLE V-1 j NORTH ANNA UNIT 1 STEAM GENERATOR TUBE RUPTURE k
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| COMPARISON TO UFSAR ACCIDENT ANALYSIS l
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| (
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| )
| |
| UFSAR Assumption Ld: ration 7/15/87 Event Or Result j I
| |
| 1 Tube rupture area Double ended rupture of Double-ended rupture a single tube. Geometry of a single tube effects reduce the discharge coefficient to less than 1.0 Lnitial break flow Approx. 550-637 gpm Approx. 710 gpm (UFSAR Fig. 15.4-32) l
| |
| [ntegrated primary- Less than 100,000 lbm Calculated: I
| |
| ;o-secondary mass 115,000 lbm
| |
| ;ransfer Used for dose calculation:
| |
| 132,000 lbm Enitial coolant .00938E-6 C1/gm 1% failed fuel =
| |
| letivity (Measured 7/14/87) 4.0E-6 Ci/gm telease path for Condenser air ejectors exhaust Main steam safety valves radionuclides to Steam driven AFU pump exhaust environment
| |
| )perator response Approx. 34 minutes 30 minutes
| |
| ;ime to identify Bccident type and
| |
| :erminate break flow
| |
| :o faulted generator
| |
| )ffsite power Available Unavailable Bite boundary dose
| |
| - whole body 5.78E-4 mren* 350 mrem **
| |
| thyroid 1.14E-3 mrem
| |
| * 370 mrem ***
| |
| : conservative estimate - no dose rate above background was actually measured 1.4% of 10 CFR 100 limits
| |
| ).12% of 10 CFR 100 limits Page 49
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| | |
| TABLE V-2 NORTH ANNA UNIT 1 STEAM CENERATOR TUBE RUPTURE l
| |
| COBRA /WRB-1 POINT ESTIMATE OF MINIMUM DNBR I l
| |
| i Parnmeter Assumption /Value Thermal Power / Heat Flux Rated Thermal Power +2% Unct.
| |
| Reactor Inlet Temperature 561 F (strip chart, SPDS) + 4 F 1 uncertainty l Hot Channel F-delta-H 1.510 (1.398 measured on 7/14 with 8% uncertainty) _
| |
| RCS Pressure 2070 psig (strip chart, 1 SPDS) - 30 psig uncertainty Reactor Coolant Flow Rate 305,500 gpm (calculated from rated thermal power of observed Thot and J
| |
| Tcold) l l
| |
| Resultant WRB-1 DNBR 1.85 WRB-1 95/95 DNBR limit (with 5% 1.23 retained margin for rod bow) ;
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| l I
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| f I
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| l i
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| Page 50
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| 4
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| )
| |
| Nominal T,y,= 586.8'T Weinal RCS flev = 289200 CFM 660 -
| |
| : 655 2400 psia 558 645 2250 yeia 848 ' 2115 psia limit 555 -
| |
| 550 525 2000 ps 623
| |
| [ s15 1860 psia i
| |
| .ia .
| |
| 685 See-595 '
| |
| 54 !
| |
| 505 - Reactor State % g I Just Prior To '
| |
| Trip 588 - >
| |
| 675
| |
| : 8. .8 .2 .5 4 .5 .s .7 .8 8l 4. 1.1 I.3 POWCE treaction or neela.i t l 1 FIGURE V-1 REACTOR CORE SAFETY LIMITS AND STATE POINT PRIOR T,0 TRIP Page 51 T-A-
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| | |
| VI. STEAM GENERATOR EVALUATION A. North Anna Steam Generator Operating Experience Since start-up in 1978, Unit I has been on All Volatile Treatment (AVT) chemistry. During February 1979, powdex resin was inadvertently introduced into the steam generators. During the resin intrusion event, the resin decomposed to form sulfuric acid, lowering the steam generator pH to 6.02 and increasing the cation conductivity to 25.0 micro-mhos. Two other less severe intrusion events occurred in July and September of 1979. During the September 1979 refueling outage, tube denting was identified in the hot leg, and a boric acid treatment was initiated. Unit I has operated in accordance with Westinghouse recommendations for use of a boric acid treatment since 1980.
| |
| Subsequent inspections have confirmed denting at the hot leg tube support plate locations. However, the denting has not resulted in extensive tube deformation. Recent comparisons of data taken in 1984 and 1987 indicate that the growth in both the number of dented intersections and extent to which intersections have been deformed has been essentially arrested.
| |
| Tube degradation involving primary side cracking of the Row I and 2 U-bends has also been observed (during the 1979 inspection on Row 1 and during the 1987 inspection on Row 2). Preventive plugging of Row 1 tubes and stress relief of unplugged Row 2 tubes in the three steam generators has been accomplished to address these issues.
| |
| The first significant indication of steam generator tube corrosion on North Anna Unit I was in December 1983. Primary to secondary leakage in Steam Generators "B" and "C" increased to a maximum value of 396 gallons per day (GFD). The unit was shut down January 10, 1984, and a visual inspection of l l
| |
| both steam generators during a hydrostatic test showed a total of five (5) leaking tubes and four (4) leaking explosive plugs between the two steam generators. A total of 579 tubes were inspected in Steam Generator "B" during the outage. Four (4) tubes were identified with greater than 40% wall indications at the tube support plates while 13 tubes were identified as having significant distortion. All leaking tubes exhibited eddy current indications, l but none of the significantly distorted indications were identified as leakers.
| |
| In Steam Generator "C", a total of 552 tubes were inspected. Four tubes
| |
| , contained deep eddy current indications and were plugged. Additionally, two l tubes were identified as having distorted indications. The unit returned to l service in February and operated until the scheduled May 1984 refueling outage.
| |
| Steam generator activities during the May 1984 refueling outage included complete eddy current inspection in all 3 steam generators and an attempted tube removal effort. A total of eight tubes in Steam Generator "A", 1 tube in Steam Generator "B", and 4 tubes in Steam Generator "C" exhibited eddy current indications greater than 40% thorough-wall. Additionally, 2 tubes in Steam Generator "A" and 1 tube in Steam Generator "C" were plugged as a preventive measure. Complex signal distortions arising from denting, copper, and magnetite were first observed during this outage, but not recorded. Two attempts were made during the outage to remove a tube from Steam Generator "C",
| |
| ! but both were unsuccessful.
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| l l Page 52 l
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| l l
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| l
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| l The unit returned to service in September 1984 and operated until August 2, 1985 when the unit was shutdown again due to primary to secondary leakage.
| |
| Primary to secondary leakage was first detected in February 1985 in trace amounts. Noticeable step changes occurred in April 1985. The leakage j gradually increased to a maximum value of 213 GPD in Steam Generator "A" in l late July. The unit came off-line on August 2, 1985, and an inspection of i
| |
| Steam Generator "A" was performed. A video inspection of the tubesheet was l performed while the steam generator was filled and pressurized. A total of 3 tubes were identified as leaking. Subsequently, a total of 830 tubes were eddy current tested, twelve (12) of which had pluggable indications (eleven of these were greater than 90% thru wall). Although distorted indications were identified in this inspection, these tubes were not separated from a larger class of 95 tubes which contained some degree of distortion. Consequent,1y, only thirteen tubes were removed from service.
| |
| l ihe unit returned to service in mid-August, 1985 with trace leakage in
| |
| , Steam Generators "B" and "C". The leakage suddenly increased after l approximately five days online to approximately 90 gpd where it remained until l the November, 1985 refueling outage. Steam generator activities performed during the November, 1985 outage included complete eddy current of all three steam generators, and the removal of two tubes containing a total of four support plate intersections. The eddy current testing program resulted in the plugging of 43 tubes for indications greater than Technical Specifications limits. Thirty additional tubes containing strong distorted indications were also removed from service for preventive purposes. The unit returned to service in January 1986 with no primary to secondary leakage. Trace, inte rmittent leakage was detected in Steam Generator "A" beginning in February j 1986. i During the November, 1985 refueling outage of North Anna Unit 1, eddy current inspection results at the hot leg support plate intersections of all three steam generators revealed tubes with eddy current signals in which the interpretation of the extent of tube wall penetration was precluded due to the level of signal distortion caused by tube denting. The eddy current analysis at that time concluded that the signal characteristics were representative of tube wall degradation but because of signal distortion, depth of penetration could not be quantified. Subsequently, in an effort to characterize the type and extent of tube wall degradation in the steam generators, Virginia Power had two tubes extracted from Steam Generator "C" and metallographically examined.
| |
| The tube examination program conducted included detailed physical measurements of the tubes at the tube support plate intersection locations, characterization of the tube metallurgical structure, chemical analysis of the tube surface deposits, and corrosion testing of the tubing. As a result of the tube examination program, the tube wall degradation has been characterized as outer diameter (OD) and inner diameter (ID) initiated stress corrosion cracking (SCC), located within and extending beyond the tube support plate thickness in conjunction with tube denting.
| |
| Unit I returned to service following the November, 1985 refueling outage and operated with only a small amount of primary-to-secondary leakage near the end cycle. The unit was shutdown for refueling in April, 1987. During the outage, an extensive 100% eddy current program was performed employing the Page 53
| |
| | |
| most advanced eddy current techniques available. In addition, a tube stress relief demonstration program was conducted at the tube support plate and all available Row 2 U-bends were stress relieved. Indications were found at the tube support plate, the anti-vibration bars in the area of the steam generator U-bends, and the tubesheet. A total of 263 tubes were plugged during .the outage as shown on Table VI-1. In addition, two tubes were removed from Steam Generator "A" for laboratory examination. Currently, the removed tubes as well as the demonstration program for tubes stress relief at the tube support plates are being evaluated.
| |
| Because greater than one percent of the tubes in the initial tube inspection sample group in the April, 1987 were found to be defective, NRC approval of the actions taken to return the steam generators to an operable status was required before returning Unit 1 to service. A meeting was held with the NRC on June 3, 1987 to discuss the results of the steam generator tube
| |
| ' inspections and Virginia Power's plans to evaluate the two tubes that were removed from steam generator "A". Virginia Power also committed at this meeting to impose a 100 gallon per day limit on primary to secondary leakage for the duration of the upcoming fuel cycle. NRC approval for restart of Unit I was given at this meeting.
| |
| A historical summary of the numbers of tubes plugged in each steam generator by outage date is provided in Table VI-2. Maps of the locations of the plugged tubes in each of the three steam generators are provided on Figures VI-1, VI-2 and VI-3 for Steam Generators "A","B" and "C", respectively.
| |
| B. Current Outage Inspection Activities j Following the steam generator tube rupture event on July 15, 1987, a i recovery organization was established to, among other activities, provide a technical evaluation of the event and determine the root cause of the rupture.
| |
| In addition, the technical evaluation function included development of steam l generator inspection plans, review of data, and development of future i inspection requirements.
| |
| As a result of the event. Steam Generator C was filled with a mixture of primary and secondary coolant. Steam generator level was approximately 63%
| |
| (vide' range). Boric acid concentration was approximately 814 ppm. Both radioactive and non-radioactive gases collected in the steam generator. Xenon was also present. Hydrogen was determined to be approximately 15% at the time of venting.
| |
| Reactor Coolant Loop C was isolated on the primary side by closing the loop stop valves. Over the course of the next several days, plans for venting Steam Generator "C" were developed and implemented. Gases were vented by the process 4 vent system via the waste gas processing system. This provided for a controlled, monitored release of gas. At no time did this controlled release l pose a threat to safe operation or to the public. Activities were conducted in i
| |
| accordance with approved procedure and were carefully controlled. Preparations were also made and completed to establish a cold shutdown, degassed condition {'
| |
| l for the rest of the reactor coolant system and place Steam Generators "A" and "B" in a wet lay-up condition.
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| Page 54
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| i The next activity was to drain the primary side of Steam Generator "C".
| |
| The purpose was twofold. One, inspection of the secondary side of the steam generator would eventually require that the primary side be drained and dry.
| |
| l Two, this draining would establish the approximate elevation of the primary-to-secondary leak and aid in further developing the inspection and recovery plan. This plan included strategies based on location, past history, j and inspection plans and techniques.
| |
| I
| |
| ! Following draining, the leak elevation was established, using wide range i
| |
| ( steam generator level, to be at approximately the 7th tube support plate. The
| |
| ]
| |
| steam generator was filled with nitrogen on the secondary side as level 4 decreased during the primary drain. Following the tube rupture event, Steam ;
| |
| Generator "C" could not be cooled down in a normal manner. This caused Steam Generator "C" to cool down at a slower rate. The proposed inspection plan included the use of eddy current and optical viewing equipment. The equipment l has specific maximum operating temperatures. Thus, it was necessary to refill i the steam generator with cool water to reduce its temperature. After the fill i to approximately 90% was complete, the secondary side water drained through the I ruptured steam generator tube. When the draining stopped, the elevation of the leak was confirmed to be in the vicinity of the 7th support plate. Meanwhile, inspection plans were being further developed including the development of the i Technical Specification required 1S sample plan. The primary side of Steam !
| |
| Generator "C" was drained and the manways removed. !
| |
| With access to the primary side row possible, the row and column location of the leak was determined by the presence of water at the tubesheet. A slight increase in the secondary side water level was made to make the location more evident. The leak was located at Row 9, Column Si. With this location established, past experience and e.xamination information were reviewed for use in revising the inspection and recovery plan for Steam Generator "C". The steam generator secondary side was then drained.
| |
| 1 Equipment needed to examine the leaking steam generator tube was placed in the primary side. The leaking tube was first inspected using an eddy current probe. This examination provided the exact location of the leak and some limited information about the size or extent of the failure. Following this activity, the tube was viewed using a fiber-optic device. The exact location and extent of the break was determined. The ruptured tube was Row 9, Column 51 at the 7th support plate on the cold leg side. The tube was broken circumferentially and the tube ends separated to approximately one-half inch above the tube support plate.
| |
| In the past this tube was inspected twice, during the refueling outages of September, 1979 and December, 1980, from the hot leg side through the 7th support plate on the cold leg side. More recently this tube has been inspected on the hot side through the 7th support plate, hot leg. No indications were found using the (standard) bobbin coil eddy current device.
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| page 55 1
| |
| | |
| 1 Inspection Plan Based on the location of the ruptured tube and the requirements of North Anna Unit 1 Technical Specifications, the following inspection plan was developed.
| |
| - Perform the-Technical Specification required sampling. program.
| |
| 4
| |
| - Perform. Eddy current examination using the standard bobbin coil of all i available U-bends and cold. legs that were not inspected during the May 1987 outage. ,
| |
| 1
| |
| - Using an 8xl ' eddy current fixture, inspect all available tubes through.
| |
| the 7th support plate on both the hot and cold leg sides' . (Note: U-bends _I are not. accessible with the 8x1 probe). !
| |
| - Use the rotating pancake coil (RPC) probe to confirm indications found by 8x1,
| |
| - Remove the cold leg side of tube R9C51.
| |
| - Locate and remove one (1) tube with circumferential indications similar to R9C51, if necessary.
| |
| - Examine pulled tubes using non-destructive and destructive techniques i similar to past tube examinations.
| |
| - Evaluate the above actions for input to the Unit 2 1987 steam generator
| |
| ! inspection plan and future Unit 1 and 2 inspections.
| |
| i l
| |
| l
| |
| , Page 56 i
| |
| | |
| l l TABLE VI-1 STEAM GENERATOR TUBES PLUGGED BY. INDICATION TYPE DURING 1987 REFUELING OUTAGE Total Tube *** Tubes Total l Clear
| |
| * Distorted ** Sheet ~ Plugged Previous IuSes Current indications Indications ndications This Outage Plugs P_ly g % Plugged i
| |
| i 25 43 15 83- 126 209 6.17 14 37 11 62 116 178 5.25-30 78 9 118**** 153 271 7.99 f Clear Indications (defective) - greater than 40 percent "thru vali" l indications.
| |
| l Distorted Indications -
| |
| tube support plate indic ations of undetermined l l "thru wall" depth.
| |
| Tubesheet indications - not evident on standard eddy current testing which were identified by the 8X1 probe and confirmed by the RPC probe.
| |
| c* Includes one tube that was erroneously plugged.
| |
| I l l
| |
| I 1
| |
| i l
| |
| l Page 57 l
| |
| . _. _-- .2
| |
| | |
| l 1 l $
| |
| i l
| |
| l
| |
| 't i
| |
| TABLE VI-2 NORTH ANNA UNIT 1 l TUBE PLUGGING
| |
| | |
| ==SUMMARY==
| |
| | |
| OUTAGE DATE STEAM GENERATOR TOTAL A_ B C TUBES i
| |
| September 1979 94 94 96 284 ;
| |
| January 1984 0 4 5 9 l May 1984 10 1 S 16 August 1985 13 0 0 13 November 1985 9 17 47 73 l April 1987 83 62 118 263 TOTAL 209 178 271 658
| |
| % (6.2%) (5.3%) (8.0%) (6.5%)
| |
| l l
| |
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| |
| Page 58
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| |
| 1 VII LESSONS LEARNED I
| |
| I LATER I
| |
| I
| |
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| |
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| Page 62 1
| |
| i__ __. _ ________._.______..J
| |
| | |
| I l
| |
| ATTACHMENT 1 l l
| |
| CHRONOLOGY ,
| |
| i STEAM GENERATOR TUBE RUPTURE EVENT 1 NORTH ANNA POWER STATION UNIT 1 JULY 15,1987 1
| |
| To support the investigation of the Steam Generator Tube Rupture the following chronology was reconstructed from the print outs of the alarm typewriter attached to the Control Room P-250 process computer, the Sequence of 4 Events Recorder (Dranetz) driven by the Hathaway annunciator system, the data j printouts extracted from the record kept by the ERF Computer in the Technical J Support Center, R0 and SRO logs and interviews, and strip charts from Control Room recorders.
| |
| l l Selected data was transmitted from the various records based on the ]
| |
| significance of each datum as it identified a sub-event or demonstrated,
| |
| {
| |
| explicitly or implicitly, a sub-event in the sequence. The intent is that this chronology can be integrated with other analysis to determine the timeliness, accuracy and effectiveness of the measures applied to mitigate the accident.
| |
| Once the data was transcribed, a review was performed to identify the synchronism for time of the various data sources. The principal item selected for synchronism uns the Automatic Pzr Lo-Lo SI. The SI action incorporates several actions including feedwater isolation and normal charging isolation that make it readily comparable over all records. The Sequence of Events Recorder logged SI at 06:35:24:805; the P-250 logg,d SI at 0639. However the earlier Reactor Manual Trip has caused the P-250 to alter its scan rates. The
| |
| ! P-250 Post Trip review logged SI at 06:35:24 plus 1012 cycles, which equates to
| |
| ( 06:35:40.86. The ERfC data set collected at 06:34: 14 records full normal I charging flow and full power feed flow to the steam generators, approximately 16 seconds after the reactor trip had been manually initiated. By 06:34:21, the ERFC data set charging flow is reduced to 82.568 gpm and feed flows are l about 600KLBH to 800KLBH. By 06:34:34, all flows had reached a stable but low level. It appears that SI occurred at or slightly before 06:34: 14. This chronology will use 06:34:14.
| |
| For automatic initiation of Safety Injection, the clock comparisons are as follows:
| |
| RECORDER TIME Sequence of Events Recorder (SER) 06:35:24:805 ,
| |
| P-250 Computer 06:35:41 ;
| |
| ERF Computer (ERFC) 06:34:14 '
| |
| l l
| |
| | |
| For Reactor Manual Trip the clock comparisons are as follows:
| |
| RECORDER TIME Ser:oence.of Events Recorder (SER) 06:35:04:548 P-250 Computer 06:35:24 ERF Computer (ERFC) 06:33:56 It is concluded that the P-250 led, the SER was within 20 seconds of the P-250 and the ERFC was about one minute behind the P-250.
| |
| The chronology that follows is annotated by clock time based on the P-230.
| |
| All the events that occurred within each minute are listed in order of occurrence as could best be determined.
| |
| I l
| |
| l l . _ _ - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ -
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| | |
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| |
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| |
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| |
| R U P D E E T - - - - - - - - - -
| |
| C S O
| |
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| |
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| |
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| |
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| |
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| M R o oo o o o o. 2 66 2 R U L LI L 1 o. L L1 O O P PP P F S O oO O O0 OO I I I l N R R R R RR RR P PP P I C S C S SC SC E EC E E
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| R U P D E 2 E T 9 4 8 C S 4 - - 62 2 - - 4 O
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| / 3 1 0 2 1 1 E E 0 05 0 R C 1 0. 1 U N . 1 2 66 2 D E 8 P E R 2 PI P PP P C E - - - I P I I I I _
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| N E _I V
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| - A R T T t ot t r oa E O N P C s a e l i
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| T N E I t z v A c S V R s yy i o T E C . u r c n G S p S d a on a i E e h t e g l rr _
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| o t ut N a d
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| e t c S n O o R c a M y n t e C
| |
| - e d c di n d g n S
| |
| t e r
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| d e
| |
| n e
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| em i r e
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| m er i e no o
| |
| o r g f e e f m d n t e r it l i e d e i s t e t p t es a e n m of m of sa M R E E N o I N o ab b
| |
| e em mi
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| _ i t t
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| _ E 4 2 0 5 6 6 3 e
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| - H 5 1 3 3 3 3 4 et l 2 3 3 3 3 3 3 t a T 1 1 1 1 1 1 1 am a i
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| ,lI fl1 j[ ,lll f l ,l ll
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| | |
| ATTACllMENT 2 NORTH ANNA UNIT 1 OPERATIONAL HISTORY PRIOR TO THE "C" S/G TUBE RUPTURE EVENT OF JULY 15, 1987 DATE TIME COMMENTS 04-19-87 0800 Commenced Unit I ramp down for refueling outage.
| |
| 1600 Unit 1 off line.
| |
| 04-20-87 0600 Commenced RCS hot hydrostatic test.
| |
| 1300 Hydrostatic pressure test terminated. Retested satisfactorily on 06-28-87.
| |
| 04-28-87 0600 Started "B" S/G eddy current inspection.
| |
| 05-05-87 1800 Completed "B" S/G inspections.
| |
| 05-08-87 0000 Started "A" S/G eddy current testing.
| |
| 05-15-87 2100 Completed "A" S/G inspections.
| |
| 05-16-87 0600 Started "C" S/G eddy current testing.
| |
| 05-23-87 1000 Completed "C" S/G inspections.
| |
| 06-08-87 0013 Commenced RCS fill and vent.
| |
| 06-10-87 2342 RCS stable at 195*F and 310 psig.
| |
| 06-11-87 1022 Established Prr bubble 06-12-87 0510 Drained, refil:ed, and added chemicals to "C" steam generator, level is 70% W.R.
| |
| 06-15-87 1135 Established containment vacuum.
| |
| 06-17-87 1244 Electrical trip on "A" Reactor Coolant rump (1-RC-P-1A), found motor dancge that required motor replacement. RCS still (200*F (mode 3).
| |
| 1832 Containment vacuum broken, and RCS depressurination underway to allow RCP motor work.
| |
| 06-28-87 0132 RCP "A" motor replaced, containment vacuum established, unit enters mode 4, heat up in progress.
| |
| 0816 Unit i enters mode 3 (>350*T)
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| Page 1
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| | |
| _DA_TE TIME COMMENTS i
| |
| 06-28-87 1450 Unit 1 0 521'F and 2296 psig for RCS hydrostatic l pressure test. ,
| |
| 1
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| {
| |
| 2053 RCS hydrostatic pressure test is completed.
| |
| 06-29-87 0310 Unit I reactor critical, unit is in mode 2.
| |
| 0900 Completed " Low Power Physics Testing" 2229 Unit 1 on-line. l l
| |
| 2248 Unit 1 Reactor and Turbine trip on 5A Feedwater heater high-high level.
| |
| 06-30-87 0321 Unit I reactor is critical. [!i 0621 Unit 1 on-line.
| |
| )
| |
| 1044 Unit 1 is 23% reactor power, holding due to inability I "C" .i to open the manual isolation valve on main feedwater regulation valve.
| |
| 07-02-87 0205 Unit 1 commenced a ramp down for maintenance on the FW manual isolation valve.
| |
| l l 0326 Unit 1 off-line.
| |
| l 0450 Shutting down Unit I reactor, entered mode 3.
| |
| l 0637 RCS cooldown and depressurization to 205*F underway.
| |
| 1445 Unit 1 is on RHR.
| |
| 07-04-87 0445 Maintenance on FW manual valve is completed. Commenced heat up of RCS.
| |
| 0501 Unit 1 in mode 3. .
| |
| 1534 l' nit I reactor critical. ;
| |
| 2123 Unit 1 on-line.
| |
| 07-05-87 0210 Unit 1 is being held < 30% power on chemistry hold to i clean up the steam generators.
| |
| 07-07-87 0158 Unit 1 is being held at 83% power for investigation of l low steam flows to the "B" Moisture Separator Reheater. !
| |
| 0240 Unit I rampdown to 50% power is commenced.
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| Page 2
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| _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ . _ _ . _ _ _ . . _ _ _ . ____ ____.______._.______.___i
| |
| | |
| i l
| |
| 1 DATE ' TIME COMMENTS t 07-10-87 2103 Commenced rampdown of Unit I to allow maintenance work on "B" MSR stop valve.
| |
| 2320- Unit 1 off-line.
| |
| 2340 Unit I reactor shutdown, entered mode 3.
| |
| 07-11-87 0056 Closed Unit 1 Main Steam Trip Valves (MSTV) for work on the "B" MSR Stop Valve. l l
| |
| 2357 Completed work on "B" Reheat Stop valve.
| |
| 07-12-87 0017 Unit I reactor critical.
| |
| 0356 Unit 1 on-line. .
| |
| I 0510 Unit 1 stabilized at 27% power for steam generator chemistry hold, 1519 Ramp up of Unit 1 to 100% power underway.
| |
| 1900 Chemistry notification of action Level 1 .on high FW cation conductivity.
| |
| 07-13-87 0809 Air ejector radiation monitor (RM-RMS-121) on Unit 1 is declared inoperable on erratic operation.
| |
| 1524 Chemistry notification of Main Steam cation conductivity out of specifications, entered Action l Level 1.
| |
| l 07-14-87 0815 Action Statement cleared on air ejector radiatien ,
| |
| monitor, the monitor has been stable over night.
| |
| 2238 Air ejector radiation monitor is erratic, and is declared inoperable. Entered Action Statement with grab (local) samples again.
| |
| 07-15-87 0630 Unit 1 Operator receives indications of a Steam l Generator Tube Rupture.
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| I t
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| i Page 3 i
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| L
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| ~ ATTACHMENT 3 SEQUENCES OF EVENTS HEALTH PHYSICS DATE TIME COMMENTS 07-13-87 0900 Air Ejector sample due'to RM-SV-121 being out of service. Power level 85.5%.
| |
| 1740 Air Ejector sample due to RM-SV-121 being out of i service. Power level 87.8%. :f d
| |
| 07-14-87 0114 Air Ejector sample due.to RM-SV-121 being'out of )
| |
| service. Power level-99%. ]
| |
| {
| |
| 2245 Air Ejector sample due to RM-SV-121 being out of I service. Power level 100%.
| |
| I
| |
| ; 07-15-87 0620 Air Ejector sample due to RM-SV-121 being out of l se rvice . Power level 100%. ,
| |
| 0650 EPIP 4.01 " Radiological Assessment Director l Controlling Procedure implemented. )
| |
| 0650 Air Ejector Sample for S/G Tube Leak Evaluation.
| |
| Reference Table III-1 for data.
| |
| )
| |
| 0652 EPIP 4.02 " Radiation Protection Controlling Procedure" implemented. j f
| |
| l 0655 HP Technician assigned to the Control Rocm to l
| |
| observe Plant Rndiation Monitors as follows.
| |
| ! Ventilation Vent A and B. !
| |
| Process Vent.
| |
| Steam Driven Aux Feed Water Pump Exhaust, j Main Steam A, B, and C. j 0658 EPIP 4.14 "Inplant Monitoring" procedure implemented. !
| |
| 0700 EPlP 4.28 "TSC/LEOF Radiation Monitoring System" procedure implemented.
| |
| Page 1
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| | |
| DATE TIME COMMENTS 07-JS-87 0705 EPIP 4.17 " Monitoring of OSC/TSC" procedure implemented.
| |
| 0708 EPIP 4.08 " Initial Offsite Release Assessment" procedure implemented.
| |
| 0710 EPIP 4.15 "Onsite Monitoring". procedure implemented.
| |
| 0712 EPIP 4.25 " Liquid Effluent Sampling During An Emergency" procedure implemented.
| |
| 0718 Condensate sample.
| |
| 0718 "A" Main Steam sample.
| |
| 0718 "C" Main Steam sample.
| |
| 0720 EPIP 4.03 " Dose Assessment Controlling Procedure" implemented.
| |
| EPIP 4.12 "Offsite Environmental Monitoring Instructions" procedure implemented.
| |
| 1 0722 Cold Lab Particulate Charcoal sample 0722 EPIP 4.27 "Use Of The Class A Meteorological And Dose Calculational Model" procedure implemented.
| |
| 0728 EPIP 4.16 "Offsite Monitoring" procedure implemented, for monitoring Teem 1. i 0729 Operations support (OSC) Particulate / Charcoal sample.
| |
| l 0730 EPIP 4.19 "Use of Radios For HP Monitoring" procedure implemented.
| |
| 0735 Technical Support Center (TSC) Particulate / Charcoal sample.
| |
| 0738 Service Building hallway, Particulate / Charcoal sample.
| |
| 0745 Access to MSHV, Turbine Building, Auxiliary Building and Steam Driven AFW Pump Building restricted.
| |
| 0807 "C" Steam Generator primary sample.
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| Page 2 l
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| l l'
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| l L_--_---_--_.-.--__--_.__--__.- _ . -
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| | |
| l DATE TIME C0!OtENTS 07-13-87 0807 Hot Leg Primary sample.
| |
| l 0813 Yard, Guard Tower 3, Particulate / Charcoal sample.
| |
| 0815 Survey of Unit 1 Mechanical Equipment Room.
| |
| 0816 Chemistry Sample Sink, Particulate / Charcoal sample.
| |
| 0823 EPIP 4.10 " Determination of X/Q" procedure implemented.
| |
| 0824 Turbine Building, Common Sump, water sample.
| |
| 0830 TSC Particulate / Charcoal sample.
| |
| 0830 EPIP 4.16 "Offsite Monitoring" procedure implemented, for monitoring Team 2.
| |
| 0846 Turbine Building, 279' Level Particulate / Charcoal l sample.
| |
| l 0852 offsite air sample / survey at access road for count l room analysis.
| |
| 0853 EPIP 4.I1 " Follow-up Offsite Release Assessment" procedure implemented.
| |
| 0858 EPIP 4.09 " Source Term Assessment" procedure implemented.
| |
| 0908 Ventilation Vent A sample.
| |
| 0910 offsite air sample / survey at Aspen Hills subdivision for count room analysis.
| |
| 0913 Ventilation Vent B sample.
| |
| 0915 OSC Particulate / Charcoal sample.
| |
| 0921 Process Vent Particulate / Charcoal / Gas sample.
| |
| 0926 TSC Particulate / Charcoal sample.
| |
| 0929 Turbine Building, 271' Level Particulate / Charcoal /Cas sample.
| |
| 0938 Turbine Euilding Basement Particulate / Charcoal sample.
| |
| Pace 3
| |
| | |
| 1 DATE TIME COMMENTS 07-15-87 0943 Steam Generator A, B, C, Blowdown sample. Sample utilized to evaluate activity from steam driven Auxiliary Feedwater Pump exhaust.
| |
| Reference Table III-2 for Data with A, B, C S/G l Steam to Pump.
| |
| Reference Table III-3 for Data with A, B S/G Steam to Pump, C S/G isolated.
| |
| 0945 "A" Cold Leg primary sample 0956 Aux Building Walkvar 272' Level, Particulate / Charcoal sample counted.
| |
| j 1004 Local Emergency Operations Facility (LE0F), y Particulate / Charcoal sample. ]
| |
| l 1010 EPIP 4.18 " Monitoring of LEOF" procedure l implemented.
| |
| i 1032 TSC, Particulate / Charcoal sample.
| |
| 1142 LEOF, Particulate / Charcoal sample.
| |
| l i 1144 TSC, Particulate / Charcoal sample, j 1205 Vegetation sampled at access road southwest of )
| |
| plant for count room analysis. I 1228 TSC, Particulate / Charcoal sample.
| |
| 1333 Aux Building, 274' Level,. Particulate / Charcoal sample.
| |
| 1440 Monitoring Team Particulate /Chsrcoal sample counted.
| |
| 1448 Monitoring Team Particulate / Charcoal oneple counted.
| |
| 1453 Monitoring Team Particulate / Charcoal sample counted.
| |
| 1629 - Soil and vegetation samples counted prior to 1941 shipment offsite for analysis Page 4
| |
| | |
| 1 j
| |
| I DATE TIME COMMENTS 07-15-87 0658 - Miscellaneous onsite surveys were performed in 1252 accordance with EPlP 4.14, "Inplant Monitoring."
| |
| )
| |
| The following areas were surveyed.
| |
| Turbine Building i Mechanical Equipment Room l- Service Building Hallways outside Areas, within the Protected Area Auxiliary Building Chemistry Labs Technical Support Center Operational Support Center Local Emergency Operations Center Aux. Feedwater Pump House Laundry and PDA Area Main Steam Valve House No.1 l.
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| l l
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| l I
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| I l
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| l l
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| l Page 5
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| = _ - - _ - - _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ -. . _ _ . __ _ - _ _ _ .
| |
| | |
| ATTACHMENT 4 EMERGENCY PLAN CllRONOLOGY TIME COMMENT 0638 EPIP-2.01, Notification of State and Local Governments, initiated.
| |
| 0639 " Notification of Unusual Event" declared.
| |
| Unit 2 SRO assumed duties as Interin Station Emergency Manager.
| |
| EPIP-1.01, Emergency Manager Controlling Procedure, initiated.
| |
| EPIP-1.02, Response to NOUE, initiated.
| |
| 0645 Security requested to dispatch officer to Control Room for access control, i
| |
| l 0650 EPIP-4.01, Radiological Assessment Director Controlling Procedure,
| |
| ! initiated.
| |
| 0651 Initial offsite notification regarding "NOUE" is transmitted to State and local authorities.
| |
| 0652 EPIP-4.02, Radiation Protection Supervisor Controlling Procedure, initiated.
| |
| 0654 Interin Station Etergency Manager upgraded event classification to
| |
| " ALERT". EPIP-1.03, Response to Alert, initiated.
| |
| Security officer established post at Control Room.
| |
| 0655 Security is notified by Control Room to initiate EPIP-3.01, Callout, EPIP-5.03, Accountability, and EPIP-5.04, Access Control.
| |
| Security is instructed by Control Roon to have officer at roadblock to allow emergency response personnel onrite.
| |
| Non-essential personnel are to be held at roadblock, and all contractors, salesmen, and vendors arriving to the . cite should be turned back.
| |
| 0700 EPIP-3.01, Activation of OSC, initiated.
| |
| Security directed to initiate CPIP-3.04, LEOF Activation.
| |
| EPIP-4.14. Inplant Monitoring, initiated.
| |
| EPIP-4.28, TSC/LEOF Radiation Monitorinc System, initiated.
| |
| 0702 Initial " ALERT" notification messages transmitted to NRC, State and local authorities. (EPIP-2.01 previously initiated.
| |
| EPIP-2.02, Notification of NRC, initiated.)
| |
| 0704 EPIP-3.02, Activation of TSC, initiated.
| |
| 0705 EPIP-4.17, Monitorine of OSC and TSC, initiated.
| |
| Page 1
| |
| | |
| '0708 Security roadblock established at Rt. 700 and station access road.
| |
| EPIP-4.08, Initial Offsite Release Assessment, initiated.
| |
| 0710 General Office Security conducts Corporate Emergency Response Team callr ut.
| |
| EPIP-4.15, Onsite Monitoring, initiated.
| |
| 0712 Security completes LEOF activation / setup.
| |
| EPIP-4.25, Liquid Effluent Sampling During an Emergency, initiated.
| |
| 0715 Supt. Operations and SRO On-Call arrive in Control Room.
| |
| HP prepares to restrict access to Turbine and Auxiliary Buildings,
| |
| * l as well as Main Steam Valve House.
| |
| Inplant monitoring teams dispatched. Radiological check-point and dosimetry issue established at entrance of Service Building.
| |
| 0718 3rd message is transmitted to State and local authorities.
| |
| 0720 Station Manager arrives in the Control Room.
| |
| EPIP-4.03, Dose Assessment Controlling Procedure, initiated.
| |
| EPIP-4.12, Offsite Environmental Monitoring Instructions, initiated.
| |
| ( 0727 EPIP-4.27, Use of Class "A" Meteorological and Dose Calculational Medel, initiated.
| |
| 0728 Station accountability complete. 10 individuals missing.
| |
| Offsite monitoring team #1 dispatched. (EPIP-4.16,-
| |
| l Offsite Monitoring, initiated).
| |
| 0730 Assistant Station Manager arrives in Control Room and initiates transition of EPIPs and communications from Control Room to TSC.
| |
| EPIP-4.19, Use of Radios for HP Monitoring, initiated.
| |
| l l 0739 Station Manager assumes Station Emergency Manager (SEM) position.
| |
| l 0745 4th message is transmitted to State and local authorities.
| |
| Security is notified by HP that access to the following areas is restricted and that the areas are roped off:
| |
| U1 - Main Steam Valve House U1 - Steam Driven Aux Feedwater Pump House Turbine and Aux Building.
| |
| A radio announcement is made to all Security personnel on restricted access areas.
| |
| EP1P-4.06, Personnel Monitoring and Decontamination, initiated.
| |
| 0747 All missing personnel located and accountability completed.
| |
| Page 2
| |
| . - _ _ _ _ _ - . _ - . _ _ _ _ .- - - - - - - - - _ _ _ _ _ _ _ _ _ -.__-..____________:-______ . _ . _ _ _ _ _ _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - . - - _ _ _ _ _ _ ______i_J
| |
| | |
| t 0752 EPIP-5.08, Damage Control Guideline, initiated.
| |
| 0753 TSC fully manned. OSC fully manned and activated.
| |
| 0757 TSC activated.
| |
| l 0800 Public Affairs inter-departmental callout conducted.
| |
| L Press release #1 is issued.
| |
| i 0801 SEM and TSC Directors provide status update to: facility.
| |
| 0805 Sth message is transmitted to State and local authorities.
| |
| 0818 Security is directed to allow access to Westinghouse j representative.
| |
| l 0820 Corporate Emergency Response Center activated.
| |
| l EPIP-2.04, Transmittal of Plant, Radiological and Emergency Status, initiated.
| |
| l l 0823 EPIP-4.10. Determination of X/Q, initiated.
| |
| l I 0828 SEM, TSC Directors provide status update to facility.
| |
| Turbine and Auxiliary Building released from radiological controls / restricted access.
| |
| Westinghouse representative onsite.
| |
| 6th message is transmitted to State and local authorities.
| |
| 0830 Corporate Public News Center operational.
| |
| Rumor Control Area operational.
| |
| HPN Communications established with NRC from TSC.
| |
| Security is notified by TSC to allow Virginia Power personnel at
| |
| ; roadblock to assemble in Warehouse #2.
| |
| Offsite monitoring team #2 dispatched.
| |
| 0843 LEOF updates Old Dominion Electric Cooperative.
| |
| 0845 Security is notified by TSC to release Virginia Power personnel to no rmal work areas. All contractors are instructed to assemble in-Project Area. Contractors that do not work in Proj ect Area are instructed to assemble in construction parking lot. All vendors and salesmen should still be turned back at the roadblock. Any news media personnel should be directed to the Local Media Center at Mineral Fire Hall.
| |
| Initiated callout of Security personnel to report at 1100 hours, ir: order to allow midnight shift personnel to leave.
| |
| 0846 7th message is transmitted to State and local authorities.
| |
| Page 3 i
| |
| | |
| 0853 CPIF-4.11, Follow-up Offsite Release Assessment, initiated.
| |
| 0858 EPIP-4.09, Source Term Assessment, initiated.
| |
| 0900 2nd press release is issued.
| |
| 0905 Security notifies Engineering and Construction that no vehicles will be allowed into Protected Area unless associated with the ;
| |
| emergency.
| |
| 0913 8th message transmitted to L'Ote and local authorities.
| |
| 4 0915 LEOF activated.
| |
| 0935 Corporate Security arrives at LEOF.
| |
| I LEOF assumes responsibility for State and local notifications'from i TSC and transmits 1st message.
| |
| f 0937 CERC weather forecast provided to LEOF.
| |
| 0955 Security is directed by TSC to allow OMNI contract personnel into the Protected Area. j 1000 Local-Media Center fully operational.
| |
| l 1009 2nd LEOF message to State and local authorities transmitted.
| |
| _1010 EPIP-4.18, Monitoring of LEOF, initiated.
| |
| i 1015 LEOF conducts conference to begin recovery planning process.
| |
| 1018 UPI arrives at Security roadblock and is directed to the Local Media Center. !,
| |
| 1035 LEOF making arrangements for media tour of site in afterncen.
| |
| 1040 Received call from Jeff Lankford, MRC Region II. Notified that F.
| |
| Cantrell, B. Revsin, L. Nicholson, and S. Gagner en route to'Sorth Anna. ETA 1300 hours.
| |
| 1043 3rd LEOF messcge transmitted to State and local authorities.
| |
| 1058 CERC provides weather forecast update.
| |
| 1100 3rd press release is issued.
| |
| 1102 State-Corporation Commission representatives arrive onsite (LEOF).
| |
| l 1115 4th LEOF message to State and local authorities is transmitted.
| |
| I l
| |
| Tage 4 i
| |
| l l
| |
| | |
| l 1139 CERC provides weather forecast update.
| |
| 1140 Mr. Lankford with NRC informed that Cantrell, Revsin, and Nicholson have a current badge at North Anna, and that access will be expedited.
| |
| 1148 5th LEOF message transmitted to State and local authorities.
| |
| 1210 CERC provides weather forecast update.
| |
| 1220 6th LEOF message transmitted to State and local authorities.
| |
| 1 j 1238 CERC provides weather forecast update.
| |
| I
| |
| ! 1241 LEOF provides update to Old Dominion Electric Cooperative.
| |
| I?52 7th LEOF message is trenecitted to State and local authorities.
| |
| I 1303 NRC team arrives onsite.
| |
| 1370 Security notified TSC that Channel 9 news was at roadblock and stated that they were sent from. Mineral Fire Hall (LMC). j 1326 8th LEOF message transmitted to State and local authorities.
| |
| )
| |
| i 1328 Security notified Channel 9 news at roadblock, as directed by TSC, j to return to the LMC and that a press conference would be held '
| |
| onsite later in the day.
| |
| 1333 EPIP-6.01, Re-entry / Recovery Guideline, initiated. 1 1335 Emergency terminated. ;
| |
| 1336 Recovery meeting is scheduled in LEOF to outline organization and long+ term plans (Recovery Organization is implemented).
| |
| NRC Communications terminated.
| |
| ! Emergency Response Facilities in process of deactivation.
| |
| 1342 Security notified to terminate TSC and Control Room posts. I 1343 9th and final (termination) message sont by LEOF to State and local authorities.
| |
| 1345 Mr. Nichoir,on (NRC) onsite. Directed to LEOF for conference. 3 l
| |
| I 1355 Security instructed by LEOF to escort media from LMC to station for press conference.
| |
| 1400 Final press release issued.
| |
| i l
| |
| Page 5 i
| |
| 1
| |
| | |
| 1 1
| |
| 1505'' . News media ensite. .
| |
| Security rondblock terminated.
| |
| '1603~ Security post at LEOF terminated, j Fecovery meetings, outage planning proceeds into the night.
| |
| t I
| |
| 'l i
| |
| f 1
| |
| s a
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| Page 6
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| ATTACIMENT 5 l
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| l f EMERGENCY OPERATING PROCEDURE l
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| I EP-0, REACTOR TRIP OR SAFETY INJECTION ,
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| f VIRGINIA POWER j
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| . NORTH ANNA POWER STATION w.au EMERGENCY PROCEDURE Procedure Title Revision Number 1.00 ;
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| I REACTOR TRIP OR SAFETY INJECTION Page 1-EP-0 (WITH TWO ATTACHMENTS) . 1 of 16 Purpose The purpose of this procedure is to verify proper response of the Reactor ,
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| Protection and Emergency Core Cooling Systems following actuation of a REACTOR l i TRIP or SAFETY INJECTION; and to assess plant conditions and identify the appropriate recovery procedure.
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| 1 i
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| User NAPS Operations Personnel j Entry Condittons u
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| j Any of the following exist: ,
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| l 1) -A-Reactor Trip has occurred as determined by NIS instrumentation, IRPI indication, or Reactor Trip breaker Status,
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| : 2) A Reactor Trip is required as determined by setpoints or other requirements being exceeded, ;
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| : 3) A Reactor Trip / Safety Injection has occurred as determined by SI p' imp status or EDG status,
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| : 4) A Reactor Trip / Safety Injection is required as determined by setpoints or other requirements being exceeded, or
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| : 5) Transition from another plant procedure. -
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| SAFT( R E _A D Rsvision Record REY. 1.01 PAGES (S) : 11, 13, Attach 2 DATE: 06-12-87 REV. PAGES(S): DATE:
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| REV. PAGES(S): DATE:
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| REY. PAGES(S): DATE:
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| REV. PAGES(S): DATE:
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| REV. PAGES(S): DATE:
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| Approval Recommended Approved Date e
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| k )
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| Chairman Station Nuc ear Safety 06-12-87 and Operating Committee
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| 1
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| ~
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| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES 1
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| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
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| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
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| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F (70*F].
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| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0,. REACTOR TRIP OR SAFETY INJECTION, Step 1, if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [c0*F].
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| O_3,
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| * PRZR level - CANNOT BE MAINTAINED GREATER TEAN 15% [50%)
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| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pu=p recire, path MOVr if - RCS PRESSURE DECREASES TO LESS THAN 1275 PSIG [1575 PSIG).
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| b) Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG.
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| : 4. RED PATH
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| ==SUMMARY==
| |
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| a) SU3 CRITICALITY - Nuclear power greater than 5" ,
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| i b) CORE COOLING - Core Exit TCs greater than 1200*F O_R, Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10" (32"] AND i
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| total feedwater flev less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100'F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
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| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
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| 1 i
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| s e7ss7:1o 1 NUCBER l'RnCEDURE TITLE REVISION l 1.00 ]
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| 1-EP-0 REACTOR TRIP'OR SAFETY INJECTION PAGE 2 of 16 I
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| - STEP ACTIONNXMCTED RESPONSE RESPONSENOTOBTAINED .i j
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| NOTE: Bracketed [ ] numbers show immediate action Steps.
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| l NOTE: Setpoints in brackets [ ] are for_ adverse containment atmosphere (20 psia containment pressure or 10 R/HR containment radiation).
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| [1.] MANUALLY TRIP REACTOR:
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| a) Open both reactor trip breakers b) Verify reactor - TRIPPED b) GO TO 1-FRP-S.1, RESPONSE TO NUCLEAR POWER -]
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| * Reactor trip and bypass GENERATION /ATWS, Step 1. i breakers - OPEN j
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| * Neutron Flux - l DECREASING TO LESS I THAN 5%
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| [2.] MANUALLY TRIP TURBINE: I a) Depress both turbine trip pushbuttons b) Verify all turbine stop b) Place both EHC pumps in !
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| valves - CLOSED P-T-L. j IF turbine still not Eipped. THEN manually runback turbine.
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| IF turbino canNOT be Enback, THEN close MSTVs l and bypass valves.
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| c) Close reheater inlet valves i by depressing reset
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| l FOLD 0DT FOR EP-0 AND ES-0 PROCEDURES
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| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
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| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
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| * RCS subcooling based on Cera Exit TCs - LESS TEAN 25*F [70*F].
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| t
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| : 2. SI REINITIATION CRITERIA :
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| Manually initiate both trains of SI g GO TO EP-0,, REACTOR TRIF OR SAFETY l INJECTION, Step 1 if either condition liste'd below occurs:
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| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F]. !
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| Oli
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| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15 [50*l.
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| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE
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| ! DECREASES TO LESS TEAN 1275 PSIG [1575 PSIG).
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| , b) Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE
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| ( INCREASES TO 2000 PSIG.
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| i l 4. RED PATH
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| ==SUMMARY==
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| a) SU3 CRITICALITY - Nuclear power greater than 5*
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| b) CORE COOLING - Core E;:it TCs greater than 1200*F
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| ! OR Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10" [30%] AND total feedvater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'T e) CONTAINMENT - Containment pressure greater than 60 PSIA
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| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%. .
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| 1
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| No. 07887810 NUMBER PROCEDURE TITLE REVISION
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| .00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 3 of 16
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| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
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| [3.] VERIFY POWER TO AC l EMERGENCY BUSSES:
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| a) AC emergency busses - AT a) Immediately restore power LEAST ONE ENERGIZED to at least one AC emergency bus.
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| E power canNOT be restored, THEN GO TO 1-ECA-0.0, LOSS OF ALL AC POWER, Step 3.
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| b) AC emergency busses - b) Try to restore power to ALL ENERGIZED deenergized AC emergency bus.
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| [4.] CHECK IF SI IS ACTUATED: Verify that none of the following conditions requiring
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| * Charging /SI pumps - RUNNING SI have occurred:
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| * LHSI pumps - RUNNING
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| * Emergency Diesel Generators -
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| * Low PRZR pressure RUNNING
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| * High containment pressure
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| * Any SI First-Out Annunciator
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| * Steamline differential LIT pressure
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| * High steamflow with low Tavg
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| * High steamflow with low steam pressure IF,SI required, THEN GO TO Step .5_.
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| IFF NOT required, THEN GO TO 1-ES-0.1, REACTOR TRIP RESPONSE, Step 1.
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| r -; , + ,y g
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| : h. [, sy- g N $ .y- s m 4s r
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| -)
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| .g
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| 's I
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| POLD0t1T FOR EP-0 AND ES-0 PROCEDURES 3
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| X t u .
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| T.
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| l -
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| : 1. .EO..
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| .- TRIP CRITERIA' Trip di RCPs if BOTH conditions listed below exist:
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| * Charging /S Fpumps - A" LEAST ONE RUNNING, re.
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| AND ,
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| * RCS subcooling based on Core Exit TCs - LESf"THAN 2{3*F, 0*F]. ,
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| : 2. SI REINITIATION CRITERIA s
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| Manually initiate bnf *Tains of SI AND GO TnZP-0,.Ri$: TOR TRIP OR SAFETY INJECTION, Step lif a{ther condition listed below occurs: ". -
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| ,ue .
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| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
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| S OR .
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| * PRZR level - CANNOT BE MAINTAINED GREATER TFJS 15% [50") .
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| Yi s - ,+
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| l ' ,Y
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| '3 <CEARGING/SI PUMP RECIRC PATH CRITERIA ,
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| ' k.1. .r a). . Isolate the Chargf.ng/SI pump'recire, path MOVs if.- RCS PRISSURE
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| .
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| * DECREASES TO LEGt,'THAN 1275 PSIG (1575 PSIG].
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| b) Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE- -;
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| INCREASES TO .2000. PSIG. ,. ym -
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| : 4. RED PATH SUMif g a) SUICRITICil.ITY y Nuclear pcuer greater than 5". .
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| , b) CORE COOLING-- Core Exit TCs greater than'1200*F' " N.
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| 4
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| % o.
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| y OR w ''
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| . , {
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| _J ,
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| " ; Corp Exit TCs greater than 700*F AND RVLI% dull range i.
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| % 91esa^than 46" with yd RCPs running. +
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| c) HEAT SINK - Narrow Range.Iavel 5.n 211 SGs less than 10" (32*.] AND total'feedvater flow ikes thru 340'gpm.
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| m d) INTEGRITY - Cold leg temperature decrease greater than 100'F in last u 60 minutes AN_D,,kcS. cold leg temperature'iess than 285'F e) CONTAINMENT - ContainGU3 pressure greater than 60 PSIA; ,,w - .
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| "6'
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| .p,
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| : 5. ECST LEVEL CRITERIA .
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| Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40'.. *
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| \-
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| .n ,Q 4
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| is_.
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| 1
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| '*l
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| ^)-1
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| ,3 $ *Q
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| .e .
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| , _ w. %'
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| ') .
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| ..Jl m'h
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| No. 97887213 NU:BER PROCEDURE TITLE REVISION 1.00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 4 of 16
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| - STEP ACTION / EXPECTED RESPONSE . PESPONSE NOTOBTAINED l
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| (5.] , MANUALLY INITIATE BOTH TRAINS OF SI l
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| -[6.] , CHECK CHARGING /SI PUMP ALIGNMENT:
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| a) RWST Suction - OPEN a) Manually open valves.
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| * MOV-1115B
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| * MOV-1115D f
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| b) VCT Suction - CLOSED b) Manually close valves. >
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| * MOV-1115C -
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| * MOV-1115E c) Normal charging - CLOSED c) Manually close valves.
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| 4
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| * MOV-1289A
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| * MOV-1289B d) Letdown isolation - CLOSED d) Hanually close valves. I
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| * HCV-1200A
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| * HCV-1200B
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| * HCV-1200C j
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| t
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| 's FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
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| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below' exist:
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| * Charging /SI. pumps - AT LEAST ONE RUNNING, t.ND
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| * RCS subcooling based on Core Exit TCs - LESS THAN 25*F [70*F].
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| i l
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| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0.. REACTOR TRIP OR SAFETY INJECTION, Step 1, if either condition listed below occurs: !
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F).
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| i pR
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| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15 [50%]. ,
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| l
| |
| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pump recire, path MOVs if - RCS. PRESSURE DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| b) Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE I
| |
| INCREASES TO 2000 PSIC.
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| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SU3 CRITICALITY - Nuclear power greater than 5: J I
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| b) CORE COOLING - Core Exit TCs greater than 1200*F OR 1
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46 with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% (32%] AND ]
| |
| total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or prsvide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
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| e i
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| i No.97887210 PROCEDURE TITI.E REVISION NUMBER 1.00 j 1-EP-0 REACTOR TRIP OR SAFETY INJECTION ^
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| PAGE 5 of 16
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| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
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| [ 7. },, VERIFY CHARGING /SI FLOW-I a) Two Charging /SI pumps - a) Manually start pumps.
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| RUNNING b) Cold leg SI flow - INDICATED b) Manually align BIT: ,
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| 1
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| * FI-1961 1) Close BIT recire
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| * FI-1962 valves:
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| * FI-1963
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| * TV-1884A
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| * FI-1943
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| * TV-1884B
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| * FI-1943-1
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| * TV-1884C
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| '2) Open BIT outlet valves: )
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| * MOV-1867C
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| * MOV-1867D
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| : 3) Open BIT inlet valves:
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| * MOV-1867A
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| * MOV-1867B
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| [8.] VERIFY FW ISOLATION:
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| l l
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| * Main FW flow control
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| * Manually close valves.
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| valves - CLOSED 1
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| * FW isolation MOVs - CLOSED
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| * Manually close valves. l
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| * Main FW pumps - TRIPPED
| |
| * Manually trip pumps.
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| * Bypass flow control valves -
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| * Manually close valves.
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| CLOSED I
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| * Standby Main FW pump (s) in -
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| * Place in-P-T-L. l P-T-L l
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| i POLDOUT FOR EP-0 AND ES-0 PROCEDURES
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| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
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| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'T (70'Fi.
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND.GO TO EP-0.. REACTOR TRIP OR SAFETY INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
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| S.E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER (50%]. TRAN 15: i j
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| : 3. .
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| Open the Charging /SI pump recire path'MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG. i
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| : 4. RED PATH SUMMART a) b) SU3 CRITICALITY - Nuclear power greater than 5:
| |
| CORE COOLING - Core Exit TCs greater than 1200*F pR Core Exit TCs greater than 100*F AND RVLIS full range c) 1ess than 46: with no RCPs running {
| |
| HEAT SINK - Narrow Range level in all SGs less than 10% [32:] AND i d) total feedwater flow less than 340 gpm INTEGRITY - Cold leg temperature decrease greater than 100'F in last.
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| e)
| |
| CONTAINMENT - Containment pressure greater than 60 PSIA I
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| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
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| I I
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| .. )
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| l m.e7es7:10 l
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| NUMBER PROCEDURE TITLE REVISION !
| |
| .00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 6 of 16
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| - STEP ACTION / EXPECTED RESPONSE RESPONSFNOTOSTAINED
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| [ 9. ]_ VERITY AFW PUMPS - RUNNING:
| |
| a) Motor-driven AFW a) Manually start pumps.
| |
| pumps - RUNNING b) Turbine-driven AFW pump - b) Manually open steam supply RUNNING valves.
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| [10.] VERIFY CONTAINMENT ISOLATION PHASE A:
| |
| a) Phase A - ACTUATED a) Manually initiate both trains of Phase A. <
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| b) Initiate Attachment 2
| |
| [11.] VERIFY LHSI PUMPS STATUS:
| |
| a) Two LHSI pumps - RUNNING a) Manually start pumps, b) RCS pressure - LESS THAN b) GO TO Step M.
| |
| 175 PSIG c) LESI pump flow - INDICATED c) Manually open LHSI valves:
| |
| * FI-1945
| |
| * MOV-1864A
| |
| * FI-1946
| |
| * MOV-1864B
| |
| * MOV-1890C
| |
| * MOV-1890D
| |
| [12.] VERIFY SERVICE WATER PUMPS - Manually start pumps as RUNNING: required. I i
| |
| * 1-SW-P-1A - RUNNING pR, 1-SW-P RUNNING
| |
| * 1-SW-P-1B - RUNNING i
| |
| - .j
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES l
| |
| l
| |
| : 1. RCP TRIP CRITERIA Trip all RC?s if BOTH conditions listed below exist:
| |
| 1
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, j AND
| |
| * RCS subcooling based on Core Exit TCs - LESS TRAV 25'F [70*F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0,. REACTOR TRIP OR SAFETY INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80'F].
| |
| E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15: [50%].
| |
| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE DECREASES TO LESS THAN 1275 PSIG [1575 PSIG].
| |
| b) Open the Charging /SI pump recire path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG.
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SU3 CRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F 8
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10" [32".] AND total feedwater flow less than 340 gpm )
| |
| d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last i 60 minutes AND RCS cold leg temperature less than 285'F )
| |
| e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| i e
| |
| | |
| i
| |
| :lD. 97887210 NUMBER PROCEDURE TITLE REVISION 1.00 i 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 7 of 16
| |
| - STEP ACTIONNXPECTED RESPONSE RESPONSENOT0 STAINED \
| |
| I (13.] CHECK IF MAIN STEAMLINES SHOULD BE ISOLATED:
| |
| a) Annunciator 2D-E3 - a) GO TO Step H.
| |
| LIT b) Verify MSTVs and b) Manually close valves, bypass valves - CLOSED
| |
| [14.] CHECK CONTAINMENT PRESSURE: j 1
| |
| a) Containment pressure - a) Perform the following: l HAS REMAINED LESS THAN l 28 PSIA ON P-LM-110B 1) Manually actuate BOTH trains of quench spray. g
| |
| ~
| |
| . 2) Initiate Attachment 1,.
| |
| : 3) Verify CCW pumps l tripped.
| |
| : 4) Stop all RCPs.
| |
| i l
| |
| b) Containment pressure - b) Verify MSTVs and bypass HAS REMAINED LESS THAN valves closed, IF~ NOT, i 17.8 PSIA ON P-LM-110B THEN manually close valves. l l
| |
| )
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist: t
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0.. REACTOR TRIP OR SAFE 1 INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F]. l
| |
| ' j OR
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [5 3.
| |
| CPJutGING/SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump rectre. path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIG (1575 PSIG).
| |
| Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE i INCREASES TO 2000 PSIG. {
| |
| 4 l
| |
| _ RED PATH
| |
| | |
| ==SUMMARY==
| |
| l a) b) SU3 CRITICALITY - Nuclear power greater than 5 CORE COOLING - Core Exit TCs greater than 1200*F i QR, Core Exit46 less than TCs greater than 700*F AND RVLIS full range c) with no RCPs running HEAT SINK - Narrow Range level in all SGs less than 10% (32%] AND d) total feedwater flow less than 340 gpm INTEGRITY - Cold leg temperature decrease greater than 100*F in last e)
| |
| CONTAINMENT - Containment pressure greater than 60 PS
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| | |
| N3 97887210 NUMBER PROCEDURE TITLE REVISION 1.00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE l
| |
| 8 of 16
| |
| - JTEP ACTION / EXPECTED RESPONSE RESPONSENOTORTAINED
| |
| -l l
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| i
| |
| : 15. VERIFY AFW FLOW:
| |
| a) AW flow - INDICATED a) Manually align AFW valves
| |
| * FI- N-100A OR.
| |
| * FI- W-100B
| |
| * FI- N-100C Manually start pumps.
| |
| b) Verify total flow - b) F SG NR level greater GREATER THAN 340 GPM than 10% [32%] in any SG, THEN control feed flow to j maintain NR level.. :
| |
| J IF NR level less than 10%
| |
| ' T32%) in all SGs, AND IF A W flow greater than 340 1 gpm can NOT be established, THEN GO TO 1-FRP-H.1, j l
| |
| i LOSS OF SECONDARY HEAT SINK, Step 1,.
| |
| : 16. DIRECT UNIT 2 OPERATOR TO INITIATE 1-AP-47, UNIT -
| |
| OPERATION WITH OTHER UNIT EMERGENCY f
| |
| l.
| |
| l
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, j AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25*F [70*F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0,, REACTOR TRIP OR SAFETY i
| |
| INJECTION, Step 1, if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exic TCs - LESS THAN 30*F [80*F].
| |
| E
| |
| * PR2R level - CANNOT BE MAINTAINED GREATER THAN 15: [50 ].
| |
| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE DECREASES TO LESS THAN 1275 PSIG [1575 PSIG).
| |
| l b) Open the Charging /SI pump recire, path MOVs if - RCS PRESSURE i INCREASES TO 2000 PSIC.
| |
| 1
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5 b) CORE COOLING - Core Exit TCs greater than 1200*F l
| |
| E Core Exit TCs greater than 700*F AND RVLIS full range less than 46: with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10: [32 ] AND
| |
| . total feedvater flev less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last !
| |
| 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| l
| |
| | |
| D.97SS7210 PROCEDURE TITLE REVISION NUCBER 1.00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 9 of 16 I 1
| |
| - STEP' ACTIONNXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 17. CHECK RCS TEMPERATURE - IF temperature less than 547'F, STABLE AT OR TRENDING TO THEN stop dumping steam.
| |
| 547'F IF cooldown continues, THEN: J a) Adjust total AFW flow to -
| |
| 340 gpm until at least one ,
| |
| SG NR level is greater than j 10% [32%].
| |
| i b) IF,cooldown continues THEN close MSTVs and bypass. >
| |
| valves.
| |
| IF, temperature greater than 547'F and increasing, THEN:
| |
| Dump steam to the condenser pR, Dump steam using SG PORVs.
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| : 2. SI REINITIATION' CRITERIA l
| |
| Manually initiate both trains of SI AND GO TO.EP-0,. REACTOR TRIP OR SA7ET INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| : g. .
| |
| * PRER level - CANNOT BE MAINTAINED GREATER THAN 15:
| |
| (50%].
| |
| 3.
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump recire, path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE ,
| |
| INCREASES TO 2000 PSIG. !
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) b) SU~ CRITICALITY - Nuclear power greater than 5% '
| |
| CORE COOLING - Core Exit TCs' greater than 1200*F E
| |
| Core Exit TCs greater than 700*F AND RVLIS full range c) less than 46% with no RC?s running HEAT SINR - Narrow Range level in all SGs less than 10% (32:] AND d) total feedwater flow less than 340 gym INTEGRITY - Cold leg temperature decrease greater than 100*F in last e)
| |
| CONTAINMENT - Containment pressure greater than 60 PSI
| |
| : 3. ECST LEVEL CRITERIA Make-up to the ECST frem the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40".
| |
| 1 l
| |
| l
| |
| | |
| M9.97887213 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 10 of 16
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSE NOT08T/.INED
| |
| : 18. CHECK PRZR PORVs AND SPRAY VALVES:
| |
| a) Master controller, a) Take manual control of PC-1-444J - NOT FAILED PC-1-444J..
| |
| b) PORVs - CLOSED b) H PRZR pressure less than 2335 psig, THEN manually
| |
| * PCV-1455C close PORVs.
| |
| * PCV-1456
| |
| _IF any PORV can NOT be !
| |
| closed, THEN manually, close its block MOV. .
| |
| IF the block MOV can NOT be H osed, THEN GO TO 1-EP-1, LOSS OF REACTOR'OR SECONDARY COOLANT, Step 1 c) PRZR spray valves - CLOSED c) g PRZR pressure less than 2260 psig THEN manually
| |
| * PCV-1455A close valves.
| |
| * PCV-1455B IF valves can NOT be H osed, THEN stop RCP(s) <
| |
| supplying failed spray valve (s):
| |
| "A" RCP - PCV-1455A "C" RCP - PCV-1455B l
| |
| l I
| |
| | |
| i l
| |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0,. REACTOR TRIP OR SAFETY INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER(50 THAN
| |
| ]. 15 3.
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| Open the Charging /SI pump recire, path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG.
| |
| : 4. _ RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) b)
| |
| SUBCRITICALITY - Nuclear power greater than 5 CORE COOLING - Core Exit TCs greater than 1200*F E
| |
| Core Exit TCs greater than 700*F AND RVLIS full range c) less than 46 with no RCPs running HEAT SINK - Narrow Range level in all SGs less than 10% (32 ] AND d) total feedvater flow less than-340 gpm _
| |
| INTEGRITY - Cold leg temperature decrease greater than 100*F in last e)
| |
| CONTAINMENT - Containment pressure greater than 60 PSI
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| u___.___. __.. ....._.___.._m_ -
| |
| _ _ _ ._ _ _ _ . _ _ ___--m-.__. __.
| |
| | |
| A 97ss7213 PROCEDURE TITLE REVISION NUMBER
| |
| .00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 11 of 16
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOT08TAINED l
| |
| CAUTION: To prevent Emergency Diesel Generator overload, the #1 or #4 PR2R heater banks should not be re-energized until the ]
| |
| respective Emergency Diesel Generator load is less than 2750 i KW.
| |
| NOTE: Loss of seal injection flow can cause RCP seal degradation.
| |
| : 19. CHECK IF RCPs SHOULD BE STOPPED:
| |
| a) Charging /SI pumps - a) GO TO Step 2J,.
| |
| AT LEAST ONE RUNNING l
| |
| b) RCS subcooling based on b) GO TO Step 21.
| |
| Core Exit TCs - LESS THAN 25'F [70*F]
| |
| c) Stop all RCPs 1 CAUTION: Charging /SI pump recire valve must be open at 2000 psig RCS pressure to prevent possible pump damage.
| |
| : 20. CHECK IF CHARGING /SI PUMP RECIRC MOV(s)
| |
| SHOULD BE CLOSED: ;
| |
| l a) RCS pressure - a) GO TO Step 21,.
| |
| LESS THAN 1275 (1575 PSIG] ,
| |
| b) Close Charging /SI pump recire MOV(s):
| |
| * MOV-1275A 1-CH-P-1A
| |
| * MOV-1275B 1-CH-P-1B
| |
| * MOV-1275C 1-CH-P-1C l
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING.
| |
| AND
| |
| * RCS subcooling based on Core Exic TCs - LESS THAN 25'T [70'F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0,. REACTOR TRIP OR SAFETY -
| |
| l INJECTION, Step 1 if either condition listed below occurs:
| |
| l
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30'F (80*F]. i 3
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%].
| |
| 3.
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump recire, path MOVs if - RCS PRESSURE b) DECREASES TO LESS TEAN 1275 PSIG (1575 PSIG).
| |
| Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIC.
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) b)
| |
| SUBCRITICALITY - Nuclear power greater than 5" CORE COOLING - Core Exit TCs greater than 1200*F E
| |
| Core Exit TCs greater than 700*F AND RVLIS full range c) less than 46 with no RCPs running HEAT SINK - Narrow Range level in all SGs less than 10" [32"] AND d) total feedwater flow less chan 340 gpm INTECRITY - Cold leg temperature decrease greater than 100*F in last e)
| |
| C0KIAINMENT - Containment pressure greater than 60 PSIA 6
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| | |
| N 97887210-NUMBER. PROCEDURE TITLE REVISION 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 12 of 16
| |
| - STEP -
| |
| ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 21. VERIFY NOTIFICATIONS:
| |
| a) STA - NOTIFIED I
| |
| b) EPIP - INITIATED ,
| |
| : 22. CHECK THAT SGs ARE NOT FAULTED: GO TO 1-EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1
| |
| * NO SG PRESSURE DECREASING IN AN UNCONTROLLED MANNER
| |
| * NO SG COMPLETELY DEPRESSURIZED
| |
| : 23. CHECK THAT SG TUBES ARE NOT GO TO 1-EP-3, STEAM GENERATOR RUPTURED: TUBE RUPTURE, Step 1.
| |
| * Condenser air ejector radiation - NORMAL
| |
| * SG main steamline radiation - NORMAL
| |
| * SG blowdown radiation -
| |
| NORMAL
| |
| : 24. CHECK THAT RCS IS INTACT: GO TO 1-EP-1, LOSS OF REACTOR OR SECONDARY COOLANT, Step 1.
| |
| * Containment radiation -
| |
| NORMAL
| |
| * Containment pressure - 1 NORMAL
| |
| * Containment Recirculation Spray sump level - NORMAL t
| |
| i
| |
| | |
| l FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, I AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| 3 2. SI REINITIATION CRITERIA i
| |
| l Manually initiate both trains of SI AND GO TO EP-0, REACTOR TRIP OR SAFETY INJECTION, Step 1 if either condition listed below occurs':
| |
| 1
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| ,O,,R,
| |
| * PR2R level - CANNOT BE MAINTAINED GREATER THAN 15" [50'].
| |
| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA I 1 a) Isolate the Charging /SI pump recire, path MOVs if - RCS PRESSURE l :
| |
| ! b) DECREASES TO LESS TEAN 1275 PSIG [1575 PSIG]. L Open the Charging /SI pump recire, path MOVs if - RCS PRESSURE '
| |
| INCREASES TO 2000 PSIG.
| |
| i
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5" i b) CORE COOLING - Core Exit TCs greater than 1200*r (
| |
| 0,,R Core Exit TCs greater than 700*F AND RVLIS full range c) less than 46% with no RCPs running HEAT SINR - Narrow Range level in all SGs less than 10" [32"] AND d) total feedwater flow less than 340 gpm INTEGRITY - Cold leg temperature decrease greater than 100*F in last e) 60 minutes AND RCS cold leg temperature less than 285'F CONTAINMENT - Containment pressure grerter than 60 PSIA
| |
| : 5. ECST LEVEL CRITERIA .
| |
| Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or ??) when ECST level decreases to 40%.
| |
| C_____________________________________________________ _ . _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _
| |
| | |
| No. 97887210 PROCEDURE TITLE REVISION NUMBER 1-EP-0 REACTOR TRIP OR SAFETY. INJECTION PAGE N 13 of 16 1
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 25. CHECK IF SI CAN BE TERMINATED:
| |
| a) RCS subcooling based on a) DO NOT STOP S1 PUMPS.
| |
| Core Exit TCs - GREATER GO TO Step E.
| |
| TRAN'30'F.
| |
| f b) Secondary heat sink: b) IFF neither condition satisfied, THEN DO NOT STOP Total AFW flow to SGs - SI PUMPS GO TO Step E.
| |
| GREATER THAN 340 GPM E
| |
| l At least one SG Narrow Range level - GREATER THAN 10%
| |
| l c) RCS pressure - STABLE c) DO NOT STOP SI PUMPS.
| |
| OR INCREASING GO TO Step 27_. j l
| |
| PRZR level - GREATER d) DO NOT STOP SI PUMPS. j' d)
| |
| THAN 15% Try to stabilize RCS l pressure with normal PRZR spray. Return To Step 25a.
| |
| : 26. GO TO 1-ES-1.1, SI TERMINATION, Step 1
| |
| : 27. INITIATE MONITORING.0F Monitor CSF Status Trees with CRITICAL SAFETY FUNCTION F-Series procedures.
| |
| STATUS TREES ON SPDS CAUTION: Alternate water sources (CST or 1-AP-22.7) to prevent loss of AFW pump suction pressure will be necessary if ECST decreases to 40%.
| |
| 2 8 . ,_, CHECK SG LEVELS:
| |
| a) Narrow Range level - a) Maintain total feed flow GREATER THAN 10% greater than 340 gpm until NR level is greater than 10%.
| |
| (STEP 28 CONTINUED ON NEXT PAGE) .
| |
| | |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, AND I
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F]. i
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0, REACTOR TRIP OR SAFETY j INJECTION, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%].
| |
| : 3. CHARGING /SI PUMP RECIRC PATH CRITERIA a) Isolate the Charging /SI pump recire, path MOVs if - RCS PRESSURE DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| b) Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE i INCREASES TO,2000 PSIG. i
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| j a) SUBCRITICALITY - Nuclear power greater than 5". j b) CORE COOLING - Core Exit TCs greater than 1200*F i E
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46: with no RCPs running ;
| |
| c) HEAT SINK - Narrow Range level in all SGs less than 10% [32%] AND total feedwater flev less than 340 gpm 3 d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last j 60 minutes AND RCS cold leg temperature less than 285'F :
| |
| e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| )
| |
| | |
| D.97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-0 itEACTOR TRIP OR SAFETY INJECTION PAGE 14 of 16 1
| |
| - STEP . ACTION / EXPECTED RESPONSE RESPONSE NOT087AINED
| |
| : 28. CHECK SG LEVELS: (CONTINUED) b) Control feed flow to b) IJF, NR level in any SG con-maintain Narrow Range tinues to increase in an ]
| |
| level between 10% uncontrolled manner, THEN ,
| |
| and 50% GO TO 1-EP-3, STEAM GENER-ATOR TUBE RUPTURE, Step 1.
| |
| J
| |
| : 29. _ RESET SI
| |
| : 30. RESET CONTAINMENT ISOLATION' PHASE A AND PHASE B
| |
| : 31. VERIFY CONTAINMENT Manually open valves. i INSTRUMENT AIR VALVES - OPEN j
| |
| * TV-IA-102A
| |
| * TV-IA-102B
| |
| : 32. CHECK SECONDARY RADIATION:
| |
| a) Reset SI and align systems for sampling as required b) Secondary radiation- b) GO TO 1-EP-3, NORMAL STEAM GENERATOR TUBE RUPTURE, Step 1.
| |
| * Condenser air ejector radiation - NORMAL
| |
| * SG main steamline radiation - NORMAL
| |
| * SG blowdown radiation -
| |
| NORMAL
| |
| | |
| l FOLDOUT FOR EP-0 AND ES-0 PROCEDURES i
| |
| : 1. RCP TRIP CRITERIA l
| |
| i Trip all RCPs if BOTH conditions listed below exist '
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, (
| |
| . 1 AND
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| : 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0, REACTOR TRIP OR SAFETY INJECTION, Step 1, if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F-[80*F].
| |
| SE
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN IS: [50:].
| |
| 3.
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA a)
| |
| Isolate the Charging /SI pump recire. path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIG (1575 PSIG].
| |
| Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG.
| |
| ! 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5 b) CORE COOLING - Core Exit TCs greater than 1200*F SE Core Exit TCs greater than 700*F AND RVLIS full range !
| |
| c) less than 46 with no RCPs running HEAT SINK - Narrow Range level in all SGs less than 10: [32 ] AND d) total feedwater flow less than 340 gpm INTEGRITT - Cold leg temperature decrease greater than 100*F in last e) 60 minutes AND RCS cold leg temperature less than 285'T CONTAINMENT - Containment pressure greater than 60 PSIA f
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| 1 i
| |
| l
| |
| | |
| we. ores 7:to NUMBER PROCEDURE TITI.E REVISION 1.00 1-EP-0 REACTOR TRIE OR SAFZTY INJECTION PAGE 15 of 16 l
| |
| - STEP ACTION / EXPECTED RESPONSE - RESPONSENOTOBTAINED 4
| |
| : 33. CHECK FOR OUTSIDE CONTAINMENT Evaluate cause of abnormal INVENTORY LOSS: conditions.
| |
| IF cause. is a loss of RCS i a) Safeguard sump Eventory outside containment, level annunciators - THEN GO TO 1-ECA-1.2, LOCA NOT LIT CUTSIDE' CONTAINMENT, Step J_.
| |
| * "E" Panel F-8
| |
| * "A" Panel C-1 )
| |
| . b) Safeguards radiation - !
| |
| ! NORMAL l
| |
| : 34. CHECK PRT CONDITIONS - NORMAL Evaluate abnormal conditions as , ,
| |
| a possible-source of RCS I inventory loss:
| |
| * Letdown relief valve l
| |
| * Sealvater return relief valve l
| |
| * Excess letdown relief valve I when aligned to VCT
| |
| * RER relief valve (Intersystem LOCA).
| |
| i l
| |
| _ _ . _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______._________..______.__-__s
| |
| | |
| I l
| |
| FOLDOUT FOR EP-0 AND ES-0 PROCEDURES 1
| |
| 1
| |
| : 1. RCP TRIP CRITERIA Trip all RCPs if BOTH conditions listed below exist:
| |
| * Charging /SI pumps - AT LEAST ONE RUNNING, k
| |
| AND j
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 25'F [70*F].
| |
| l 2. SI REINITIATION CRITERIA Manually initiate both trains of SI AND GO TO EP-0 . REACTOR TRIP OR SAFE INJECTION, Step 1 if either condition" listed below occurs: *
| |
| * RCS subcooling based on Core Exte TCs - LESS THAN 30*F [80*F].
| |
| O_3
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15" [50:
| |
| 3.
| |
| CHARGING /SI PUMP RECIRC PATH CRITERIA I
| |
| a)
| |
| Isolate the Charging /SI pump recire, path MOVs if - RCS PRESSURE b) DECREASES TO LESS THAN 1275 PSIC (1575 PSIG).
| |
| Open the Charging /SI pump recire. path MOVs if - RCS PRESSURE INCREASES TO 2000 PSIG.
| |
| : 4. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) b) SU3 CRITICALITY - Nuclear power greater than 5%
| |
| CORE COOLING - Core Ex1e TCs greater than 1200*F EE Core Exit TCs greater than 700*F AND RVLIS full range
| |
| ! c) less than 46: with no RC?s running
| |
| ! HEAT SINK - Narrow Range level in all SGs less than 10" (32%] AND d) total feedwater flow less than 340 gpm INTEGRITY - Cold leg temperature decrease greater than 100*F in last e)
| |
| CONTAINMENT - Containment pressure greater than 60 PS
| |
| : 5. ECST LEVEL CRITERIA i
| |
| Make-up to the ECST from the CST or provide alternate sources for AFW j suction (SW or FP) when ECST level decreases to 40".
| |
| i i
| |
| 1
| |
| | |
| ' No',97887210 NUMBER PROCEDURE TITI.E REVISION
| |
| .00 1-EP-0 REACTOR TRIP OR SAFETY INJECTION PAGE 16 of 16 STEP ACTION / EXPECTED RESPONSE Rett 0NSE NOTOBTAINED
| |
| -o**************************************
| |
| CAUTION: If RCS pressure decreases to less than 250 psig, then the LHSI pumps i must be manually restarted to supply water to the RCS.
| |
| 1 o************************************** .
| |
| l
| |
| : 35. CHECK IF LHSI PUMPS SHOULD BE STQPPED:
| |
| a) Check RCS pressure:
| |
| : 1) Pressure - GREATER 1) GO TO 1-EP-1, LOSS OF THAN 250 PSIG REACTOR OR SECONDARY COOLANT, Step 1_.
| |
| : 2) Pressure - STABLE OR ~ 2) GO TO Step 36,.
| |
| INCREASING j
| |
| b) Stop LHSI pumps and place in Auto-Standby l
| |
| : 36. CHECK IF DIESEL GENERATORS ]
| |
| SHOULD BE STOPPED:
| |
| a) Verify AC emergency busses - a) Initiate 1-AP-10.1 to ENERGIZED BY OFFSITE POWER restore offsite power.
| |
| b) Stop any unloaded diesel
| |
| ! generator as per 1-0P-6.1 and/or 1-OP-6.2
| |
| : 37. RETURN TO STEP E END 4
| |
| | |
| N3.97887220 I
| |
| NUMBER ATTACHMENT TITLE REVISION 1-EP-0 1.00 A TTACHMENT PAGE PHASE B ISOLATION 1
| |
| 1 of 4 ,
| |
| i
| |
| : 1. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS ON "H" j
| |
| SAFEGUAES PANEL:
| |
| a) Verify Phase "B" isolation valves - CLOSED (located on the lower-right corner of the panel)
| |
| TRIP VALVES (TV) - CLOSED (GREEN)
| |
| CC- CC- CC- CC-105A 102A 104A-1 101A CC- CC- CC- CC-105B 102C 104B-1 103A ,
| |
| . 1 CC- CC- CC- IA- )
| |
| 105C 102E 104C-1 102A l b) Verify Quench Spray - ALIGNED AND RUNNING RUNNING (RED) OPEN (RED) 1-QS-P-1A ,___ MOV-QS101A MOV-QS100A i c) Verify SW - ISOLATED TO CCW HEAT EXCHANGERS CLOSED (GREEN)
| |
| MOV-SW-108A l
| |
| d) Verify SW - ALIGNED TO RS HIs l
| |
| OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED) l MOV-SW-103A MOV-SW-103D MOV-SW-104A MOV-SW-104D MOV-SW-101A MOV-SW-101C MOV-SW-105A MOV-SW-105C e) Verify Recire Spray - ALIGNED AND RUNNING (Note time delays)
| |
| RUNNING (RED) RUNNING (RED) OPEN (RED)
| |
| MOV-RS156A 1-RS-P-1A (3\ min. T.D.) 1-RS-P-2A (3h min. T.D.) MOV-RS155A '
| |
| - - - _ _ _ __a
| |
| | |
| No.97887220 NUMBER A TTACHMENT TITLE REVISION 1-EP-0 1.00 A TTACHMENT PAGE PHASE B ISCLATION 1 2 of 4
| |
| : 2. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS ON "J" SAFEGUARDS PANEL:-
| |
| a) Verify Phase "B" isolation. valves . CLOSED (located on the lower right corner of the panel)'
| |
| TRIP VALVES (TV) - CLOSED (GREEN)
| |
| CC- CC-: CC- CC-100A 1023 104A-2 101B CC- CC- CC- CC-100B 102D 104R 4. 103B CC- CC- CC- IA-100C 102F _ 104C-2 102B b) Verify Quench Spray - ALIGNED AND RUNNINGt RUNNING (RED) OPEN (AED) 1-QS-P-1B MOV-QS101B MOV-QS100B c) Verify SW - ISOLATED TO CCW HEAT EXCHANGERS CLOSED (GREEN)
| |
| MOV-SW-108B d) Verify SW - ALIGNED TO RS IUts OPEN (RED) OPEN (RED) OPEN (RED) OPEN (RED)
| |
| MOV-SW-103B MOV-SW-103C MOV-SW-104B MOV-SW-104C MOV-SW-101B MOV 37-101D MOV-SW-105B. MOV-SW-105D e) - Verify Recirc Spray - ALIGNED 43 RUNNING (Note time delays)
| |
| RUNNING (RED) RUNNING (RED) OPEN (RED).
| |
| _ MOV-RS156B 1-RS-P-1B (3k min. T.D.) _ 1-RS-P-23 (3b min. T.D.-) MOV-RS1553
| |
| | |
| i +, '.
| |
| Mo.97887880 9 <
| |
| w_
| |
| NUMBER ' A TTalCHMENT177LE REVISION 1-EP-0 s1.00 A TTACHMENT PAGE PHASE B ISOLATION 1 , ;- 3 of 4
| |
| . 2..
| |
| 6
| |
| : 3. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS ON THE UNIT 1 VENTILATION PANEL.
| |
| a) Verify all Containment Air Recire Fans TRIPPED (AMBER) OR OFF (GREEN),
| |
| 1 1-HV-F-1A 1-HV-F-1C _ _ .J-HV-F-1B 1-HV-F-1C b) Verify all shroud cooling fans OFF (GR EN 1-HV-F-37A 1-HV-F-37B 1-HV-F-37C 1-HV-F 37D _-HV-F-37E 1-HV-F-37F c) Verify Air Clrs E=er Supply Valves - 30 SED (GREEN)
| |
| CLOSED (GREEN) CLOSED (GREEN)
| |
| SW-TV-101A-1 SW-TV-101A-2 SW-TV-101B-1 SW-TV-101B-2 ,
| |
| d) Verify that the following filters - DIVERTED TO THE 10 DINE FILTEit FILTER (RED) FILTEP.,,(RED)
| |
| A0D-HV-107A1,2,3,4 A0D-HV-107B1,2,3,4 A0D-HV-128-1,2,3,4 A0D-HV-128-1,2,3,4
| |
| | |
| N r. 91987223 NUMBER A TTACHMENT TITLE REVISION 1-EP-0 1.00 VERIFICATION OF A TTACHMENT pgggg 3 339gg339y PAGE 1 4 of 4
| |
| : 4. VERIFY OR MANUALLY PERFOPJ4 THE FOLLOWING AUTOMATIC OPERATIONS ON THE UNIT l EDIATION MONITORING PANEL:
| |
| a) Verify the following sample pump's RED " Low Flow" light - NOT LIT (Note : 2 minute Ti=e Delay-)
| |
| NOT LIT NOT LIT NOT LIT NOT LIT 1-SW-P-5 1-SW-P-8 1-SW-P-6 1-SW-P-7
| |
| : 5. R MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS ON THE VERIFY O_R, UNIT 1 CASING COOLING PANEL LOCATED BEHIND THE UNIT 1 BACKBOARDS:
| |
| a) Verify Casing Cooling Pumps - ALIGNED AND RUNNING "H" PANEL (left)
| |
| OPEN (RED) OPEN (RED) RUNNING (RED)
| |
| MOV-RS1013 MOV-RS100A 1-RS-P-3A "J" PANEL (Right)
| |
| OPEN (RED) OPEN (RED) RUNNING (RED)
| |
| MOV-RS101A MOV-RS100B 1-RS-P-3B
| |
| \
| |
| | |
| Mo.97887820 NUMBER A TTACHMENT TITLE REVl50N 1-EP-0 1.00 A TTACHMENT PAGE ISOLATION 2 1 of 4
| |
| : 1. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS ON THE UNIT 1 "H" SAFEGUARDS PANEL:
| |
| l CLOSED (GREEN) OPEN (RED)
| |
| TV-1884C MOV-1867C OPEN (RED) OPEN (RED) CLOSED (GREEN) l TV-1884-A MOV-1867A MOV-SW-121A MOV-SW-122A MOV-SW-123A 0FF (GPEEN) CLOSED (GREEN) 1-CV-P-3A MOV-1380 Trip Valve (TV) - CLOSED (GREEN)
| |
| BD- CV- DA- LM- LM- RM- SI- SS- SS- SS-100A 150A 100A 100C 101A&C 100A 100A 100A 103A 112A BD- CV- DG- LM- VG- RM- MS- SS- SS-100C 150C 100A 100E 100A 100B 109A 101A 104A BD SI- LM- LM- MS- TV- TV- SS- SS- TV-100E 101 100A 100G 110A 1204A 1859 102A 106A 1519A SV- SV-102-1 102-2
| |
| | |
| N3.9783722)
| |
| . NUMBER A TTACHMENT TITLE REVISION i 1-EP-0 1.00 VERIFICATION OF PHASE A A TTAC# MENT PAGE !
| |
| ISOLATION l
| |
| 2 2 of 4 ;
| |
| l
| |
| : 2. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERAIIONS ON THE l
| |
| . UNIT 1 5" SAFEGUARDS PANEL: 1 CLOSED (GREEN) OPEN (RED) OPEN (RED) OPEN (RED) CLOSED ' (GREEN)
| |
| MOV-1867D SW-MOV-121B SW-MOV-122B SW-MOV-123B TV-1884B MOV-1867B 0FF (GREEN) CLOSED (GREEN) i 1-CV-P-3B MOV-1381 Trip Valves (TV) - CLOSED (GREEN)
| |
| BD- CV- DA- LM- LM- RM- SI- SS- SS- -SS-100B 150B 1003 100D 101B&D 100C 1003 100B 103B 1123 BD- CV- DG- LM- VG- RM- MS- SS- SS-100D 150D 100B 100F 100B 100D 109 1013 ,104B BD- HCV- LM- LM- MS- TV- TV- SS. SS-100F 1936 100B 100H 110 1204B 1842 102) . 106B SV-
| |
| _ 103
| |
| : 3. VERIFY CLOSED OR MANUALLY CLOSE THE FOLLOWING VALVES ON THE POST ACCIDENT -.
| |
| MONITORING (PAE7 PANEL: .
| |
| CLOSED (GREEN) CLOSED (GREEN)
| |
| TV-DA103B TV-DA103A
| |
| -l
| |
| | |
| No.97847220
| |
| * I 1
| |
| NUMBER ATTACHMENT TITLE REVISION )
| |
| 1-EP-0 1.00 l ATTACHMENT PAGE 3 o 2 3 of 4 1
| |
| 1 4
| |
| : 4. VERIFY OR PERFORM THE FOLLOWING AUTOMATIC OPERATIONS LOCATED AT THE BOTTOM Of THE UNIT 1 VENTILATION PANEL:
| |
| l CLOSED (GREEN 1 0FF (GREEN) ,
| |
| CLOSED (GREEN)
| |
| HV-AOD-160-1 1-HV-F-15 HV-AOD-161-1 .,
| |
| l
| |
| : 5. Secure Waste Gas Releases:
| |
| _ a) FCV-GW 101 - CLOSED
| |
| _ b) Inform unaffected unit to secure containment purge or hogging operations.
| |
| : 6. VERIFY OR MANUALLY PERFORM THE FOLLOWING AUTOMATIC OPERATIONS LOCATED AT THE BOTTOM OF THE UNIT 2 VENTILATION PANEL:
| |
| CLOSED (GREEN) CLOSED (GREEN)
| |
| HV-AOD-160-2 HV-AOD-161-2 MOD MOD
| |
| | |
| Nr.97887220 NUMBER A TTACHMENT TITLE REVISION 1-EP-0 1.00 VERIFICATION OF PHASE A ATTACHMENT PA M zgnyg379y 2
| |
| 4 of 4
| |
| : 7. VERIFT OR MANUALLY PERFORM THE FOLLOWING AUTOYATIC OPERATIONS ON THE UNIT 2 SAFEGUARDS PANELS:
| |
| "H" Panel "J" Panel
| |
| , RUNNING (RED) RUNNING (RED) 2-S'4-P- 1 A 2-SW-P-1B Z
| |
| 2-SW-P-4 ON (RED) (5 nin T.D.) ON (RED) (5 min T.D.)
| |
| H Analyzer Heat H., Analyzer Heat 2
| |
| Tracing Train A Tracing Train B
| |
| : 8. VERIFY AUTOMATIC INITIATION OF BOITLED FRESH AIR SUPPLY TO THE CONTROL ROOM:
| |
| PI-HV-1311 PI-HV-2311 (Located behind the Unit 1 Post Accident Monitoring (PAM) Panel).
| |
| : 9. VERIFY OR PLACE HYDROGEN ANALYZER IN SERVICE AS PER 1-OP-63.2.
| |
| : 10. NOTIFY THE STA TO REVIEW 1.97 VARIABLES ON GROUPS 36 AND 37 0F THE SPDS.
| |
| l l
| |
| | |
| ATTACHMENT 6 t
| |
| l l EMERGENCY OPEPATING PROCEDURE l
| |
| EP-3, STEAM GENERATOR TUBE RUPTURE
| |
| | |
| l VIRGINIA POWER
| |
| % e7ees72o i
| |
| <ess cas NORTH ANNA POWER STATION ,
| |
| l EMERGENCY PROCEDURE I
| |
| Procedure Title Revision !
| |
| Number l
| |
| 1.00 j l
| |
| STEAM GENERATOR TUBE RUPTURE Page 1-EP-3 (WITH ONE ATTACHMENT) 1 of 26 l
| |
| Purpose The purpose of this procedure is to provide instructions for identifying a l ruptured SG(s). The procedure includes isolating the ruptured SG(s) followed l j
| |
| by RCS cooldown and depressurization to the point of SI termination.
| |
| User NAPS Operations Personnel Entry Conditions This procedure is entered from: - -aJ
| |
| : 1) 1-EP-0, REACTOR TRIP OR SAFE *Y INJECTION,
| |
| : 2) 1-EP-1, LOSS OF REACTOR OR SECONDARY COOLANT,
| |
| : 3) 1-EP-2, PAULTED STEAM GENERATOR ISOLATION,
| |
| : 4) 1-FRP-H.3, RESPONSE TO STEAM GENERATOP. HIGH LEVEL,
| |
| : 5) 1-ES-1.2 POST LOCA COOLDOWN AND DEPRESSURIZATION, -
| |
| : 6) 1-ES-3.1, POST-SGTR C00LDOWN USING BACKFILL,
| |
| : 7) 1-ES-3.2, POST-SGTR C00LDOWN USING BLOWDOWN,
| |
| : 8) 1-ES-3.3, POST-SGTR C00LDOWN USING STEAM DUMP,
| |
| : 9) 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED,
| |
| : 10) 1-ECA-3.2, SGTR WITH LOSS OF REACTOR COOLANT - SATURATED RECOVERY DESIRED,
| |
| : 11) 1-ECA-3.3, SGTR WITHOUT PRESSURIZER PRESSURE CONTROL, or
| |
| : 12) 1-EP-1 SERIES FOLDOUT PAGE.
| |
| SAEETY TE_A Rsvision Record D
| |
| REV. 1.01 PAGES(S): 2,9,10,11,13,14,17,18,23,att 1 DATE: 06-12-87 REV. PAGES(S): DATE:
| |
| REV. PAGES(S): DATE:
| |
| REV. PAGES(S): DATE:
| |
| REV. PAGES(S) : DATE:
| |
| REY. PAGES(S): DATE:
| |
| Approval Ret.ommended Approved Date l f
| |
| * l Chairman Station NucMar Safety 06m12-87 and Operating Committee .
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO i
| |
| ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs: ,
| |
| 1
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F (80*F].
| |
| I J
| |
| sE!
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%).
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5 b) CORE CCOLING - Core Exit TCs greater than 1200*F 0,,,R Core Exit TCs greater than 700*F AND R7LIS full range less than 46% vich no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10" (32%] AND total feedwater flev less than 340 gym d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F.
| |
| e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LIG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FF) when ECST level decreases to 40%.
| |
| | |
| E 97887210 !
| |
| REVISION PROCEDURE TITLE NUMBER 1.00 1 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 2 of 26 i
| |
| ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP NOTE: Setpoints in brackets [ ] are for adverse containment atmosphere (20 psia containment pressure or 10 R/ER containment radiation.)
| |
| NOTE: Personnel should be available for sampling during this procedure.
| |
| NOTE: Seal injection flow should be maintained to all RCPs.
| |
| : 1. CHECK IF RCPs SHOULD BE STOPPED: ,
| |
| l a) Charging /SI pumps - a) GO TO Step j[.
| |
| AT LEAST ONE RUNNING {
| |
| l b) RCS subcooling based on b) GO TO Step j[. l Core Exit TCs - LESS THAN 25*F [70*F) c) Stop all RCPs j i
| |
| CAUTION: Charging /SI pump recirc valve must be open at 2000 g.41g RCS l pressure to prevent possible pump damage. )
| |
| : 2. CHECK IF CHARGING /SI PUMP RECIRC MOV(s) SHOULD BE CLOSED:
| |
| a) RCS pressure - a) GO TO Step j[.
| |
| LESS THAN 1275 PSIG
| |
| [1575 PSIG]
| |
| l l b) Close Charging /SI pump l recire MOV(s):
| |
| * MOV-1275A 1-CH-P-1A
| |
| * MOV-1275B 1-CH-P-1B
| |
| * MOV-1275C 1-CH-P-1C
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERLi Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT 'SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs: l
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F (80*F].
| |
| E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%). a
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F i S
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running a c) HEAT SINK - Narrow Range level in all SGs less than 10% (32"] AND total feedwater flow less than 34,0 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F e) CONTAINMENT - Contain=ent pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled =anner or has completely depressurized, and has not been isolated, unless needed for RCS cooldewn.
| |
| 4.' COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| | |
| i 00,97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RLTTURE PAGE 3 of 26
| |
| . ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| - STEP -
| |
| : 3. IDENTIFY RUPTURED SG(s): IF, ruptured SG(s) can NOT be immediately identified, THEN Unexpected increase in continue with Steps 6 thru 13.
| |
| SG Narrow Range level WHEN ruptured SG(s) identified, THEN do Steps 4 and 5.
| |
| pR High radiation from any SGBD monitor:
| |
| "A" SG RM-SS-122 "B" SG RM-SS-123 "C" SG RM-SS-124 9.R High radiation from any SG steamline:
| |
| "A" SG RI-MS-170 i "B" SG RI-MS-171 .
| |
| "C" SG RI-MS-172 'I
| |
| ~
| |
| OR l
| |
| l High radiation from any SG sample 1
| |
| i
| |
| | |
| f FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA t
| |
| Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs: !
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| S.R
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%].
| |
| t
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F pR Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10" (32".] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA l 3. SECONDARY INTEGRITY CRITERIA i
| |
| GO TO EF-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has ccepletely depressuri:ed, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 297..
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST frem the CST or provide alternate sources for AFW suction (SW or FF) when ECST level decreases to 40".
| |
| 1 l
| |
| 1
| |
| ~
| |
| l 3
| |
| l
| |
| | |
| XD. 97887210 PROCEDURE TITLE REVISION NUMBER 1.00 l
| |
| \
| |
| 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 4 of 26
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED CAUTION: If the Turbine-Driven AFW pump is the only available source of feed flow, then steam supply to the Turbine-Driven AFW pump l must be maintained from at least one SG.
| |
| CAUTION: To maintain secondary heat sink, at least one SG must be maintained available for RCS cooldown.
| |
| : 4. ISOLATE FLOW FROM RUPTURED -
| |
| SG(s): I a) Verify ruptured SG(s)
| |
| PORV controller setpoint -
| |
| AT 1025 PSIG (Pot setting at 5.3) b) Check ruptured SG(s) b) WHEN ruptured SG(s) ,
| |
| PORVs - CLOSED pressure less than 1025 psig, THEN verify SG PORV(s) closed.
| |
| IF,NOT closed. THEN manually close.
| |
| F PORV(s) canNOT be I_F_
| |
| closed, THEN locally close PORV.
| |
| c) Close ruptured SG(s) steam supply valve to Turbine Driven AFW pump:
| |
| "A" SG 1-MS-18 "B" SG 1-MS-57 l "C" SG 1-MS-95 l
| |
| l d) Check decay heat release e) Manually close valve.
| |
| I valve - CLOSED (Step 4 CONTINUED ON NEXT PAGE)
| |
| | |
| )
| |
| l FOLDOUT FOR EP-3 AND ES-3 PROCEDURES I
| |
| l
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1 SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| O,,R,
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%) .
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| l b) CORE COOLING - Core Exit TCs greater than 1200*F
| |
| _OR Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10% [32%) AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'T e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldovn.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| l S. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECSI level decreases to 40%.
| |
| | |
| 29 97887210 1
| |
| REVISION NUMBER PROCEDURE TITLE 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 5 of 26 RESPONSE NOTOBTAINED
| |
| - STEP ACTION / EXPECTED RESPONSE
| |
| : 4. ISOLATE FLOW FROM RUPTURED SG(s) (CONTINUED):
| |
| e) Verify ruptured SG blow- e) IJ[ NOT open for sampling down TVs - CLOSED procedure, THEN manually close ruptured SG blow-down trip valves.
| |
| f) Close ruptured SG(s) f) Close ruptured'SG(s) NRV(s)
| |
| MSIV(s) and bypass and bypass NRV(s). IF I valve (s) NRV(s) can NOT be closed, j THEN: i
| |
| : 1) Close intact SG(s) ':
| |
| MSTVs and' bypass !
| |
| valves (s).
| |
| : 2) Use intact SG(s) PORVs ;
| |
| for steam dump. l IJ[ any ruptured SG can NOT be isolated from at least one intact SG, THEN GO TO I
| |
| l-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT -
| |
| SUBC00 LED RECOVERY DESIRED, Step 1,.
| |
| oo**************** *********************
| |
| s CAUTION: If any ruptured SG is faulted, then feed flow to that SG should remain isolated during subsequent recovery actions unless needed for RCS cooldown.
| |
| oo**************** *********************
| |
| So CHECK RUPTURED SG(s) LEVEL:
| |
| a) Narrow Range level - a) Maintain feed flow until GREATER THAN 10% [32%] ruptured SG NR level greater than 10% [32%).
| |
| b) Control feed flow to maintain Narrow Range level between 10%
| |
| [32%) and 50%
| |
| | |
| L FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REI" "IATION CRITERIA Manually operate Charging /SI pumps and align BIT.as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, '
| |
| Step i if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F'[80*F].
| |
| _OR .-
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%].
| |
| [
| |
| 2.' _.
| |
| RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SU3 CRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs. greater than 1200*F PR, Core Exit TCs greater than 700*F AND RVLIS full range less than 46% vith no RCFs running !
| |
| c) HEAT SINK - Narrow Range level in all SGs less than.10% [32!).AND.
| |
| total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in .last 60 minutes AND RCS cold leg temperature less than 285*F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step p if any-SG pressure is decreasing in an uncontrolled manner or has ec=pletely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST. level decreases to less than 29%.
| |
| : 5. ECST CVEL CRITERIA i
| |
| i Make-up to the ECST from the CST or provide alternate sources for AFW
| |
| ~
| |
| suction (SW or FP) when ECST level decreases to 40%.
| |
| -j
| |
| | |
| l Do.97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 6 of 26 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP CAUTION: If any PRZR PORV opens because of high PRZR pressure, then Step.
| |
| 6b should be repeated after RCS pressure decreases to less than 2335 psig.
| |
| : 6. CHECK PRZR PORVs AND BLOCK VALVES:
| |
| a) Power to block valves - a) Restore power to block AVAILABLE MOVs.
| |
| b) PORVs - CLOSED b) IF PRZR pressure less than 2335 psig, THEN manually j
| |
| ~
| |
| * PCV-1455C close PORVs.
| |
| * PCV-1456 IF any valve can ,NOT be closed, THEN manually close its block valve.
| |
| IF block valve can NOT be !
| |
| closed, THEN GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR C001 ANT -
| |
| SUBC00 LED RECOVERY DESIRED, ,
| |
| Step 1.
| |
| I c) PORV block valves - c) Open one block valve unless j AT LEAST CNE OPEN it was closed to isolate an open PORV.
| |
| * MOV-1535
| |
| * MOV-1536
| |
| }
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDUIES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, )
| |
| Step 1 if either condition listed below occurs:
| |
| t
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| 0,,R,,
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%].
| |
| l 2. RED PATH SLWJGY a) SUBCRITICALITY - Nuclear power greater.than 5 b) CORE COOLING - Core Exit TCs greater than 1200*F 0,, R, Core Exit TCs greater than 700*F AND R7LIS full range
| |
| ! less than 46" with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10" [32"] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Contaianent pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA .
| |
| i GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled nanner or has cenpletely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step.1, if RWST level decreases to less than 29".
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40".
| |
| i 1
| |
| 1 k
| |
| l
| |
| --- -_ _]
| |
| | |
| 33.S74e7210 l
| |
| 4 PROCEDURE TITI.E REVISION NUMBER 1.00 l
| |
| 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 7 of 26 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED f STE?
| |
| ]
| |
| l
| |
| : 7. CHECK THAT SGs ARE NOT Verify all faulted SGs I FAULTED: isolated unless required i for RCS cooldown: /
| |
| * NO SG PRESSURE DECREASING IN AN
| |
| * Steamlines UNCONTROLLED MANNER
| |
| * Feedlines
| |
| * NO SG COMPLETELY IF NOT, THEN GO TO DEPRESSURIZED 1-EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1.
| |
| oo**************** *********************
| |
| l CAUTION: Alternate water sources (CST or 1-AP-22.7) to prevent loss of AFW pump suction pressure will be necessary if'ECST decreases to 40%.
| |
| o o * * * * * * * *-* * * * * * * * *********************
| |
| : 8. CHECK INTACT SG LEVEL:
| |
| a) Narrow Range level - a) Maintain total feed flow GREATER THAN 10% [32%] greater than 340 gpm until NR level in at least one SG is greater than 10% (32%].
| |
| b) Control feed flow to b) F NR level in any intact IF
| |
| ' maintain Narrow Range SG continues to increase level between 10% (32%) in an uncontrolled manner, and 50% THEN return to Step 1.
| |
| l
| |
| : 9. RESET SI
| |
| : 10. RESET CONTAINMENT ISOLATION PHASE A AND PHASE B
| |
| : 11. VERIFY CONTAINMENT Manually open valves.
| |
| INSTRUMENT AIR VALVES - OPEN
| |
| * TV-IA-102A
| |
| * TV-IA-102B
| |
| /
| |
| | |
| l l
| |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA l
| |
| l Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, l
| |
| l Step 1 if either condition listed below occurs:
| |
| l
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| _OR t
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%).
| |
| I I 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exic TCs greater than 1200*F pR, Core Exit TCs greater than 700*F AND RVLIS full range less than 46: with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10% [32%) AND total feedvater flov less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F l e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA i GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure ;
| |
| is decreasing in an uncontrolled manner or has completely depressurized, !
| |
| and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA l
| |
| Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| i i
| |
| | |
| E.97447210 I PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 8 of 26 1
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED }
| |
| i l
| |
| 12.- VERIFT ALL AC Initiate 1-AP-10.1 to restore BUSSES - ENERGIZED offsite power.
| |
| BY OFFSITE POWER IF, required, THEN manually load the following. equipment on the AC Emergency Busses: ]
| |
| /
| |
| * Shroud Cooling fans. .)
| |
| * Containment /.ir Recire fans.
| |
| * PR2R heaters.
| |
| ****************** ************6**** **** )
| |
| CAUTION: If RCS pr' essure decreases to'less than 250 psig [525 psig],
| |
| then the LESI pumps must be manually restarted to supply water
| |
| -to the RCS.
| |
| : 13. CHECK IF LHSI PUMPS SHOULD BE STOPPED:
| |
| a) RCS pressure - a) GO TO Step M .
| |
| GREATER THAN 250 PSIG [525 PSIG]
| |
| b) Stop LHSI pumps and place in Auto-Standby
| |
| : 14. VERIFY FLOW FROM RUPTURED SG(s) - ISOLATED a) Step 4_ - COMPLETE a) Return to Step 4_, UNLESS ruptured SG recuired for RCS cooldown.
| |
| b) Check ruptured SG(s) b) GO TO 1-ECA-3.1, SGTR WITH pressure - GREATER LOSS OF REACTOR COOLANT -
| |
| THAN 350 PSIC SUBC00 LED RECOVERY DESIRED,.
| |
| Step 1.
| |
| l l
| |
| | |
| 1 l
| |
| 1 FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESI)a Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F (80*F].
| |
| S
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%] .
| |
| i
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% [32%] ffEL total feedwater flow less than 340 gpm d) INTEGRITT - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F !
| |
| e) CONTAIN>ENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2 FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, i and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION Go TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW cr ??) when ECST level decreases to 40%.
| |
| j l
| |
| 1 l
| |
| | |
| h 97887210 REVISION NUMBER PROCEDURE TITLE 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 9 of 26 RESPONSE NOTOBTAINED
| |
| - STEP ACTION / EXPECTED RESPONSE NOTE: Low PRZR pressure SI should be blocked when PRZR pressure is less than 2000 psig.
| |
| NOTE: Low steam line pressure SI should be blocked when PRZR pressure is less than 2000 psig and RCS Tavg is less than 543*F.
| |
| : 15. INITIATE RCS C00LDOWN:
| |
| a) Determine required Core Exit temperature based on closest SG pressure:
| |
| INTACT LOOP RUPTURED SG PRESSURE (psig) CORE EXIT TEMPERATURE (*F) 1100.............. 510 (460]
| |
| 1000.............. 495 [445]
| |
| 900.............. 485 [435]
| |
| 800.............. 470 [420]
| |
| 700.............. 455 (405]
| |
| 600.............. 440 (390]
| |
| l 500.............. 420 [370]
| |
| 400...............400 (350]
| |
| 340.............. 385 (335]
| |
| b) Dump steam to main b) Manually or locally condenser from intact dump steam at maximum SG(s) at maximum rate from SG(s):
| |
| controllable rate Intact SG FORV(s)
| |
| Decay heat release valve IF no intact SG available, TEEN perform the following:
| |
| Use faulted SG S
| |
| GO TO 1-FCA-3.1, SGTR WITH LOSS OF REACTOR COOLANT- SUBC00 LED RECOVERY DESIRED, Step 1.
| |
| (STEP 15 CONTINUED ON NEXT PAGE) u - - _ - - _ - _ ___
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO
| |
| ~
| |
| ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, l
| |
| ! Step 1 if either condition listed below occurs:
| |
| l
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| pR, l
| |
| )
| |
| j
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN.15% [50%].
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F l
| |
| _OR Core Exit TCs greater than 700*F AND RVLIS full range less than 46% vich no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% [32%] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater'chan 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDALT INTEGRITY CRITERIA i GO TO E?-2, FAULTED STEAM GENERATOR ISOLATION, Step-1, if any SG pressure l 1s decreasing in an uncontrolled. manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION j GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| I 4
| |
| e e
| |
| | |
| %.97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE AL7TURE PAGE 10 of 26 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP
| |
| : 15. INITIATE RCS C00LDOWN: (CONTINUZD):
| |
| c) Core Exit TCs - LESS c) Return to Step 15b. '
| |
| l THAN REQUIRED TEMPERATURE I
| |
| d) Stop RCS cooldown and maintain required RCS temperature
| |
| : 16. CHECK RUPTURED SG(s) IF pressure continues to PRESSURE - STABLE OR decrease to less than 100 psi INCREASING above intact SG(s) pressure, THEN GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1.
| |
| : 17. CHECK RCS SUBC00 LING GO TO 1-ECA-3.1, SGTR WITH LOSS BASED ON CORE EXIT TCs - 0F REACTOR COOLANT - SUBC00 LED GREATER THAN 50*F [100*F] RECOVERY DESIRED, Step 1.
| |
| : 18. DEPRESSURIZE RCS TO MINIMIZE BREAKFLOW AND REFILL PRZR:
| |
| a) Normal PRZR spray - a) GO TO Step H. OBSERVE AVAILABLE CAUTIONS AND NOTE PRIOR TO STEP H.
| |
| b) Spray PRZR with maximum i
| |
| available spray l
| |
| (STEP 18 CONTINUED ON NEXT PAGE)
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES 1
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO.
| |
| ECA-3.1, SGTR WITH. LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERT DESIRED, Step i if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*? [80*F]~.
| |
| 0,, R,
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER TRAN 15% [50%]. ]
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5% !
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F j 1
| |
| 0,,R Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% [32%) AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has cenpletely depressurized, and has not been isolated, unless needed for RCS cooldown. ,
| |
| l
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION i
| |
| GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| 1 l
| |
| | |
| .r---- -
| |
| 00,97887210
| |
| %t PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3' STEAM GENERATOR TUBE RUPTURE PAGE 11 of 26 l__
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 18. DEPRESSURIZE RCS TO MINIMIZE BREAKFLOW AND REFILL PR2it: -(CONTINUED) c)2 Verify PRZR pressure c) GO TO Step H. OBSERVE satisfactorily decreasing CAUTIONS AND NOTE PRIOR until ANY of the following TO STEP J9_.
| |
| satisfied:
| |
| PRZR level - GREATER THAN .
| |
| 70%~[65%]
| |
| pR RCS subcooling based on' <
| |
| Core Exit TCs - LESS THAN -
| |
| 30*P (80*P] k
| |
| )
| |
| _OR BOTH of the following:
| |
| : 1) RCS pressure -
| |
| LESS THAN RUPTURED SG(s) PRESSURE AND
| |
| : 2) PRZR level -
| |
| GREATER THAN 15% [50%)
| |
| d) Close spray valve (s):
| |
| : 1) Normal spray valve (s) 1) Stop RCP(s) supplying failed spray valve (s):-
| |
| PCV-1455A "A" RCP
| |
| (
| |
| PCV-1455B "C" RCP
| |
| : 2) Auxiliary spray valve 2) Isolate normal j HCV-1311 letdown and charging.
| |
| e) GO TO Step g. OBSERVE CAUTION PRIOR'I'0 STEP g.
| |
| s ao
| |
| | |
| i i
| |
| l t
| |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| .j
| |
| : 1. SI REINITIATION-CRITERIA l
| |
| Manually operate Charging /SI pumps and align BIT as required and GO.TO~
| |
| ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT .- SUBC00 LED RECOVERY DESIRED,"
| |
| Step 1 if either condition listed below occurs: .{
| |
| I'
| |
| * RCS sube,o. sling based on Core Exit TCs - LESS TRAN.30*F [80*F).
| |
| l.
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER TEAN 15% [50%] .
| |
| : 2. RED PATH SIPotARY !
| |
| a) SU3 CRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F 3R, Core Exit TCs greater than 700*F AND RVLIS full range ,
| |
| less than 46% with no RCFs running. !
| |
| c) HEAT SINK - Narrow Range level in.all SGs less than 10% [32%] AND total feedwater_ flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F '
| |
| e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for'RCS cooldown. (
| |
| l
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level ]
| |
| decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST frem the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%. l l
| |
| l j ..,7,,
| |
| \ _ _i___
| |
| E- -__
| |
| ._______.______________;_______________________ + , , . . .
| |
| | |
| No,97Q87210 l NUMBER PROCEDURE TITLE REVISION l 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 12 of 26 l'
| |
| - STEP ACTIONEXPECTED RESPONSE RESPONSENOTOBTAINED oo**************** *********************
| |
| CAUTION: The PRT may rupture if a PRZR PORV is used for RCS depressurization. This may result in abnormal containment conditions. -
| |
| oo**************** *********************
| |
| CAUTION: To prevent possible valve failure, cycling of PRZR PORVs should be minimized.
| |
| oo**************** *********************
| |
| NOTE: The upper head region may void during RCS depressurization if RCPs are not running. This will result in a rapidly increasing PRZR level.
| |
| : 19. DEPRESSURIZE RCS USING PRZR PORV TO MINIMIZE BREAK FLOW AND REFILL PRZR:
| |
| a) PRZR PORV - AT LEAST a) Establish auxiliary spray ONE AVAILABLE and return to Step 18b.
| |
| E auxiliary spray can NOT be established, THEN GO TO 1-ECA-3.3, SGTR WITHOUT PRES $URIZER PRESSURE CONTROL, Step 1.
| |
| (STEP 19 CONTINUED ON NEIT PAGE) 1
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOS3 OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs: ]
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| 1 p
| |
| _ R_
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%].
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F f.R, Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10% [32"] AND_
| |
| total feedvater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less chan 285'T e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERAT6R ISOLA!!ON, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA l
| |
| Make-up to the ECST from the CST or provide alternate sources for AFW j suction (SW or FP) when ECST level decreases to 40%. ]
| |
| j 1
| |
| i
| |
| | |
| No.97887210 NUMBER PROCEDURE TITLE REVISION l 1.00 .
| |
| 1-EP-3 STEAM GENERATOR TUBE RUPTURE PA GE 13 of 26
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| : 19. DEPRESSURIZE RCS USING PRZR PORV TO MINIMIZE BREAK FLOW AND REFILL PRZR (CONTINUED) :
| |
| b) Open one PRZR PORV until ANY of the following conditions satisfied:
| |
| PRZR level - GREATER THAN 70% [65%]
| |
| 2R RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F]
| |
| pR BOTH of the following:
| |
| : 1) RCS pressure -
| |
| LESS THAN RUPTURED SG(s) PRESSURE AND
| |
| : 2) PRZR level -
| |
| GREATER THAN 15% [50 ]
| |
| c) Close PRZR PORV c) Close PRZR PORV block MOV.
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Cha.ging/SI pumps and align BIT as required and GO TO' .;
| |
| ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT'- SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| .OR !
| |
| 1
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%) .
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| 1 a) SUBCRITICALITY - Nuclear power greater than 5% l b) CORE COOLING - Core Exit TCs greater than 1200*F pR, Core Exit TCs greater than 700*F AND RVLIS fullLrange less than 46 vich no'RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10: (32 ] AND total feedwater flov less than 340 gpm-
| |
| : 4) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 3 60 minutes'AND RCS cold leg temperature less than 285*F l e) CONTAINMENT - Containment pressure greater than 60 PSIA 'l
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure j is decreasing in &n uncontrolled manner or has completely depressurized, )
| |
| and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANS/ER TO COLD LEG RECIRCULATION, Step ~1 if RWST level decreases to less than 29%.
| |
| : 5. ECST LE7EL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| 1 l
| |
| 1
| |
| | |
| No. 97887210 NUMBER PROCEDURE TITLE REVIS N 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 14 of 26 L
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINEO
| |
| : 20. CHECK RCS PRESSURE - Close PRZR PORY block MOV.
| |
| INCREASING IF, pressure continues to decrease, THEN perform the following:
| |
| : 1) Monitor the following for indication of leakage from PRZR PORV:
| |
| * PORV discharge temperature.
| |
| * PRT pressure.
| |
| * PRT temperature.
| |
| : 2) GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR C00LAh7 -
| |
| SUBC00 LED RECOVERY DESIRED, Step 1 CAUTION: SI must be terminated when termination criteria are satisfied to prevent overfill of the ruptured SG(s).
| |
| : 21. CHECK IF SI FLOW SHOULD BE TERMINATF9:
| |
| a) RCS subcooling based on a) DO NOT STOP SI PUMPS. GO Core Exit TCs - GREATER TO 1-ECA-3.1, SGTR WITH TRAN 30*F [80'F] LOSS OF REACTOR COOLANT -
| |
| SUBC00 LED RECOVERY DESIRED, Step 1 (STEP 21 CONTINUED ON NEXT PAGE)
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1. SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| l
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F]. j
| |
| /
| |
| 3
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15" (50").
| |
| l
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5" b) CORE COOLING - Core Exit TCs greater than 1200*F 8
| |
| l l
| |
| Core Exit TCs greater than 700*F AND RVLIS full range I less than 46" with no RCPs running l c) HEAT SINK - Narrow Range level in all SGs less than 10% [32".] AND cotal feedvater flow less than 340 gpm l
| |
| ! d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last l
| |
| 60 minutes g RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has ce=pletely depressuri::ed, and has not been isolated, unless needed for RCS cooldown.
| |
| l 1
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION Go TO ES-1.3, TRANSFER TO COLD LEG RECIRCL1ATICN, Step 1, if RWST level decreases to less than 29I.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST frem the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40".
| |
| | |
| 30.978B7210 PROCEDURE TITLE RE IS N NUMBER 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 15 of 26 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP
| |
| : 21. CHECK IF SI FLOW SHOULD BE TERMINATED (CONTINUED):
| |
| b) Secondary heat sink: b) _IF neither condition is satisfied, THEN DO,NOT STOP Total feed flow to SI PUMPS. GO TO 1-ECA-3.1, SG(s) - GREATER THAN SGTR WITH LOSS OF REACTOR 340 GPM AVAILABLE COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1,.
| |
| O_E Narrow Range level in at least one intact l
| |
| SG - GREATER THAN 10% [32%) l i
| |
| c) RCS pressure - STABLE OR c) DO NOT STOP SI PUMPS. GO INCREASING TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED '
| |
| RECOVERY DESIRED, Step 1,.
| |
| d) PR2R level - GREATER d) DO NOT STOP SI PUMPS.
| |
| THAN 15% (50%] IIe~ turn to Step M.
| |
| : 22. STOP ALL BUT ONE CHARGING /SI PUMP AND PLACE IN STANDBY i
| |
| : 23. VERIFY CHARGING /SI PUMP RECIRC:
| |
| l a) Check MOV-1373 - OPEN a) Manually open valve. f b) Check recire valve for b) Manually open valves.
| |
| Charging /SI pumps - OPEN
| |
| * MOV-1275A 1-CH-P-1A
| |
| * MOV-1275B 1-CH-P-1B {
| |
| 1
| |
| * MOV-1275C 1-CH-P-1C
| |
| | |
| e I
| |
| FOLDOUT FOR EP-3 AND'ES-3 PROCEDURES
| |
| ~l I
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as :eysired and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED 10ZCOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Dcit TCs - LESS THAN 30*F [80'F] .
| |
| S
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15: [50").
| |
| : 2. RED PATH SUMMAPT a) SUBCRITICALITT - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F Corc Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all'SCs less than 10" [32"] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LFG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreas'es to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW.
| |
| suction (SW or FP) when ECST level decreases to-40".
| |
| | |
| f G.97887210 i 1
| |
| 1 REVISION l PROCEDURE TITl.E NUMBER 1.00 i 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 16 of 26 l
| |
| ACTION / EXPECTED RESPCNSE RESPONSENOTOBTAINED
| |
| - STEP I
| |
| I
| |
| : 24. ISOLATE BIT:
| |
| a) Close inlet isolation valves:
| |
| * MOV-1867A
| |
| * MOV-1867B b) Close outlet isolation valves:
| |
| * MOV-1867C
| |
| * MOV-1867D
| |
| : 25. ESTABLISH CHARGING:
| |
| a) Place controll a for FCV-1122 in manual and close FCV-1122 b) Check auxiliary spray, b) Manually close valve.
| |
| BCV-1311 - CLOSED c) Open charging line isolation valves:
| |
| * MOV-1289A
| |
| * HCV-1310
| |
| * MOV-1289B d) Open charging flow control valve, FCV-1122, to establish 25 gpm charging flow e) Adjust seal injection as required
| |
| : 26. CONTROL CHARGING FLOW TO MAINTAI.4 PR2R LEVEL
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR C00LANT'- SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| E
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%) .
| |
| : 2. RED PATH SUMMARi a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F S
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46I with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% [32%] AND total feedwater flow less than 340 gpm l d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has cenpletely depressurized, ;
| |
| and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION j GO TO ES-1.3, TRANSFER TO COLD LEG REC!RCULAIICN, Step 1, if RWST level l decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA-Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| l l
| |
| \
| |
| | |
| 30.97887210 PROCEDURE TITI.E REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 17 of 26 j
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 27. VERIFY SI FLOW NOT REQUIRED:
| |
| a) RCS subcooling based a) Manually operate on Core Exit TCs - Charging /SI pumps and align GREATER THAN 30*F [80*F] BIT as necessary. GO TO l-ECA-3.1, SGTR WITH LOSS OF REACTOR' COOLANT -
| |
| SUBC00 LED RECOVERY DESIRED, j Step 1,. 'j b) PRZR level - GREATER b) Increase charging flow- *!
| |
| THAN 15% (50%] and operate Charging /SI 1 pumps as necessary.
| |
| IF level canNOT be
| |
| ~
| |
| E intained, THEN manually operate Charging /SI pumps and align BIT as necessary.
| |
| GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT -
| |
| SUBC00 LED RECOVERY DESIRED, Step 1,.
| |
| : 28. CHECK VCT MAKEUP CONTROL Adjust controls as necessary.
| |
| SYSTEM:
| |
| a) Makeup set at cold shutdown concentration b) Makeup set for automatic operation
| |
| | |
| FOLD 00T FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERT DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*7 [80*F].
| |
| E l
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15: [50%).
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greate chan 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F 3
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% [32"] AND !
| |
| total feedvater flov less than 340 gpm i d) ' INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2. FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has conpletely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| l
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step ~1, if RWST level decreases to less than 29" I
| |
| S. ECST LEVEL CRITERIA l Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| l l
| |
| l-1
| |
| | |
| N.97887210.
| |
| NUMBER PROCEDURE TITLE REVISION 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 18 of 26
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTO8TAINED \
| |
| l 1
| |
| : 29. CHECK IF LETDOWN CAN BE ESTABLISHED:
| |
| a) PRZR level - GREATER a) Continue with Step 30_. .
| |
| THAN 30% [55%] WHEN PRZR level increases ,
| |
| to greater than 30% [55%],
| |
| THEN perform Step 29b.
| |
| b) Establish letdown: b) Establish excess letdown as per 1-OP-8,5.
| |
| : 1) Place PCV-1145 in '
| |
| manual and open to 50%
| |
| : 2) Open the following:
| |
| * TV-1204A i
| |
| * TV-1204B l
| |
| .
| |
| * LCV-1460A
| |
| * LCV-1460B i
| |
| : 3) Open HCV-1200A, 1200B, or 1200C j
| |
| : 4) Adjust PCV-1145 to obtain 300 psig letdewn pressure and place PCV-1145 in auto '
| |
| -l
| |
| : 5) Place additional orifice '
| |
| valves in service as required i i
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA i
| |
| Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| OR l
| |
| i d
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%] .
| |
| : 2. RED PATH SU)LS.ARY a) SU3 CRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F 9.E Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% (32 ] AND total feedwater flov less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg te=perature less than 285'F e) CONTAINMENT - Contain=ent pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has ccupletely depressurized, and has not been isolated, unless needed for RCS cocidown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION j I
| |
| GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA 1
| |
| l Make-up to the ECST from the CST or provide alternate sources for AFW I suction (SW or FP) when ECST level decreases to 40%.
| |
| I l
| |
| | |
| 00.97887210 PROCEDURE TITLE REVISION NU4' id 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PA GE 19 of 26 w
| |
| ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP -
| |
| : 30. ALIGN CHARGING /SI PLHP SUCTION TO VCT:
| |
| l a) Open VCT Suetion valves:
| |
| * MOV-1115C )
| |
| * MOV-1115E -l b) Close RWST Suction valves: 1
| |
| * MOV-1115B
| |
| * MOV-1115D CAUTION: To prevent Emergency Diesel Generator overload, the #1 or #4 PRZR heaters should not be re-energized until the respeccive Emergency Diesel Generator load is less than 2750 KW.
| |
| 1
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /S'I pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1,if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F (80*F].
| |
| _O R_
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15'. [50%).
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SU3 CRITICALITY - Nuclear power greater than 5" l b) CORE COOLING - Core Exit TCs greater than 1200*F 0,R, Core Exit TCs greater than 700*F AND RVLIS full range less than 46I with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10" [32"] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA ,
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GE'ERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an unecuerolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCTLATICN, Step 1, if RWST level decreases to less than 29". I i
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40I.
| |
| l l
| |
| l !
| |
| | |
| l l
| |
| h 97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 20 of 26 I ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINEO
| |
| - STEP
| |
| }
| |
| l co*************************************
| |
| CAUTION: RCS and ruptured SG(s) pressures must be maintained less than 1025 psig to prevent lifting ruptured SG(s) PORV. ,
| |
| oo************************************* f l
| |
| j
| |
| : 31. CONTROL RCS PRESSURE AND MAKEUP FLOW TO MINIMIZE RCS-TO-SECONDARY LEAKAGE: l a) Perform appropriate action (s) from table:
| |
| PRZR LEVEL RUPTURED SG(s) LEVEL INCREASING DECREASING OFFSCALE HIGH LESS THAN
| |
| * Increase RCS Increase RCS
| |
| * Increase RCS 30% [55%] Makeup Flow Makeup Flow Makeup Flow
| |
| * Depressurize
| |
| * Maintain RCS-RCS Using and Ruptured Step 31b SG(s) Pressure Equal BETWEEN Depressurize Turn On Maintain RCS 30% [55%] RCS Using PRZR and Ruptured AND 60% Step 31b Heaters SG(s) Pressure Equal BETWEEN
| |
| * Decrease RCS Turn On Maintain RCS 60% AND Makeup Flow PRZR and Ruptured 70% [65%] Heaters SG(s) Pressure
| |
| * Depressurize Equal RCS Using Step 31b GREATER THAN Decrease RCS Turn On Maintain RCS 70% [65%] Makeup Flow PRZR and Ruptured Heaters SG(s) Pressure Equal b) Use normal spray b) E letdown is in service, THEN use auxiliary spray. IF NOT, THEN use one PRZR PORV.
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| S
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%).
| |
| j
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| l a) SU3 CRITICALITY - Nuclear power greater than 5: I' b) CORE COOLING - Core Exit TCs greater than 1200*F 8
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46I with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10" (32%) AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F e) C0KIAINMENT - Contain=ent pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has ce=pletely depressutized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION j
| |
| GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level j decreases to less than 291.
| |
| : 5. ECST LEVEL CRITERIA f i
| |
| Make-up to the ECST frem the CST or provide alternate sources for AFW ;
| |
| suction (SW or FP) when ECST level decreases to 40%. 1
| |
| (
| |
| 1
| |
| -_________a
| |
| | |
| l J
| |
| 00,97887210 e
| |
| PROCEDURE TITLE
| |
| (
| |
| REvtS ON NUMBER 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE l 21 of 26 i l
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED .
| |
| I
| |
| : 32. CHECK IF DIESEL GENERATORS SHOULD BE STOPPED. 4 a) Verify AC emergency a) Initiate 1-AP-10.1 to l busses - ENERGIZED BY restore offsite power. ~);
| |
| 0FFSITE POWER b) Stop any unloaded diesel generator as per 1-OP-6.1 and/or 1-0P-6.2 33.
| |
| MINIMIZE SECONDARY SYSTEM CONTAMINATION:
| |
| * Isolate Hotvell high level divert
| |
| * Bypass Powdex System
| |
| * Transfer Auxiliary Steam to Unit 2 2E Place auxiliary boiler in service
| |
| : 34. TURN ON PR2R HEATERS AS NECESSARY TO SATURATE PRZR WATER AT RUPTURED SG(s) -
| |
| PRESSURE
| |
| | |
| l FOLDOUT FOR EP-3 AND ES-3 PROCEDURES l
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| OR l
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THM 15% [50%) .
| |
| : 2. RED PATE
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SU3 CRITICALITY - Nuclear power greater than 5:
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F SE, Core Exit TCs greater than 700*F AND R7LIS full range less than 46 with no RCPs running c) HEAT SINE - Narrow Range level in all SGs less than 10% [32%] AND total feedwater flow less than 340 gpm l d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285*F l e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has cenpletely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION Go TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| | |
| Co.97847210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 22 of 26 '
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSENOTOBTAINED
| |
| : 35. CHECK RCP COOLING FLOW - Establish flow as required.
| |
| NORMAL
| |
| * Upper Bearing )
| |
| * Lower Bearing
| |
| * Stator
| |
| : 36. CHECK RCP SEAL COOLING: IF either flow indicated, THEN re-establish the other flow, a) Thermal barrier - E no flow indicated, THEN FLOW INDICATED initiate 1-AP-33.2 to '
| |
| re-establish seal cooling.
| |
| b) Seal injection -
| |
| FLOW INDICATED
| |
| : 37. CHECK IF RCP SEAL WATER RETURN CAN BE ESTABLISHED:
| |
| l a) RCS pressure - GREATER a) GO TO Step 38_. )
| |
| TRAN 100 PSIG ;
| |
| I b) Verify seal injection b) GO TO Step 38.
| |
| flow - INDICATED c) Open seal water return MOVs
| |
| * MOV-1380
| |
| * MOV-1381 l
| |
| i
| |
| | |
| I i
| |
| l i
| |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES I
| |
| : 1. SI REINITIATION CRITERIA Mahually ope} ate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SITBC00 LED RECOVERY DE1 Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| pR
| |
| * PR2R level - CANNOT BE MAINTAINED GREATER THAN IS: (50 ].
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5" b) CORE COOLING - Core Exit TCs greater than 1200*F E
| |
| _ R.
| |
| Core Exit TCs greater than 700*F AND RVLIS full range c) less than 46" with no RCPs running I HEAT SINK - Narrow Range level in all SGs less than 10:
| |
| [32%] AND d) total feedwater flow less than 340 gym INTEGRITY - Cold leg temperature decrease greater than 100'F in last e) 60 minutes AND RCS cold leg temperature less than 285*F CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pre is decreasing in an uncontrolled manner or has cenplately deprescurized, and has not been isolated, unless needed for RCS cooldown.
| |
| 4.
| |
| COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%. -
| |
| : 5. ECST LEVEL CRITERIA t
| |
| Make-up to the ECST from the. CST or provide alternate sources for AFW l suction (SW or FP) when ECST level decreases to 40".
| |
| 1
| |
| | |
| 30.97887210 PROCEDURE TITLE REVISION NUMBER 1.00 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 23 of 26 l
| |
| l ACTION / EXPECTED RESPONSE RESPONSENOTO8TAINED
| |
| - STEP CAUTION: If all seal cooling had been previously lost, then to prevent any further seal damage, the affected RCP(s) should not be started without prior status evaluation.
| |
| l CAUTION: Due to reduced RTD bypass flow on natural circulation, RTD bypass temperatures and associated interlocks will be inaccurate.
| |
| ****************** **************w* *****
| |
| CAUTION: If an Emergency Diesel Generator is paralleled to the same RSS bus as a Reactor Coolant pump, then to prevent voltage oscillations, that Reactor Cool.2nt pump should not be started.
| |
| NOTE: Normal PRZR spray should be islocated from any RCP that is stopped.
| |
| NOTE: RCPS should be run in order of priority to provide normal PRZR spray.
| |
| : 38. CHECK RCP STATUS:
| |
| a) At least one RCP - a) Try to start one RCP:
| |
| - RUNNING
| |
| : 1) H RVLIS upper plenum range indication less than 95%, THEN perform the following:
| |
| * Increase PRZR level
| |
| ( to greater than 72%-
| |
| [60%)
| |
| * Increase RCS sub-cooling based on Core Exit TCs to greater than 55'F (105'F]
| |
| : 2) Try to start one RCP as per 1-0P-5.2. IJ[
| |
| t an RCP canNOT be started, THEN initiate Attachment 1.
| |
| l 1
| |
| b) Stop all but one RCP l
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. - SI REINITIATION CRITERIA Manuelly operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| _OR
| |
| '* PRZR level - CANNOT BE MAINTAINED GREATER TEAN 15% [50%).
| |
| : 2. RED PATH SIMMARY a) SUBCRITICALITY - Nuclear power greater than 5".
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F l pR, Core Exit TCs greater than 700*F AND RVLIS full range less than 46 :with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less than 10: (32"] AND total feedwater flow less than 340 gym d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285"F e) CONTAINMENT - Containment pressure greater than 60 P.SIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressuri::ed, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1,,, if RWST level decreases to less than 29".
| |
| : 3. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AW suction (SW or FP) when ECST level decreases to 40".
| |
| l l .
| |
| | |
| i l
| |
| G 97887810 PROCEDURE TITLE REVISION
| |
| ' NUMBER 1.00 STEAM GENERATOR TUBE RUPTURE PAGE 1-EP-3 24 of 26
| |
| - ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED STEP l
| |
| : 39. CHECK INTERMEDIATE RANGE FLUX:
| |
| ~U a) Continue with Step ,4_0_.
| |
| 0 a) Flux - BELOW 5x10 ON N-35 AND N-36 WHEN both conditions satisfied, THEN AND do Step 39b and c.
| |
| P-6 (Both channels) - NOT LIT
| |
| * IL-F1
| |
| * IL-F2 ,
| |
| l b) Verify both N-31 and N b) Initiate 1-AP-4.1.
| |
| RE-ENERGIZED c) Transfer recorder NR-45 to .
| |
| N-31 and N-32
| |
| : 40. ALIGN RUPTURED SG(s) FOR SAMPLING:
| |
| a) Rack in one Main Feedwater pump breakers to the test position b) Place the Main Feedwater pump switches in the start position
| |
| {
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| gR l
| |
| * Pl<2R level - CANNOT BE MAINTAINED '3REATER THAN 15% [50") . ,
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5:
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F pR Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no RCPs running c) HEAT SINR - Narrow Range level in all SGs less than 10" [32".] AND total feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS ' cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA .
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled =anner er has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LIG RECIRCULATION, Step 1, if RWST level decreases to less than 29".
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40".
| |
| | |
| . \.
| |
| Do. 97887210 PROCE00RE TITLE REVISION .
| |
| NUMBER ,
| |
| 1.00 1-EP-3' STEAM GENERATOR TUBE RUPTURE PAGE 3
| |
| 25 of 26 RESPONSENOTOBTAINED
| |
| - STEP ACTION / EXPECTED RESPONSE ' l
| |
| : 41. SHUT DOWN UNNECESSARY PLANT EQUIPMENT:
| |
| a) Check station service NOT a) Establish load shedding as being supplied from per 1-0P-26.7.
| |
| reserve station service b) Check main generator removed from service as per 1-OP-15.2 c) Check main turbine:
| |
| : 1) Turning gear oil pump 1) Manually start pump.
| |
| starts at 12 psig oil pressure
| |
| : 2) Seal oil back-up pump 2) Manually start pump, starts at 12 psig oil pressure d) All turbine drains - OPEN d) Manually open valves, e) Two main feed pumps -STCPPED e) Manually stop all but one Pump.
| |
| f) Two main condensate pumps - f) Manually stop all but one STOPPED Pump.
| |
| g) Low pressure heater drain g) Manually stop pumps.
| |
| pumps - STOPPED h) High pressure heater drain h) Manually stop pumps.
| |
| pumps - STOP?ED'
| |
| | |
| t FOLDOUT FOR EP-3 AND ES-3 PR0b2DURES
| |
| - z
| |
| : 1. SI REINITIATION CRITERIA s )
| |
| Manually operate Charging /SI prmps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS 011 REACTOR COOLANT - SUBCOOLED' REC 3VERY DESIRE Step 1,if either conditioO listid below occurs:
| |
| * RCS subevaling based en Core Exit TCs - LESS TEAN 30*F [80*Fl .
| |
| OR
| |
| . l i
| |
| * PRZR level - CANNOT BE MADTADED GREATER THAN '15% [50:1 l
| |
| : 2. RED PATH SLTMRT a) SUBCRITICALITI - Nuclear pown greater than 5 b) C GE COOLING - Care Exit TCs greaterichan 1200*F l
| |
| .OR..
| |
| s Core Exit TCs great.cr than 700*F AND RVLIS full rarge !
| |
| less than 46: with no RCPs running c) HEAT S'2E - Narrow Range level in all SGs less than 10% { 32 ] AND cotal feedvater' flow less than 340 gpw.
| |
| d) DiTEZRTTT - Cold leg temperature decrease greater than 100*F in last l j
| |
| 60 minutes ,AZJ,RCS cold leg temperature lass then 285*F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| \
| |
| : 3. SECONDARY INTEGRITYiCRIT2RIA GO TO EP-2, FAULTED STEAM GENCIATOR ISCLATION, Step 1, if any SG pressure is decreasing in an kneontrolled manner or has completely' depressuri:dd, and has not been Bolated, unless needed for RCS cooldovn.
| |
| : 4. COLD LEG RECIRCULATION % SWITCROVEs CRITERION s <
| |
| Cb TO ES-1.3, TI.tNCFIS 40 COLD LEG RECIRCULATION, Steg.,, if RWST level -
| |
| t dacrease.s'to less than 29%.'
| |
| 1 2
| |
| : 5. ECST LEVEL CRI*I*ERIA'
| |
| \ ,,
| |
| .19, Make-up to the fm r from the CST or provide alternate sources for AFW s g
| |
| suction (SW ct FP) chen ECST level decreases to 40%.
| |
| t . 1A x
| |
| k ,
| |
| . 1 L u i '
| |
| , ,i Y y Q )
| |
| N' 4 s o
| |
| .1-L
| |
| | |
| m.stsento NUMBER PROCEDURE TITLE REflSgN 1-EP-3 STEAM GENERATOR TUBE RUPTURE PAGE 26 of 26 ACTION / EXPECTED RESPONSE . RESPONSE 70TOBTAINED
| |
| - STEP
| |
| : 42. GO TO APPROPRIATE POST-SGTR C00LDOWN METHOD:
| |
| GO TO 1-ES-3.1, POST-SGTR C00LDOWN USING BACKFILL Step 1 pR GO TO 1-ES-3.2, POST-SGTR COOLDOWN USING BLOWDOWN Step 1 j
| |
| ~
| |
| OR l
| |
| GO TO 1-ES-3.3, POST-SGTR )
| |
| C00LDOWN USING STEAM DUMP l Step 1 l 1
| |
| I ;
| |
| I i
| |
| J
| |
| }
| |
| a _ _ _ _ _ _ _ _ - ..
| |
| | |
| i No.97887223 i
| |
| i NUMBER ATTACHMENT TITLE REVISION 1-EP-3 1.00 NATURAL CIRCULATION A TTACHMENT VERIFICATI0y PAGE 1
| |
| 1 of 1 l
| |
| i STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED l
| |
| NOTE: The following conditions support or indicate natural circulation flow. 3 I
| |
| : 1. VERIFY NATURAL CIRCULATION IF natural circulation NOT FLOW: E rified, THEN notify the Shift Supervisor.
| |
| l a) RCS subcooling based on Core Exit TCs - GREATER j
| |
| ,l THAN 30*F [80*F] 4 b) SG pressures - STABLE OR DECREASING c') RCS hot leg temperatures -
| |
| - j STABLE _OR DECREASING d) Core Exit TCs - FTABLE OR j DECREASING 1
| |
| e) RCS cold leg temperatures - ;
| |
| AT SATURATION TEMPERATURE FOR SG PRESSURE i
| |
| i l
| |
| I I
| |
| l e
| |
| | |
| 4 i
| |
| I l
| |
| l l
| |
| ATTACHMENT 7
| |
| \
| |
| i i
| |
| EMERGENCY OPERATING PROCEDURE l ES-3.1, POST STEAM GENERATOR TUBE RUPTURE COOLDOWN USING BACKFILL j
| |
| .)
| |
| )
| |
| i
| |
| | |
| VIRGINIA POWER erwen NORTH ANNA POWER STATION EMERGENCY PROCEDURE Procedure Title Revision Number 1.00 1-ES-3.1 POST-SGTR C00LDOWN Page
| |
| .USING BACKFILL (WITH NO ATTACHMENTS) 1 of 6 Purpose To provide guidance for operations personnel to cooldown and depressurize the i
| |
| RCS to CSD conditions following a SGTR. This recovery method depressurizes the ruptured SG(s) by draining through the ruptured SG(s) tube (s) into the RCS.
| |
| User NAPS Operations Personnel Entry Conditions This procedure is entered from:
| |
| : 1) 1-EP-3, STEAM GENERATOR TUBE' RUPTURE, or
| |
| : 2) 1-ES-3.2, POST-SGTR C00LDOWN USING BLOWDOWN.
| |
| SAEY R _~ATU Revision Record REV. 1.01 PAGES (S): 2&5 DATE: 06-12-87 REV. PAGES(S): DATE:
| |
| REV. PAGES(S): DATE:
| |
| REV. PAGES(S): DATE:
| |
| REV. PAGES (S): DATE:
| |
| REV. PAGES(S): DATE:
| |
| Approval Recommended Approved Date e V O Chairman Station Nuclear Safety 06-12-87 and Operating Committee h
| |
| | |
| FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO.TO ,
| |
| I ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs: i
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F (80*F].
| |
| S.R,
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% (50%] .
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than.5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200'F 2R.
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46" with no ECPs running c) HEAT SINK - Narrow Range level in all SGs less than 10% (32"] AND total feedvater flov less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| 4, COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULCION, Step 1,1f RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA.
| |
| Make-up to the ECST from the CST or provide alternate sources for AFW ]
| |
| suction (SW or FP) when ECST level decreases to 40%. {
| |
| I i
| |
| | |
| Co.97887210
| |
| ~
| |
| )
| |
| PROCEDURE TITLE NUMBER REflS}N l
| |
| ~
| |
| i 1-ES-3.1 POST-SGTR C00LDOWN ' {
| |
| I USING BACKFILL 2 of 6 l
| |
| ACTION / EXPECTED RESPONSE RESPONSENOT OBTAINED
| |
| - STEP NOTE: Setpointsinbrackets[]areforadversecontginment atmosphere (20 psia containment pressure or 10 R/HR containment radiation.)
| |
| : 1. TURN ON PRZR HEATERS AS .
| |
| NECESSARY TO SATURATE PRZR WATER AT RUPTURED SG(s) ,
| |
| PRESSURE .{
| |
| i
| |
| : 2. CHECK IF SI ACCUMULATORS SHOULD BE ISOLATED:
| |
| a) Check the following: a) IF either conditien NOT satisfien, THEN GO TO
| |
| * RCS subcooling based 1-ECA-3.1, SGTR WITH LOSS on Core Exit.TCs - 0F REACTOR COOLANT - SUB- l GREATER THAN 30*F [80*F] COOLED RECOVERY DESIP.ED, ]
| |
| Step J,. .l
| |
| * ?RZR level-GREATER j THAN 15% [50%)
| |
| b) Accumulator discharge b) Restore power'to discharge MOVs - ENERGIZED MOVs.
| |
| c) Close all SI accumulator c) Vent any accumulator that discharge MOVs can NOT be isolated as per.
| |
| 1-0P-7.3.
| |
| : 3. VERIFY ADEQUATE SHUTDOWN MARGIN AS PER 1-PT-10: !
| |
| a) Sample ruptured SG(s) b) Sample RCS c) Shutdown Margin- c) Borate as required.
| |
| ADEQ" ATE
| |
| | |
| FOLD 00T FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based .cnt Core Exit TCs - LESS THAN 30*F (80*F] . l 1
| |
| Slh
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%]. j
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| I a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F i
| |
| S.R Core Exit TCs greater than 700*F AND RVLIS full range j less than 46 with no RCPs running j c) HEAT SINK - Narrow Range level in all SGs less than 10% (32%l AND ;
| |
| i
| |
| ' total feedveter flow less than 340 gpm d) INTEGRITT - Cold leg ten;peratura decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature less than 285'F e) CONTADIMENT - Containmc.nt pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GDIERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40".
| |
| l
| |
| -)
| |
| | |
| Co.97887210 PROCEDURE TITLE REV/SION NUMBER 1.00 1-ES-3.1 POST-SGTR C00LDOWN PAGE USING BACKFILL 3 of 6 l
| |
| ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED
| |
| - STEP CAUTION: Alternate water sources (CST or 1-AP-22.7) to prevent less of AFW pump suction pressure will be necessary if ECST decreases to 40%.
| |
| : 4. CHECK INTACT SG(s) LEVEL:
| |
| a) Narrow Range level - a) Maintain total feed flow GREATER THAN 10% (32%) greater than 340 gpm until NR level in at least one SG is greater than 10% (32%1 b) Control feed flow to b) I_F,NR level in any maintain Narrow Range intact SG continues to level between increase in an uncontrolled 10% [32%] and 50% manner, THEN GO TO 1-EP-3, STEAM GENERATOR TUBE RUPTURE, Step 1.
| |
| NOTE: Since ruptured SG(s) may continue to depressurize to less than the minimum RCS pressure required for RCP operation, ccoldown to CSD should be completed as quickly as possible within Tech l Spec cooldown restrictions.
| |
| l
| |
| : 5. INITIATE RCS C00LDOWN TO COLD SHUTDOWN:
| |
| a) Maintain cooldown rate in RCS cold legs - LESS THAN 100'F/HR b) Use RHR system if in service (STEP 5 CONTINUED ON NEXT PAGE)
| |
| | |
| l l
| |
| i l
| |
| l
| |
| , FOLDOUT FOR EP-3 AND ES-3 PROCEDURES i
| |
| 1
| |
| : 1. SI REINITIATION CRITERIA !
| |
| Manually opera c Jharging/SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED REC 0 VERT DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| S
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%).
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F Core Exit TCs greater than 700*F AND RVLIS full range less than 46 with no RCPs. running c) HEAT SINK - Narrow Range level in all SGs less than 10% (32%) AND total feedvater flev less than 340 gpm d) - INTEGRITY - Cold leg temperature decrease greater than 100*F in last 60 minutes AND RCS cold leg temperature lest than 285"F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITT CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled =anner or has completely depressurized, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION l
| |
| GO TO ES-1.3, TRANSFER TO COLD LEG RECIRCULATION, Step 1, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level c'.ecreases to 40"..
| |
| | |
| = _ . . - _ . - _ _ - . ____________ __________________
| |
| e.97sa7810 ;
| |
| PROCEDURE TITLE REVIS ON NUMBER 1-ES-3.1 POST-SGTR COOLDOWN PAGE USING BACKFILL 4 of 6 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| . STEP
| |
| : 5. INITIATE RCS C00LDOWN TO COLD SHUTDOWN: (CONTINUED) c) Dump steam to c) Manually or locally dump condenser from intact steam from intact SG(s):
| |
| SG(s)
| |
| SG PORV(s) g 4 Decay heat release valve. ;
| |
| E no intact SG available, THEN perform the following: -
| |
| Use faulted SG S
| |
| GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED,~ Step 1,.
| |
| : 6. CHECK RUPTURED SG(s) Refill ruptured SG to NARROW RANGE LEVEL - 75% [60%] using feed flow.
| |
| CREATER THAN 10% [32%] IF,,either of the following occurs, THEN stop feed flow to the ruptured SG:
| |
| Ruptured SG pressure decreases in an uncontrolled manner E
| |
| Ruptured SG pressure increases to 1000 psig.
| |
| l
| |
| -- __ __ - -__ _ _ ____--_____ _.-_.___m-_
| |
| | |
| I 1
| |
| ]
| |
| 70LDOUT FOR EP-3 AND ES-3 PROCEDURES I
| |
| J
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1 if either condition listed below occurs:
| |
| * RCS subcooling based on Core Exit TCs - LESS THAN 30*F [80*F].
| |
| gR,
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%).
| |
| : 2. RED PATH SU! GARY 1
| |
| l a) SUBCRITICALITY - Nuclear pcwer greater than 5%
| |
| b) CORE COOLING - Core Exit TCs greater than 1200*F pR, Core Exit TCs greater than 700*F AND RVLIS full range less than 46: with no RC?s running c) HEAT SINK - Narrow Range level in all SGa less than 10% [32%] AND total feedwater flov less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than,10D*F in last 60 minutes flip, RCS cold les toeperature less than 285*F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECCNDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1, if any SG pressure is decreasing in an uncontrolled manner or has conpletely depressuri::ed, and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCUIJ. TION SWITCHOVER CRITERION 1
| |
| GO TO ES-1.3, TRALSFER TO COLD LEG RECIRCUI.ATION, Step 1, if RWST level decreases to less than 291.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| - _ . - _ _ ~ . _ _ - _ _ . . . . _ _ . . . _ _ _ . _ _ . _ _ _ _
| |
| | |
| j i
| |
| & 97887810 PROCEDURE TITLE REVISION NUMBER 1.00 ;
| |
| l 1-ES-3.1 POST-SGTR C00LDOWN FA CE USING BACKFILL 5 of 6 l 1
| |
| l l
| |
| I
| |
| - STEP ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINEO \
| |
| : 7. CONTROL RCS MAKEUP FLOW TO MAINTAIN PRZR LEVEL:
| |
| I n) PRZR level - GREATER a) Increase RCS makeup flow as l THAN 30% [55%] required. GO TO Step 8.
| |
| b) PRZR level - LESS b) Decrease RCS makeup flow as THAN 70% [65%] required. GO TO Step 9.
| |
| 1 NOTE: The upper head region may void during RCS depressurization if l RCPs are not running. This may result in a rapidly increasing PRZR level. l
| |
| : 8. DEPRESSURIZE RCS TO BACK -
| |
| FILL FROM RUPTURED SG(s):
| |
| a) Use normal PRZR spray a) IF, letdown in service, THEN use auxiliary spray I_(
| |
| NOT, THEN use one PRZR PORV.
| |
| b) Turn on PRZR heaters as necessary c) Maintain RCS subecoling based on Core E::it TCs -
| |
| GREATER THAN 30 'E [80 *F]
| |
| : 9. CHECK IF RHR CAN BE PLACED IN SERVICE:
| |
| a) Check the following: a) GO TO Step 10.
| |
| * RCS WR T -
| |
| LESS IIIAN 350*F
| |
| * RCS pressure -
| |
| LESS THAN 400 PSIG
| |
| [225 PSIG]
| |
| b) Place RHR in service j as per 1-0P-14.1 l
| |
| | |
| l FOLDOUT FOR EP-3 AND ES-3 PROCEDURES
| |
| : 1. SI REINITIATION CRITERIA Manually operate Charging /SI pumps and align BIT as required and GO TO ECA-3.1, 3GTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1,i! aither condition listed below occurs:
| |
| * RCS subcooling based on Core Exir TCs - LESS TRAN 30*F [80*F].
| |
| I 0R
| |
| * PRZR level - CANNOT BE MAINTAINED GREATER THAN 15% [50%) .
| |
| : 2. RED PATH
| |
| | |
| ==SUMMARY==
| |
| | |
| a) SUBCRITICALITY - Nuclear power greater than 5% j CORE COOLING - Core Exit TCs greater than 1200*F b)
| |
| O,R,,,
| |
| Core Exit TCs greater than 700*F AND RVLIS full range less than 46% with no RCPs running c) HEAT SINK - Narrow Range level in all SGs less chan 10% [32%] fyg[
| |
| cotal feedwater flow less than 340 gpm d) INTEGRITY - Cold leg temperature decrease greater than 100'F in last j 60 minutes AND RCS cold leg temperature 1ess than 285'F e) CONTAINMENT - Containment pressure greater than 60 PSIA
| |
| : 3. SECONDARY INTEGRITY CRITERIA GO TO EP-2, FAULTED STEAM GENERATOR ISOLATION, Step 1,, if any SG pressure )
| |
| is decreasing in an uncontrolled manner or has completely depressurized, j and has not been isolated, unless needed for RCS cooldown.
| |
| : 4. COLD LEG RECIRCULATION SWITCHOVER CRITERION GO TO ES-1.3, TRANSFER TO COLD LEG REC 1RCULATION, Step 1,, if RWST level decreases to less than 29%.
| |
| : 5. ECST LEVEL CRITERIA Make-up to the ECST from the CST or provide alternate sources for AFW suction (SW or FP) when ECST level decreases to 40%.
| |
| l L_._______
| |
| | |
| be.97887210 PROCEDURE TITLE REVIS N NUMBER 1-ES-3.1 POST-SGTR C00LDOWN PA GE USING BACKFILL 6 of 6 ACTION / EXPECTED RESPONSE RESPONSE NOTOBTAINED
| |
| - STEP l
| |
| : 10. CHECK IF RCPs MUST BE STOPPED:
| |
| a) Monitor the following: a) GO TO Step 11,.
| |
| i Number 1 seal differ- l ential pressure - l
| |
| : LESS THAN 210 PSID l l 1 SE Number 1 seal leakoff flow - LESS THAN 0.3 GPM b) Stop affected RCP(s)
| |
| Return to Step 3. I
| |
| < 11. CHECK RCS WR T -
| |
| LESSTHAN200*!
| |
| i
| |
| { 12. EVALUATE LONG TERM PLANT STATUS:
| |
| a) Maintain cold shutdown b) Consult TSC personnel END
| |
| | |
| W'ginia Bectric and Power Company 4$$ggoa e e ba MA w5 $%4:
| |
| s IIh y-4 I
| |
| sg hahby w$!\
| |
| $g y ig ((k( $,3!hg y 'k i h North Anna Unit 1 July 15,1987 Steam Generator Tube Rupture Event Presentation July 29,1987 Revision 0
| |
| | |
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| NORTH ANNA UNIT 1 -
| |
| 9 STEAM GENERATOR TUBE RUPTURE l l
| |
| | |
| ,y
| |
| ''' \ ........... . . . . .
| |
| MGlE N .. , .... . .DN y
| |
| llNTRODUCTib N. ,!!
| |
| . !:i' i$EMUENCEMEJEVEN31S]
| |
| i j l
| |
| LTECH N.lCAMEVlf04IlON?
| |
| 1
| |
| ::CONCLUSidN l
| |
| }
| |
| | |
| 4 e a. ,
| |
| INITIAL INDICATIONS AND ACTIONS SAFETY INJECTION IDENTIFYING THE EVENT ISOLATING THE "C" STEAM GENERATOR .l l TERMINATING THE BREAK FLOW COOLDOWN AND DEPRESSURIZATION EVENT TERMl4ATIO4 AND RECOVERY l-
| |
| | |
| 3,s s 15EM$nSMEfMniisMGBMMi , a
| |
| :!:!*isi!!
| |
| JULY 14,1987 2238 Air Ejector R.M. Inoperable JULY 15,1987 0630 Main Steam R.h4. Alarms 0631 Przr. Level Decreasing Rapidly 0632 Manual Control Charging Maximum Przr. Lo Pressure Alarm (2135 psig) 0633 Letdown Isolatec Charging to RWST Turbine Ram o Down l
| |
| t
| |
| | |
| l 7x is i
| |
| MEMbinI9MnidninMnM$Mi: ,
| |
| y .m
| |
| ! JULY 15,1987 (cont.) i 0635 Manual Reactor Trip Initiate Emergency Procedures -
| |
| 1 Przr. Level 45% l Przr. Pressure 2100 psig SI (Przr. _' o _o)
| |
| Auxilary Feed Pum as Start 0636 Przr. Level 2.7%
| |
| Przr. 3ressure 1700 asig l
| |
| | |
| x' T( '
| |
| mu ,
| |
| , isEN- e[$M5$dRisjMi$i@:nMi l=
| |
| JULY 15,1987 (cont.)
| |
| 0639 NOUE Declared 80640 Move to SGTR Emergency Procedure -
| |
| 0642 S/G 'C' N.R. Level Increasing (18%)
| |
| 0644 Si Reset LHSI Pumps Secured 0645 S/G 'C' N.R. Level increasing ,
| |
| (25%)
| |
| l I i
| |
| ( - - _ _ _ _ _ - _ .
| |
| | |
| i s,. ,
| |
| I 5 '
| |
| *# l[SM@$M[$$d$le$$$$
| |
| . . .m ns . )
| |
| JULY 15,1987 (cont.)
| |
| l *0646 SS Confirmed SGTR 'C' Auxilary Feed to S/G 'C' Isolated Main Steam Trip Valve 'C' Closed 0648 S/G 'C' Steam Supply to Terry Turbine Isolated RCS Pressure & Przr. Level increasing 0649 Ra old Cooldown on Stea m Dumps
| |
| *0650 Przr. Level Off Scale Low ,
| |
| | |
| s
| |
| .i i \
| |
| inEMRBHMEddBleMnNinhji ,
| |
| . a_.
| |
| iJULY 15,1987 (cont.)
| |
| 0651 Notifications to State / Local 0654 Upgraded to Alert
| |
| '*0658 Przr. Level on Scale and Increasing l 0700 Przr. Level > 15%
| |
| 0702 Notifications to State / Local ;
| |
| l and NRC of Alert
| |
| | |
| 1
| |
| . . s s. .
| |
| s... .,.v .,,v ...............,n.
| |
| ' \ . . . . . . , . . . . , . . . .. .,:,. .. . . , . . ,
| |
| <., $ EQRERRElQE.$.,- .... .....,.,.,....w..,v.,....
| |
| p ..
| |
| a w . .-
| |
| l JULY 15,1987 (cont.)
| |
| ; 0704 S/G 'C' Level Stable i Initiate RCS Depressurization <
| |
| on (1) PORV l 0713 RCP's 'C' and 'B' Secu red 0714 RCS Pressure Control in Manual 0718 Initiate S/G Cooldown and j l
| |
| l Backfill l 0722 Przr. Level Decreasing 73%
| |
| 80727 Initiate RCS Cooldown i
| |
| | |
| j
| |
| # !$$nidnNM$$dni$@[$M$Mj!
| |
| \ -
| |
| JULY 15,1987 (cont.)
| |
| 0739 Station Manager Assumes SEM l 0756 Air Ejector Manual Diverted .
| |
| to Containment 0757 TSC Activated 0820 CERC Activated l 0900 Loose Parts Alarm S/G 'C' 0915 LEOF Activated i
| |
| | |
| /
| |
| y .
| |
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| |
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| |
| ;~
| |
| age.w
| |
| ~ . -
| |
| um m *-
| |
| e ma~ r
| |
| ~~
| |
| JULY 15,1987 (cont.)
| |
| 1108 Mode 4 1219 RHR in Service 1221 RCP 'A' Secured 1312 Air Ejector to Atmosphere 1330 Mode 5 1335 Emergency Terminated 1336 Notifications to State / Locals -
| |
| and NRC of Termination implemented Recovery Organization ;
| |
| 1 l
| |
| l l
| |
| l
| |
| | |
| ^g 'a ''h,
| |
| .E M E RG E N 'C. W_Pbn. N..N. IN G-m cu
| |
| -z jQ
| |
| /
| |
| Emerg ncy(Plan:li dalemerdedl a
| |
| 'i
| |
| ::Successfu lyl< .
| |
| l
| |
| : . J Emergency FacilitiesLand "
| |
| l 1
| |
| O rg.a ni~zationsL Effectivec .,
| |
| r a
| |
| 1 MdY
| |
| | |
| t'
| |
| '1T'.$1 '
| |
| AEMMMEn$[UnGANM%hldNi m
| |
| , 8E M8s '
| |
| iS_tatiorn __ Recovery JNRCjWUgg Manager ;M a na g.e_____rg)! .Inspp!Jeacht E
| |
| ' lCornrnd ndi;..
| |
| 3 ChirMbEOM 1
| |
| l i
| |
| w ,g _ a.= = .
| |
| i I I M ainti1 Tech: NucinaW '
| |
| lindustFft LiiHealt;hl;i
| |
| &l:Relidiri 1Eva l.y iS & Li .I'nterf66e? lP hysibst l
| |
| 1
| |
| | |
| ^
| |
| s
| |
| .'' NN '
| |
| ilMPACEGRiSAMEIM EBoundidLbpjUFSAR oMCo re3la f st y j lfi m'it s N otsC h a l le;n;;g ed l to Shutdowr1 and Thermalj Mar lg)inss
| |
| ; Maintained f: Jo:
| |
| E R a d iola ctLivei R ele a sle s3MMdidF R$ :
| |
| 100 Li~mits- .
| |
| [
| |
| 4
| |
| | |
| i' '
| |
| ENGib EERINWAN AlWSIS :
| |
| ![dn!EMENT!
| |
| EVENS MFSX$
| |
| Maximum- .560-63%GPM; RdO;GPM Lsak; Rate Mass MMOO,000 Ibm: #3b[OOOillimi Transfer Cods! Safety Closed Lopeni Valve Position- i
| |
| | |
| i l
| |
| 'i T '
| |
| (R M I M Diliil M E M A M E ni M 0!!
| |
| !RE(GIEQ%iM]$MMR$$i SITE SOURCE CURIES % OF T.S. BOUNDARY
| |
| ! DOSE Em. Evt. O.265 0.62% 3.43 E-4 Assess. mrem RMS Mon. O.159 0.33% 5.78 E-4
| |
| & Samples mrem Envr. TLDs, No Detectable Activity Above Soil, Veg., Normal Background Levels Air Part., Noted Rad. Sam :). ;
| |
| | |
| t s
| |
| ~ ilNidM[s#R8i@]sEnMRMnll
| |
| . ma .
| |
| are qs ,
| |
| :8hs ssd,!NS ssgs l o NEWS MEDIA )
| |
| - 5 Press Releases
| |
| - 8 Press Conferences 1'
| |
| - Numerous Individual Interviews
| |
| - Tou rs
| |
| . l l
| |
| l o ORGANIZATION BRIEFINGS
| |
| - DAILY UPDATES
| |
| - State Corporation Commission
| |
| - D.epartment of Emergency Services
| |
| - Old Dominion Electric Cooperative i
| |
| | |
| j s=
| |
| llNDQ$ilMMilN!sen$Dh o INSTITUTE OF NUCLEAR POWER .
| |
| OPERATIONS (IN PO)
| |
| Ooerating Experience Review Team 3 Members 8 Nuclear Network Entries o UTILITY BRIEFINGS Direct Utility Contact .
| |
| Foreign Uti ity Contact Nationwide Meetings o LOCAL GOVERNMENT OFFICIALS BRIEFING 54 Individuals from State & Local Governments i l
| |
| l
| |
| | |
| .vu .w. i w.
| |
| 'Ne'e' /
| |
| STEAM NN.'
| |
| GENERATOR ~'
| |
| p.-
| |
| L. HISTORY
| |
| */.%'."
| |
| | |
| 'i .- . .. .
| |
| 1ti~lBjEMRQBilidRlEl
| |
| \' jlif # RGEWE%. . K
| |
| ; %lEMEBi$$KlissM !
| |
| o POINT BEACH 1 1 2/75 ~125 g p m I l
| |
| Axial Failure at Top of Hot Leg Tube Sheet Due To Wastage /
| |
| SCC o SURRY 2 !
| |
| 9/76 ~330 g pm Axial Failure at U-Bend Apex Due to PWSCC :
| |
| l l
| |
| u-- - - - - - _ - _ - - - - - - - - - - - - - - -
| |
| | |
| \'
| |
| !$$RGEl!![IlEWs%IllMEl![RMMIDIfiEl
| |
| $$$$!Nis!38!Ii5$6M o DOEL 2 6/79 ~150 g pm Axial Failure at U-Bend Apex l
| |
| Due To PWSCC o PRAIRIE ISLAND 1 10/79 ~390 gpm Axial Failure at Top of Hot Leg Tube Sheet Due to Loose Parts Fretting j
| |
| | |
| , o a* * ,,, ,, , _ > < * - * ,,a l\ , J,,. ,e aARG5, yt!EfAR/$$8plRlil@@$E!
| |
| [ ,
| |
| (EMENKil![ljsiliRRM o GINNA 1/82 ~760 gpm Axial Failure at Top of Hot Leg Tube Sheet Due to Loose Parts Fretting l
| |
| | |
| s
| |
| . a.:,
| |
| STEAM: GE VE RATOMI.lSIORM
| |
| .:o : LWg: isoo py:PW R3w it h%t eliMSMs1
| |
| : o. Mill Annealed AIloyL60dllubes! ,
| |
| . . a"
| |
| : o. Carbon St' eel?Tu b;e;Sd'apor;tiP lates::
| |
| : oJW.e xt e xs Ex p;lo s ive l yi Ex pa riEle[d
| |
| :Tu bel Ends "
| |
| o Began:0 aerationMunel, #978) a
| |
| :oMAVT Chemistryisince StaMUS i
| |
| : o ; sPowd ext R es i n::I nt ru s io n D Q r:i rig;: y 1st1 Cycle' j
| |
| | |
| f l
| |
| k' {h''f. , ,. , , , , .: ,, ,
| |
| ,s ,.- iMyEHMROM[$$jjK@MMj[$Mi DUTAGE STEAM TOTAL DATE GENERATOR TUBES l
| |
| A B C l9-79 94 94 96 284 1-84 0 4 5 9 5-84 10 1 5 16 8-85 13 O O 13 11-85 9 17 47 73 l4-87 83 62 118 263 TOTAL 209 178 271 658 (6.2%) (5.3%) (8.0%) (6.5%) ; .
| |
| 1 I
| |
| l l
| |
| l
| |
| | |
| t STEA u ;GENERAIO;RlHISTO RY fo@98'5l lip O L L E DiT U B. E:S E X AMI N ED?
| |
| Discovered ID & LOD 3AttackiciM;TSPL RSubmittedLW Ca stoLN RC JMletJwithENRC in March,J198W 1^
| |
| o.19870 REFUELING OUTAGEL I -
| |
| Li n s pect ed J A IIT St ea m L Ge n;e ra~t;o rs ,,
| |
| Indluding 8:ix:1, RPClanct Profilornetry
| |
| | |
| mmm-A STEAM:lGEN E RAIQRAHl%31DRM WM8%!!REEOELINGIOOTAMEl!
| |
| - Found TSP, AVBrandii: Tubs'
| |
| !;$heetElndiccition'st :
| |
| -iP lu g g ed ?AllA$ P1 D ist o r;t era n ell QMO%$1ndications
| |
| - TC o n fi r m ed? D e n t s L N ott: G row i nlgj L
| |
| 4Dernonst ratedsTSWHl eat lStrsssj RelinMn4BTSteamTGenerafor: 1
| |
| | |
| 4 LSTEAWGENERATORMISIO;RM ddil;!%8%REEUELINnUOIXG$
| |
| :- PerformeWRowi2?UMBeniiNeati U ist resstReliedin!All1@Sfsani Gen 6ratdrst
| |
| :P.ulled321:iTu bsssf romi::MSteam:
| |
| Generator?
| |
| L Pibgg;ed LA-IIEu be:iSaest;
| |
| : Ini3ications;
| |
| -: M et::' wit h1N RC in4J u ne,11987s ;
| |
| | |
| 1 1
| |
| % %%)*.%
| |
| # N%%'.*
| |
| .w.v :
| |
| ,'/d'e%' ,%'/,'" * ,
| |
| 8%%*,8,,,1,P,*6 FC STEAM :
| |
| 4NfS GENERATOR i i
| |
| 1 YN's HISTORY ,
| |
| l I
| |
| i l
| |
| l i.v ,
| |
| 1 I
| |
| i l
| |
| l l
| |
| 1
| |
| __------_______m_m
| |
| | |
| ,,.,i, , , , , , .
| |
| iG STEAR ! GEN ERXTDRill MIS $R$
| |
| ldid9M
| |
| -M8dhtissflns pe:ctdEij,jii%ws1]
| |
| .= ..
| |
| Esbkh56tTubss:nRIU$gydf 161!!!982 l: 3-j:igniTUbesihs :>eeted,Mone1
| |
| !Ridgg;s:&
| |
| l i
| |
| | |
| WSTEAMiGEN ERAIDRsHISTORM sady8$
| |
| _7:5520;;Tu bssiins aecte6,jsXbah Identified,f51;]Tu bss Pliaggedi o 1984:: j f328%Tu be~sf;lnspeeted!,7 1N;o;i:lh e a k k q L5?Tu bes; Plugged,Profil6metrf; j T;on) ~12DlTub;es: .i i
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| , a
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| LCdSIEA WGEN ERXTOLRMIS$OR$
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| W M9l8l5i
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| - 032823 u bls sil n s psctedj,j#dkaikk
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| ;l d;e n tif ie d , M7)nu tss sjP;Tugg;eBI Y%ubss Pulled; u
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| ;o: A987)
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| -13235?Iu';tse;sil n s p e et ed (516i!!FU 112 isnythM24RThruWB;snW242%
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| !HcitMeg l Side) .witESta'nda ral j Bobbin Coil <
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| W.
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| C!STE4 mig lEN ERAIORMinIGR$
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| W 9871 M23511ns decstediisyf8sx@an Hot?De;gUnbe SHsetti
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| '-MTilns psoted%yTRPChtitM6t!
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| , Lbgflu bESheet?
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| - M83Tu bestPlugged!
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| 4-m:
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| Rows 2kU-lBeriWHeat!StrsssMelish
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| _ :;3dya s sfS l u Elfg sfa n c el (1.6Mb]O b s) ,
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| 1 1
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| t x -
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| MSIEAMj!GE'N ERATdR!!iHISl[ORM Lol M987 r- JHifunWa n&$ tea mi Drd mi ins ~p;eWti6;ni Meco nlib ry$A n n u l u sl:ll nls pectibn El;dw$ldtMhOtog;ral;ihn u
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| s '
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| sx MSIEAMMEN ERATORiliHISLTORM (6bDESCRIPTIDWOF1FAlEORET
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| -qRowj9LColu m nl5%anTogiondSRl
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| :Coldileg; Side"
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| ?36.O?Dey rss s :iC i'rdu m f e rs ritii;a lt r LBr;eakitat Toa ofsTSP
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| --Probab e;CausedisiFatiguet
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| :!AssistedFSt ress!Corros?idn1
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| ; Cracking:
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| 1
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| s' 'd :; MOD;Els!NG3DE iFALl6UREMEdnAWISMil
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| ;*;iArialyfical!:Model; L*TsstingfModel; L:
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| I
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| i i % .
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| \' , ,:? , . . . .
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| . ... . ... - . . . . . . .- - , . . . wn :..
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| l$s;Sili E %,, ,. .:. . .. ,.. . . . . . A '. . . . . ' -:@,
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| s-
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| - . . . . !. .~ .hCGENER iQRifflSKQ .
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| i 1
| |
| ROW 9 COLUMN 51 INSPECTIONS 9/79 Hot Leg Through 7th Cold Leg
| |
| ; Support Plate l 12/80 Hot Leg Through 7th Colc Leg Support Plate 5/84 Hot Leg Through 7th Support Plate 1
| |
| i i
| |
| l l
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| | |
| %!$!EEXLijneNERAkURilMididAMs sr , ,
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| ROW 9 COLUMN 51 INSPECTIONS (cont.)
| |
| 11/85 Hot Leg Through 7th -
| |
| Support Plate 4/87 Hot Leg Through 7th Support Plate l
| |
| | |
| - I I
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| e' e
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| k (ElDDnCURRE3MNinMlWM- O$S] ,
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| e m
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| :o BOBBIN COIL
| |
| - Sta ndard
| |
| - ID and OD Axial Indications
| |
| - Good Resolution for Axial Indications
| |
| ;o 8 x 1 PROBE
| |
| - ID and OD Circumferential Indications ,
| |
| - ID and OD Axial Indications
| |
| - Good to Fair Resolution
| |
| !o ROTATING PANCAKE COIL (RPC) PROBE l
| |
| - Similar to 8 x 1
| |
| - Slower S aeed !
| |
| | |
| I ''
| |
| !;$IERMMEiNEin%@R)
| |
| . llR$RFMWWNil!Rl!RN!
| |
| o ALL 3 STEAM GENERATORS
| |
| - Standard Bobbin of All Tube Portions Not inspected on Previous Inspection
| |
| - 8 x 1 of AII Hot Legs and Cold Legs -
| |
| Through 7 TSP
| |
| - RPC of All 8 x 1 Indications '
| |
| - Profilometry of Selected Intersections '
| |
| o C STEAM GENERATOR
| |
| - Technical Specifications Required Standard Bobbin Inspection, Tube Sheet to Tube S ,eet :
| |
| l L
| |
| | |
| n m's
| |
| .;EDDECORREb n p- lllN SBEClELOM RESULTS ~W DATE
| |
| .t Y
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| =
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| g..
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| -.a _ _ . . . . . . .
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| | |
| I a('4 x V ''
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| TUBE RETRIEVAMERGM
| |
| , n. .. SI.EXW.G EM.. ER.. AT. ..U. .n. i o: REMDVEER9tST HDevelopingjStabi ization;; Method;;
| |
| -ibbtainingnProfile;f:ofLoweLr?
| |
| intersections to:DetermineJExtent?
| |
| :of; Denting L
| |
| -LPlanL;to Pull Thru :ColdLL;eg; Tu belshe~et; a
| |
| fMM N M - - - - - - - ' - - ' ' - -
| |
| | |
| l tM .IUB EL RETRIEVAMEROM
| |
| \ CnSTEAMiiGEN ERATOR!
| |
| $Wils@PEREGRMMOMPI3ERE!
| |
| EVAliU ATibN?dF# TUB E!
| |
| - Determine: Fracture Mor ah0logyi
| |
| : Determine::Initicitionf!Siteg L (OD or.lD) ;
| |
| - Determine: Pro aogation,
| |
| ; Mechanism o MAY P.ULLLADDITION ALJTUBES2WITH.-
| |
| l SikILAR DEFECTS, IF IDENTIFIED y . - . - - . . .
| |
| | |
| s '
| |
| s 's x 'N ~' \ .' .
| |
| JU B E P AU.GGIM. . . . . . . . REAN; TdLPldy$11M25f%f(TschWpedh Requi~r emerit) o oPlugjAllLCorsfirmsdi4xialflridicationsi onTWhich! Death Can N6t. . . Be Dstermih'sdi .
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| a 1
| |
| 4:oTPlug All:ConfirrnedfCircumferential:;
| |
| : Indications 1
| |
| | |
| l
| |
| ' y' '
| |
| x .
| |
| ~~
| |
| RONGilOSIONS?
| |
| LOperlationst Res aons~sExce.Ilsnh iS d f et y[iS,y s te m s1Ple r fo r mi%sDesig ri et E m e rg ;e n c y hl m a le m e rit a t;io nlxce l le n t?
| |
| EventDWithin Design LBasissoffPlantii 1
| |
| | |
| 1, '-
| |
| '' \
| |
| % '\ .. . . . .... . . .. . . ,
| |
| !C.ONCISUSIO iS1 a
| |
| lEOTOR$iAQTIONS!
| |
| LN RChncu rrence Jodlustryli;B riefings; nom aletenedhnicalhEvalOOfioni MPR Independent Review; "
| |
| Mariagemeint Revie.wtPriorMoi n
| |
| , Sta rtlU. 3;
| |
| ~1 PRO'JECTED START U? 9/25/87
| |
| _ _ _ . A
| |
| | |
| 1 1
| |
| NAPS UFSAR 5.5-14 f
| |
| vary in depth by plus or minus 3/8 in, without affecting the weld. This is measured by a stainless steel dipstick.
| |
| h 1
| |
| 5.5.1.4.2.4 Weld Cross-Section Configuration. The higher the current or )
| |
| {
| |
| heat input and the lower the heat output, the greater the dilution of weld i i
| |
| 1 metal with base metal, causing a more round barrel-shaped configuration as '
| |
| compared to welding with less heat input and higher heat. output. This would cut the amount of dilution to provide a more narrow barrel-shaped' configuration. This is also a function of section thicknesses; the thinner g
| |
| the section, the more round the pattern that is produced. l 4
| |
| 1 5.5.1.4.3 Welder Qualification !
| |
| Welder qualification in accordance with ASME Code, Section IX, 1965, is required, using transverse side bend test specimens per Table Q.24.1.
| |
| 5.5.2 STEAM CENERATOR l
| |
| )
| |
| 5.5.2.1 Design Bases Steam generator design data are given in Table 5.5-3. The design sustains transient conditions given in Section 5.2.1. Estimates of $
| |
| radioactivity levels expected in the secondary side of the steam generators during normal operation, and the bases for the estimates, are given in Chapter 11.
| |
| The rupture of a steam-generator tube is discussed in Chapter 15.
| |
| The internal moisture separation equipment is designed to ensure that ;
| |
| moisture carryover does not exceed 0.25% by weight 'under the following conditions:
| |
| l 1.
| |
| Steady-state operation up to 100% of full-load steam flow, with water at the normal operating level.
| |
| 2.
| |
| Loading or unloading at a rate of 5% of full-power steam flow per min in the range from 15% to 100% of full-load steam flow.
| |
| g y -.f")-N$
| |
| | |
| NAPS UFSAR 5.5-15 ,
| |
| I
| |
| : 3. }
| |
| \ A step-load change of 10% of full power in the range from 15% to 100%
| |
| full-load steam flow. .
| |
| f The steam-generator tubesheet complex meets the stress limitations and $
| |
| fatigue criteria specified in the ASME Code, Section III, as well as emergen condition limitations specified in Section 5.1. j Codes and materials !
| |
| requirements of the steam-generator are given in Section 5,2.
| |
| I The steam generator design maximizes integrity against hydrodynamic !
| |
| excitation and failure of the tubes for plant life.
| |
| O q
| |
| a The water chemistry in the l necessary boron content reactor side is selected to provide the k for reactivity control and to minimize the corrosion i of reactor coolant system surf aces. i given in Table 5.5-4. The water chemistry of the steam side is 1 i
| |
| 5.5.2.2 Design Description
| |
| (
| |
| The steam generator shown in Figure 5.5-3 is a vertical shell and eU tub -
| |
| evaporator with integral moisture separating equipment.
| |
| The reactor coolant flows through the inverted U-tubes, entering and leaving throughz es .
| |
| rhe noz l {
| |
| located in the hemispherical bottom head of the steam generator.
| |
| The head is l divided into inlet and outlet chambers by a vertical partition plateending ext from the head to the tubesheet.
| |
| Manways are provided for access to both sides of the divided head.
| |
| Steam is generated on the shell side and flows upward through the moisture separators to the outlet nozzle at the top of th e vessel.
| |
| The unit is primarily carbon steel. ,
| |
| The heat-transfer tubes and the divider plate are Inconel and the interior surfaces of the reactor coolant channel i heads and nozzles are clad with austenitic stainless steel. side The primary of the tubesheet is weld clad with Inconel.
| |
| Feedwater flows from a feedring into the annulus formed by the shell and tube bundle wrapper before entering the boiler section of the ste am generator.
| |
| Subsequently, a water-steam mixture flows upward through the tube e and bundl I into the steam drum section. A set of centrifugal moisture separators, located above the tube bundle, removes most of the entrained water f rom the
| |
| * NAPS UFSAR 5.5-16 steam. Steam dryers increase the steam quality to a minimum of 99.75%
| |
| (0.25% moisture). The moisture separators recirculate water.that mixes with '
| |
| feedwater as it passes through the annulus formed by the shell and ' tube bundle wrapper.
| |
| The steam drum has two bolted and gasketed access openings for inspection and maintenance of the dryers, which can be (i.sassembled and removed through ;
| |
| the opening.
| |
| 5.5.2.3 Design Evaluation 3
| |
| 5.5.2.3.1 Forced Convection The limiting case for heat-transfer capability is the " nominal 100%
| |
| design" case. The steam generator ef fective heat-transf er coefficient is based on the coolant conditions of temperature and flow for this case, and includes a conservative allowance for tube fouling. Enough tube area is relected to ensure that the full design heat removal rate is achieved.
| |
| O 5.5.2.3.2 Natural-Circulation Flow The steam generators that provide a heat sink are at a higher elevation than the reactor. core, which is the heat source. Thus, natural circulation of reactor coolant is ensured for the removal of decay heat.
| |
| 5.5.2.3.3 Tube and Tubesheet Stress Analyses Tube and tubesheet stress analyses of the steam generator are given in Section 5.2.
| |
| Calculations confirm that the steam-generator tubesheet will withstand the loading (which is quasistatic rather than a shock loading) caused by a loss of reactor coolant.
| |
| I t
| |
| | |
| NArS UFSAR 5.5-17 I l
| |
| 1 5.5.2.3.4 Corrosion No significant general corrosion of the Inconel tubing is expected during the life of the unit. Corrosion tests show a worst-case rate of 15.0 mg/dm2 l in the 2000-hr test under simulated reactor coolant chemistry conditions. The conversion of this rate to a 40-year plant -life gives a corrosion loss of 1.3 x 10 ~3 in., which is insignificant compared to the minimum wall thickness.
| |
| Comparab3e tests with Inconel-600 exposed to simulated steam-generator !
| |
| water chemistry have shown equally low general corrosion rates. Testing to investigate the susceptibility of heat exchanger construction materials to stress corrosion in caustic and chloride aqueous solutions has indicated that inconel-600 has excellent resistance to general and pitting-type corrosion in severe operating water conditions, hence its selection for use in the steam generator.
| |
| ! 5.5.2.3.5 Flow-Induced Vibration i O In the design of Westinghouse steam generators, consideration has been given to the possibility of vibratory failure of tubes due to mechanical or flow-induced excitation. This consideration includes a detailed analysis of the tube supporting system as well as an extensive research program with tube i vibration model tests at the Westinghouse Research and Development Laboratories. .
| |
| 1 k
| |
| The major cause of tube vibratory failure in heat exchanger components is l that resulting from hydrodynamic excitation by the fluid outside the tube.
| |
| Consideration is given by Westinghouse to the following three regions where the possibility of flow-induced vibration may exist:
| |
| f l
| |
| 1.
| |
| At the entrance of downcomer feed to the tube bundle (cross flow).
| |
| l 2. Along the straight sections of the tube (parallel flow).
| |
| 3.
| |
| In the curved tube section of the U-bend (cross flow).
| |
| L From the description of these regions, it is noted that two types of flow exist: cross flow and parallel flow. {
| |
| For the case of parallel flow, analysis
| |
| 'l
| |
| | |
| ' l I
| |
| 1 is done to determine the vibratory deflections. The analysis of the '
| |
| steam-generator tubes indicates the flow velocities to be sufficiently below h that required for damaging fatigue or impacting vibratory amplitudes.
| |
| Therefore, the support system is deemed adequate to preclude par'allel-flow ,
| |
| excitation. For the case of cross-flow excitation, it is noted in the t literature that several techniques for the analysis of the tube vibration exist. The design problem is to ascertain that the tube natural frequency is {
| |
| well above the vortex shedding frequency. In order to avoid resonant l vibration, adequate tube supports are provided. l Because the problem of cross-flow-induced vibration was of major concern in the design of shell and tube heat exchangers, Westinghouse has given consideration to the experimental evaluation of the behavior of tube arrays under cross flow. While consideration was given to instrumentation of actual units in service, the hostile environment would limit the amount and quality {
| |
| t of information obtained therefrom. As a result, it was deemed prudent to undertake a research program that would allow the study of fluid elastic vibration behavior of tubes in arrays. A wind tunnel was built specifically for this purpose and Westinghouse has invested approximately 3 years of research into the study of this problem. The research facilities for the tube j vibration study have expanded with the construction of a water tunnel I facility..
| |
| The resultv of this research confirm the vortex shedding mechanism. More significant, hotever, is the evaluation of a fluid elastic mechanism not i
| |
| associated with vortex shedding. This is not commonly understood from the literature and could be a source of vibration failure. Westinghouse steam generators are evaluated on this basis in addition to the aforementioned l techniques and were found to be adequately designed. Testing has also been i
| |
| conducted using specific parameters of the steam generator and the results l show the support system to be adequate.
| |
| Summarizing the results of analysis and tests of steam-generator tubes for flow-induced vibration, it can be stated that a check of support adequacy has been made using all published techniques believed appropriate to heat exchanger tube support design. In addition, the tube support system is
| |
| | |
| NAPS UFSAR 5.5-19 i
| |
| t
| |
| .- consistent with accepted standards of heat exchanger design used throughout the industry (spacing, clearance, etc.). Furthermore, the design technique.
| |
| are supplemented with a continuing research and development program to f understand the complex mechanism of concern t
| |
| Service experience of Westinghouse PWR steam generators shows that flow-induced vibration and cavitation effects do not cause tube thinning.
| |
| Preliminary estimates of tube degradation from erosion / corrosion mechanisms indicate that approximately 2.5-mils wall thinning (2-mile primary, 0.5-mil secondary side) will result over the 40-year lifetime.
| |
| The effects of vibration, erosion, and cavitation have been given consideration, and the stress limitations for each category have been met.
| |
| The analysis of LOCA blowdown forces on as-f abricated U-tubes has shown that the maximum bending load elastic stress intensity is well below the faulted condition limit. The maximum bending load elastic stress intensity (based on the nominal tube wall thickness) would still be below the f aulted condition p limit. Therefore, as a minimum, at least 2.5-mils (per wall) thinning can be
| |
| !O tolerated without exceeding the allowable stress limits. Vibration effects are eliminated during normal operation by the supporting system. Under LOCA conditions, vibrat, ion is of a short duration and there is no endurance problem.
| |
| 5.5.2.4 Tests and Inspections The steam generator quality assurance program is given in Table 5.5-5.
| |
| I Radiographic inspection and acceptance standard are in accordance with the requirements of Section III of the ASME Code, 1968.
| |
| Liquid penetrant inspection is performed on weld-deposited tubesheet cladding, channel head cladding, tube-to-tubesheet weldments, and weld-deposited cladding. Liquid-penetrant inspection and ecceptance standards are in accordance with the requirements of Section III of the ASME Code, 1968.
| |
| O i l _ _ _ _ _ _ _ _
| |
| | |
| I l ,
| |
| NAPS UFSAR 5.5-20 i
| |
| Magnetic particle inspection is performed on the tubesheet forging, channel head casting, cozzle forgings, and the following weldments-
| |
| : 1. Nozzle to shell.
| |
| {
| |
| I 2. Support brackets.
| |
| : 3. Instrument connections (primary and secondary).
| |
| ]
| |
| : 4. Temporary attachments af ter removal.
| |
| : 5. All accessible pressure containing welds after hydrostatic test. )
| |
| Magnetic particle inspection and acceptance standards are in accordance with the requirements of Section III of the ASME Code, 1968. {
| |
| l An ultrasonic test is performed on the tubesheet forging, tubesheet cladding, secondary shell, and head plate and nozzle forgings.
| |
| i The heat-transfer tubing is subjected to eddy current test.
| |
| Hydrostatic tests were performed in accordance with Section III of the i ASME Code, 1968. In addition, the heat-transfer tubes are subjected to a
| |
| \
| |
| hydrostatic test pressure, before installation into the vessel, which is not I less than 1.25 times the primary-side design pressure multiplied by the ratio of the material allowable stress at the testing and design temperatures.
| |
| l j
| |
| Manways are to provide access to both the primary and secondary sides. j
| |
| ! A specific plan for inservice inspection of steam generator tubes is not cvailable. Because of the activity in the channel head and the large number of tubes involved, tube testing is done on a per plant basis. The extent of tube testing planner' in any particular plant will depend on tube performance to date, the channel head activity, and the results of tube sample testing.
| |
| j Inservice inspection of the steam-generator tubes is not planned at this time, '
| |
| but the eddy current testing method is available.
| |
| O .
| |
| | |
| , NAPS UFSAR 5.2-30 All parts of the reactor coolant pump in contact with the reactor coolant are austenitic stainless steel except for seals, bearings, and special parts. ) I
| |
| . 1
| |
| (
| |
| i The portions of the steam-generator in contact with the reactor coolant
| |
| ! water are weld overlay clad with austenitic stainless steel. The i
| |
| l steam-generator tubesheet is weld clad with Inconel; and the heat-transfer tubes are made of Inconel. Table 5.2-20 summarizes the materials of construction of these components. ]
| |
| The reactor coolant piping and fittings that make up the loops are austenitic stainless steel. All smaller piping that comprises part of the reac-tor coolant' system boundary, such as the pressurizer surge line, spray and relief lines, loop drains, and connecting lines to other systems are also austenitic stainless steel. '
| |
| 1 All valves in the reactor coolant system that are in contact with the coolant are constructed primarily of stainless steel. Other materials in l contact with the coolant, such as materials for hard surfacing and packing, are special materials.
| |
| (
| |
| The welding materials used for joining the ferritic-base materials of the reactor coolant boundary conform to or are equivalent to ASME Material Specifications SFA 5.1, 5.2, 5.5, 5.17, 5.18, and 5.20. They are tested and I qualified to the requirements of the ASME Code, Section III. The welding materials used for joining the austenitic stainless' steel base materials of the reactor coolant boundary conform to ASME Material Specifications SFA 5.4 and 5.9. They are tested and qualified according to the requirements stipulated in Section 5.2.5.
| |
| The welding materials used for joining nickel-chronium-iron alloy in similar base material combination and in dissimilar ferritic or austenitic base material combination of the reactor coolant boundary conform to ASME Material Specifications SFA 5.11 and 5.14. They are tested and qualified _to the requirements of ASME Code, Section III, and.are used only in procedures that have been qualified to these same rules.
| |
| J
| |
| | |
| NAPS UFSAR- 5.5-81' i Table 5.5-3 9 '
| |
| STEAM GENERATOR DESIGN DATA {
| |
| 1 Parameter Value
| |
| ' i I
| |
| Design pressure, reactor coolant side, psig 2485 -
| |
| Design pressure, steam side, psig 1085 l Design temperature, reactor coolant side, 'F 650- .
| |
| Design temperature, steam side, 'F 600 l Total heat transfer surface area, ft 51,500 l Maximum moisture carryover, we % 0.25 Overall height, ft-in. 67-8 Number of U-tubes 3388 l U-tube o.d., in.
| |
| 0.875 l Tube wall thickness, nominal, in. l 1
| |
| 0.050
| |
| ! Number of manways 4 !
| |
| i I.d. of manways, in. 16 ;
| |
| Number of handholes 2 1.d. of handholes, in. 6 i
| |
| Table 5.5-4 STEAM-GENERATOR WATER (STEAM-SIDE) CHEMISTRY. SPECIFICATION (BL l
| |
| l Parameter Value pH (normal operation) 25'c i.
| |
| 8.5 to 9.0 * '
| |
| Cation conductivity, maximum micrombos at 25'C 2.0
| |
| )
| |
| | |
| f 1
| |
| NAPS UFSAR 5.5-82 l / l 1
| |
| Table 5.5-5 1 STEAM GENERATOR QUALITY ASSURANCE PF0 GRAM 4! i l
| |
| Examination RT UT PT MT ET Tube sheet l
| |
| Forging Yes Yes Cladding Yes* Yes Channel head i ,, Casting Yes Yes ~
| |
| Cladding Yes Secondary shell and head Plates Yes i Tubes Yes Yes Nozzles (forgings) Yes Yes Weldments Shell, longitudinal I Yes Yes !
| |
| Shell, circumferential Yes Yes Cladding (channel head-tubesheet joint cladding restoration) Yes Steam and feedwater nozzle to shell Yes Yes K_ey: RT = radiographic UT = ultrasonic PT = dye penetrant MT = magnetic particle ET = eddy current I
| |
| a Flat surfaces only.
| |
| b Weld deposit areas only.
| |
| * NAPS UFSAR 5.5-83 Table 5.5-5 (continued)
| |
| STEAM GENERATOR QUALITY ASSURANCE PROGRAM Examination l RT UT PT MT , ET Weldments (continued)
| |
| Support brackets Yes Tube to tubesheet Yes Instrument connections (primary and secondary)- Yes Temporary attacha nts after removal Yes After hydrostatic te t (all welds and comple':e channel head, where accessible) Yes Nozzle safe ends Yes
| |
| ] (if forgings) Yes
| |
| .V Nozzle safe ends (if weld deposit) Yes Key: RT = radiographic
| |
| ; UT = ultrasonic PT = dye penetrant MT = magnetic particle ET = eddy current t
| |
| I
| |
| ^
| |
| | |
| I NAPS UFSAR 5.2-95 .
| |
| I Table 5.2-20 I REACTOR COOLANT PRESSURE BOUNDARY MATERIALS i
| |
| i Component _ Tm Reactor vessel components Shell and head plates (other SA 533 Grade A, B, or C; Class I or 2 ,
| |
| than core region) (vacuum treated)
| |
| Shell, flange and nozzle forgings SA 508 Class 2 or 3 )'
| |
| nozzle safe ends SA 182 Type F304 or F316 weld buildup e
| |
| CRDM appurtenances - upper head SB 166 or 167 and SA 182 Type F304 Instrumentation tube SB 166 or 167 and SA 182 Type F304, j F304L, or F316 l appurtenances - lower head Closure studs, nuts, and washers SA 540 Class 3 Grade B23 or B24 i Core support pads SB 166 with carbon less than 0.10% )
| |
| Monitor tubes and vent pipe SA 312 or 376 Type 304 or.316 or SB 167 Vessel supports, seal ledge SA 516 Grade 70 quenched and tempered or ,
| |
| D SA 533 Grade A, B, or C; Class I or 2 (vessel supports may be of weld metal
| |
| )
| |
| )
| |
| buildup of equivalent strength)
| |
| Cladding Stainless steel weld metal analysis A-7 and Ni-Cr-Fe weld metal F-Number 43 ,
| |
| 1 Heat lifting lugs SA 212 Grade B i
| |
| Steam-generator components Pressure plates SA 533 Grade A, B, or C; Class 1 or 2 Pressure forgings SA 508 Class 2 or 3 Nozzle safe ends Stainless steel weld metal analysis A-7 Channel heads SA 216 Grade WCC or SA 533 Grade A, B, or C; Class 1 or 2 Tubes SB 163 Ni-Cr-Fe, annealed Cladding Stainless steel weld metal analysis A-7 ,
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| and Ni-Cr-Fe weld metal F-Number 43
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| ( Closure bolting SA 540
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| NAPS.UFSAR Figure 5.5-3 STEAM OUTLET N0ZZLE UPPER HEAD I !
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| l l UPPER SHELL l
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| lC l
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| l l l
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| l l FEEDWATER RING --w A
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| l o
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| X ;- p'll l
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| [ TRANSITION COME J
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| WRAPPER TUBE % ,j ANT l-V IBRAT 10N BARS l,
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| l[ LOWER SHELL
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| ! , i f
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| I n - , ih TUBE SUPPORTS i
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| I I l
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| '' STAYR00 O TUBE SHEET SUPPORT PAD ann fif PRIMARY INLET N0ZZLE KANNEL HEAD 0 51 SERIES STEAM GENERATOR
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| | |
| I )
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| l FIGURE 2
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| )
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| Tube Support Plate /Tubesheet Nomenclature
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| = I
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| - -. - , l I
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| 8 8 2AVB SAVB 1AVB 4AVB 7H % -' -70 6H % \\ - \\I
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| - 6C
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| % - c7 7o l 5H 50 T'! r 4H ll lil .
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| 4C l l
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| 3H 3C
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| ~^
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| LK S 2H lll lll 2C L, ......... ,c ,
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| 1H l
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| \
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| M 2 . .4 8 . :' . ~~ ~~
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| HTS e,
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| 6, ,,.g. " 'g^ Westinghouse
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| ~
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| Series 51 Inlet i Outlet
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| __}}
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