ML19329D652
| ML19329D652 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/02/1976 |
| From: | Rodgers J FLORIDA POWER CORP. |
| To: | Darrin Butler Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8003160275 | |
| Download: ML19329D652 (12) | |
Text
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NRC nlm RIBt! TION FOR PART 50 DOCN~ f.iATERIAL
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(FEMPOR ARY I:ORIA)
CONTROL NO:_los2 FILE:
FROM: FLORIDA POUER CORP DATE OF DOC DATE REC'D LTR TWX RPT OTIlER ST PETERSBURG. FL 2-2-76 2 4-76 XXXX
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DR D A BUTLER 1 SIGNED 20 SENT LOC /sl PDR XXXX CLASS UNCLASS PROP!NFO INPUT NO CYS REC'D DOCKET NO:
XXXXXX 21 50-302 DESCRIPTION:
ENCLOSURES:
LTR RE OUR 11-11-75 TRANS THE FOLLOWING.....
RESPONSE TO REVIEW OF BAW-10003 QUALIFICATION TESTING OF PROTECTION SYSTEM INSTRUMENTATION FOR CRYSTA1. RIVER UNIT #3........
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Dr. D. A. Butler, Chief Light Water Reactors Branch #4 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Wa"shington, D.C.
20555 In Re:
Florida Power Corporation Crystal River Unit #3 Docket No. 50-302
Dear Dr. Butler:
Attached are twenty (20) copies of our response to Section 6 of your staff's evaluation of topical report BAW-10003, Revision 4, " Qualification Testing of Protection System Instrumentation,"
contained in Mr. A.
Schwencer's letter of November 11, 1975.
It is hoped that our response to your staff's concerns will allow them to complete their review of BAW-10003 and its applicability to RPS equipment quallrication testing for Crystal River Unit #3.
Please contact this office if additional discussion-is-
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required.
Very truly yours,
.c J.
. Rodgers Asst. Vice President JTR/iw Attachments.
1052 General Office 3201 Thirty-fourin street soutn. P.O. Box 14042. St Petersburg Florida 33733 813--866-5151
QUESTION 6 a.:
The still air temperature of the equipment must be maintained within a range of 40 to 110* F at relative humidities of 50%.
RESPONSE
The complete' control building system, during normal and post accident operation, maintains 75'F t 2*F and limits the relative humidity to maximum of 50%.
No humidity is added to the control complex air so that during periods of low outside air moisture the relative humidity may fall well below the 50% level.
QUESTION 6 b.:
The interconnecting wiring and connectors must be qualified to meet the environmental and seismic plant design criteria.
RESPONSE
The internal wiring and connectors used with the Bailey Meter Company systems supplied by B6W were qualified during environ-mental and seismic qualification tests.
Test fixtures used during the tests included wiring and connectors used on Bailey Meter Company systems (see Page 6-4 of BAW-10003).
All interconnecting cable specified by GAI is qualified for the environmental conditions which will be encountered.
More details of cable testing is given in FSAR Table 8-2.
None of the GAI specified cable used with safety related instrumentation or logic modules was supplied with connectors.
All of the GAI cables were hardwired in the respective panels, cabinets, and relay racks and seismically qualified (Reference " Control Boards G Engineered Safeguards Actuation Relay Cabinets Seismic Test Report", by Environmental Engineering Laboratory, Report No. 6984 and 6984 Supplement 1.)
Therefore, all cable and connections within the GAI scope meet plant environmental and seismic conditions.
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QUESTION 6 c.:
The electrical' power sources for the instrumentation shall maintain to the cabinets of 107 to 127 V.AC at 58 to 62 Hz.
RESPONSE
..All of the power sources for safety-related instrumentation for Crystal River #3 are within the range identified above.
QUESTION 6 d.:
The Applicant's Chapter 15 analysis shall be based on assumed errors for the following equipment which are not less than those listed therein.
(1)
Reactor Coolant Pressure Foxboro Equipment
+ 1.85%
Motorola Equipment
[ 3.69%
(2)
Reactor Coolant Flow
+ 3.0%
(3)
Reactor Coolant Temperature
+ 1.0%
(4)
Reactor Coolant Pump Status
-+ 0.88%
Monitor (5)
Reactor Building Pressure
+ 1.0%
-50%
(6)
Neutron Monitoring Total Flux
+ 4.0%
Flux Imbalance
[5.0%
RESPONSE
The following are corrections made to the above question prior to. answering:
4 (1)
The Safety Analysis.for Crystal River #3 is contained in Chapter 14 of-the FSAR.
(4)
Preamplifier 100 ns rise time (5)
Count rate amplifier 400 s 0 0.'1 H:
1s@ 106 H:
(6)
Rate-of-change amplifier
.c10s (7)
Buffer amplifier 50 ms, 1s (8)
Linear amplifier 50 ms
-11 (9)
Logarithmic amplifier 10 s @ 10 A
0.1 s @ 10-3 A (10)
Square root extractor Incr 250 ms
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Dect 500 ms (ll.)
Linear bridge 10 ms (12)
Signal converter 300 ms (temp output)
(13)
Signal converter 300 ms (press. output)
(14)
Sum / difference amplifier E
10 ms Scaled E g (15)
Function generator Slope 15 ms (16)
Bistable 100 ms (17)
Reactor Trip 30 ms (interlock or test trips)
(18)
Contact Buffer Modules energize 40 ms de-energize 100 ms (19)
Logic Buffer Module 10 ms (20)
Unit. Control Module Trip 50 ms e
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(2)
The Motorola Equipment listed in item 1 of the question should be replaced with Westinghouse Equipment.
Item 4'- Reactor Coolant Pump Status Monitor -
is not applicable to Crystal River #3.
The Chapter 14 analysis is based on assumed "end-to-end" instrument strings rather than individual equipment.
The string errors assumed are greater than or equal to those errors listed in Question 6 d.
The flow rate assumed in the Chapter 14 Accident Analysis is less than the predicted actual Reactor Coolant flow rate by a margin that is greater than the flow measurement error listed in Question 6 d.
QUESTION 6 e.:
If Motorola pressure sensors are used, the technical sp.ecifica-tions shall require calibration not less frequently than once every four months.
RESPONSE
There are no Motorola pressure sensors in use at Crystal River Unit #3.
In use at Crystal River Unit #3 are Westinghouse Model #59PH4 Pressure Sensors which are certified to an accuracy of span of 0.5% and a 0.1% span repeatability accuracy.
These sensors are calibrated as part of the channel calibration once every refueling as required by STS Surveillance Requirement 4.3.1.1.1.
,-QUESTION 6 f.:
The analysis in Chapter.15 of the Safety Analysis Report shall be based on assumed response times which are equal to or greater than the following times required for a response to 99% of the differential change from a step input:
(1)
Pressure Detectors 250 msec.
(2)
Differential Pressure 250 msec.
Detectors (Flow)
(3)
-Temperature Detectors Models 5 sec*
177 GX and-JD M,odel-177 H 3 sec*
- time to change 63.2% of applied step change in input
RESPONSE
'The response time values as stated in Criteria 6.f. represent time constants for sensors and step response times for in-dividual modules as performed during laboratory tests and reported in BAW-10003.
However, safety analysis uses end-to-end string responses which cannot be calculated by summation of the individual module response times listed in Criteria 6.f.
In-
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stead of such a summation, the safety analysis in Chapter 14 is ba. sed on the addition of the sensor's time constant, used as 1 pure delay, which is then added to the cabinet mounted modules' string response time (also assumed as a pure delay).
The time constant is used as a pure delay since the dynamic response of the plant variable has a time constant greater than the instru-ment string.
Only those modules utilized in trip strings are tested for string response.
Other equipment, while housed in the same cabinet but whose functions are not utilized in safety analysis, is not tested for string response.
The re-qui'ed trip string response for cabinet mounted modules is r
specified to the equipment vendor and is verified prior to shipment of every system leaving the factory.
The actual values obtained during such response tests are documented in the Q.A.
data manual of every contract.
Table 1 entitled " Sensor Time Constants", tabulates the required values along with typical values obtained during tests of the sensors.
Table 2 entitled " Cabinet Equipment String Response" tabulates the required values along with typical values of string response obtained during checkout of each trip string of the cabinet mounted modules.
A simulated step input is utilized with the starting value teing representative of a normal operating plant variable and the end value above (or below) the trip setpoint value.
The string response (from input terminal to reactor trip module output) is recorded and documented in the Q.A. data manual.
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9 TABLE 1 SENSOR TIME CONSTANTS PLANT VARIABLE REQUIRED TYPICAL SENSOR SIMULATION VALUE TESTED VALUE Pressure Decrease Pressure 220 ms*
25 ms Temperature Increase Temp.
177 GW 3.0 sec.**
3.25 sec.
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17 HW 5.0 sec.
5.25 sec.
R.C. Flow Decrease Flow 240 ms*
110 ms R. B. Pressure ***
Increase Pressure 240 ms 40 ms' (Mercoid)
Neutron Flux Increase Flux 10 ms*
(Ionic Delay not' sted)
- Required values assumed in Chapter 14 Safety Analysis.
- The required value is based on a 60 ft/sec flow, whereas the
-tested value is based on 3 ft/sec. flow which is_ conservative.
- The R.B. pressure switch provides a digital (on-off) signal.
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TABLE 2 CABINET EQUIP' MENT STRING RESPONSE TRIP STRING PLANT VARIABLE REQUIRED TYPICAL FUNCTION SIMULATION VALUE TEST VALUE Ove rpowe r*
Increase Flux 150 ms 105 + 10 ms Power / Delta Flux /*
Increase Flux 150 ms 105 + 20 ms Flow Decrease Flow 240 ms 160 +.20 ms High Pressure
- Increase Pressure 150 ms**
85 + 10 ms Low Pressure
- Decrease Pressure 150 ms**
85 + 10 ms High Temperature Increase Temperature 150 ms 105 + 10 ms Pressure / Temperature Decrease Pressure 150 ms 80 + 10 ms R.B.
Pressure Increase Pressure 150 ms 90 + 15 ms
- TripstringfunctionsassumedinChafter14SafetyAnal[hs.
- The required value is 140 ms if mono ilar stators are u e O
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QUESTION 6 g.:
The applicant shall provide reactor coolant pump status monitors and detectors which shall have a combined accuracy of 0.88%
and a response time of 6 ms or less.
RESPONSE
Question g is not applicable to Crystal River Unit #3, as reactor coolant pump status monitors are not used.
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QUESTION 6 h.:
The applicant shall provide a description of and procedures for testing to assure that the equipment is more accurate than and responds in a shorter time than the limits presented in require-ments (d), (f), and (g) above.
RESPONSE
Item (d) - During test procedure TP-305-1, RPS Pre-Operational Calibration, the string accuracies of the Reactor Protection System are established.
The string accuracy is determined from the transmitter (where possible) to the actuation device.
This accuracy is then compared with the allowable values con-tained in the Crystal River #3 Standardized Technical Specifica-tions for Acceptance.
These procedures and test results are reviewed on site by inspectors from Region II of the Inspection 6 Enforcement Branch of the NRC.
Item (f) - Field tests for detectors require that the detectors be tested on an individual component basis for obtaining a time constant from a simulated input.
Field testing of installed detectors is difficult to accomplish for the following reasons:
(1)
Limited plant variable simulation using present state-of-the-art on-line or in-situ testing techniques.
(2).
Inability to simulate a step change in the plant variable compatible to the laboratory test simulation.
However, the RPS detectors were furnished as part of the B6W NSSS package and certification as to the acceptability of the response times of these detectors is contained in our Quality
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files.
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' During test procedure TP-600-23, RPS Functional Test, the response times of all strings and combinations of trips within the string of the Reactor Protection System are verified.
During-this test the cabinet response times (excludes the detector and control rod drive breakers) of the RPS functional channels will be verified.
Specifically, a controlled simulated trip signal is injected to the input of the RP system.
The response time is measured from the point in time the limit is exceeded until power to the AC-CRD under voltage coils is interrupted.
The measured RPS cabinet response will be compared against the response time values listed in the Crystal River #3 FSAR for acceptance.
The Standardized Technical Specifications (STS) for Crystal River *# 3 require that the response times of the following RPS channel functions be measured according to the STS definition of response time as these channel func-tions were used in the Safety Analysis.
(1)
Nuclear Overpower Based on RCS Flow.
(2)
RCS Pressure - Low.
(3)
RCS Pressure - High.
(4)
Nuclear Overpower.
(5)
Nuclear Overpower Based on Imbalance.
The STS definition requires that the RPS respor se time be measured from the time the monitored parameter exceeds its trip setpoint at the channel detector until power interruption at the control rod drive breakers.
During TP-6'0-23, the response times of the RCS Flow and Pressure de ectors and the response times of the control rod breakeIJ will be measured.
These response t?.mes will be added to the RPS cabinet response times pre tously deter-mined in TP-600-23 to obtain the total channel response time.
The total response time will then be compared against the values listed in Table 3.3-2 of the STS for Nuclear Overpower Based on RCS Flow, RCS Pressure-Low, and RCS Pressure-High for acceptance.
The Crystal River #3 STS exempt neutron detectors from response time testing.
Therefore, the response times for Nuclear Overpower and Nuclear Overpower based on Imbalance will consist of the sum of the RPS cabinet responet time and the control rod breaker response time which will be measured in TP-600-23 for these RPS channel functions.
The total response time will then be compared to the values listed in Table 3.3-2 of the Crystal River #3 STS for Nuclear Overpower aid haclear Overpower Based on Imbalance for Acceptance.
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- n Each of the four (4) RPS channels will be tested in-dividually'during TP-600-23 while blocking the other three (3) channels.
The response time testing of the RPS will be performed during Hot Functional Testing.
The procedures used for response time testing of the Reactor Protection System and the results of these tests are reviewed on site by Region II of~the Inspection 6 Enforcement Branch of the NRC.
Item (g) - Reactor Coolant Pump Status Monitors are not applicable to Crystal River #3.
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