ML021050038
| ML021050038 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/02/2002 |
| From: | Barron H Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML021050038 (29) | |
Text
Duke Energy.
Duke Energy Corporation McGuire Nuclear Station 12700 Hagers Ferry Road Huntersville, NC 28078-9340 (704) 875-4800 OFFICE (704) 875-4809 FAX H. B. Barron Vice President April 2, 2002 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Re:
McGuire Nuclear Station Unit 1 Docket No.
50-369 McGuire Nuclear Station Unit 2 Docket No.
50-370 Changes to Emergency Plan Implementing Procedures Attached to this letter are a revised Emergency Plan Implementing Procedure (EPIP) Index, a notice to delete three (3)
Emergency Plan Implementing Procedures, and a notice of revision to one (1)
Emergency Plan Implementing Procedure.
These procedure deletions and revision were evaluated pursuant to the requirements of 10 CFR 50.54 (q).
These changes do not constitute a reduction in the effectiveness of the emergency plan and the plan continues to meet the requirements of 10 CFR 50.47 (b) and 10 CFR 50 Appendix E.
Duke implemented these changes on March 13, 2002.
A copy of these changes is also being sent to the NRC Office of Nuclear Material Safety and Safeguards as per 10 CFR 72.44 (f).
Revision bars in the procedure index indicate the procedure deletions. The following procedure index changes, deletions, and revision have been implemented:
EPIP Index Page 1 Dated 3/13/2002 EPIP Index Page 2 Dated 3/13/2002 EPIP Index Page 3 Dated 3/13/2002 DELETE the following three procedures:
OP/0/B/6200/090 Dated 3/13/2002 HP/l/B/1009/015 Dated 3/13/2002 HP/2/B/1009/015 Dated 3/13/2002 REVISION to the following procedure:
RP/0/A/5700/019 Dated 3/13/2002 (revision 04)
There are no new regulatory commitments in this document.
Duke is also supplying two copies of this submittal to the Regional Administrator of Region II.
Questions on this document should be directed to Kevin Murray at (704) 875-4672.
Very truly yours,
/t/i1 'W H.
B.
Barron Attachments
/4 095
U.S. Nuclear Regulatory Commission April 2, 2002 Page 2 xc:
(w/attachment)
Mr.
Luis Reyes, Regional Administrator U.S. Nuclear Regulatory Commission Region II 61 Forsyth St.,
SW, Suite 23T85 Atlanta, Georgia 30303 (w/attachment)
Mr. Martin J. Virgilio, Director Office of Nuclear Material Safety and Safeguards Mail Stop T-8A23 Washington, D.C. 20555-0001 R.
E. Martin, USNRC U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D.C.
20555 (w/o attachment)
NRC Resident Inspector McGuire Nuclear Station E.M. Kuhr (EC050)
M.T. Cash, Manager NRIA (EC050)
Electronic Licensing Library (EC050)
EP File I11
DUKE McGUIRE NUCLEAR SITE EMERGENCY PLAN IMPLEMENTING PROCEDURES APPROVED:
ýW6WdA64,,,
SA4TYASSURANCE MANAGER DATE APPROVED EPIP Index Page EPIP Index Page EPIP Index Page 1
2 3
Dated Dated Dated 3/13/2002 3/13/2002 3/13/2002 DELETE the following three procedures:
OP/o/B/6200/090 HP/I/B/1009/015 HP/2/B/1009/015 Dated Dated Dated 3/13/2002 3/13/2002 3/13/2002 REVISION to the following procedure:
RP/0/A/5700/019, (Revision 04)
Dated 3/13/2002
EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX PROCEDURE #
RP/0/A/5700/000 RP/0/A/5700/001 RP/0/A15700/002 RP/0/A/5700/003 RP/0/A/5700/004 RP/0/A/5700/05 RP/0/A/5700/006 RP/0/A/5700/007 RP/O/A/5700/008 RP/0/A/5700/009 RP/0/A/5700/010 RP/0/A/5700/011 RP/0/A/5700/012 RP/0/A/5700/013 RP/0/A/5700/14 RP/0/A/5700/015 RP/0/A/5700/16 RP/O/A/5700/17 RP/0/A/5700/018 RP/0/A/5700/019 RP/0/A/5700/020 RP/0/A/5700/21 RP/0/A/5700/022 RP/0/A/5700/024 RP/0/A/5700/026 RP/0/B/5700/023 OP/O/B/6200/090 TITLE Classification of Emergency Notification of Unusual Event Alert Site Area Emergency General Emergency Care and Transportation of Contaminated Injured Individual(s) From Site to Offsite Medical Facility Natural Disasters Earthquake Release of Toxic or Flammable Gases Collisions/Explosions NRC Immediate Notification Requirements Conducting a Site Assembly, Site Evacuation or Containment Evacuation Activation of the Technical Support Center (TSC)
Activation of the Emergency Operations Facility (EOF)
Emergency Telephone Directory Notifications to the State and Counties from the EOF EOF Commodities and Facilities Procedure Emergency Data Transmittal System Access Notifications to the State and Counties from the TSC Core Damage Assessment Activation of the Operations Support Center (OSC)
EOF Access Control Spill Response Procedure Recovery and Reentry Procedure Operations/Engineering Technical Evaluations in the Technical Support Center (TSC)
Community Relations Emergency Response Plan PALSS Operation for Accident Sampling March 13, 2002 Rev. 27 REVISION NUMBER Rev. 008 Rev. 016 Rev. 016 Rev. 016 Rev. 016 DELETE Rev. 009 Rev. 007 Rev. 004 Rev. 001 Rev. 013 Rev. 005 Rev. 019 DELETE DELETE DELETE DELETE DELETE Rev. 010 Rev. 004 Rev. 011 DELETE Rev. 009 Rev. 002 Rev. 002 Rev. 002 DELETED I
I I
EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX PROCEDURE #
HP/0/B/1009/002 HP/O/B/1009/003 HP/0/B/1009/05 HP/0/B/1009/006 HP/O/B/1009/010 HP/l/B/1009/015 HP/2/B/1009/015 HP/0/B/1009/016 HP/0/B/1009/020 HP/0/B/1009/021 HP/0/B/1009/022 HP/0/B/1009/023 HIP/O/B/1009/024 HP/O/B/1009/029 SH/0/B/2005/0 1 SH/0/B/2005/002 SR/0/B/2000/01 SR/0/B/2000/002 SR/O/B/2000/003 SR/O/B/2000/004 TITLE Alternative Method for Determining Dose Rate Within the Reactor Building Recovery Plan Initial Evaluation of Protective Action Guides Due to Abnormal Plant Conditions Procedure for Quantifying High Level Radioactivity Releases During Accident Conditions Releases of Radioactive Effluents Exceeding Selected Licensee Commitments Unit 1 Nuclear Post-Accident Containment Air Sampling System Operating Procedure Unit 2 Nuclear Post-Accident Containment Air Sampling System Operating Procedure Distribution of Potassium Iodide Tablets in the Event of a Radioiodine Release Manual Procedure for Offsite Dose Projections Estimating Food Chain Doses Under Post-Accident Conditions Accident and Emergency Response Environmental Monitoring for Emergency Conditions Personnel Monitoring for Emergency Conditions Initial Response On-Shift Dose Assessment Emergency Response Offsite Dose Projections Protocol for the Field Monitoring Coordinator During Emergency Conditions Standard Procedure for Public Affairs Response to the Emergency Operations Facility Standard Procedure for EOF Commodities and Facilities Activation of the Emergency Operations Facility Notification to States and Counties from the Emergency Operations Facility March 13, 2002 Rev. 27 REVISION NUMBER Rev. 002 Rev. 003 DELETED Rev. 005 Rev. 006 DELETED DELETED Rev. 002 DELETED Rev. 001 Rev. 003 Rev. 004 Rev. 001 Rev. 005 Rev. 001 Rev. 002 Rev. 003 Rev. 002 Rev. 008 Rev. 004 2
EMERGENCY PLAN IMPLEMENTING PROCEDURES INDEX PROCEDURE #
McGuire Site Directive 280 EP Group Manual MNS RP Manual:
PT/0/A/4600/088 TITLE Site Assembly/Accountability and Evacuation/Containment Evacuation Section 1.1 Emergency Organization Section 18.1 Accident and Emergency Response Section 18.2 Environmental Monitoring for Emergency Conditions Section 18.3 Personnel Monitoring for Emergency Conditions Section 18.4 Planned Emergency Exposure Functional Check of Emergency Vehicle and Equipment March 13, 2002 Rev. 27 REVISION NUMBER DELETED Rev. 017 DELETED DELETED DELETED DELETED Rev. 007 3
(R04-01)
Duke Power Company PROCEDURE PROCESS RECORD (1) ID No. RP/0/A/5700/019 Revision No. 004 PREPARATION (2) Station MCGUIRE NUCLEAR STATION (3) Procedure Title Core Damage Assessment (4) Prepared By Date______
(5) Requires NSD 228 Applicability Determination?
Li Yes (New procedure or revision with major changes)
El No (Revision with minor changes)
[I No (To incorporate r ously approved changes)
(6) Reviewed By e-(QR)
Date 5/O0 Cross-Disciplinar;<eview By (QR)
Date a
Reactivity Mgmt. Review By (QR)
NA Date
______2 Mgmt. Involvement Review By (Ops Supt.) g Date 3/oIdZ (7) Additional ReviewsQ, Reviewed By Q
,I-Date
! i/
Reviewed By Date (8) Temporary Approval (if necessary)
By (OSM/QR)
Date By (QR)
Date (9) Approved By i5.
Date3 PERFORMANCE (Compar with Control Copy every 14 calendar days while work is being performed.)
(10) Compared with Control Copy Date Compared with Control Copy Date Compared with Control Copy Date (11) Date(s) Performed Work Order Number (WO#)
COMPLETION (12) Procedure Completion Verification El Yes El NA Check lists and/or blanks initialed, signed, dated, or filled in NA, as appropriate?
El Yes 0l NA Required enclosures attached?
E0 Yes El NA Data sheets attached, completed, dated, and signed?
El Yes El NA Charts, graphs, etc. attached dated, identified, and marked?
El Yes El NA Procedure requirements met?
Verified By Date (13) Procedure Completion Approved Date (14)Remarks (Attach additional pages, if necessary)
Duke Power Company McGuire Nuclear Station Core Damage Assessment Multiple Use Procedure No.
RP/O/A/5700/019 Revision No.
004 Electronic Reference No.
MC0048MM
RP/O/A/5700/019 Page 2 of 11 Core Damage Assessment
- 1.
Symptoms NOTE:
This procedure will normally be performed by Nuclear Engineers while in the Technical Support Center (TSC) to provide a means of determining the status of the core based on various parameters.
1.1 1(2) EMF 51, "Containment Radiation Monitor" in alarm.
1.2 High Core Exit Thermocouple (CET) readings.
1.3 Low Reactor Vessel Level Indication System (RVLIS) levels.
1.4 Any condition in which failed fuel is suspected.
- 2. Immediate Actions None
- 3. Subsequent Actions NOTE:
- 1. Perform each of the following sections as data is available.
- 2. Test Coordinator may elect to perform some or all of the following sections in the order that best suits the situation.
- 3. Since the plant status may change during the performance of this procedure it may be desirable to repeat certain parts of this procedure.
- 4. Results of the various instrument readings are estimates. The results from different methods may disagree.
- 5. Some of the key OAC points are listed in Enclosure 4.5.
3.1 Initial Estimate of Core Conditions 3.1.1 Figure 1 and Figure 2 (and associated tables).
Figure 1 Containment Radiation Level vs. Time for RCS Release RP/O/A/5700/O1 9 Page 3 of 11 1 1 1 1 1 1 1 1 1 1 1 f "r as
- ibl 14]);lg No r
(;Jri
- ny ag T ``.
200 300 400 Time After Shutdown (hrs) 500 600 700 Containment Radiation Level vs.
r 1 Time for RCS Release Time After Containment Dose Rate Shutdown (hrs)
(Rad/hr) 0.5 9.1808 1
8.8621 2
8.3792 8
7.0574 16 6.2611 24 5.7672 100 3.8545 240 2.3002 720 0.41169 I.OOE+01 "a)
I1.00E+00 1.00E-01
-I 0
100 800
RP/O/A/5700/019 Page 4 of 11 Figure 2 Containment Radiation Level vs. Time for 1% Fuel Overtemperature Release 100 200 300 400 Time After Shutdown (Hrs) 500 600 700
$- RCS Pressure < 1600 PSIG w/o Spray (Rad/Hr) -- W-RCS Pressure > 1600 PSIG w/o Spray (Rad/Hr)
A RCS Pressure > 1600 PSIG w/ Spray (Rad/Hr) ----
RCS Pressure < 1600 PSIG w/ Spray (Rad/Hr)
Figure 2 1 Containment Radiation Level vs. Time for 1% Fuel I_
Overtemperature Release Time After RCS Pressure RCS Pressure RCS Pressure RCS Pressure
<1600 PSIG
> 1600 PSIG
>1600 PSIG
<1600 PSIG Shudw w/o Spray w/o Spray w/Spray w/ Spray (hrs)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr) 0.5 5.30E+03 2.86E+03 2.19E+03 5.03E+03 1
4.39E+03 2.37E+03 1.82E+03 4.17E+03 2
3.16E+03 1.71E+03 1.29E+03 2.99E+03 8
1.13E+03 6.33E+02 4.20E+02 1.05E+03 16 6.45E+02 3.71E+02 2.21E+02 5.88E+02 24 5.15E+02 2.98E+02 1.78E+02 4.72E+02 100 2.90E+02 1.70E+02 1.04E+02 2.73E+02 240 1.54E+02 9.56E+01 4.88E+01 1.43E+02 720 3.49E+01 2.68E+01 4.37E+00 3.08E+01 L
1.OOE+04 1.OOE+03 1.OOE+02 U
1.OOE+01 1.OOE+00 0
800 I
RP/O/A/5700/019 Page 5 of 11 3.1.2 Complete the following table based on Current Plant Data (OAC PID's can be found on Enclosure 4.5).
3.1.3 3.1.4 Use the above data in 3.1.2 the figures in 3.1.1 to determine the condition that best represents the core.
No Core Cladding Fuel Overtemp Damage Damage Damage CET OF
< 700
< 2000
>1400 RVLIS (%)
> 55
< 55 to 40
< 40 L EMF (R/hr)
< Figure 1
< Figure 2
> Figure 2 Condition Best Representing Core Date/Time IF 'NO CORE DAMAGE' is found, exit this procedure and continue to monitor plant conditions. If conditions warrant, re-run this procedure at any time.
IF 'CLADDING DAMAGE' is found, proceed to Step 3.2.
IF 'FUEL OVERTEMPERATURE DAMAGE' is found, proceed to Step 3.3.
3.2 Clad Damage Assessment NOTE:
EMF 51 may not be useful to assess core damage for containment bypass sequences (e.g.,
S/G tube ruptures).
3.2.1 Record 1(2)EMF 51 reading:
R/hr at hours after shutdown.
Containment Spray:
RCS Pressure:
D ON 0 OFF psig Time of data, (mm/dd/yy hh:mm)
Time of Reactor Shutdown, To Hours Core Exit Thermocouples (CET)
Deg F RVLIS Lower Range Containment Radiation Monitors R/hr l(2)EMF 51A or B OAC PID A0829 or A0835
RP/O/A/5700/019 Page 6 of 11 3.2.2 Figure 3 (and associated table)
Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage Release 100 200 300 400 500 600 700 Time After Shutdown (Hrs)
RCS Pressure < 1600 PSIG w/o Spray (Rad/Hr) ----
RCS Pressure > 1600 PSIG w/o Spray (Rad/fHr)
RCS Pressure < 1600 PSIG w/ Spray (Rad/Hr)
Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage Release RCS Pressure RCS Pressure RCS Pressure RCS Pressure Shutdown
<1600 PSIG
> 1600 PSIG
> 1600 PSIG
<1600 PSIG Shudw w/o Spray w/o Spray w/ Spray w/ Spray (hrs)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr) 0.5 3.37E+04 1.15E+04 1.10E+04 2.91E+04 1
2.78E+04 9.53E+03 9.14E+03 2.42E+04 2
2.03E+04 6.80E+03 6.51 E+03 1.75E+04 8
8.09E+03 2.28E+03 2.13E+03 6.26E+03 16 4.96E+03 1.23E+03 1.12E+03 3.48E+03 24 3.98E+03 9.90E+02 9.02E+02 2.68E+03 100 2.19E+03 5.75E+02 5.28EE+02 1.30E+03 240 1.29E+03 2.81 E+02 2.49E+02 6.28E+02 720 4.22E+02 3.96E+01 2.43E+01 6.88E+01 100000 10000 1000
.0 U
100 10 0
800
RP/0/A/5700/019 Page 7 of 11 3.2.3 Using RCS Pressure from Step 3.2.1 and Figure 3 (and associated table), find the "Predicted Containment Radiation Level at 100% Clad Damage" Predicted Containment Radiation Level at 100% Clad Damage:
% Clad Damage =Current Containment Radiation Level Predicted Containment Radiation Level for 100% Clad Damage 3.2.4
% Clad Damage x100 =
3.2.5 Estimate Fuel Clad Damage based on Core Exit Thermocouple Readings.
A. Complete Enclosure 4.1.
B.
RCS Pressure GREATER THAN 1600 PSIG (If Applicable):
Number of CET's > 1400*F from Enclosure 4.1:
CET's Number of Operable CET's from Enclosure 4.1:
CET's
" Clad Damage = [ Number of CET' s > 1400°F ]x 100 Total Operable CET's
"% Clad Damage =
x100=
____C. RCS Pressure LESS THAN 1600 PSIG (If Applicable):
Number of CET's > 1200'F from Enclosure 4.1:
CET's Number of Operable CET's from Enclosure 4.1:
CET's
% Clad Damage = [.Number of CET's > 1200°F ]x100 Total Operable CET's
% Clad Damage =
x100=
3.2.6 Record Results From Steps 3.2.4 and 3.2.5 onto Enclosure 4.4.
3.2.7 Confirm Reasonableness of Clad Damage Estimates using Expected Responses below. Record any Comments on Enclosure 4.4.
RP/O/A/5700/019 Page 8 of 11 A. Expected Responses
"* RVLIS Less Than 53.5 % and Greater Than 39 %
"* Hot Leg RTD Greater Than Tsat and Less Than 650'F
"* Source Range Monitor Greater Than le+5 cps
"* Difference in clad damage estimates from Containment Radiation Monitors (CRM) and CET's less than 50%, using:
ABS[ %CladDamage * - %CladDamage 1r L
%CladDamagec-
]
B.
If the expected response is not obtained, determine if the deviation can be explained from the accident progression
- Injection of water to the RCS
- Bleed Paths from the RCS
- Direct radiation to the containment radiation monitors or If the expected response is not obtained, determine if the deviation can be explained from conservatism in the predictive model
'fuel burnup
'fission product retention in the RCS
'fission product removal from containment 3.2.8 IF time permits, complete Enclosure 4.2, Source Range Detector Response.
3.2.9 IF time permits, complete Enclosure 4.3, Movable Detector Surveillance in an Accident Condition 3.3 Fuel Overtemperature Damage Assessment 3.3.1 Record 1 (2)EMF 51 reading:
R/hr at hours after shutdown.
Containment Spray:
L] ON 0i OFF RCS Pressure:
psig
RP/O/A/5700/019 Page 9 of 11 3.3.2 Figure 4 (and associated table)
Figure 4 Containment Radiation Level vs. Time for 100% Fuel Overtemperature Release 0
100 200 300 400 500 600 700 Time After Shutdown (Hrs)
RCS Pressure < 1600 PSIG wlo Spray (Rad/Hr) -- U-- RCS Pressure > 1600 PSIG w/o Spray (Rad/Hr)
A,-RCS Pressure> 1600 PSIG w/ Spray (Rad/-r) ----
RCS Pressure < 1600 PSIG w/ Spray (Rad/Hr)
Containment Radiation Level vs. Time for 100% Fuel Overtemperature Release Time RCS Pressure RCS Pressure RCS Pressure RCS Pressure After
> 1600 PSIG < 1600 PSIG Shutdown w/o Spray w/o Spray w/ Spray w/Spray (hrs)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr)
(Rad/Hr) 0.5 5.30E+05 2.86E+05 2.19E+05 5.03E+05 1
4.39E+05 2.37E+05 1.82E+05 4.17E+05 2
3.16E+05 1.71 E+05 1.29E+05 2.99E+05 8
1.13E+05 6.33E+04 4.20E+04 1.05E+05 16 6.45E+04 3.71E+04 2.21E+04 5.88E+04 24 5.15E+04 2.98E+04 1.78E+04 4.72E+04 100 2.90E+04 1.70E+04 1.04E+04 2.73E+04 240 1.54E+04 9.56E+03 4.88E+03 1.43E+04 720 3.49E+03 2.68E+03 4.37E+02 3.08E+03 1.00E+06 1.00E+05 1.00E+04 1 00E+03 1.OOE+02 800
RP/0/AI5700/019 Page 10 of 11 3.3.3 Using RCS Pressure from Step 3.3.1 and Figure 4 (and associated table), find the "Predicted Containment Radiation Level vs. Time for 100% Fuel Overtemperature Release" Predicted Containment Radiation Level For 100% Fuel OT Release:
% Core Damage =
Current Containment Radiation Level Predicted Containment Radiation Level for 100% OT Release 3.3.4
% Core Damage =
-xlO0 =
3.3.5 Estimate Fuel OT Damage based on Core Exit Thermocouple Readings.
A. Complete Enclosure 4.1.
B. Number of CET's > 2000'F from Enclosure 4.1:
CET's Number of Operable CET's from Enclosure 4.1:
CET's
% Core Damage = [ Number of CET's > 2000'F ] x 100 Total Operable CET's
% Core Damage =
X100=
Record Results From Steps 3.3.4 and 3.3.5 onto Enclosure 4.4.
Confirm Reasonableness of Expected Core OT Damage Estimate using Predicted Responses below. Record any Comments on Enclosure 4.4.
A. Expected Responses for Fuel Overtemperature Damage
"* RVLIS Less Than 39 %
Hot Leg RTD Greater than 650'F Source Range Monitor Greater Than le+5 cps Difference fuel overtemperature estimates from Containment Radiation Monitors (CRM) and CET's less than 50%, using:
ABS[ % Core DamagecRm - % Core DamagecEr L%
Core Damagecp m 3.3.6 3.3.7
3.3.8 3.3.9 RP/O/A/5700/019 Page 11 of 11 B. If the expected response is not obtained, determine if the deviation can be explained from the accident progression
- Injection of water to the RCS
- Bleed Paths from the RCS
- Direct radiation to the containment radiation monitors or If the expected response is not obtained, determine if the deviation can be explained from conservatism in the predictive model efuel burnup
- fission product retention in the RCS efission product removal from containment IF time permits, complete Enclosure 4.2, Source Range Detector Response.
IF time permits, complete Enclosure 4.3, Movable Detector Surveillance in an Accident Condition.
- 4. Enclosures 4.1 Core Exit Thermocouple Analysis 4.2 Source Range Detector Response 4.3 Movable Incore Detector Response 4.4 Results and Comments 4.5 OAC Points
- 5. References 5.1 Duke Power Calculation DPC-1507.03-00-0001, "Core Damage Assessment Guidelines.",
Rev 0, 6/30/92.
5.2 "Accident Response of Instrumentation" by Westinghouse.
.1 Core Exit Thermocouple Analysis RP/O/A/5700/019 Page 1 of 3 4.1.1 IF possible, obtain a printout of OAC Program Nuclear, Incore Thermocouple Map and attach to this enclosure.
4.3.2 IF desired, use the manual toggle switch panel for CET (in the back of the Control Room on the Moveable Incore Detector panel) to record the non-safety grade thermocouple readings on Page 3 of this enclosure.
NOTE:
Each reference junction box, located in containment, has 2 RTDs which feed into the OAC.
Either RTD can be used for step 4.3.2.
4.3.2 Obtain the reference junction box temperature from the OAC (A 1265 or A 1271) and record below.
A. Reference Junction Box Temperature =
Deg F B.
Verify Reference Junction Box Temperature is reading 160 +/- 100 Deg F. If not, record the difference between the actual junction box reading and 160 Deg F at the Top of Page 3.
Actual Reading - 160'F =
OF
_ F - 160'F =
OF 4.3.3 Complete the CET Readings table on Page 3 of this enclosure to record the readings from available CETs.
NOTE:
- 1. Safety grade thermocouples do not need any temperature correction but non-safety grade thermocouples need temperature correction since they are still tied into the reference junction.
- 2. OAC Program Nuclear, Incore Thermocouple Map, will show both safety and non-safety grade thermocouples.
- 3. Thermocouple calibration may be incorrect if they experience temperatures in excess of 1650 Deg. F.
- 4. The calibrated range of the CET system is 32 Deg F to 2300 Deg F. Once core uncovery occurs, CET readings will be considerably lower than the average maximum fuel temperature.
T-62*
T-51 T-26*
2 3
4 5
6 7
8 900 10__
11 12 13 14__
15__
--F I
R P.1 Core Exit Thermocouple Analysis CET Core Locations 1800 T-53 T-48 T-43#
T-11*
T-21
- T-10*
T-25*
T-20*
T-15*
RP/0/A/5700/019 Page 2 of 3 NORTH "
ZI8
_J__
T-49 T-39__I___P I__
T-34*
T-12*
T-22*
T-14 T-24*
11 T-56 T-j j
46 T-42 jT-38*
T-57 T-52 T-47* _
I I
I I
I I
I I
I III I
I I
I I
I I
I N
M L
K J
H G
F E
D C
00 I
I B
A XX - Thermocouple #
T-XX* Safety Grade T-XX# These thermocouples are temporarily located in the upper head on U-1 only.
T-3*
T-6*
T-29*
T-6 T-30
-- F I
T-63 T-54 T-44*
T-35 T-31*
T-27*
T-17*
T-8*
T-4*
T-1*
T-64 T-59#
T-50 TAO T-32*
T-23*
T-18*
T-13#
T-5*
T-65 T-55 T-45*
T-36 T-33*
T-28*
T-19*
T-9*
T-2#
T-60 T-51 T-41 T-37 2700 T-16*
T-7*
.1 Core Exit Thermocouple Analysis RP/O/A/5700/019 Page 3 of 3 Non-Safetv Grade ThermocouDles CET Readings Reference Junction Difference from M60TF (Sten 4.3.2) =
Thermocouple Temperature Readings 1 Corrections Number Core Location OAC Point Incore Inst OAC Corrected Corrected ID Panel Deg F.
Panel OAC Deg_ F.
Deg F Deg F T35 B-5 A1275 T36 B-9 A0113 T37 B-11 A1281 T39 D-3 A0131 T40 D-7 A0137 T41 D-11 A0143 T42 D-13 A1287 T43 F-i A0155 T46 F-13 A0173 T48 H-1 A0185 T49 H-3 A0191 T50 H-7 A0197 T51 H-11 A1311 T52 H-15 A0209 T53 K-1 A1317 T54 K-5 A1323 T55 K-9 A0227 T56 K-13 A1326 T57 K-15 A1335 T58 M-3 A0245 T60 M-11 A1341 T61 M-13 A1348 T63 P-5 A1360 T64 P-7 A0281 T65 P-9 A1366 NOTE:
I.
To make correction add the reference junction difference to the recorded readings.
Prepared By/Date:
.2 Source Range Detector Response RP/O/A/5700/019 Page 1 of 1 4.2.1 Obtain a plot of the Source Range Detectors (OAC points Al 177 and A1206) response since they were energized following the shutdown. If available attach a hard copy of the plots to this Enclosure.
For a detailed description of the possible accident response of the excore instrumentation see "Accident Response of Instrumentation" of the Westinghouse Advanced Station Nuclear Engineer Training.
105 104 103 102 10 Counts/
104 102 10 1
2
- 3.
4 5
Time After Shutdown (Hours)
Second 1
2 3
4
.5 Time After Shutdown (Hours) 4.2.2 Using the plots above as a guide, examine the Source Range plots from Step 4.2.1 to determine if the core experienced voiding and for how long.
Core Experienced Voiding IF Core Experienced Voiding, What was the Duration of the Voiding:
0 YES IZ NO hrs NOTE:
.3 Movable Incore Detector Response RP/O/A/5700/019 Page 1 of 5 NOTE:
IF the Keithley picoammeter is not able to be set up or if it is not desirable to set it up, the ENA system can still be used to estimate the 3-D location of core damage by attempted insertion of the detectors.
4.3.1 Set up ENA System for special low level measurement per OP/0/A/6150/007 (Incore Instrumentation), Section 8.
4.3.2 Select thimble locations trying to put at least one detector in each quadrant and one near the center of the core.
4.3.3 Record the selected thimble locations in the table in step 4.3.8 (Page 2 of this enclosure).
NOTE:
Operate the ENA System in MANUAL LOW SPEED.
4.3.4 Insert the detectors to top of core.
4.3.5 IF any detector will not insert to top of core complete the following steps in this section.
- 1. Record actual distance reading the detector(s) were able to be inserted in the table in step 4.3.8 (Page 2 of this enclosure).
- 2.
Record expected Top Of Core, from Enclosure 4.3, Table 2.7 of the applicable Units Data Book in the table in step 4.3.8 (Page 2 of this enclosure).
- 3.
Calculate the difference in inches on 4.3.8 (Page 2 of this enclosure) and record on the Core Layout in step 4.3.9 (Page 3 of this enclosure).
NOTE:
For detailed description of the possible accident response of the incore instrumentation see "Accident Response of Instrumentation" of the Westinghouse Advanced Station Nuclear Engineer Training.
4.3.6 Obtain a flux trace and attach to this enclosure.
4.3.7 Compare flux traces with the anticipated traces on Pages 4 and 5 of this enclosure
("Anticipated Low Level Trace With Partial core Voiding" and "Anticipated Low Level Trace With Significant Core Damage) and identify/record indications of core voiding and/or core damage.
.3 Movable Incore Detector Response RP/O/A/5700/019 Page 2 of 5 4.3.8 Complete the Following Table
[PASS I
Drive A Drive B I
Drive C
[
Drive D I Drive E Drive F 1
Thimble Location
______[
Expected T.O.C.
Actual Reading Difference (inches) 2 Thimble Location Expected T.O.C.
Actual Reading Difference (inches) 3 Thimble Location Expected T.O.C.
Actual Reading Difference (inches) 4 Thimble Location Expected T.O.C Actual Reading Difference (inches) 5 Thimble Location Expected T.O.C Actual Reading Difference (inches) 6 Thimble Location Expected T.O.C.
Actual Reading Difference (inches)
NOTE:
- 1)
T.O.C. = Top of Core nixie tube readout
- 2)
Difference = (Expected T.O.C.) - (Actual Reading)
This difference represents where core damage impairs incore movement relative to the T.O.C.
- 3)
Expected T.O.C. can be found in Enclosure 4.3, Table 2.7 of the applicable Unit's Data Book.
.3 RP/0/A/5700/019 Movable Incore Detector Response Page 3 of 5 4.3.9 Core Layout NORTH "4 1800 10 2
3 4
5 6
7 9Q0 2700 9
10 11 12 13 1441 15 I
1 R
P N
M L
K J
H G
F E
D C
B A
00 NOTE:
Enter difference between expected and measured detector insertion value in appropriate thimble location.
I
.3 Movable Incore Detector Response ANTICIPATED LOW LEVEL TRACE WITH PARTIAL CORE VOIDING (Gamma Response)
Pico Amps 60 50 40 30 20
- 10.
0 RP/O/A/5700/019 Page 4 of 5 Note H-lghar Leva(
of damma Response Core Position Onsotof
/
Note VoiAding -J,,
of Gdds
/
r More Subtle /
TOP Flux Decay-of Care j
ýOff Bottom of Core jLeakaga
.3 Movable Incore Detector Response ANTICIPATED LOW LEVEL TRACE WITH SIGNIFICANT CORE DAMAGE (Gamma Response)
RP/O/A/5700/019 Page 5 of 5 Pico Amps 60 50 40 30 20 Botto of Co 10 aR L~eak, 0
Core Positiofi Loa-Peft
.4 Results and Comments RP/O/A/5700/019 Page 1 of 1 RESULTS El Possible Fuel Rod Clad Damage (Step 3.2)
El Possible Fuel OT Damage (Step 3.3)
% Damage Based on CRM's (Step 3.2.4)
% Damage Based on CRM's (Step 3.3.4)
% Damage Based on CET's (Step 3.2.5)
% Damage Based on CET's (Step 3.3.5)
Comments from Step 3.2.6 or Step 3.3.6/Additional Comments:
.5 RP/O/A/5700/019 OAC Points Page 1 of 1 "The points below make up the NUC01 group on the OAC.
This can be displayed by typing "GD NUCO 1" to bring up the group display.
OAC Point Ids Description P0828 5 Highest CET's (Deg F)
P1457 Burnup, EFPD A1047 Containment Pressure A (psig)
A0665 Containment Pressure B (psig)
A1041 Containment Sump Level A (Ft)
A0671 Containment Sump Level B (Ft)
A0829 EMT 51 A (R/hr)
A0835 EMF 51B (R/hr)
A0848 Hydrogen Concentration A (%)
A0854 Hydrogen Concentration B (%)
P0755 Lower Cont Weighted Avg Temperature (Deg F)
P1461 NC Loops Temp AVG-AVG From OK Loops, (Deg F)
P0829 NC Pressure, Best Estimate (psig)
P1470 Pressurizer Level, Best Estimate (%)
A1306 RVLIS Low Range A (%)
A1330 RVLIS Low Range B (%)
Al 177 Source Range Level Channel 1 (cps)
A1206 Source Range Level Channel 2 (cps)
A1204 Upper Containment Ambient Air Temp. A A1210 Upper Containment Ambient Air Temp. B