ML20215L087
| ML20215L087 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/18/1987 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20215L065 | List: |
| References | |
| NUDOCS 8706250540 | |
| Download: ML20215L087 (6) | |
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. LIMITING SAFETY SYSTEM SETTINGS-f,&
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BASES q
Intermediate and Source Range, Neutron Flux
'The Intermediate and Source Range, Neutron Flux trips provide core i
protection during reactor startup to mitigate the consequences of an uncon-4 trolled rod cluster control assembly bank withdrawal from a suberitical condition. These trips provide redundant protection to the. Low Setpoint-trip f
of the Power Range, Neutron Flux channels. The Source Range channels will.
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initiate a Reactor trip at about 108 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually; blocked whe' P-10 becomes active.
a n
i Overtemperature af i
The Overtemperature AT trip provides core protection to prevent DNS.
l for all combinations of pressure, power, coolant temperature, and axial power
. distribution, provided that the transient is slow with respect to piping' transit delays from the core to the temperature detectors (about 4. seconds),
1
~
,N and pressure is within the range between the Pressurizer High and. Low Pressure i
j trips. The Setpoint is automatically varied with: (1) coolant temperature to j
7 correct for temperature induced changes in density and heat capacity of water:
and includes dynamic compensation for piping delays from the core to the. loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-l
' tion. With normal axial power distribution, this Reactor trip limit is always below the core' Safety Limit as shown in Figure 2.1-1.. If axial peaks are l
greater than design, as indicated by the difference between top and bottom I
power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
g T d A
- Overpower ai i
i The Overpower aT trip provides assurance 'of fuel integrity (e.g.,'no f uel pellet melting and less than 1% cladding strain) under all possible'
)
overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is auto-J J
matica11y varied with:
(1) coolant temperature to correct for temperature k
induced changes in' density and heat capacity of water,'and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the l
loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in
{
WCAP-9226 " Reactor Core Response to Excessive Secondary Steam Releases."
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w TABLE 4.3-1-(Continued).
i 1
TABLE NOTATIONS "Only if the-Reactor Trip System breakers happen-to be closed and the Con-trol Rod Drive System is capable of rod withdrawal.
- The specified 18 month frequency may be waived for Cycle I provided the surveillance is perfomed prior to restart following the ~(f rst refueling-outage or June 1,1986, whichever occurs first. The provisions of.
Specification _4.0.2 are reset from performance of this surveillance.
L
' NBelow P-6 (Intermediate Range Neutron Flux interlock)' Setpoint.
N#8elow P-10 (Low Setpoint Power Range Neutron Flux interlock) Setpoint.
(1) If not performed in previous 31 days.
(2) ' Comparison of calorimetric to excore power' indication above 15% of RATED i
THERMAL POWER., Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 21 The provisions of Specification 4.0.4 are not applicable for entry into M00E.2 or 1.
'(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15%
t of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 31 The provisions of Specification 4.0.4 are not appli-cable for entry into MODE 2 or 1.
. (4) _ Neutron detectors may be excluded fros' CHANNEL CALIBRATION.
(5) Detector plat' eau turves shalT be obtained, evaluated and compared to manu-facturer's-data. _ For the Intermediate Range and Power Range Neutron Flux channels ~ the provisions of Specification 4.0.4 are not applicable for entry l
into MODE 2 or 1.
~
Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The proyi-(6) sions of ' Specification 4.0.4 are not applicable for entry into MODE 2 or 1. TMB }
('7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall' independently verify the OPERASILITY of the Undervoltage and Shunt Trip Attachments of the Reactors j
Trip Breakers.
(8) Deleted I
. (9) Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Quarterly i
- survei.11ance shall include verification of the Baron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minuta period.
a (10) Setpoint verification is not required.
~
(lli Following maintenance or adjustment of the Reactor trip breakers, the
~
l TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent veri-
- fication of the Undervoltage and Shunt trips.
l (12) At least once per 18 months during shutdown, verify that on a simulated
~ Boron Dilution Doubling test' signal the normal CVCS discharge valves will close and the centrifugal charging pumps suction valves from the RWST will open within 30 seconds.
- CALLAWAY - UNIT 1 3/4 3-12 Amendment No. 1.9 et *# FW M.q M M pee 4 ggggap.y g
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-Insert A
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Delta-T, as used in the Overtemperature and Overpower Delta-T trips, Eepresents the 100% RTP value as measured-by the plant for
-l each loop.
This normalizes each lcop's AT trips to.the actual operating conditions existing at the. time of measurement, thus
'l forcing the trip to reflect the equivalent' full power conditions as-assumed in the accident analyses.
These differences in vessel j
AT can arise due to several factors, the most prevalent being.
measured RCS loop flows greater.than Minimum. Measured Flow, and slightly asymmetric power distributions between quadrants.
While RCS loop flows are-not expected to change with cycle life, radial
. power redistribution between quadrants.may occur, resulting in small changes in. loop specific vessel AT values.
Accurate determination of the loop specific vessel AT value should be made when performing the Incore/Excore quarterly recalibration and
l under steady state conditions (i.e., power distributions not affected by Xe or other transient conditions).
Insert B
-Determination of the loop specific vessel AT value should be made when performing the Incore/Excore quarterly recalibration, under i
steady state conditions.
I i
l l
i Insert Page
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