ML17258B001
| ML17258B001 | |
| Person / Time | |
|---|---|
| Site: | Ginna  (DPR-18-A-038, DPR-18-A-38) |
| Issue date: | 03/26/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17258B002 | List: |
| References | |
| NUDOCS 8105050123 | |
| Download: ML17258B001 (18) | |
Text
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UNITEDSTATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
E.
GINNA NUCLEAR POMER PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE 4
WE'"
4 Amendment No. 38 License No.
DPR-18 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Rochester Gas and Electric Corporation (the licensee) dated September 9,
1980.(transmitted by letter dated September 15, 1980),
as supported by filings dated July 21, 1977, November 21,
- 1977, December 16, 1977, (transmitted by letter dated December 22, 1977), July 31, 1979 (transmitted August 3, 1979 - two separated submittals) and December 19, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the appli-cation, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be.inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
C
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Provisional Operating License No.
DPR-18 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
38
, are hereby incorporated in the license.
The licensee 'shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications nnis M. Crutchfie d, Ch' Operating Reactors Branch
~5 Division of Licensing Date of Issuance:
March 26, 1981
I 1
ATTACHMENT TO LICENSE AMENDMENT NO. 38 PROVISIONAL OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages contain the captioned amendment number and a vertical line which indicates the area of change.
REMOVE 2.3-4 2.3-5 2.3-6 2.3-7 "2.3-8 2.3-9 3.5-4a 4.1-7 INSERT 2.3-4 2.3-5 2.3-6 2.3-7 (reformat only) 2.3-8 2.3-9 2.3-10 3.5-4a 4.1-7
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V f.
Low reactor coolant flow >
90% of normal indicated flow.
g.
Low reactor coolant pump frequency
> 57.5 Hz.
2.3.1.3 Other reactor tri s 2.3.2 2.3.2.1 a.
High pressurizer water level -
92% of span b.
Low-low steam'enerator water level 5% of narrow range instrument span Protective instrumentation settings for reactor trip interlocks shall be as follows:
Remove bypass of "at power" reactor trips at high 'power (lor. pressurizer pressure and low reactor coolant flow) for both loops:
Power range nuclear flux -
8..5% of rated power V
(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.
Neasured 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.
38
0
~ I 7
I
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.
Ileasured values shall fall at least 5% below the theoretical limit.
This 5% margin is sho~n 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)
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 AT trips.
The high and low pres'ure reactor trips limit the pressure range in wh-'ch 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 Amendment No. 38
q I 1
1
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), (4) and (2) pressure is within the range between the high and low pressure reactor trips.
Nith normal axial power distribution, the reactor trip limit, with allowance for errors, is always (2) 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 d.'stribution, change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.
The specified set points meet this requirement and include allowance for instrument errors.
The overpower and (2) 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 Amendment No. ~, 38
~l
in the accident analysis.
The underfrequency 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 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 f'ow 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 starting delays for the auxiliary feedwater system.
The specified reactor trips are blocked at
- ow 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 needed.
Operation with one pump will not be permitted above 130 11NT (8.5%).'n orderly power reduction to less than 130 KPZ (8.5%)
will be accomplished if a pump is lost while operating between 130 NNT (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 lo'wer flow 2
~ 3 7
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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 significan loss of design life. All control circuitry or safety related control centers and load centers, except for motor control centers N and L, are d.c.
Therefore, degraded grid voltages do not affect these control centers and load centers.
Notor control centers N 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. 3-8 Amendment No. 38
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. Mhite, Jr. to A. Schwencer, NRC, dated September 30, 1977 (10) Letter from L. D. Mhite, Jr. to A. Schwencer, NRC, dated September 30, 1977
,(11) Letter from L. D. 4'hite, Jr. to D. Ziemann NRC, dated July 24, 1978 2.3-9 Amendment No. 38
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' t LOSS -4 D "ECOND LEVEL
( DEG.":-.DED)
UXDERVOL' GE RELAY 0?E.=PATING P'-..'GES 1400 FIG. 2. 3-1
'.1200 1000 I,I 1
/
/
I 800 i
30 20 1 0 8.5 8;0
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~ l 60 240 52~o 70 LOSS OF VOLTAGE OPERATING RANGE
I I
80 320 70%
(curve' 100 103.5 llo 414 90%
90 92 368 80~a
...I
(
1Second leve operat1ng region l
(eauipment thermal capability
~
curve)
I 5%
600 P P 400 P
i r
1
.200 tolerance i
0 120 480 1 04'R O 5 CO N fg
~ Ci I
I Vl=
o~
N Ul
'w rw
~E 0 SAF" GUARDS ~US VOLTAGE 2.3-10 Amendment No.
38
I
Page 3 of 3 1
2 3
4 5.
6 NO. of MIN.
MIN.
PERMISSIBLE OPERATOR ACTION NO. of CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS COLUMN 3 or 5
CANNOT BE MET
- 17. Circulating Water Flood Protection a.
Screenhouse 2
b.Condenser 18.
Loss of Voltage/
Degraded Voltage 400 Volt Safe-guards Bus 4/bus 2/bus 2/bus Power operation may be continued for a period of up to 7 days with 1 channel inoperable or for a period of 24 h~
with two channels in~
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.
NOTE 1:
When block condition exists, maintain normal operation.
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.
TABf,E 4.1-1 (CONTINUED}
4 C}rarrrre 1 f)r:scr'.i nti o>>
A 25.
Corrtair>>nent Pressure Check Calibrate Test R
M Remarks Narrow range containment. pressure
(-3.0,
+3 psig excluded}
26.
Steam Generator Pressure S
27.
Turbine First Stage Pressure S
28.
Emergency Plan Radiation Instruments R
e 29.
Errviror>>nental Monitors N.A.
N.A.
30.
f.os of Voltage/Degraded Vol tage 400 Volt Safe-guards Bus N.A.
S Eaclr Strift 0
f)@ily B/8 Biweekl y Q
Qrrar Lerly M
Monthly P
Prior to each startup if not done previous week R
Each Refueling Slrutdown N.A.
Not applicable