ML17256B180
| ML17256B180 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/09/1982 |
| From: | ROCHESTER GAS & ELECTRIC CORP. |
| To: | |
| Shared Package | |
| ML17256B178 | List: |
| References | |
| NUDOCS 8208170181 | |
| Download: ML17256B180 (19) | |
Text
Attachment A
1.
Replace page l-l with the enclosed replacement pages 1-1 and l-la.
2.
Replace page 2.3-1 through 2.3-9a with enclosed replacement, pages.
3.
Replace page 3.5-5 with the enclosed replacement page.
4.
Replace page 3.5-7 with the enclosed replacement.
page.
8208170 i8i 820809 PDR ADOCK 05000244 P
TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined for uniform interpretation of the specifications.
Thermal Power The rate that the thermal energy generated by the fuel is accumulated by the coolant as it passes through the reactor vessel.
Reactor 0 eratin Modes Mode Refueling Cold Shutdown Hot Shutdown Reactivity bk k'10 1
Coolant Temperature oF T
< 140 T
< 200 T
540 for operation at, 2250 psia or at 2000 psia primary system pressu're Operating 0
T
'573.5'or operation at 2250 psia primary system pressure av for operation at 2000 psia primary system pressure Proposed
le3 Refuelin Any operation within the containment involving movement of fuel and/or control rods when the vessel head is un-bolted.
~Oerab le Capable of performing all intended functions in the in-tended manner.
1-la Proposed
2.3 2.3.1 Limitin Safet S stem Settin s, Protective Instrumentation
~11'pplies to trip settings for instruments monitoring reactor power; reactor coolant pressure, temperature,
'and flow; and pressurizer level.
~b'o provide for automatic protective action in the event that the principal process variables approach a safety limit.
~f'rotective instrumentation for reactor trip settings shall be as follows:
2.3.1.1 Startu Protection 2.3.1.2 High flux, power range (low set point) -
25% of rated power.
Core Protection a.
High flux, power-range (high set point) -
109% of i
rated power.
b.
c ~
High pressurizer pressure
< 2385 psig.
Low pressurizer pressure
> 1865 psig for operation at 2250 psia primary system pressure 1715 psig for operation at, 2000 psia primary system pressure 2
~ 3 1
Amendment No. gg
~
Proposed
d.
Overtemperature bT 1 + xlS
< ~ T
[Kl + K (P-P')
K3(T-T') (1 +
S)
-f(~I)]
where 6, T
= indicated b, T at. rated power,
'F 0
T T1 p
p 1 average temperature,
'F 573.5'F for operation at 2250 psia primary s'stem pressure 558.5'F for operation at 2000 psia'primary system pressure pressurizer
- pressure, psxg 2235 psig for operation at 2250 psia primary system pressure 1985 psig for operation at 2000 psia primary system pressure K2 0.0007356 for operation at 2250 psia primary system pressure 0.000656 for operation at 2000 psia primary system pressure K3
= 0.01577 for operation at 2250 psia primary system pressure
= 0.0136 for operation at 2000 psia primary system pressure
= 25 sec 5 sec and f (hI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests where qt and qb are the percent power in the top and bottom halves of the core respectively, and gt + gb is the total core power in percent of rated power such that:
2
~ 3 2
Amendment No.~
Proposed
(i) for qt qb within -18,
+10 percent, f (b,I) = 0 (ii) for each percent that the magnitude of qt-qb exceeds
+10 percent, the bT trip set point shall be automatically reduced by an equivalent of 2.7 percent of rated power.
(iii) for each percent that the magnitude of qt-qb exceeds
-18 percent, the bT set, point, shall be automatically reduced by an equivalent of 2.0 percent of rated power.
Overpower hT 0
4 5
6 S+1 6 T
[K - K (T-T
) - K
- f (b,I)]
'3 where T
= indicated bT at rated power,
'F 0
T Tl K4 K5 K6 f(b,I )
average temperature,
'F indicated T avg at nominal conditions at rated power, 'F 1.083 for operation at 2250 psia primary system pressure 1.09 for operation at, 2000 psia primary system pressure 0
0 for T
< T 0.001 for T
> T 0.0262 for increasing T
0.0 for decreasing T
10 sec.
as defined in 2.3.1.2.d.
2 ~ 3 3
Amendment, No. g Proposed
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 2.3.2.1 2.3.2.2 a.
High pressurizer water level -
88% of span b.
Low-low steam generator water level 6% of narrow range instrument span Protective instrumentation settings for reactor trip interlocks shall be as follows:
I 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.)
Remove bypass of single loss of flow trip at high I
power:
2.3.3 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 bel'ow:-
2.3.3.1 Loss of voltage relay operating time
< 8.5 seconds for 480 volt safeguards bus voltages
< '68 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 Amendment No.
Proposed
2.3.3.2 Acceptable degraded voltage relay operating times and setpoints, for 480 volt safeguards bus voltages
< 414 volts and
> 368 volts are 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 Figure 2.3-1.
- Basis:
The high flux reactor trip (low set point) provides redundant protection in the power range for a power excursion beginning from low power.
This trip value was used in the safety analysis.
~
(1)(13)
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 I
melting would occur.
The reactor is prevented from reaching the overpower limit condition by action of the nuclear overpower and overpower hT 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.
- Basis applies to operation at nominal (2250 psia) and reduced (2000 psia) pressure 2.3-5 Proposed
Thh 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 Figure 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 center-f line melting would occur 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 detector.
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.
2.3-6 Proposed
~
~
Two sets of overpower and overtemperature setpoints are specified.
One set provides protection when operating at a primary system pressure of 2250 psia.
The other set provides protection when operating at reduced pressure of 2000 psia.
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 con-sistent with the value used in the accident analysis.
The underfreguency reactor trip protects against a
decrease in flow caused by low electrical frequency.
The specified set point assures a reactor trip signal before the low flow trip point is reached.
The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief.
Approximately 700 ft.. of water corresponds to 92% of I
span.
A trip at this set,point contains margin for both normal instrument error and transient overshoot of level beyond this trip setting.
An additional 4%
instrument error has been assumed to account for the effects of elevated temperatures on level measurement
'n accordance with IE Bulletin 79-21.
Therefore a
trip setpoint of 88% prevents the water level from reaching the safety valves.
2
% 3 7
Proposed
The low-low-steam generator water level reactor trip protects against loss of feedwater flow accidents.
A set point of 5% is equivalent to at least 40,000 lbs.
of water and assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.
An additional 11% has been added to the set, point to account. for error which may be introduced into the steam generator level system at a containment, temperature of 286'F as determined by an evaluation performed for temperature effects on level measurements required by IE Bulletin 79-12.
The specified reactor trips are blocked at low power where they 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 I
needed.
Operation with one pump will not, be permitted above 130 MWT (8.5%).
An orderly power reduction to less than 130 MWT (8.5%) will be accomplished if a pump is lost while operating between 130 MWT (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 2.3-8 Proposed
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.
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 long periods of time at degraded conditions above the trip set points without significant loss of design life. All control circuitry or safety releated control centers and load centers, except for motor control centers M and L, are d.c.
Therefore, degraded grid voltages do not affect these control centers and load centers.
Motor control centers M and L, which supply the Standby Auxiliary Feedwater
- System, are fully protected by.the undervoltage set points.
- Further,
.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.
2.3-9 Amendment No.
Proposed
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 (11) Letter from L. D. White, Jr. to D. Ziemann, NRC, dated July 24, 1978 I(
(12) Letter from L. D. White, Jr. to B. Grier, USNRC, dated September 14, 1979 (13) ZN-NF-82-57 "Plant Transient Analysis for Operations of the R. E. Ginna Unit 1 Nuclear Power Plant at Reduced Pressure and Temperature",
Exxon Nuclear Company, Inc.,
June 1982 2.3-9a Proposed
TABLE.3.5-2 EMERGENCY COOLING NO.
FUNCTIONAL UNIT 1.
SAFETY INJECTION a.
Manual NO. of CHANNELS 2
3 NO. of MIN.
CHANNELS OPERABLE TO TRIP~
CHANNELS MIN.
PERMISSIBLE DEGREE OF BYPASS REDUNDANCY CONDITIONS Primary pressure less than 2000 psig for operation at 2250 psia 6
OPERATOR ACTION IF CONDITIONS OF COLUMN 3 OR 5 CANNOT BE MET
- b. High Containment 3
Pressure Primary pressure less than 1800 psig for operation at 2000 psia c.
Iow Steam Pressure/Loop
- d. Pressurizer Iow Pressure 2.
Manual an
- b. Hi-Hi Containment 2 sets Pressure (Contain-of 3 ment Spray) 2of3 in ea.
set 2 per 1/set set Must actuate 2 switches simultaneously.
3.5"5 Proposed
E
~
~
Attachment B
The R. E. Ginna nuclear plant was originally designed and operated at 2250 psia.
Primary system operating temperature and pressure was reduced to approximately 566'F and 2000 psia during Cycle 3. in order to extend the time to clad collapse associated with the fuel densification problem.
Justification for operation at 2000 psia was contained in the Westinghouse report WCAP-8153 "Fuel Densification R. E. Ginna Nuclear Plant Low Pressure Analysis,"
July 1973.
Operation at 2000 psia was approved with the issuing of change No. 10, October, 1973 to the Ginna Technical Specifications.
Return to a primary system operating temperature and pressure of 573.5'F and 2250 psia was approved with the issuing of Amendment No. 10, March 30, 1976 to the Ginna Technical Specifications.
Return to 2250 psia resulted from improved clad collapse models and improved fuel rod design.
I Because steam generator tube intergranular attack may be reduced at reduced primary system temperatures, it may be advantageous to operate at reduced primary system temperature and pressure.
Therefore, Rochester Gas and Electric requests a
Technical Specification change that would allow operation at normal temperature and pressure or reduced temperature and pressure.
The Ginna core currently consists of 117 fuel assemblies built Exxon Nuclear Company and 4 mixed oxide fuel assemblies with fuel rods fabricated by Westinghouse.
0 1
~
I
The current. fuel supplier for Ginna, Exxon Nuclear, has reviewed the safety implications of reduced temperature and pressure operation.
The results of the review are contained in three enclosed reports:
1.
ZN-NF-82-45 "Plant Transient Analysis for Operation of the R. E. Ginna'nit 1 Nuclear Power Plant at Reduced Pressure and Temperature",
June 1982.
2.
XN-NF-82-57 "R. E. Ginna Nuclear Plant Cycle 12 Safety Analysis Report for Low Temperature and Pressure",
July 1982.
3.
ZN-NF-82-26 "R. E. Ginna Revised LOCA ECCS Analysis for Nominal and Reduced Temperature and Pressure Operation",
April 1982.
The setpoints presented in Attachment A are the same as those previously approved for reduced temperature and pressure operation (Change 10).
These setpoints, including appropriate conservatisms for instrument error, were used in the Exxon Nuclear analyses.
The existing analyses at nominal conditions and the Exxon Nuclear analyses at reduced conditions provide the support for operation at nominal or reduced temperature and pressure.
TABLE 3.5 Continued TABLE.NOTATION ACTION STATEMENTS 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.
This start signal is required only during power operation above 5%.
ACTION 1 With the number of operable Channels one less than the Total Number of Channels, restore the inoperable channel to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION.2 With the number of operable channels one less than the Total Number of Channels, operation may proceed until performance of the next, required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ACTION 3 With the number of operable channels one less than the minimum operable channels and the permissible bypass conditions not met, be in at least hot shutdown within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and in cold shutdown within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.5-7 Amendment No. pl+
Proposed