ML20209F914
| ML20209F914 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/27/1985 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | Butcher E Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.2, TASK-TM LIC-85-211, NUDOCS 8507120580 | |
| Download: ML20209F914 (60) | |
Text
I 4
i Omaha Public Power District 1623 Harney Omaha, Nebraska 68102 402/536 4000 June 27, 1985 LIC-85-211 Mr. Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear regulatory Commission Washington, DC 20555
References:
(1) Docket No. 50-285 (2) Letter NRC (J. R. Miller) to OPPD (W. C. Jones) dated July 11, 1984 (3) Letter OPPD (R. L. Andrews) to NRC (J. R. Miller) dated September 28, 1984 (LIC-85-323)
Dear Mr. Butcher:
NUREG-0737, Item II.F.2 Inadequate Core Cooling Instrumentation (ICCI)
Reference (2) requested the Omaha Public Power District supply a schedule for various milestones associated with the ICCI System. Accordingly, Reference (3) provided June,1985 as a milestone date for submittal of the final design de-scription for the ICCI, including the documentation requirements of NUREG-0737, Item II.F.2.
Attached is the District's final design description for the sub-ject item.
Please note that this document is currently undergoing an internal ir: dependent review.
If any additional information is deemed necessary as a re-sult of this review, it will be submitted by July 19, 1985.
l Sincerely, G 3.
M R. L. Andrews Division Manager Nuclear Production l
RLA/DJM/dao
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l Attachment V
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cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, DC 20036 Mr. E. G. Tourigny, NRC Project Manager Mr. L. A. Yandell, NRC Senior Resident Inspector 8507120580 850627 PDR ADOCK 05000285 P
PDR i
45 sea Employmen h Qua Oppor:unny
(E4) 4 BACKGROUND On October 31, 1980, the Nuclear Regulatory Commission (NRC) issued HUREG-0737,
" Clarification of TMI Action Plan Requirements."
(1.1).
Item II.F.2 of thi s document required the installation of additional instrumentation for the detec-tion of inadequate core cooling (ICC).
In September and November of 1981 and September of 1982 the Combustion Engineering Owners Group (CE0G) transmitted to the NRC staff for review the following generic Combustion Engineering ICCI documents:
(a) CEN-181:
Generic Responses to NRC questions on the C-E Inadequate Core Cooling Instrumentation (3.1)
(b) CEN-185: Documentation of Inadequate Core Cooling Instrumentation for Combustion Engineering Nuclear Steam Supply Systems (3.2)
(c) CEN-185, Supplement.1: Heated Junction Thermocouple Phase I Test Report (3.3)-
(d) CEN-185, Supplement 2: Heated Junction Thermocouple Phase II Test Report (3.4)
(e) CEN-185, Supplement 3: Heated Junction Themocouple Phase III Test Report (3.6)
In March of 1982, NUREG/CR-2627 (0RNL/TM-8286), " Inadequate Core Cooling Instru-mentation using Heated Junction Thermocouples for Reactor Yessel Level Measure-ment" (1.2) was issued. This report documented NRC reviews of the CE ICCI system.
On December 10, 1982, per Generic Letter 82-28 (1.3), the Commission requested certain plant specific information required for review of ICCI systems.
Gen-eric Letter 82-33 (dated December 17, 1982) (1.4) provided additional clariff-cation to NUREG-0737 in the fom of Supplement 1.
This letter also asked utilities to define their implementation schedules for the NUREG-0737 itens.
Omaha Public Power District (0 PPD) responded to these letters on March 8,1983 (LIC-83-056) (2.2) and April 15, 1983 (LIC-83-093) (2.3), respectively.
In these letters OPPD identified that the CE HJTC/QSPDS system would be used for inadequate core cooling detection at the Fort Calhoun Station and referenced the previously submitted CE0G reports.
l In a letter dated July 11, 1984 (1.7) the NRC issued the safety evaluation (SE) on OPPD's ICCI system. This SER was based on our March 8,1983 submittal (2.2). A request for additional infomation was included with the SE.
This l
additional information was provided per LIC-84-323, dated September 28, 1984 l
(2.6).
Item 1 of Attachment 2 to Reference 2.6 provided a June 1985 date for the following request:
"1.
Submit final design description (by licensee) (complete documenta-tion requirements of NUREG-0737, Item II. F.2, including all plant-specific infomation items identified in applicable NRC evaluation reports for generic approved systems)."
l l
l 1
l
(E4) l REFERENCES 1.0 NRC Documents & Letters 1.1 NUREG-0737, " Clarification of TMI Action Plan Requirements",
October 31, 1980.
1.2 NUREG/CR-2627 (ORNL/TM-8268)," Inadequate Core Cooling Instru-mentation Using Heated Junction Thermocouples for Reactor Vessel Level Measurement", March 1982.
1.3 Generic Letter 82-28, " Inadequate Core Cooling Instrumentation System", Decenber 10, 1982.
1.4 Generic Letter 82-33, " Supplement 1 to NUREG-0737; Requirements for Emergency Response Capability", December 17, f
1982.
i
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1.5 Generic Letter 83-37, November 1,1983.
1.6 Emergency Response Capabilities Confirmatory Order, February 22, 1984.
1.7 Fort Calhoun Station ICCI Safety Evaluation, July 11, 1984.
1.8 Request for Additional Information, July 24, 1984.
2
(E4) 2.0 OPPD LETTERS 2.1
" Implementation and Schedule for NUREG-0578," December 31, 1979.
2.2 L IC-83-0 56,
" Response to Generic Letter 82-28,"
March 8,
1983.
.2.3 LIC-83-093, " Response to Generic Letter 82-33," April 15, 1983.
' 2.4 L IC-83-276,
" Safe ty Parameter Di splay Systen
. Safety Analysis," October 28, 1983.
2.5 LIC-84-071, "NUREG-0737 Technical Specification," March 27, 1984.
2.6 LIC-84-323, "ICCI Systen Evaluation," September 28, 1984.
2.7 L IC-84-327,
" Safety Parameter Display System, Response to Request for Additional Infonnation," December 7,1984.
P. 8 LIC-85-020, " Safety Parameter Display System," March 5,1985.
2.9 LIC-85-117, " Fort Calhoun Station Compliance with Regulatory Guide 1.97, Rev. 2," April 1, 1985.
2.10 LIC-85-181, " Detail ed Control Room Design Revi ew," May 2,
1985.
2.11 LIC-85-233, " Detailed Control Room Design Review," June 15, 1985.
3
I
-(E4) 4 3.0 CE/CEOG LETTERS & DOCUMENTS 3.1 CEN-181, " Generic Responses to NRC questions on the C-E Inadequate Core Cooling Instrumentation", September 1981.
3.2 CEN-185, " Documentation of Inadequate Core Cooling Instrumen-tation for Combustion Engineering Nuclear Steam Supply Systems", September 1981.
3.3 CEN-185, Supplement 1, " Heated Junction Themocouple Phase I Test Report", November 1981 3.4 CEN-185, Supplement 2, " Heated Junction Themocouple Phase II Test Report", November 1981.
3.5 CE0G Letter to NRC, June 1,1982.
3.6 CEN-185, Supplement 3, " Heated Junction Thermocouple Phase III Test Report", September 1982.
3.7 CEN-152, Rev. 02, " Combustion Engineering Emergency Procedures Guidelines", May 1984.
3.8 Appendix A to CEN-152, Rev. 02, " Response to NRC questions on CEN-152 Rev. 02", November 1984.
3.9 CE0G Letter to NRC re. RVLMS Technical Specifications, February 19, 1985.
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(E4) 1.
Description of the proposed final system including:
1.a. a final design description of additional instrumentation and displays; OPPD utilizes the Combustion Engineering Heated Junction Thenno-couple System (HJTCS) and Qualified Safety Parameter Display System (QSPDS) for the detection of and processing signals of Inadequate Core Cooling (ICC).
The Fort Calhoun Emergency Response Facilities Computer System (ERFCS) and the QSPDS together form the plant's SPDS which is the primary display for ICCI.
An overview of the complete system is shown in figure 1.
A generic description of the HJTCS was provided per references 3.1, 3.2, 3.3', 3.4 and 3.6.
The following is a description of the QSPDS:
Qualified Safety Parameter Display System. As depicted in figure 1, the QSPDS is a two-channel system which displays the inadequate core cooling instrument (saturation margin monitor, heated junction thenn-ocouple, and core exit thermocouple system) outputs to the. control room.
The QSPDS uses a microprocessor-based design for the signal processing equipment in conjunction with an alphanumeric display and associated keyboard for each of the two channels.
Each channel will accept and process inadequate core cooling input signals and trans-mit its output to the Emergency Response Facilities Computer System (ERFCS)
The two QSPDS channels are powered by Channel A and B station vital busses.
Each 'QSPDS is electrically independent and physically sep-a rated.
The QSPDS is designed to meet Class 1E isolation require-ments.
Any non-1E signal inrM ta the QSPDS is isolated before it enters the QSPDS processor-
- 1 particular, the QSPDS provides for digital signals to be e M U ly isolated, thermocouples isolated with the "flyi ng captci p technique, and the high level anal og signals protected with ctive ow pass filters.
The QSPDS is envi-ronnentally and seismically qualified per testing conducted in accor-dance with the methodology and guidelines outlined in IEEE-323-1974 and IEEE-344-1975.
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3
u (E4)
The QSPDS displays present the most reliable basic infomation for each of the inadequate core cooling instrument systems.
The QSPDS displays are designed:
To give primary instrument indications in the remote chance that the Emergency Response Facility Computer System display becomes inoperable To provide confimatory indications to the Emergency Response Facility Computer System display To aid in surveillance tests and diagnostics The QSPDS displays (located in the control room) also incorporate human factors engineering.
The alphanumeric display is a liquid plasma display with paging capabilities to display all the inadequate core cooling in-strument outputs.
The two channels of QSPDS display present direct and continuous safety-grade indication of the inadequate core cooling detec-tion parameters.
The Emergency Response Facility Computer System pro-vides trendi ng capability for all inadequate core cooling parameters.
The QSPDS displays the following types of information:
Safety parameters according to safety function Addi tional safety parameter information on other pages (such as additional Reg. Guide 1.97 variables)
Alam indication Saturation Margin Monitoring System.
The saturation margin monitoring system processing equipment performs the following functions:
A.
Calculates the saturation margin The saturation temperatur: is calculated from the pressure input and the saturation pressure is calculated from the maximum temperature input.
The temperature saturation margin is the di fference between saturation temperature and the maximum temperature input.
The pressure saturation margin is the difference between saturation pressure and the pressure input.
B.
Processes all outputs for display.
C.
Provides an alarm output when saturation margin reaches a preselected setpoint.
8
s (E4)
The following information is presented on the QSPDS displays:
Pressure margin to saturation Temperature margin to saturation for each temperature source (i.e.,
resistance temperature detectors, heated junction thermocouples, or core exit themocouple)
Temperature inputs Pressure inputs Heated Juntion Thermocouple System.
The processing equipment fo r the heated junction thermocouples perfoms the following functions:
A.
Detennines if liquid inventory exists at the heated junction thermocouple positions.
The heated and unheated themocouples in the heated junction thermocouple are connected in such a way. that absolute and di f ferential temperature signals are available.
The exact value for the differential temperature ( T) and the unheated junction temperature. (T ) level setpoints are based on test U
results and a setpoint calculation.
The T setpoint is 200*F.
It is selected to ensure that covered and uncovered conditions can be disti nguished from each other unambigu-ously.
The unheated junction (T ) setpoint is 700F.
The TU V
setpoint is used to ensure' a continuous indication of sensor uncovery in high temperature enviroments when the applied heated j unction he.ter power is cut back to prevent overheating the heated junction themocouples.
B.
Detennine the maximum upper head fluid temperature from the top three unheated themocouples for use as an input to the saturation margin monitor.
C.
Process all inputs and calculated outputs for display.
D.
Provide an alam output when any of the heated junction thennocouples detects the absence of liquid level.
E.
Provide control of heater power for proper heated junction themocouple output signal level.
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9
(E4) o The following information is displayed on the QSPDS displays:
A.
Liquid inventory level above the fuel core.
B.
Unheated junction temperature at the eight positions.
C.
Heated junction temperature at the eight positions.
D.
Temperature differential at the eight positions.
The following plant-specific information is hereby provided:
i)
Paragraph 3.1.1 of Reference 3.2 listed C-E's recommended instrument ranges for certain SMM inputs:
RECOMMENDED FC1 INPUT RANGE RANGE Pressurizer Pressure 0-3000 psia 0-2500 psia Cold Leg Temperature 60-710"F 465-615*F Hot Leg Temperature 60-710"F 515-665*F UHJTC Temperature 100-1800*F 32-2300'F Pressurizer pressure indication provided is adequate to nonitor all anticipated pressure events analyzed in the USAR.
For those analyses the pressure will not exceed the 2500 psia end point.
A commitment has been made (Ref. 2.6) to upgrade the qualification and range of the hot and cold leg RTD's.
ii) Paragraph 3.2 of Ref. 3.2 stated that alanns for the following ICC variables will be provided with setpoints to be predetermined on plant-speci fic bases.
For Fort Cal houn, these setpoints are as follows:
VARIABLE SETPOINT Temperature Subcooled Margin
< 10*F subcooled Core Exit Temperature T 670*F Reactor Vessel Level T 100%
10
l (E4) 1.b.
detailed description of existing instrumentation systems.
Reference the District's submittal of December 31, 1979.
Implementation ~ has been completed as planned.
Range upgrade and LOCA qualification of the primary system hot and cold leg RTD's will be per the schedule supplied in reference 2.6.
11
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(E4) 1.c.
description of completed or' planned undifications.
A description of, and schedule for changes to the SMM RTD's was provided in reference 2.6.
Addi tionally, a modification to the range of the pressurizer pressure transmitters will be perfonned if the resolution of the ATWS issue indicates the necessity for such an upgrade (reference 2.9).
O i
-4 12
+ ~
(E4) 0 2.
A design analysis and evaluation of inventory trend instrumentation, and test data to support design in item 1.
The design analysis and evaluation of the ICC detection instrumenta-tion is discussed in sections 2.0 and 4.0 and Appendix A of refer-ence 3.2.
Additional information on the design analysis and evalu-ation is provided in reference 3.8.
Reference 3.1 provides answers to previous NRC questions regarding the HJTCS.
Test data to support the, design of the HJTCS is provided in refer-ences 3.3, 3.4, and 3.6.
NRC analysis of these evaluations is provided in reference 1.2.
9
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v (E4) o 3.
Description of tests planned and results of tests completed for evaluation, qualification, and calibration of additional instrumen-tation.
System qualification and system verification testing is discussed in sections 5.0 and 6.0 of reference 3.2.
Additional details and test documentation is provided in references 3.3, 3.4, and 3.6.
A functional test procedure, designed to verify perfonnance of the QSPDS, includi ng sof tware routines and proper di splay functioning,
has been prepared and pe rfo rmed.
Addi tion ally, a hardware test procedure, designed to ensure proper function and calibration of the QSPDS microprocessors, has been perfo rmed.
Portions of these tests will be incorporated into a routine system test procedure to demon-strate and assure continued system performance and calibration of the ICCI associated instrumentation.
14 i
(E4) 4.
Provide a table or description covering the evaluation of confor-mance with NUREG-0737: II, F.2, Attachment 1, and Appendix B (to be reviewed on a plant specific basis) 4.1.
Provide diagram of core exit thermocouple locations or reference the generic description if appropriate A diagram showing relative core exit thermocouple locations was provided as figure 1 of Attachment 1 to reference 2.6.
15
(E4) 4.2.
Provide a description of the primary operator displays including:
4.2.a A diagram of the display panel layout for the core map and description of how it is implemented, e.g.,
hardware or CRT display.
Primary operator displays for the core exi t thennocouple systen are the plant SPDS CRT's located in the main control room.
Attached are the two primary display pages showing the CET information.
Display pages from the backup display are included in the response to item 4.3.
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(E4) 4.2.b. Provide the range of the readouts The range for all core exit thermocouples is 32*F to 2300'F.
19
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(E4) 4.2.c. Describe the alam system The ERFCS alarms specific information which is of importance to the operators.
For ICC instrumentation, the alam func-tions are as follows:
When the infomation is received from the QSPDS micro-processor, it is checked for High-High, High, Low, Low-Low, and out-of-range (failed sensor) and an alam is provided.
Alam f s provided by (1) an audible alann in the control roon CRT function keyboard; (2) 111umi-nating the keyboard pushbutton associated with the top level (critical function) di splay; (3) changing the color of the associated critical function alam box (located in lower left corner of the SPDS di splay pages) from the normal green to yellow (alert) to magenta ( for alam).
This alann method provides for the directing of the operator to the appropriate di splay page which contains the information on the parameter that has gone into alarm.
On this page, the alanned variable is distinguishable by a flashing value (as opposed to the normal constant display) and the numbers are displayed in the appropriate color (yellow or magenta).
I 20
(E4) 4.2.d. Describe how the ICC instrumentation readouts are arranged with respect to each other To ef fectively organize the information presented by the plant SPDS, both the QSPDS and the ERFCS utilize a top-down, multi-level heirarchy of displays.
The ERFCS display pages are arranged in a five level heirarchy which consists of:
(i)
Level 1 display pages which include the primary Safety Function Matrix and the Display Directory pages.
(ii) Level 2 di splay pages which include the primary safety function overviews (i.e.,
Reactivi ty Control,
Core Heat Removal,
Reactor Yessel Integrity, etc.)
(iii) Level 3,
4, and 5 display pages which provide more detailed information including P&ID's, etc.
The heirarchy of ICC related di splays in the ERFCS is as follows:
210 Core Heat Removal 310 Reactor Core (Includes Core Map) 311 Reactor / Core Exit T/C's 220 RCS Pressure & Inventory Control 320 Primary System (P&ID)
Display pages on the same level and pages one level up or down can be accessed by a single keystroke at the CRT func-tion keyboard.
The display heirarchy for the QSPDS is much the same as that for the ERFCS.
At the top of the three level system are the system overview pages.
Level two displays provide increased detail for the ICC variables and level three displays provide detail s on the HJTC and CET temps.
QSPDS ICC related displays have the following heirarchy:
101 Reactor Core (overview) 211 Saturation Margin 212 Reactor Vessel Level 321 HJTC Temperatures 213 Core Exit Thennocouples 331 Core Map Di splay pages on the same level and pages one level up or down can be easily accessed by a minimum number of keystrokes at the display unit keypad.
21
~_
(E4) 4.3. Describe the implementation of the backup display (including the subcooling margin monitors), how the thermocouples are selected, how they are checked for operability, and the range of the display.
The backup displays for ICC i nstrumentation are the QSPDS plasma display units.
These plasma display units (PDU's) are installed in the control room in the QSPDS panels.
The two display units, and the two QSPDS sys tems, are redundant to each other and are com-pletely electrically isolated from each other.
Each QSPDS train processes and displays one-half of the total number of CET's (14 of 28 total).
A minimum of 3 CET's in each core quadrant are processed and dis-played by each QSPDS channel.
Out-of-range checks are perfonned on all analog inputs to the QSPDS to determine operabili ty.
Addi tion ally, iterative statistical analyses are perfonned on the CET inputs and any CET indications that do not f all within an acceptable temperature band are flagged as " suspect."
Temperature ranges for CET's are 32 F to 2300'F.
Attached are the QSPDS display pages associated with the ICC instrumentation.
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(E4) 4.4.
Describe the use of the primary and backup displays.
What
-training will the operators have in using the core exit thermocouple instrumentation?
How will the operator know when to use the core exit thermocouples and when not to use them?
Reference appropriate emergency operating quideline where applicable.
The primary ICCI displays are the plant SPDS CRT's located in the control room.
The backup displays are the QSPDS Plasma Display Units (PDU's) mounted in the QSPDS panels in the con-trol room.
The backup displays are required for operator use only in the event the primary SPDS displays are not operable.
Emergency Operating Procedures (E0P's) consistent with the standard C-E EPG's are currently being devel oped.
Training in the use of the CET's will be covered when these plant-specific E0P's are finalized.
31
(E4) 4.5.
Confirm completion of control room design task analysis applicable to ICC, instrumentation.
Confirm that the core exit thermocouples meet the criteria of NUREG-0737, Attachment 1
and Appendix B,
or identify and justify deviation.
Reference 2.11 stated that a task analysis (in accordance with the schedule for implementing the upgraded E0P's) will be completed prior to Cycle 10 startup.
The core exit thermocouples meet the criteria of Attachment 1 and Appendix B to NUREG-0737 as described in the responses to the various items contained in this submittal.
32
u-(E4) 4.6.
Describe what parts of the system are powered from the 1E power sources used, and how isolation from non-1E equipment is provided.
Describe the power supply for the primary display.
Clearly delineate in two catagories which hardware j
is included up to the isolation device and which is not.
i All portions of the QSPDS microprocessors, the heaters for the HJTC's, the QSPDS PDU's, and the isolation system, up to a nd includi ng the QSPDS end of the fiber optic link is powered fra the station vital buses.
Isolation of the class 1E QSPDS from the non-1E portion of the SPDS is accmplished by a fiber optic data link which provides complete electrical i solation between the two sys tems.
See figure 1 for a clear delineation of.1E/non-1E portions of the system.
Power supply for the primary display system is offsite power, backed up by a highly reliable uninterruptible power supply (U PS) and an emergency diesel ge nerato r, sepa rate fran the station diesel generators and other plant power supplies.
i l
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r c., s 4.7.
Confirm the environmental qualification of the core exit thermocouple instrumentation up to the isolation device.
The core exi t thermocouples, their as sociated cabli ng, the QSPDS microprocessors and displays and the isolation device in the QSPDS (i.e., the fiber optic modem) are qualified for use in their respective environnents.
See response to iten 1 of Appendix 8 to NUREG-0737.
l 1
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(E4) 5.
Describe computer, software and display functions associated with ICC monitoring in the plant.
For detailed infonnation regarding ICCI signal processing and dis-play functions see responses to questions 1.a, 4.2a, 4.2d, and 4.3 of this submittal.
Additional information is provided in section 3 of the reference 3.2.
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(E4) 6.
Provide a proposed schedule for installation, testing, and calibra-tion and implementation.of any proposed new instrumentation or-displays. to reference 2.6 provided a schedule for the completion of the installation, testing, calibration and implementation of the ICCI system.
36
u (E4) 7.
Describe quidelines for use of reactor coolant inentory tracking system, and analysis used to develop procedures.
Guidelines for the use of the ICC detection instrumentation are dis-cussed in section 2.0 and 7.0 of reference 3.2.
Additional guidance is provided in references 3.7 and 3.8.
4 l
37 i
v (E4) 8.
Operator instructions in emergency operating procedures for ICC and how these procedures will be modified when final monitoring system is implemented. to reference 2.3 provided a schedule for the implementa-tion of the upgraded Emergency Operating Procedures (EOP's).
These procedures are being developed based on the generic Combustion Engi-neering Emergency Procedure Guidelines (EPG's, reference 3.7.)
38
(E4) 9.
Provide a schedule for additional submittals required.
9.1.
Discuss the spacing of the sensors from the core alignment plate to the top of the reactor vessel head.
How would the decrease in resolution due to the loss of a single sensor affect the ability of the system to detect an approach to ICC7 Sensor placement is discussed in paragraph 3.1.2 of reference 3.2 and again, in more detail, in paragraph 4.5 of reference 3.8.
The actual spacing for the Fort Calhoun sensors is as follows:
INCHES INCHES AB0VE BELOW LEVEL SENSOR N0.
FUEL ALIGN.
TOP 0F INDICATION (%)
PLATE HEAD (Top of Head) 194 23/32 0
1 180 7/32 14 1/2 100 2
140 11/32 52 3/8 83 (Upper Guide Structure) 128 5/8 67 3/32 3
104 17/32 90 3/16 63 (Top of }lot Leg) 4 66 5/8 128 3/32 43 (Center of Hot Leg) 5 50 5/8 144 3/32 29 (Bottom of Hot Leg) 6 34 5/8 160 3/32 21 7
22 11/32 172 3/8 14 8
10 3/32 184 5/8 8
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194 23/32 For addi tional infonnation regardi ng decrease in resolution due to loss of a single sensor, see reference 3.1, page 2.8-1.
A schedule for addi tional document submi ttal s has been provided per attachment 2 of reference 2.6.
39
(E4)
NUREG-0737, Appendix B Confirs explicitly the conformance to the Appendix B item listed below for the ICC instrumentation, i.e., the SM, the reactor coolant inventory tracking system, the core exit thennocouples and the display systems.
1 Environmental qualification The CET's and HJTC sensors have been designed such that the ex-vessel portions of the sensors and the electrical connecto rs, including the in-containment cables, will continue to function during and after design bases accident condi tions in a degraded containment environment resulting from a LOCA/MSLB.
The conclusion is based on the results of a qualification test program conducted in accordance with the methodology and guidelines outlined in IEEE-323-1974 and IEEE-344-1975.
Test results are documented in the following C-E reports:
CE-NPSD-230-P Class 1E Qualification of the CET-MI Cables CE-HPSD-251-P Cla ss 1E Qualification Test of the Ex-vessel Portion of the HJTC-MI cable Interface Through a G&H Connector CE-NPSD-253-P Class 1E Qualification Test of the HJTC-MI Cable and Litton Connectors CE-NPSD-271-P Class 1E Qualification Test of the Litton Connectors CE-NPSD-272-P Class 1E Qualification Test of the Retrofitted G&H Connectors TR-ESE-460 Vibration and Aging Seismic Qualifica-tion of the CET and Cable Test Report The in-vessel portions of the CET and HJTC sensors have been t
designed and proven tc perfonn in an extremely harsh enviroment.
Environmental qualification testi ng has not been performed other than nonnal proof-of-operation testing.
The remainder of the ICCI system is located in mild environment and this equipment (up to and including the fiber optic modem isolation devices) is quali fied environmentally to IEEE-323-1974 and seis-mically qualified according to IEEE-344-1975.
40
u (E4) 2.
Single failure analysis Two physically separate, electrically isolated trains of ICC instru-mentation is provided.
No single failure within either the instru-mentation, its auxiliary supporting features, or their power sources concurrent with the failures that are a condition or result of a spect fic accident will prevent the operators fra being presented the ICC information.
In the remote possibility that one cmplete channel of instrumenta-tion is lost, there is sufficient diversity of information presented by the surviving channel to prevent infomation ambiguity.
41
(E4) 3.
Class 1E power source Each channel of the QSPDS, including HJTC heater control and HJTC, SMM and CET signal processing, the backup displays, and the class 1E to non-1E i sola tion device are powered fron the station's vi tal instrumentation buses.
These buses are energized from the station standby power sources and are backed up by batteries.
42
v (E4) 4.
Availability prior to an accident ICC i nstrumentation channels are available prior to an accident.
Availability will be addressed in the technical specifications.
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v (E4) 5.
Quality Assurance The District's position on the various QA requirements applicable to Fort Calhoun is delineated -in the Fort Calhoun QA program as described in the Fort Calhoun USAR ( Appendix A), and the Quality Assurance Plan.
Nuclear Safety Related components and services which fall within the scope of.10 CFR 50, Appendix B are classified as CQE (Critical Qual-ity Elements).
CQE components are procured and maintained in accor-dance with the District's QA Plan.
The ICC instrumentation desig-nated class 1E in figure 1 falls within the CQE definition and is procured and maintained in accordance with the applicable QA Plan requi rements.
The District believes that compliance with the requirements as out-lined in our QA Plan is adequate to meet the intent of item 5 of Appendix B to NUREG-0737.
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(E4) 6.
Continuous indications Continuous indication of ICC instrumentation is avail able, upon demand, at the SPDS CRT's located in the control room.
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45
7 (E4) 7.
Recording of instrument outputs Recording of ICCI outputs is available in the ERF computer system.
Data is stored and displayed continuously on demand.
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(E4) 8.
Identification of instruments The ICC indication in the control room is provided by the SPDS dis-plays.
Adequa te training is provided to the ope rato rs to. ensure that they are aware that these displays are intended for use under accident conditions.
47 m,
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go (E4) 9.
Isolation Isolation of ICCI signals from the class 1E portion of the SPDS to the non-1E portion is accomplished by the fiber optic data link which.is designated as part of the monitoring instrumentation and meets the provisions of Appendix B to NUREG-0737.
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