ML13302B100
| ML13302B100 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/21/1981 |
| From: | Rood H Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8105070398 | |
| Download: ML13302B100 (23) | |
Text
APR sa mt APRt2i2 1981 Docket Nos.:
50-361/362 APPLICANTS: SOUTHERN CALIFORNIA EDISON COMPANY (SCE)
S* DIEGO GAS AND ELECTRIC MPANY (SDG&E)
FACILITY:
SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3
SUBJECT:
SUMMARY
OF SAN ONOFRE MEETING ON ICC PROCEDURE On April 10, 1981, members of the NRC staff met with representatives of the applicants, in Bethesda, Maryland, to discuss the above subject. Attendees at the meeting are given in Enclosure 1. The objective of the meeting was to discuss the San Onofre 2 and 3 ICC!procedure, and the modifications to it that the applicants plan to make in response to the staff's letter of March 6, 1981 to the CE Owner's Group (P. Check -o K. Baskin) regarding ICC procedure guidelines.
At the April 10, 1981 meeting, the applicants presented their proposed response to each of the issues raised in the March 6, 1981 letter. These are given in Enclosure 2 (a 4,morandum dated April 8, 1981).
Also discussed at the April 10, 1981 meeting was the,San -Onofre 2 and 3 ope,ating instruction S023-3-2.30, which is given :in "Enclosure 3. This ICC procedure reflects the proposed responses to the staff concerns that are described in Enclosure 2.
At the conclusion of the mee Jng, it was agreed that additional changes would be made in items 2b, 3, 4, and 7 of Enclosure 2, and that these changes would be reflected, as appropriate, in the ICC procedure. The applicants stated that he revised responses would be submitted for staff review by April 17, 1981.
Harry Rood, Project Manager Licensing Branch No. 3 Division of Licensing
Enclosures:
A stated cc:
See next page.
OFFICE DL: LB#3 D
- 3 SURNAME)
H.!ood jb FJ lia I
DATE)..4/f1/ 81...
.4.
81 1050 0 NRC ORM31o0oM0o240 OFFICIAL RECORD COPY USGPO:1980-329-82
MEETING SUMM4ARY DI STRI BUT ION
-Docket File G. Lear NRC POR Local PDR S. Pawlicki NSIC V. Boniaroya TERA Z. Rosztoczy LBli3 Reading W. Haass H. Denton D. ul leir E. Case R. Ballard D. Eisenhut W. Regan R. Purple R. Mattson B. J. Youngblood P. Check A. Schwencer F. Miraglia
- 0. Parr J. Miller F. Rosa G. Lainas W. Butler R. VolImer W. Kreger J. P. Knight R. Houston R. Bosnak T. Murphy F. Schauer L. Rubenstein R. E. Jackson T. Speis Project Manager H. Rood W. Johnston Attorney, OELD J. Stolz J.. Lee S. Hanauer OIE (3)
W. Gammi 11 ACRS (16)
T. Murley R. Tedesco T. Sch ruoder D. Skovhol t M. Ernst NRC
Participants:
R. Baer C. BerlI i nger J. Clifford K. Kniel Mark Rubin K.'i'l(K Brian Sheron Va n D. Tond i J. Kramer
.. Vassallo P. Collins
- 0. Ziemann E. Adensam bcc:
Applicant & Service List
S REG&(
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 APR 21 1981 Docket Nos,:
50-361/362 APPLICANTS: SOUTHERN CALIFORNIA EDISON COMPANY (SCE)
SAN DIEGO GAS AND ELECTRIC COMPANY (SDG&E)
FACILITY:
SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3
SUBJECT:
SUMMARY
OF SAN ONOFRE MEETING ON ICC PROCEDURE On April 10, 1981, members of the NRC staff met with representatives of the applicants, in Bethesda, Maryland, to discuss the above subject, Attendees at the meeting are given in Enclosure 1, The objective of the meeting was to discuss the San Onofre 2 and 3 ICC procedure, and the modifications to it that the applicants plan to make in response to the staff's letter of March 6, 1981 to the CE Owner's Group (P, Check to K. Baskin) regarding ICC procedure guidelines.
At the April 10, 1981 meeting, the applicants presented their proposed response to each of the issues raised in the March 6, 1981 letter, These are given in Enclosure 2 (a Memorandum dated April 8, 1981).
Also discussed at the April 10, 1981 meeting was the San Onofre 2 and 3 operating instruction S023-3-2,30, which is given in Enclosure 3. This ICC procedure reflects the proposed responses to the staff concerns that are described in Enclosure 2.
At the conclusion of the meeting, it was agreed that additional changes would be made in items 2b, 3, 4, and 7 of Enclosure 2, and that these changes would be reflected, as appropriate, in the ICC procedure, The applicants stated that the revised responses would be submitted for staff review by April 17, 1981.
Harry Rood, Project Manager Licensing Branch No. 3 Division of Licensing
Enclosures:
As stated cc; See next page,
Mr. Robert Dietch Vice President Southern California Edison Company 2244 Walnut Grove Avenue P. 0. Box 800 Rosemead, California 91770 Mr. D. W. Gilman Vice President -
Power Supply San Diego Gas & Electric Company 101 Ash Street P. 0. Box 1831 San Diego, California 92112 cc:
Charles R. Kocher, Esq.
James A. Beoletto, Esq.
Southern California Edison Company 2244 Walnut Grove Avenue P. 0. Box 800 Rosemead, California 91770 Chickering & Gregory ATTN:
David R. Pigott, Esq.
Counsel for San Diego Gas &.Electric Company &
Southern California Edison Company 3 Embarcadero Center -
23rd Floor San Francisco, California 94112 Mr. George Caravalho City Manager City of San Clemente 100 Avenido Presidio San Clemente, California 92672 Alan R. Watts, Esq.
Rourke & Woodruff Suite 1020 1055 North Main Street Santa Ana, California 92701 Lawrence Q. Garcia, Esq.
California Public Utilities Commission 5066 State Building San Francisco, California 94102 Mr. V. C. Hall Combustion Engineering, Incorporated 1000 Prospect Hill Road Windsor, Connecticut 06095
Mr. Robert Dietch
- 2 Mr. D. W. Gilman cc:
Mr. P. Dragolovich Bechtel Power Corporation P. 0. Box 60860, Terminal Annex Los Angeles, California 90060 Mr. Mark Medford Southern California Edison Company 2244 Walnut Grove Avenue P. 0. Box 800 Rosemead, California 91770 Henry Peters San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Ms. Lyn Harris Hicks Advocate for GUARD 3908 Calle Ariana San Clemente, California 92672 Richard J. Wharton, Esq.
Wharton & Pogalies University of San Diego School of Law Environmental Law Clinic San Diego, California 92110 Phyllis M. Gallagher, Esq.
Suite 222 1695 West Crescent Avenue Anaheim, California 92701 Mr. A. S. Carstens 2071 Caminito Circulo Norte Mt. La Jolla, California 92037 Resident Inspector, San Onofre/NPS c/o U. S. Nuclear Regulatory Commission P. 0. Box AA Oceanside, California 92054
ENCLOSURE 1 ATTENDEES 4-10-81 MEETING SAN ONOFRE 2/3 ICC PROCEDURE NAME ORGANIZATION H. Rood NRC-DL Gordon Bis'choff CE Van B. Fisher SCE Frederick R. Nandy SCE J. Clifford NRC-PTRB-DHFS Mark Rubin NRC-RSB Brian Sheron NRC-RSB
ENCLOSURE 2 SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 and 3 MEMORANDUM April 8, 1981 DETERMINATION OF ADEQUATE CORE COOLING RSB has reviewed S023-3-2.30 for operator response to an inadequate core cooling situation.
When resolutions for the following exceptions to the pro cedure are reached, RSB will find the procedure acceptable.
The purpose of this memorandum is to provide tentitive responses for each exception.
la.
More direct indications of ICC should be utilized in the section dealing with loss of heat sink.
Response: Section 6.1 of procedure has been revised to more clearly indicate that both Calculation Sheets are to be used.
Calculation Sheet 1 has also been revised to include more direct indication of actual ICC.
lb. Instruments such as core exit thermocouples and hot leg RTD's should be considered for use.
Response: Calculation Sheet 1 has been revised to improve the usage of Core Exit T/C and hot leg RTD's.
1c.
The ability of the Core Exit Thermocouples to identify the onset of ICC should be demonstrated.
Response: CE has performed studies to show that CET indications can be related to clad temperatures. Work is continuing in the area and will be reflected in the June submittal of CE Guidelines.
2a.
All criteria in the section dealing with loss of heat sink specified as symptoms of ICC rely on Steam Generator instrumentation.
Response: Section 6.1 of the procedure has been revised to more clearly indicate that both Calculation Sheets 1 and 2 are to be used to assess core cooling. Calculation Sheet 1 uses RCS parameters.
DETERMINATION OF ADEQUATE CORE COOLING April 8, 1981 Page -2 2b. It should be demonstrated by the applicant that no single failure could eliminate the operators capacity to identify the listed criteria for confirming ICC.
Response: All indications are from.safety grade instrumentation having redundant channels powered from separate power supplies ex cept the following:
RCP motor amps (only one indicator per pump).
Th & Tc (only one indicator is listed but redundant safety grade channels are available).
Incore T/C (displayed on Plant Monitoring Computer - computer has two power supplies and by using a volt meter and calibra tion data, the temperature can be obtained during a loss of computer event).
HPSI flow (only one indicator available per loop).
Main Steam Generator Feedwater Flow (only one control grade indi cator per S/G provided).
Auxiliary Feedwater Flow (only one safety grade indicator pro vided per steam generator but backup is provided by the redundant safety grade S/G level indicator).
Based on the above, indications are available to monitor each ICC criteria, in light of single failures.
2c.
Criteria #3 should be modified to include a minimum acceptable flow rate.
Response: Section 3.1 has been revised to include a minimum acceptable flow rate.
2d. Additional Criterion of system repressurization should be added unless justification for its omission is provided.
Response: Section 3.1.1 has been revised to add system repressurization as a possible indicator of loss of secondary heat sink.
NOTE:
System repressurization will not occur if a LOCA exists of sufficient size to adequately remove core heat.
DETERMINATION OF ADEQUATE CORE 0OLING April 8, 1981 Page -3
- 3. The procedures should be modified to include the criteria and recommenda tions for restarting the Reactor Coolant Pumps during an inadequate core cooling event.
Response: The LOCA procedure will be revised to add the following:
"If RCS cooldown and depressurization is impeded due to non condensible gasses blocking natural circulation, bump RCP's and use the reactor head vent to vent the gasses".
- 4. More specific directions should be provided regarding alternate pumping and/or piping arrangements available to supply emergency feedwater if the primary sources of auxiliary feedwater are not available.
Response: The emergency procedures have been revised to include alternate sources of emergency feedwater including use of the third emer gency feedwater pump and supplying the emergency feedwater suction tank with water from the fire main.
- 5. The criteria of Th not less than 620*F is considered unacceptable and should be revised.
Response: The procedure has been revised to indicate that Th greater than 620 0F is just a symptom of the approach to ICC and not an indi cator of actual ICC. The procedure has been further revised to indicate that the onset of clad damage must be assumed when more than 1 Core Exit T/C indicates greater than 1200'F.
- 6. A means by which the operators can vent steam from the primary system in order to depressurize it below the shutoff head of the HPSI pumps should
.be evaluated and the results forwarded to the NRC.
Response: The third emergency feedwater pump was added to provide re dundant supplies to maintain the secondary heat sink.
- 7. The need to specify a minimum cooldown rate under severe conditions in which significant core damage could occur if depressurization and-cool down did not proceed rapidly should be evaluated.
Response: The LOCA procedure will be revised to add the following:
In order to obtain maximum HPSI-flow and initiate discharge of the SIT's, the maximum rate of cooldown during post accident conditions does not apply above 475 0 F. Therefore, the RCS can be cooled down to 475'F and depressurized to the 50aF subcooled pressure as fast as possible. If the RCS is saturated, this will result in SIT discharge (saturation pressure for 475*F =
540 psia) and if the RCS is subcooled, this will result in near maximum HPSI flow (500F subcooled pressure for 4750F 850 psia).
VBF/sa
ENLOSURE 3 SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Revision 2 Page 1 of 3 DETERMINATION OF ADEQUATE CORE COOLING 1.0 OBJECTIVE 1.1 This instruction provides supplemental methodogy for determining the adequacy of the existing mode of core cooling, during post trip conditions.
1.2 This instruction provides supplemental methodogy for determining the adequacy of the existing mode of core cooling, during power operation with indications of a core anomaly.
1.3 This instruction provides requirements to ensure that the Watch Engineer is informed concerning the status of core cooling.
2.0 REFERENCES
2.1 The following Emergency Operating Instruction require utilization of this instruction:
2.1.1 5023-3-5.1, "Emergency Plant Shutdown".
2.1.2 S023-3-5.6, "Loss of Coolant Accident".
2.1.3 S023-3-5.9, "Steam Line Rupture".
2.1.4 S023-3-5.29, "Steam Generator Tube Rupture".
2.1.5 5023-3-5.30, "Loss of Feedwater".
2.2 The following provides the basis for this instruction.
2.2.1 CE Operational Guidance for Inadequate Core Cooling.
3.0 PREREQUISITES 3.1 A licensed operator not fulfilling a licensed operator requirement for the involved unit is available to perform this instruction.
3.2 Utilization of Check-Off Lists 1 and 2 is required per one or more of the referenced instructions during post trip conditions.
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Revision 2 Page 2 of 3 3.0 PREREQUISITES (Continued) 3.2 Utilization of Check-Off List 3 is required upon receipt of one or more of the following alarms or indications during power operations.
3.2.1 Nuclear Power Deviation Hi/Lo.
3.2.2 Pre-trip or trip alarms on DNBR or Local High Power Density.
3.2.3 Azimuthal Tilt.
3.2.4 Kw/ft. Power Operating Limit Margin.
3.2.5 DNB Operating Limit Margin.
3.2.6 Letdown Process Radiation High.
4.0 PRECAUTIONS 4.1 Operational decisions shall not be based solely on a single plant parameter or instrument when more than one confirmatory indication is available.
4.2 If high energy fluid is released into the containment, readings from non qualified instruments must be used with caution and verified against qualified instrument readings.
5.0 CHECK-OFF LISTS 5.1 Calculation Sheet 1, Reactor Coolant Performance Evaluation.
5.2 Calculation Sheet 2, Secondary Heat Sink Performance Evaluation.
5.3 Calculation Sheet 3, Localized DNB Evaluation.
6.0 PROCEDURE 6.1 Complete both Calculation Sheet 1, Reactor Coolant Performance Evaluation and Calculation Sheet 2, Secondary Heat Sink Performance Evaluation, when directed to determine the adequacy of core cooling per any one of the Emergency Operating Instructions listed below.
6.1.1 S023-3-5.1, "Emergency Plant Shutdown".
6.1.2 S023-3-5.6, "Loss of Coolant Accident".
6.1.3 S023-3-5.9, "Steam Line Rupture".
6.1.4 S023-3-5.29, "Steam Generator Tube Rupture".
6.1.5 S023-3-5.30, "Loss of Feedwater".
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Revision 2 Page 3 of 3 6.0 PROCEDURE (Continued) 6.2 The Calculation Sheets 1 and 2 contain the step by step procedure to collect and evaluate data, and contain requirements to inform the Watch Engineer of the status of core cooling.
6.3 Complete Calculation Sheet 3, Localized DNB Evluation, to evaluate core cooling due to receipt of one or more of the following anomaly indications.
6.3.1 Nuclear Power Deviation Hi/Lo.
6.3.2 DNBR Pre-trip or Trip Alarms.
6.3.3 LPD Pre-trip or Trip Alarms.
6.3.4 Azimuthal Tilt.
6.3.5 Kw/ft Power Operating Limit Margin.
6.3.6 DNB Operating Limit Margin.
6.3.7 Letdown Process High Radiation.
6.4 Calculation Sheet 3 contains the step by step procedure to collect initial and subsequent data, evaluate initial and subsequent data, and contains required notification actions if localized DNB is identified.
7.0 RECORDS 7.1 Place the completed Calculation Sheets in the Outage Package after review by the Watch Engineer.
8.0 ATTACHMENTS 8.1 Calculation Sheet 1 (4 pages).
8.2 Calculation Sheet 2 (3 pages).
8.3 Calculation Sheet 3 (4 pages).
H. E. MORGAN Superintendent Units 2 and 3 APPROVED:
J. M. CURRAN Plant Manager VBF/sa
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 1 Page 1 of 4 Revision 2 REACTOR COOLANT PERFORMANCE EVALUATION 1.0 SCOPE 1.1 The intent of this calculation and evaluation data sheet is to provide means to assess reactor cooling.
2.0 DATA COLLECTION 2.1 Record the following data:
2.1.1 RCS Subcooled Margin Initial 10 min.
20 min.
2.1.1.1 TI 911-1 2.1.1.2 TI 921-2 2.1.2 Pressurizer Pressure 2.1.2.1 PI 102-1 2.1.2.2 PI 102-2 2.1.2.3 PI 102-3 2.1.2.4 PI 102-4 2.1.3 Steam Generator #1 AP 2.1.3.1 PDI 0979-1.
2.1.3.2 PDI 0979-2 2.1.3.3 PDI 0979-3 2.1.3.4 PDI 0979-4 2.1.4 Steam Generator #2 AP 2.1.4.1 PDI 0978-1 2.1.4.2 PD1 0978-2 2.1.4.3 PD1 0978-3 2.1.4.4 PDI 0978-4
SAN ONOFRE NUCLEAR GE*ATING STATION OPERA@G INSTRUCTION 5023-3-2.30 UNITS 2 and 3 Check-Off List 1 Page 2 of 4 Revision 2 2.0 DATA COLLECTION (Continued)
Initial 10 min.
20 min.
2.1.5 RCP Motor Amps 2.1.5.1 P001 2.1.5.2 P002 2.1.5.3 P003 2.1.5.4 P004 2.1.6 RCS Temperature Loop 1 2.1.6.1 Th TR-0111 2.1.6.2 Tc TR-0115 2.1.6.3 AT (Th -
Tc) 2.1.7 RCS Temperatures Loop 2 2.1.7.1 Th TR-0121 2.1.7.2 T
TR-0125 2.1.7.3 AT (Th - Tcd 2.1.8 Incore Thermocouples (random 21% sample) 2.1.8.1 TRC 18 2.1.8.2 TRE 04 2.1.8.3 TRE 16 2.1.8.4 TRG 06 2.1.8.5 TRG 03 2.1.8.6 TRL 09_
2.1.8.7 TRL 13 2.1.8.8 TRR 09 2.1.8.9 TRR 16 2.1.8.10 TRT 06 2.1.8.11 TRT 18 2.1.8.12 TRW 12
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION SQ23-3-2.30 UNITS 2 and 3 Check-Off List 1 Page 3 of 4 Revision 2 2.0 DATA COLLECTION (Continued)
Initial 10 min.
20 min.
2.1.9 HPSI Flow 2.1.9.1 F1 0311-2 2.1.9.2 FI 0321-1 2.1.9.3 FI 0331-1 2.1.9.4 Fl 0341-2 3.0 INITIAL EVALUATION INITIALS 3.1 Compare the collected data to the following acceptance criteria:
3.1.1 RCS Subcooling greater than 100F.
3.1.2 Pressurizer pressure greater than 1900 psia.
3.1.3 If RCP's are running, Steam Generator AP greater than 20 psi.
3.1.4 RCP motor amp greater than 540 amps.
3.1.5 Th less than 620 0 F.
3.1.6 Tc less than'56 0 *F.
3.1.7 AT less than 60*F.
3.1.8 Incore T/C lessthan 6200F.
3.1.9 If pressurizer pressure is less than 1250 psia, HPSI flow greater than 415 gpm.
3.2 If one or more of the applicable parameters violates the accep tance criteria, immediately notify the Watch Engineer that an approach to inadequate core cooling (ICC) may exist, and advise the Watch Engineer that an independent verification should be made to confirm that all specified actions in Emergency Oper ating Instruction S023-3-5.1, "Emergency Plant Shutdown" are being carried out.
3.3 If an approach to ICC has been tentitively identified, use the following to positively identify an approach to ICC.
3.3.1 Th not decreasing with time.
3.3.2 If RCP's are running, RCP motor amps decreasing with time or indicating below 400 amps.
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 1 Page 4 of 4 Revision 2 3.0 INITIAL EVALUATION (Continued)
INITIALS 3.3.3 More than one incore T/C indicating superheat conditions.
NOTE:
If more than one incore T/C indicates super heat conditions, the core must be assumed to be uncovered.
3.3.1 If any of the random sample T/C's indicate super heat conditions, then print out all incore T/C's and all incore flux detectors.
3.4 If an approach to ICC has been postively identified, immediately notify the Watch Engineer and advise him that an independent ver ification should be made to confirm that all specified actions of Step 4.2 in Emergency Operating Instruction 5023-3-5.6, "Loss of Coolant Accident" are being carried out.
3.5 If more than one incore T/C indicates a temperature in excess of 1200 0F, immediately notify the Watch Engineer and advise him that ICC exists and that the onset of clad failure must be assumed.
3.6 If ICC has been identified, advise the Watch Engineer that an in dependent verification should be made to confirm that the Con tainment Integrity, Control Room Habitability and Radioactive Effluent Control safety functions are being carried out, per Emergency Operating Instruction S023-3-5.6, "Loss of Coolant Accident".
4.0 TREND EVALUATION 4.1 Take subsequent data at a 10 minute intervals until conditions stabilize within the acceptance criteria provided in Step 3.1, or until long term cooling is established.
4.2 Compare the collected data to the acceptance criteria provided in Step 3.1.
4.3 If a trend develops such that the acceptance criteria is being approached or violated, perform all applicable substeps in section 3.0.
Completed By Reviewed By:
FILE DISPOSITION:
Place in Outage Package after review.
VBF/sa
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 2 Page 1 of 3 Revision 2 SECONDARY HEAT SINK PERFORMANCE EVALUATION 1.0 SCOPE 1.1 The intent of this calculation and evaluation data sheet is to provide an independent means to assess secondary heat sink performance.
2.0 DATA COLLECTION 2.1 Record the following data:
Initial 10 min.
20 min.
2.1.1 Steam Generator #1 Wide Range Level 2.1.1-.1 LI 1115-1 2.1.1.2 LI 1115-2 2.1.2 Steam Generator #1 Pressure 2.1.2.1 PI 1013A1 2.1.2.2 PI 1013A2 2.1.2.3 PI 1013A3 2.1.2.4 PI 1013A4 2.1.3 Steam Generator #1 Feedwater Flow 2.1.3.1 FR 1011 2.1.3.2 F1 4725-2 2.1.4 Steam Generator #2 Wide Range Level 2.1.4.1 LI 1125-1 2.1.4.2 LI 1125-2 2.1.5 Steam Generator #2 Pressure 2.1.5.1 PI 1023A1 2.1.5.2 Pl 1023A2 2..1.5.3 PT 1023A3 2.1.5.4 PI 1023A4.
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 2 Page 2 of 3 Revision 2 2.0 DATA COLLECTION (Continued)
Initial 10 min.
20 min.
2.1.6 Steam Generator #2 Feedwater Flow 2.1.6.1 FR 1021 2.1.6.2 FI 4720-1 3.0 INITIAL EVALUATION INITIALS 3.1 Compare the collected data to the following acceptance criteria:
3.1.1 At least one steam generator indicates greater than 50% wide range level.
NOTE:
If loss of adequate level in both steam gen erators occurs, RCS temperature and pressure may increase and pressurizer level may rap idly increase when subcooling is lost.
3.1.2 Steam Generator pressure greater than Tc saturation pressure.
NOTE:
If steam generator pressure is less than the corresponding Tc saturation pressure, dry out.has occured.
3.1.2.1 If steam generator dry out is detected, advise the Watch Engineer that the provisions and limitations for refilling a dry steam generator contained in Emergency Operating Instruction 5023-3-5.30, "Loss of Feedwater" apply.
3.1.3 Main feedwater flow greater than 3% or auxiliary.
feedwater flow greater than 350 gpm.
NOTE:
If steam generator level is greater than 30% narrow range level, and is stable or increasing, the acceptance criteria for minimum feedwater flow does not apply.
3.2 If one or more of the applicable parameters violates the acceptance criteria, 'immediately notify the Watch Engineer that secondary heat sink performance may be degraded and advise the Watch Engineer that an independent verification should be made to confirm that all specified actions in Emergency Operating Instruction S023-3-5.1, "Emergency Plant Shutdown" are being carried out.
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 2 Page 3 of 3 Revision 2 3.0 INITIAL EVALUATION (Continued)
INITIALS 3.3 If degraded secondary heat sink performance has been tentitively identified, use the following criteria to positively identify an approach to loss of secondary heat sink.
3.3.1 Wide range level not increasing with time.
3.3.2 Steam Generator pressure below T saturation pressure.
3.3.3 No feedwater flow to either steam generator.
3.4 If an approach to loss of secondary heat sink has been posi tively identified, immediately notify the Watch Engineer and advise him that an independent verification should be made to confirm that all specified actions in Emergency Operating Instruction S023-3-5.30, "Less of Feedwater" are being carried out.
4.0 TREND EVALUATION 4.1 Take subsequent data at 10 minute intervals until conditions stabilize within the acceptance criteria provided in Step 3.1, or until long term cooling has been established.
4.2 Compare the collected data to the acceptance criteria provided in Step 3.1.
4.3 If a trend develops such that the acceptance criteria is being approached or violated, perform all applicable substeps in section 3.0.
Completed By:
Reviewed By:
FILE DISPOSITION:
Place in outage package after review.
VBF/sa
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 3 Page 1 of 4 Rtevislon 2 LOCALIZED DNB EVALUATION 1.0 SCOPE 1.1 The intent of this calculation and evaluation data sheet is to provide an independent means to assess localized DNB.
2.0 DATA COLLECTION 2.1 Record the following data:
Initial 10 min.
20 min.
2.1.1 Azimuthal Tilt 2.1.2 LPD Margin 2.1.2.1 Ch A 2.1.2.2 Ch B 2.1.2.3 Ch C.
2.1.
- 2. 4 Ch D 2.1.3 DNBR Margin 2.1.3.1 Ch A 2.1.3.2 Ch B 2.1.3.3 Ch C 2.1.3.4 Ch D 2.1.4 Incore Thermocouples 2.1.4.1 Display the following incore T/C's and record below any T/C's which exceed RCS hot leg temperature by more than 200F.
2.1.4.1.1 TRA 08, 14 2.1.4.1.2 TRC 04, 06, 09, 13, 18 2.1.4.1.3 TRE 02, 04, 06, 09, 13, 18, 20
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 3 Pag 2 of 4 Revision 2 2.0 DATA COLLECTION (Continued) 2.1.4.1.4 TRG 02, 04, 06, 09, 13, 18, 20 2.1.4.1.5 TRL 02, 04, 06, 09, 13, 18, 20 2.1.4.1.6 TRR 02, 04, 06, 09, 13, 18, 20 2.1.4.1.7 TRT 02, 04, 06, 09, 13, 18, 20 2.1.4.1.8 TRW 04, 06, 09, 13, 18 2.1.4.1.9 TRY 08, 14 2.1.4.2 Initial 10 min.
20 mtn.
ID Value ID Value ID Value 2.1.5 Incore flux Monitors.
2.1.5.1 Record the following SPND's and record below any SPND's which exceed the planar average by more than 1.35.
2.1.5.1.1 CSAO8, 14 (-1 to -5)
CSC 04, 06, 09, 13, 16, 18 (-1 to -5)
CSE 02, 04, 06, 09, 13, 16, 18, 20 (-1 to -5)
CSG 02, 04, 06, 09, 13, 16, 18, 20 (-1 to -5)
CSL 02, 04, 06, 09, 13, 16, 18, 20 (-1 to -5)
CSR 02, 04, 06, 09, 13, 16, 18, 20 (-1 to -5)
CST 02, 04, 06, 09, 13, 16, 18, 20 (-1 to -5)
CSW 04, 06, 09, 13, 16, 18 (-1 to -5)
CSY 08, 14 (-1 to -5)
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTION S023-3-2.30 UNITS 2 and 3 Check-Off List 3 Page 3 of 4 Revision 2 2.0 DATA COLLECTION (Continued)
Initial 10 min.
20 min.
ID Value ID Value ID Value 3.0 INITIAL EVALUATION 3.1 If the collected data violates the following acceptance criteria, immediately notify the Watch Engineer that localized DNB may exist, and advise the Watch Engineer that reactor power should be reduced, until the data falls within the acceptance criteria.
3.1.1 Azimuthal Tilt less than.01.
3.1.2 LPD Margin greater than (later).
3.1.3 DNBR Margin greater than (later).
3.1.4 Incore 7/C less than 58*F above h*
3.1.5 Incore SPND's less than 2 times the midplane average.
4.0 TREND EVALUATION 4.1 Take subsequent data at 10 min. intervals until conditions stabilize within the acceptance criteria.
SAN ONOFRE NUCLEAR GENERATING STATION OPERATING INSTRUCTIONSO23-3-2.30 UNITS 2 and 3 Check-Off List 3 Page 4 of 4 Revision 2 4.0 TREND EVALUATION (Continued) 4.2 If a trend develops such that the acceptance criteria is being approached or violated, immediately-notify the Watch Engineer.
Completed By:
Reviewed By:
FILE DISPOSITION:
Place in outage package after review.
'BF/sa