ML20041F975
| ML20041F975 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/09/1982 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20041F969 | List: |
| References | |
| TAC-48020, NUDOCS 8203170610 | |
| Download: ML20041F975 (6) | |
Text
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TECHNICAL SPECIFICATION CHANGE REQUEST NO. 94 Replace Appendix A pages 2-6, B2-7 and 3/4 3-6 with the attached revised pages 2-6, B 2-7 and 3/4 3-6.
Reason for Proposed Changes A.
Pages 2-6 and B 2 Florida Power Corporation agreed to take credit for the pump 120% overpower trip to resolve NRC concern that faulted conditions in the Reactor Coolant Pump Power Monitors (RCPPM) could mask loss of flow conditions. These changes address those concerns.
B.
Page 3/4 3 The change to Item 8 is proposed to allow the installation of additional time delay relays in the RCPPM circuitry to prevent spurious trips.
Furthermore, this change corrects a previous error supplied by Florida Power Corporation based on a misunderstanding of Babcock and Wilcox supplied data.
Technical Specifications Change Request No. 77 proposed an RCPPM response time of 0.470 seconds which was based on BAW - 1684 " Crystal River Unit 3 Cycle and Reload Report". This total was comprised of the following components:
.240 Sensor (CT, PT, WT, etc.)
.150 Reactor Protective System
.080 CRD Breaker Delay 0.470 TOTAL This was that set of equipment that Babcock and Wilcox felt could and should be tested in place. This was never utilized as an analytical assumption, nor did it reflect corrections due to the "RPS Instrument Inaccuracy Concern." Two other changes to this page were part of this error re-analysis but not the RCPPM value.
The changes to the other channels (Items 2, 5 and 6) are proposed to resolve a difference between the definition of response time between the FSAR and Technical Specifications. The FSAR analyses include as response time, the time betvieen the monitored parameter change and rod insertion.
The Technical Specifications only include change in parameter through opening of the control rod drive breakers. The remainder of the trip string (CRDM release delay and rod insertion are addressed by Technical Specification 3.1.3.4, (ROD DROP TIME).
This difference of 0.060 seconds was not correctly reflected in the Technical Specification tables.
l 8203170610 820309 PDR ADOCK 05000302 i
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J Safety Analysis The safety analyses used the following information for response time.
TABLE 1 Dedicated Delay Time Sensor &
Breaker CRD Release Margin for Assumed for Functional Unit RPS Delay Delay Delay Uncertainty FSAR Anal.
Nuclear Overpower 0.186 0.080 0.060 0.000 0.326 Nuclear Overpower based on RCS flow and Axial Power imbalance 1.690 0.080 0.060 0.020 1.850 RCS Pressure High and Low 0.261 0.080 0.060 0.099 0.500 Pump Status Based on (RCPPM's) 0.480 0.080 0.060 0.000 0.620 Using the Technical Specification definition, response times assumed in the FSAR supporting analyses should be decreased by.060 seconds to the values on the attached revised Table 3.3-2.
Operation with these time delays is, therefore, supported by the safety analyses and will not constitute an unreviewed safety 7
question.
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TABLE 2 EVALUATION OF RCPPM RESPONSE TIMES ORIGINAL DESIGN POST CHANGE POST CHANGE CURRENT COMPONENT RESPONSE TIME ADDING OPTIC ISOLATORS ADDING TIME DELAY RELAYS RESPONSE TIME CT/PT Included w/ Watts 16 ms 16 ms (1) 20 ms (2)
Transducer Watts Transducer (3) 240 ms (4) 72 ms (5)
Allocated 72 ms 72 ms
- 2t40 ms.
Results in Time Delay Relay 2 ms.
150 ms (4) 139 ms (6) margin Optic isolator 150 ms (4) 106 ms (9) 106 ms RPS 150 ms (4) 131 ms (7) 131 ms
-104 ms (8)
CRD Breaker 80 ms (4) 80 ms 80 ms Subtotal 470 ms 451 ms 555 ms 441 ms Margin 90 ms 109 ms 5 ms 119 ms Total 560 ms 560 ms 560 ms 560 ms Release Delay 60 ms (4) 60 ms 60 ms 60 ms Safety Analysis Assumption 620 ms 620 ms 620 ms 620 ms NOTES (1)
GE Documentated Analysis Results Available (2)
Allocated Response Time (G.E. Documentated Analysis Results Available)
(3)
Includes Watts Transducer, Bistable and Output Relay.
(4)
Allocated Response Time (5)
B&W Test (By Rochester). Recent Test Date Indicate Actual Response Time of 22 ms WT only).
72 ms. Represents Worst Case Including: Watts Transducer, Bistable and Clare Output Relay.
(6)
Tested at FPC. Maximum Delay on Any Relay Was 139 ms. Each Relay Tested Five Times.
(7)
Specification Requirements Supported by Type Test Data (See BAW - 10003 and/or references therein).
(8)
SP-ll5 Step 6.5.1.1 Includes Entire Channel From Output of Flux Monitor through CRD Breaker Opening, (i.e., RPS & CRD Breaker &
Additional Components).
104 ms. is Worst Case. Test is Periodically Repeated.
(9)
Type Test Data at 650F at 95% Confidence Level. Testing at Ambient Temperatures (740F) Yielded Even Lower Values.
TABLE 2.2-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTION UNIT TRIP SETPOINT ALLOWABLE VALUES I
S.
Pump Status Based on More than one pump drawing More than one pump drawing Reactor goolant Pump Power i3900 KW or 19000 KW L 3900 KW or 19000 KW Monitors I) 9.
Reactor Containment Vessel Pressure High i 4 psig i 4 psig u
E (1)
Trip may be manually bypassed when RCS pressure 11720 psig by actuating Shutdown Bypass provided that:
a.
The Nuclear Overpower Trip Setpoint is 15% of RATED THERMAL POWER b.
The Shutdown Bypass RCS Pressure - High Trip Setpoint of 11720 psig is imposed, and c.
The Shutdown Bypass is temoved when RCS Pressure > 1800 psig.
LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Containment Vessel Pressure - High The Reactor Containment Vessel Pressure-High Trip Setpoint d 4 psig, provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the containment vessel or a loss-of-coolant accident, even in the absence of a RCS Pressure - Low trip.
Reactor Coolant Pump Power Monitors In conjunction with the power / imbalance / flow trips, the Reactor Coolant Pump Power Monitors trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to more than one reactor coolant pump not operating.
A reactor coolant pump is considered to be not operating when the power required by the pump is z.120% or is s 70% of the nominal operating power. The nominal operating power decreases from when a pump is first started during heatup and is pumping dense fluid (typically 7500 KW) to when a pump is operating at full reactor power and is pumping less dense fluid (typically 5500KW).
In order to avoid spurious trips during normal operation, the 120% trip setpoint (900GKW)is based on the nominal operating power for a pump during heatup and the 70% trip setpoint (3900KW) is based on the nominal operating power for a pump cperating at full reactor power. Florida Power has agreed to take credit for the pump overpower trip in order to assure that certain potential faults such as a seismically induced fault high signal will not prevent this instrumentation from providing the protective action (i.e., a trip signal).
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TABLE 3.3-2 REACTOR PROTECTION SYSTEM INSTRUMENTATION RESPONSE TIMES Functional Unit Response Times 1.
Manual Reactor Trip Not Applicable 2.
Nuclear Overpower *
< 0.266 seconds 3.
RCS Outlet Temperature - High Not Applicable 4.
Nuclear Overpower Based on RCS Flow and 3
~ l.79 seconds AXIAL POWER IMBALANCE
- v E
.< 0.44 seconds 5.
RCS Pressure - Low w
6.
RCS Pressure - High
< 0.44 seconds 7.
Variable Low RCS Pressure Not Applicable 8.
Pump Status Based on RCPPMs
< 0.56 seconds 9.
Reactor Containment Pressure - High Not Applicable Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.