ML20072F253
| ML20072F253 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/22/1983 |
| From: | Hering R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM AEP:NRC:0761A, AEP:NRC:761A, GL-82-28, NUDOCS 8306270312 | |
| Download: ML20072F253 (19) | |
Text
m lNDIANA & MICHIGAN ELECTRIC COMPANY P. O. BO X 18 BOWLING GREEN ST ATION N EW Y ORK, N. Y.10004 June 22, 1983 AEP:NRC:0761A Donald C. Cook Nuclear Plant Unit Nos. I and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 INADEQUATE CORE COOLING INSTRUMENTATION (GENERIC LETTER NO. 82-28)
Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Lear Mr. Eisenhut:
This letter and its Attachments provide a response to your letter of December 10, 1982 (Generic Letter No. 82-28), regarding Inadequate Core Cooling (ICC) instrumentation systems. In particular, Generic Letter No. 82-28 requested three items of information to help complete the NRC's review of the subject systems at the Donald C. Cook Nuclear Plant. .
Item 1 of your letter requested that we submit the design for the Reactor Vessel Level Instrumentation System (RVLIS) and schedules for ics engineering, procurement, and installation. Attachment I to this letter provides the requested information for our Westinghouse (W)
Item 2 of your letter requested that we' review the status of all ICC instrumentation system coc ponents' conforrance to NUREG-0737 Item II.F.2, and submit a report on such status. This information is included in Attachments 1 and 2 to this letter for the RVLIS and the subcooling margin monitor, respectively. Attachmenc 3 provides information on the current incore thermocouple system and our plans to upgrade the system to NUREG-0737 requirements.
Item 3 of your letter requested that we present a schedule for implementation of ICC instrumentation system improvements for review by the assigned NRC Project Manager. Where applicable, this 8306270312 830622 fO0 PDR ADOCK 05000315 l' p PDR
.Mr. DIrrall G. Eiecnhut 2 AEP:NRC:0761A information is contained in the Attachments to this letter. We are willing to discuss these improvements with members of the NRC staff at their convenience.
As requested by Generic Letter No. 82-28, the information contained in this letter and its Attachments is affirmed in accordance with 10 CFR 50.54(f) .
Very truly yours, W
RFH/os R. F. Hering Vice President cc: John E. Dolan - Columbus R. S. Hunter M. P. Alexich R. W. Jurgensen W. G. Smith, Jr. - Bridgman R. C. Callen G. Charnoff NRC Resident Inspector at Cook Plant - Bridgman p
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. e STATE OF NEW YORK )
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COUNTY OF NEW YORK)
R. F. Hering, being duly sworn, deposes and says that he is a Vice President of Licensee Indiana & Michigan Electric Company, that he has read the foregoing response to Generic Letter No. 82-28 (AEP:NRC:0761A),
and knows the contents thereof; and that said contents are true to the best of his knowledge and belief.
T/C /
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- Subscribed end sworn to before me this c7o? day of @ w g , 19 PJ .
NA /E w cv (Notary Public)
WILLIAM J. PROCHASKA Notary Pub'ic, Sta'e cf New York No. 43-4536690 Qualified m Rahmcnd County Ceri.saa e e .. n f4ew fork Count .
Comminkn Empires March 30,19
. ATTACHMENT NO. 1 TO AEP:NRC:0761A REACTOR VESSEL LEVEL INSTilUMENTATION SYSTEM DONALD C. COOK NUCLEAR PLANT UNIT NOS. 1 AND 2 Int.oduction Genetic Letter No. 82-28 requested two items of information with regard to the Donald C. Cook Nuclear Plant Reactor Vessel Level Instrumentation System (RVLIS). More specifically, the subject letter requested that we identify the RVLIS design and. submit schedules for its engineering, procurement, and installation. We were also requested to report on the status of RVLIS conformance to NUREG-0737.
This Attachment provides the requested information. The RVLIS derign is first identified, and then RVLIS conformance to NUREG-0737 Item II.F.2 and Appendix B of NUREG-0737 is outlined.
Additional information regarding the location of RVLIS equipment and the use of hydraulic isolators and sensors in the impulse lines, as requested on page 4 of the checklist attached to Generic Letter No.
82-28, is also included herein.
Where applicable, reference is made to one or more documents which we understand are either in the NRC's possession or to be trancmitted directly to the NRC by W in the future. To the best of our
! knowledge, documents already submitted to the NRC are still valid in response to NUREG-0737 documentation requirements. The list of references is provided at the end of this Attachment. Some RVLIS information has also been provided in the form of notes for ease of reference. The notes are also provided at the end of this Attachment.
RVLIS Design The RVLIS design selected for the Donald C. Cook Nuclear Plant is the W differential pressure (d/p) syetem. This system has been previously described in References (1), (2), and (3). With the exception of those items identified in Note (iii), this system has been physically installed in both Units of the Donald C. Cook Nuclear Plant.
We understand that the RVLIS cannot be operated until the task analysis portion of the Control Room Design Review is completed and NRC approval of the RVLIS is granted. It is anticipated, however, that the additional RVLIS work will be completed without causing further delay of RVLIS operation.
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_RVLIS Conformance to NUREG-0737 Itest II.F.2 This section addresses the nine items of documentation identified on NUREG-0737 pages II.F.2-3 and II.F.2-4. For each item of required documentation identified in NUREG-0737, the information requested by the checklist attached to Generic Letter No. 82-28 is provided. As noted above, references and notes are contained at the end of this Attachment.
Item (1)(a):
References:
(1) through (3) .
Deviations: Functionally none.
Schedule: Note (i).
- Item (1)(b):
References:
(1) through (3) .
Deviations: Note (ii).
'l Schedule: Note (ii).
Item (1)(c): Not applicable to RVLIS. No modifications planned.
Item (2):
References:
(1) through (7) .
Deviations: Functionally none.
Se!edule: Note (1) .
Item (3):
References:
(4) through (7).
Deviations: None.
Schedule: Note (1) .
Item (4):
References:
(2), (3), (8), and (10) through (13).
Deviations: None.
Schedule: Note (1), except for Reference (13).
Item (5):
References:
(1) through (3).
t Deviations: None.
l Schedule: Note (1) .
- Note: Documentation of Item (5) is applicable to display function only.
j Item (6):
References:
Note (iii).
Deviations: Note (iii) .
Schedule: Note (iii).
Item (7):
Reference:
(9).
Deviations: None.
Schedule: Note (i).
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1-3 Item (8):
Reference:
Note (iv).
Deviations: None anticipated.
Schedule: Note (iv).
Item (9): See Note (1). _
RVLIS Conformance to NUREG-0737 Appendix B
[
As requested by the checklist attached to Generic Letter No.
82-28, RVLIS conformance to the first nine items of NUREG-0737 Appendix B is outlined below. Since the checklist did not address Appendix B Items 10 through 18, it is assumed that NRC approval of the generic }[
d/p system connotes approval of these items.
Item (1): See References (10) through (13). No deviations.
Item (2): See Reference (8). No deviations.
Item (3): The instrumentation ie energized from Class IE i power source.
! Item (4): One channel operable prior to an accident.
Item (5): Quality Assurance per 10 CFR 50, Appendix B.
Item (6): Continuous indication is provided.
Item (7): Recording of instrument outputs is provided (single train).
. Item (8): See Reference (1) .
Item (9): Isolation devices are utilized.
Additional Information Page 4 of the checklist attached to Generic Letter No. 82-28 requested additional information from users of the }[ d/p system. In response to the three questions contained therein, the following l information is provided.
(1) The effect of instrument uncertainties on the measurement of level has been described in References (2) and (3) . See also Note (ii).
(2) Differential pressure transmitters are located outside containment. '
l (3) Hydraulic isolators and sensors are included in the impulse lines.
1-4 r
NOTES (1) No further work is currently planned.
(ii) W indicated, via letter No. AEP-83-522, that the existing instrumentation used by RVLIS does not functionally deviate from NUREG-0737 Item II.F.2 requirements. However, the W 1etter transmitted additional docucantation which stated that the interfacing of wide range pressure transmitters, as
, originally designed by W, is still under study and as such is .
not recommended. In document No. NS-EPR-2586, which }[ *
. attached to the above referenced letter, concern was expressed about post-accident environmental effects on i in-containment wide range pressure transmitters and subsequent effects on RVLIS accuracy. Since the wide range ;
transmitters in use at the Donald C. Cook Nuclear Plant have been qualified for use in accordante with the DOR Guidelin2s (except for aging and a surveillance / maintenance / replacement program), this situation is presently under discussion with W. Therefore, with the possible exception of the wide range
- pressure transmitter, the description of existing instrumentation documented in Reference (1) through (3) is still applicable. We will resolve the wide range preseure transmitter application with 1[ and will, if any changes to the RVLIS description are necessary, prepare a schedule for the NRC.
(iii) Work is continuing on the items previously identified in our submittal letter No. AEP:NRC:0398H, dated November 9,1982 (e.g., implementing calibration correction f actors, reviewing the environmental qualification of system cable routing, and obtaining equipment qualification reports from W) . In addition, the panel recorder for the Donald C. Cook Nuclear Plant Unit No. 1 RVLIS remains to be installed. We understand that the RVLIS cannot be operated until the task analysis portion of the Control Room Design Review is completed and NRC approval of the RVLIS is granted. It is anticipated, however, that the additional RVLIS work will be completed without causing further delay of RVLIS operation.
l . (iv) We will incorporate the Westinghouse Owners Group Emergency Response Guidelines (ERGS) in the Donald C. Cook Nuclear Plant operating instructions, as previously explained in our response to Generic Letter No. 82-33 (AEP:NRC:0773, dated
. April 15, 1983). It is anticipated that this work will be completed without causing further delay of RVLIS operation.
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1-5 REFERENCES (1) Letter No. AEP:NRC:0398D, dated March 31, 1981, R. S. Hunter (I&MECo) to H. R. Denton (NRC). Refer to " Summary Report, Westinghouse Reactor Vessel Level Instrumentation System for '
Monitoring Inadequate Core Cooling (7300 System), December, 1980."
Refer also to Letter No. NS-TMA-2357, dated December 23, 1980, T.
M. Anderson (W) to D. G. Eisenh':t (NRC).
(2) Letter No. AEP:NRC:0398F, dated September 17, 1981, R. S. Hunter (I&MECo) to H. R. Denton (NRC) .
, (3) Letter No. AEP:NRC:0398G, dated November 20, 1981, G. P. Maloney
( (I&MECo) to H. R. Denton (NRC).
l (4) Letter No. NS-EPR-2579, dated March 19, 1982 E. P. Rahe (W) to L.
E. Phillips (NRC).
(5) Letter No. NS-EPR-2526, dated December 9,1981, E. P. Rahe (jd) to L. E. Phillips (NRC).
i l (6) Letter No. NS-EPR-2542, dated January 13, 1982, E. P. Rahe (W) to I
L. E. Phillips (NRC).
(7) Letter No. SED-SA-0081, dated June 28, 1982, E. P. Rahe (W) to L.
E. Phillips (NRC).
(8) Letter No. NS-EPR-2597, dated May 14, 1982, E. P . Raha (]d) to D.
M. Crutchfield (NRC).
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(9) Westinghouse Owners Group Emergency Response Guidelines, Volumes l 1, 2, and 3. Transmitted to the NRC via letters No. OG-64, dated November 30, 1981, and No. OG-83, dated January 4,1983.
-**(10) Letters No. NS-EPR-2513 and No. NS-EPR-2600, transmitted to the NRC on September 30, 1981, and May 19, 1982, respectively, from E.
P. Rahe (W) . See equipment Qualification Data Package No. ESE-3, D/P Transmitters -- Qualification Group A. Note: We understand that W, in the future, will reference these transmitters specifically to the RVLIS.
- (11) Equipment Qualification Data Package No. ESE '3, 7300 Procesa Electronics. Transmitted to the NRC via Letter No. NS-TMA-2380, dated Janucry 28, 1981, from T. M. Anderson (W), and by Letter No.
NS-EPR-2482 dated August 3,1982, from E. P. Rahe (W) . Note: We understand that W, in the future, will reference the process electronics specifically to the RVLIS.
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- (12) Equipment Qualification Data Packages ESE-42 (Strap -on RTDs),
ESE-48 (High Volume Pressure Sensor), and ESE-49 (Hydraulic Isolator) . Submitted to the NRC via NS-EPR-2725, dated February 28, 1983, from E. P. Rahe (W).
(13) Equipment Qualification Data Package No. ESE-50 (Entire RVLIS Package). Note: We understand that W will transmit this package to the NRC by June 30, 1983.
Note: P4ferences (10) through (12) were submitted as supplements to the environmental qualification topical reports WCAP-8587 (non-proprietary) and WCAP-8687 (proprietary).
ATTACHMENT NO. 2 TO AEP:NRC:0761A SUBC00 LING MARGIN MONITOR DONALD C. COOK NUCLEAR PLANT UNIT NOS. 1 AND 2 Introduction Generic Letter No. 82-28 requested that the Donald C. Cook j Nuclear Plant subcooling margin monitor (SMM) be reviewed for l
conformance to NUREG-0737 Item II.F.2. This Attachment provides the results of that review.
Information on the Babcock & Wilcox (B&W) saturation meter and the P250 Plant computer is provided first. Then a response is presented to the nine items of required documentation identified on NUREG-0737, pages II.F.2-3 and II.F.2-4. Finally, conformance to NUREG-0737, Appendix B requirements are outlined.
Reactor Coolant Margin to Saturation Meter The B&W saturation meter is a device which is connected to
, reactor coolant temperature and pressure signals and displays the margin to saturation temperature or to saturation pressure. The meter provides an on-line aid to the operator to help him assure that an i adequate saturation margin is maintained at all times.
(
Pressure signals f rom Loop 1 and Loop 2 hot legs as well as a third signal, which is the lower of the two, are fed to the P1/P2/P3 inputs of the meter. Seven individual and the highest of eight temperature signals from the vide range reactor coolant RTDs and seven individual and the highest of eight temperature signals from eight of the incore thermocouples (two per reactor quadrant) provide a tntal of sixteen temperature inputs.
l The front panel display has four switches. The switch labelled P1/P2/P3 selects one of the three pressure sources described above. The meter determines the saturation temperature of the selected pressure. An eight position rotary switch in conjunction with the "INCORE/RTD" toggle switch selects one of the sixteen temperature l inputs. The meter finds the saturation pressure for this temperature.
l The derived saturation temperature is compared with the selected temperature, and the derived saturation pressure is compared with the selected pressure. Thus, saturation margins for both temperature and pressure are determined. Margin to saturation temperature is displayed l
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l l _ _ _ _ _ , __ ._ _ _ _ _ _
2-2 on the digital panel meter. Margin to saturation pressure can be momentarily displayed using the spring-loaded "T-sat /P-sat" to3Sle switch.
The alsrm circuitry examines all of the input temperature margins to derived saturation temperature. If any of these margins is less than the setpoint temperature margin, an alarm will agnunciate on the, Reactor Coolant Panel. The setpoint can be set from 0 F to 199 F by a thumbwheel switch located at the equipment rack. Note -
that the alarm function is contingent on the selected pressure.
A test display module is provided for use in calibrating the analog portion of the saturation meter.
Table 1 provides additional information on the subcooling meter.
Plant Computer Subcooling Program The P250 computer will calculate the subcooling of the reactor coolant in degrees Fahrenheit. There are five calculated values which are available for analog or digital trending on the P250 computer output devices. The calculations are based on the two wide range pressure inputs f rom Loops 1 and 2, eight thermocouple inputs (two per quadrant), and eight RID wide range temperature inputs (two per Loop).
From these inputs, calculated values are obtained for saturation temperature based on lower system pressure, and die margin to saturation temperature based on: (1) the hottest RTD, (2) the hottest thermocouple, (3) the average of the RTDs, with the hottest and coldest RTDs excluded, and (4) the average of the thermocouples, with l
the hottest and coldest thermocouples excluded.
The computer checks each temperature input signal as a function of its deviation from its associated average. Signals which deviate by more than a specified amount are determined to be unreliable. Nevertheless, all inputs are used for calculational purposes because a hot spot could exist in the core as a result of flow tiockage. There are two alarm conditions which will annunciate and cause a message to be printed on the trend typewriter. The message l includes all input and calculated values. The first alarm results from l a saturation margin less than the setpoin . The second results from an ,
l unreliable input.
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2-3 Table 2 provides additional information on the P250 Plant computer. It should be noted that it is our intention to retain the P250 subcooling program as a back-up display / alarm to calculate subcooling margin.
Subcooling Margin Monitor Conformance to NUREG-0737 Itam II.F.2 This section addresses the nine items of documentation identified on NUREG-0737, pages II.F.2-3 and II.F.2-4. For each item of required documentation identified in NUREG-0737, the information requested by the checklist attached to Generic Letter No. 82-28 is provided.
Item (1)(a): Not appliccble.
Item (1)(b):
References:
This Attachment.
Deviations: Not Applicable.
Schedule: Completed herewith.
Item (1)(c): Not applicable.
Item (2):
References:
Item 2.1.3.b of Attachment 6 to our letter dated January 10, 1980 (AEP:NRC:0334).
Deviations: None.
Schedule: Complete.
Item (3): Not applicable.
Item (4):
References:
Item 2.1.3.b of Attachment l to our letter dated March L
10,1980 (AEP:NRC:0334B).
Deviations: Yes.
Schedule: See following section on conformance to NUREG-0737 Appendix B.
Iten (5):
References:
Item 2.1.3.b of our letter dated December 19, 1979 (AEP:NRC:0253B), updated per this Attachment.
Deviations: None.
l Schedule: Completed herewith.
l Item (6): Not applicable.
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2-4 2 Item (7): Not applicable.
Item (8):
References:
See Reference for Item (2) above. See also Item d of Attachment 1 to letter No.
AEP:NRC:0773, dated April ,
15, 1983.
Deviations: None.
Schedule: Item IV of Attachment 3 to Letter No. AEP:NRC:0773, dated April 15, 1983.
Item (9):
References:
Item IV of Attachment 3 to Letter No. AEP:NRC:0773, dated April 15, 1983.
Deviations: Not applicable.
Schedule: Not applicable.
Subcooling Margin lionitor Conformance to NUREG-0737. Appendix B As requested by the checklist attached to Generic Letter No.
82-28, SMM conformance to the first nine items of NUREG-0737 Appendix B is outlined below.
Item (1): See Attachments 1, 2, and 4 to letter No.
, AEP:NRC:0775C, dated May 20, 1983. These l Attachments identify deviations and the i proposed schedule for completion of I
environmental qualification for wide range reactor coolant temperature RTDs (NTR-110, 120, 130, 140, 210, 220, 230, 240) and pressure transmitters (NPS-121, 122).
Item (2): See Notes (i) and (ii). No deviations.
l Item (3): See Note (ii). No deviations.
Item (4): See Technical Specifications 3.3.3.8 (Unit 1) and 3.3.3.6 (Unit 2) . No deviations.
Item (5): 10 CFR 50, Appendix B. No deviations.
Item (6): See this Attachment and Note (iii). No deviations.
1 l Item (7): See this Attachment and Note (iii). No -
deviations.
M 2-S Item (8): See Note (iv). No deviations.
Item (9): See Item 2.1.3.b of Attachment to letter No.
AEP:NRC:0334B, dated March 10, 1980. No deviations.
NOTES (i) A review of the design of the Donald C. Cook Nuclear Plant saturation monitoring systems has shown that no single failure within the accident monitoring instrumentation, auxiliary supporting feetures, or power sources, will prevent the operator from determining margin to subcooling. There are at least two physically and electrically separated channels for reactor coolant tempcrature and pressure information available in the Control Room.
(ii) The wide range reactor coolant temperature and pressure-channels, including their isolation devices, are energized from Class IE power sources. The B&W subcooling meter and its associated input signal conditioning, isolation, and selection logic modules are also energized from a Class IE power source.
(iii) Reactor coolant temperature and pressure are continuously recorded. Subcooling margia is continuously displayed by the B&W subcooling meter.
(iv) Subcooling margin display, wide range reactor coolant temperature' recorders, and wide range reactor coolant pressure recorder and indicator are all specifically identified on the control panels.
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2-6 TABLE 1 Subcooling Meter Infounation Display Information displayed: Margin to saturation temperature or saturation pressure.
Display Type: Digital, Continuous, Single.
Location: Control Room.
Alarms and overall uncertainty: See letter No. AEP:NRC:0346, Range of display: datgdJanuary 35 F superheat29,to19200 80.,F subcooled; O to 500 psi margin.
Calculator:
Type: Dedicated digital.
Single calculator.
Selection logic: Highest temperature; lowest pressure.
Calculational technique: Steam Tables Input:
Temperature: RTDs and thermocouples.
Number of sensors and locations: See (*) below.
Range of Temperature sensors: RTDs -- 0 F to 70g F; thermocouples -- 0 F to 2500 F.
Pressure: Wide range reactor coolant system.
Number of sensors and locations: 2 RCS Loops 1 & 2.
Range of pressure sensors: 0 to 3000 psi.
I Uncertainty of temperature and
- pressure sensors
- See letter No. AEP:NRC:0346, dated l January 29, 1980.
l Qualifications: RTDs -- See (**) below; thermocouples --
control grade; pressure -- See (**) below.
t l Back-up Capability:
Availability of Temp. and Press.: In Control Room --
recorders and indicators.
Availability of Steam Tables: Pasted to Control Room panel.
Trainingsof Operators: Yes.
Procedures: Yes.
(*) Sensors selected are 4 hot leg RTDs, 4 cold leg RTDs, and 8 core exit thermocouples which are input to the calculator through hi-l select devices.
l (**) See letters No. AEP:NRC:0578B, dated June 11, 1982, and No.
l AEP:NRC:0775C, dated May 20, 1983.
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2-7 TABLE 2 PRODAC Computer Information Display:
Information displayed: T - T(sat); temperature; pressure.
Display type: Digital, automatic on alarm or on demand (temperatures and pressures available for trend, with updated values every 4 seconds; T - T(sat) margin available for trend, with updated values every 64 seconds),
single but back-up for output.
Location: Control Room.
Alarms and overall uncertainty: See letter No. AEP:NRC:0346, dated January 29, 1980.
Range of display: Adjustable.
Calculator:
Type: Process Cosputer.
Availability: Approximately 95%.
Single calculator.
Selection logic: Highest temperature; lower pressures.
~ Calculational technique: Functional fit agrees with steam tables to 1 part in 1000.
Input See Table 1 of this submittal for pertinent information.
Back-up Capability i
1 Availability of Temp. & Press.: Procedures and saturation I curves.
Availability of Steam Tables: Pasted to Control Room panel.
Training of operators: Yes, Procedures: Yes.
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ATTACHNENT NO. 3 TO AEP:NRC:0761A !
INCORE THERMOCOUPLE SYSTEM , !
DONALD C. COOK NUCLEAR PLANT UNIT NOS. 1 AND 2_ l Introduction Primary indication of ICC conditions at the Donald C. Cook Nuclear Plant will be provided by the RVLIS and the subcooling margin monitor. These systems are described in Attachments 1 and 2 to this letter, respectively. The incore thermocouples, on the other hand, represent a supplemental or back-up instrumentation system for ICC indication purposes. For this reason, we do not believe that the incore thermocouple system should be classified as a Class IE system.
A program has been undertaken, however, to perform studies in an attempt to upgrade the incore thermocouple system to NUREG-0737 requirements.
This Attachment presents a description of the currently installed thermocouple system, a status report with regard to system .
upgrade to NUREG-0737 requirements, and the presently planned '
implementation schedule. Due to the ongoing nature of the program, l responses in the form of the checklist attached to Generic Letter No. i 82-28 are not being provided at this time.
Current Syetem Design Description The current'incore thermocouple system is a standard E system, provided as original equipment with the Donald C. Cook Nuclear Plant.
Each Unit was provided with sixty-five incore thermocouples which are routed through reactor vessel head incore columns (13 thermocouples per column) . The thermocouples are type "K" chromel-alumel, ungrounded junction,1/8" in diameter, supplied with a Thermo-Electric mating connector, type "KK" at the top. Each tharmocouple is enclosed in a 3/16" inside diameter guide tube which l extends from the instrument column to the bottom of the reactor upper j internals.
Thermocouple extension wire runs from the Thermo-Electric mating connectors to another set of Maelin mating connectors in the disconnect boards (disconnect boards are points at which all instrument and electrical cables to the reactor head are disconnected prior to lifting the head). Thermocouple extension wire runs from the
, 3-2 o
disconnect boards to two Whittaker thermocouple reference junction i boxes located inside containment.
The junction boxes are heated to 160 F and permit transition from chromel-alumel thermocouple wiring to copper field wiring. Each junction box also contains three platinum Resistance Temperature Detectors (RIDS). Two RTDs in each junction box are connected to the Plant computer for monitoring junction reference temperature. The third RID in each junction box is a spare.
Primary thermocouple readout capabigity is prgvided by the P250 Plant computer, which has a range of 160 F to 2500 F. The computer prints the temperature values on the printer, or displays them on the console, upon request.
Back-up readout capability is provided by a Honeywell Precision Indicator mounted in the Flux Mapping Control Console. This instrument is supplied with a double-range mgasuring gircuit which permitg measurement within the ranges of 100 F to 400 F or 400 F to 700 F. Selection of a single thermocouple to be monitored is made via use of non-locking key switches on the front of the indicator. The switch must be manually held in position (lef t or right) to monitor the desired thermocouple. The switch returns to neutral (center) position when released.. Since the thermocouple input signal to the indicator is in parallel with the Plant computer, a contact closure signal informs the computer when any thermocouple is being monitored.
Furthermore, thermocouple signals are sent to the' Technical Support Center Computer, and eight thermocouple signals are input to the subcooling margin' monitor.
The Donald C. Cook Nuclear Plant operators are trained to use this thermocouple system.
System Upgrade Status Report Because of the high failure rate of Unit No. 1 incore thermocouples, we have undertaken a program with three Nuclear Steam Supply System (NSSS) vendors to either repair or replace the incore thermocouple system. A goal of this program is to upgrade the system to conform as closely as possible to NUREG-0737 requirements.
Currently planned improvements are as follows:
(1) Replacement of existing incore thermocouples, if required.
We have been notified by the NSSS vendors that no known qualified incore thermocouple systems have been installed at other nuclear generating stations. Furthermore,
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only limited experience exists where incore thermocouples have been recovered and replaced for a W Pressurized Water Reactor. State of the art technology is still in a developmental stage, and at this time a successful incore thermocouple replacement program cannot be guaranteed.
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(2) Replacement of connectors and cabling.
l (3) Replacement and/or relocatior of junction boxes outside containment, or replacement of junction boxes inside containment.
l (4) Separation of thermocouples by columns into two independent groups, and routing of each group's cabling separate from the other, to the extent possible. It is noted that separation criteria deviations will be necessary in the vicinity of the reactor vessel head.
(5) Termination of two incore thermocouples from each group in l the reactor vessel head to monitor fluid temperature there.
l (6) Use of independent Class IE power sources.
(7) Provisions for independent indication for each group in the Control Room.
l (8) Testing and calibration to be conducted after completion of installation.
We note that these planned modifications may be subject to change as a result of the ongoing review process.
Upgrade-Implementation Schedule l Barring unforeseen complications, repair or replacement of l
the incore thermocouple system should commence for Unit No. 1 during a 1984-1985 refueling outage. Following this upgrade, a report will be made to the NRC stating our progress and experience. Thereafter a plan will be presented for Unit No. 2.
I
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