ML17250A594
| ML17250A594 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/09/1980 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17250A593 | List: |
| References | |
| NUDOCS 8009170306 | |
| Download: ML17250A594 (14) | |
Text
f.
Low reactor coolant flow -
90% of normal indicated flow.
2.3.1.3 g.
Low reactor coolant pump frequency -
> 57.5 Hz.
Other reactor tri s 2.3.2 a.
High pressurizer water level -
92% of span b.
Low-low steam generator water level -
5% of narrow range instrument span Protective instrumentation settings for reactor trip interlocks shall be as follows:
2.3.2.1 Remove bypass of "at power" reactor trips at, high power (low pressurizer pressure and low reactor coolant flow) for both loops:
Power range nuclear flux -
< 8.5% of rated power (1)
(Note:
During cold rod drop tests, the pressurizer high level trip may be bypassed.)
2.3.2.2 Remove bypass of single loss of flow trip at high 2.3.3 2.3.3.1 power:
Power range nuclear flux <
50% rated power Relay operating will be tested to insure that they perform according to their design characteristics which must fall in within the ranges defined below:
Loss of voltage relay operating time
< 8.5 seconds for 480 volt safeguards bus voltages
< 368 volts.
Measured values shall fall at least 5% below the theoretical limit.
This 5% margin is shown as the 5% tolerance curve in Figure 2.3-1.
- 2. 3-4
2.3.3.2 Basis:
Acceptable degrad'ed voltage relay operating times and s
setpoints, for. 480 volt safeguards bus voltages
< 414 volts and
> 368 voltsare defined by the safeguard equipment thermal capability curve shown in Figure 2.3-1.
Measured values..!shall fall at least 5% below the theoretical limit.
This 5% margin is shown as the 5% tolerance curve in Fi.'jure, 2.3-1.
The high flux reactor trip (low set point.) provides redundant protection in the power range for a power excursion beginning 1
from lower power.
This trip value was used in the safety analyst.s.
/
In the power range of operation, the overpower nuclear. flux reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.
The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur.
The reactor is prevented from reaching the overpower limit, condition by action of the nuclear overpower and overpower bT trips.
The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted.
The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set, pressure for these valves (2485 psig).
The low pressurizer pressure reactor trip also trips the reactor in the unlikely event of a loss of coolant accident.
2.3-5
The overtemperature hT reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that:
(1) the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),
and (2) pressure is within the range between the high and low pressure reactor trips.
With normal axial power distribution, the reactor trip limit, with allowance for errors, is always below the core safety limit. as shown in Fgure 2.1-1.
If axial peaks are greater than design, as indicated by difference between top and bottom power range nuclear detectors, the reactor trip limit is automatically reduced.
The overpower bT reactor trip prevents power density anywhere in the core from exceeding a
value at which fuel pellet centerline melting would occur as described in Section 7.2.3 of the FSAR and in WCAP-8058, "Fuel Densification, R. E. Ginna Nuclear Power Plant Unit 1, Cycle 3" and includes corrections for axial power distribution, change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to a loop temperature detectors.
The specified set points meet this requirement and include allowance for instrument errors.
The overpower and overtemperature protection set points include consideration of the effects of fuel densification on core safety limits.
The low flow reactor trip protects the core against, DNB in the event of a sudden loss of power to one or both reactor coolant pumps.
The set point specified is consistent with the value used 2.3-6
in the accident analysis. 'he underfrequency reactor trip protects against a decrease in flow caused by low electrical I
frequency.
The specified set p'oint assures a reactor trip signal before the low flow trip point is.reached.
The high pressurizer water level rea'ctor trip protects the pressurizer safety valves against water relief.
Approximately 700 ft.
of water corresponds to 92% of span.
The specified set point contains margin for both instrument. error and transient overshoot, of level beyond this trip setting, and therefore the trip function prevents
- the water level from reaching the safety valves. (2)
The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents.
The specified set.
point equivalent to at least. 40,000 lbs. of water assures that there will be sufficient. water inventory in the steam generators at the time of trip to allow for stari ing delays for the auxiliary feedwater system.
The specified reactor trips are blocked at low power where th y are not, required for protection and would otherwise interfere with normal plant operations.
The prescribed set point above which these trips are unblocked assures their availability in the power range where needed.
Operation with one pump will not be permitted above 130 NWT (8.5%).
An orderly power reduction to less than 130 NWT (8.5%)
will be accomplished if a pump is lost while operating between 130 NWT (8.5%)
and 50%.
Automatic protection is provided so that a power-to-flow ratio is maintained equal to or less than one, which insures that the minimum DNB ratio increases at. lower flow 2.3-7
because the maximum enthalpy rise does not increase.
For this reason the single pump loss of flow trip can be bypassed below 50% power.
The loss of voltage and degraded voltage trips ensure operability of safeguards equipment during a postulated design basis event concurrent'with a degraded bus voltage condition. (9)(10)(11)
The undervoltage set. points have been selected so that safeguards motors will start and accelerate the driven loads (pumps) within the required time and will be able to perform for'ong periods of time at. degraded conditions above the trip set points without significant. loss of design life. All control circuitry or safety related control centers and load centers, except for motor control centers M and L, are d.c.
Therefore, degraded grid voltages do not effect, these control centers and load centers.
Motor control centers M and I which supply the Standby Auxiliary Feedwater
- System, are fully protected by the undervoltage set points.
'urther, the Standby System is normally not in service and is manually operated only in total loss of feedwater and auxiliary feedwater.
The 5% tolerance curve in Figure 2.3-1 and the requirements of specifications 2.3.3.1 and 2.3.3.2 include 5% allowance for measurement error.
Thus, providing the measurement error is less than 5%, measured values may be directly compared to the curve.
If measurement error exceeds 5%, appropriate allowance shall be made.
2.3-8
References:
(1)
FSAR 14.1.1 (2)
- FSAR, Page 14-3 (3)
FSAR 14.3.1 (4)
FSAR 14.1.2 (5)
FSAR 7.2, 7.3 (6)
FSAR 3.2.1 (7)
FSAR 14.1.6 (8)
FSAR 14.1.9 (9)
Letter from L. D. White, Jr. to A. Schwencer, NRC, dated September 30, 1977 (10) Letter from L. D. White, Jr. to A. Schwencer, NRC, dated September 30, 1977 (ll) Letter from L. D. White, Jr. to D. Ziemann NRC, dated July 24, 1978 2.3-9
lo00 LOSS AND SECOND LEVEL '(DEGR.= DED)
UNDERVOLTAGE RELAY. OPERATING
~ RANGES FIG. 2.3-1
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1000 ti:
t=.
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t 800 t=
t=.t 30 20 10 8.5 8.0 I
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, LOSS OF VOLTAGE OPERATING RANGE (equi pment
(
thermal capability
(
curve)
'I I
5'g'toleiance
/curve t=>t 600 C
t I
400 t-.
I 200 c0 CL 0
60 240 52%
70 80 320 70~o 90 92 368 80%
100 103.5 110 414 90%
120 480 104%
SAFEGUARDS BUS VOLTAGE M =
4 O Q m 2.3-10
Page 3 of 3 NO. of CHANNELS 2
3 4
5 6
NO. of MIN.
MIN.
PERMISSIBLE OPERATOR ACTION CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF TO TRIP CHANNELS REDUNDANCY CONDITIONS COLUMN 3 or 5 CANNOT BE MET 4J Vl I
R0
- 17. Circulating Water Flood Protection a.
Screenhouse 2
b.Condenser 18.
Loss of Voltage/
Degraded Voltage 480 Volt Safe-guards Bus 4/bus 2/bus 2/bus NOTE 1:
When block condition exists, maintain normal operation.
Power operation may be continued for a perio~
of up to 7 days with~
channel inoperable or for a period of 24 hrs.
with two channels in-operable.
Power operation may be continued for a period of up to 7
days with 1 channel inoperable or.for a period of 24 hrs.
with two channels inoperable.
Maintain hot shut-down or place bus on diesel generator.
0 0
Cda F.P.
Full Power Not Applicable, If both rod misalignment monitors (a and b) inoperable for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or more, the nuclear overpower trip shall be reset to 93% of rated power in addition to the in-creased surveillance noted.
If a functional unit, is operating with the minimum operable
- channels, the number of channels to trip the reactor will be column 3 less column 4.
A channel is considered operable with 1 out of 2 logic or 2 out of 3 logic.
TABLE 4. 1-1 (CONTINUED)
Channel Check Calibrate
- Test, Remarks 25.
Containment Pressure 26.
Steam Generator Pressure R
M Narrow range containment pressure
(-3.0,
+3 psig excluded) 27.
Turbine First Stage Pressure S
28.
Emergency Plan Radiation Instruments 29.
Environmental Monitors N.A.
N.A.
30.
Loss of Voltage/Degraded Voltage 480 Volt, Safe-guards Bus N.A.
S Each Shift D
Daily B/W Biweekly Q
Quarterly M
Monthly P
.Prior to each startup if not done previous week R
Each Refueling Shutdown N.A.
Not applicable
Attachment B
By letter dated June '3, 1977; the NRC requested that RGEE assess the susceptibility of safety related electrical equipment with regard to (1) sustained degraded voltage conditions at the offsite power sources and (2) interaction between the offsite and onsite emergency power sources.
An analysis of undervoltage protection at Ginna Station was performed and submitted to the NRC in a letter dated July 21, 1977.
This analysis reviewed the current undervoltage protection and described the basis for a modification which would reduce the Station's susceptibility to a sustained degraded voltage.
These Specifications will provide protection for a complete loss of 480 volt bus voltage as well as for degraded voltage conditions.
Both relaying systems will assure that assumptions of all safety analysis are met.
Specifically, equipment will be loaded onto the diesel generators within the time assumed in the safety analyses.
The undervoltage relay protection, which was identified as "second-level protection" in the June 3,
1977 NRC letter and in our two subsequent.
letters, will protect equipment against a bus voltage which is greater than the loss of voltage relay setpoint but less than the voltage guaranteed by equipment manufacturers for continuous duty for Ginna safeguards equipment.
The proposed undervoltage setpoints will provide the required protection while also assuring that. spurious trips will not occur while equipment, is being sequenced onto the diesel generators.
The NRC Staff requested that the proposed system be modified to
provide "coincident logic" as described in position 1, part b of their June 3,
1977 letter.
Provision for "coincident. logic" does not effect the setpoint information contained in the December 16, 1977 Application for Change to Operating License and resubmitted herewith.
However, it has changed the reguired number of relays necessary to trip.
The coincident logic scheme is described in our July 24, 1978 letter (ref.
11 to the Technical Specification Basis).
Loss of voltage and degraded voltage conditions will be simulated during monthly tests performed to verify system performance.
Relay calibration will be performed during refueling shutdown.
Based on the analyses provided in our letters of July 21,
- 1977, September 30,
- 1977, and July 24, 1978 the proposed Technical Specification will provide protection against, 480 volt bus under-voltage.
Attachment C
The Technical Specification change proposal revises a
proposal submitted July 31, 1979.
The re=vision is necessary to incorporate a
change in design which>. was request..d by the NRC Sta ff.
The design change was described in our submittal of July 24, 1970.
Because it is a revision which the Staff requested, and is not a
new request, no fee is required.
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