ML13331B144
| ML13331B144 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 03/11/1989 |
| From: | Baskin K Southern California Edison Co |
| To: | NRC/IRM |
| Shared Package | |
| ML13331B145 | List: |
| References | |
| NUDOCS 8903170013 | |
| Download: ML13331B144 (14) | |
Text
Southern California Edison Company P.O. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 KENNETH P. BASKIN TELEPHONE VICE PRESIDENT March 11, 1989 818-302-1401 U. S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Supplement 3 to Amendment Application No. 164 "RCP Bus Undervoltage Trip" San Onofre Nuclear Generating Station Unit 1 By letter dated March 4, 1989, SCE submitted Supplement 2 to Amendment Application No. 164 (Revision 2 of Proposed Change 204) to include, among other things, the undercurrent and overcurrent RCP breaker trips in Appendix A, Technical Specifications, Table 2.1, "Maximum Safety System Settings," to prevent core damage from the seized shaft and sheared shaft events. The purpose of this supplement is to provide an additional safety setpoint in Table 2.1 to include the RCP bus undervoltage trip. Technical Specifications 3.5.1, "Reactor Trip System Instrumentation," and 4.1.1, "Reactor Trip System Instrumentation Surveillance Requirements," have also been changed to reflect the addition of this trip.
In loss of flow events caused by a loss of RCP bus, the low flow trip and the undervoltage trip provide diverse and redundant reactor protection. RCP bus undervoltage trip is currently maintained administratively.
If you have any questions regarding this matter, please contact me.
Respectfully submitted, 890:3170013 890311 PDR ADOCK 050002'0(By:
P N1IC.
By:
Kenneth P. Baskin Vice President Subscribed and sworn to before me this day of
~OFF CIAL SEAL CAOROL A.
GOAEZ NOTARY PUBUC - CALFORNIA LOS ANGELES COUNTY MY Com. Expires Feb. 26, 1993 ary Public in and for th County of Los Angeles, State of Cali nia cc: 3. B. Martin, Regional Administrator, NRC Region V F. R. Huey, NRC Senior Resident Inspector, San Onofre Units 1, 2 and 3
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of SOUTHERN CALIFORNIA EDISON COMPANY
)
and SAN DIEGO GAS & ELECTRIC
)
Docket No. 50-206 COMPANY (San Onofre Nuclear
)
Generating Station Unit No. 1 CERTIFICATE OF SERVICE I hereby certify that a copy of Amendment Application No. 204, Supplement 3, was served on the following by deposit in the United States Mail, postage prepaid, on the 13th day of March
, 1989.
Benjamin H. Vogler, Esq.
Staff Counsel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 David R. Pigott, Esq.
Samuel B. Casey, Esq.
Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 L. G. Hinkleman Bechtel Power Corporation P.O. Box 60860, Terminal Annex Los Angeles, California 90060 Michael L. Mellor, Esq.
Thelen, Marrin, Johnson & Bridges Two Embarcadero Center San Francisco, California 94111 Huey Johnson Secretary for Resources State of California 1416 Ninth Street Sacramento, California 95814 Janice E. Kerr, General Counsel California Public Utilities Commission 5066 State Building San Francisco, California 94102
-2 C. J. Craig Manager U.S. Nuclear Projects I, ESSD Westinghouse Electric Corporation Post Office Box 355 Pittsburgh, Pennsylvania 15230 A. I. Gaede 23222 Cheswald Drive Laugna Niguel, California 92677 Frederick E. John, Executive Director California Public Utilities Commission 5050 State Building San Francisco, California 94102 Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Southern California Edison Company
0 0
DESCRIPTION OF SUPPLEMENTAL CHANGE TO PROPOSED CHANGE NO. 204 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-13 SUPPLEMENT 1 TO REVISION 2 The following is a supplemental request to revise Section 2.1, "Reactor Core Limiting Combination of Power, Pressure and Temperature," and Table 3.5.1-1, "Reactor Trip System Instrumentation," and Table 4.1.1, "Reactor Trip System Instrumentation Surveillance Requirements," of the Appendix A, Technical Specifications for San Onofre Nuclear Generating Station, Unit 1 (SONGS 1).
Description of Supplemental Change The bus undervoltage trip of RCP breakers provides a diverse and redundant reactor protection against potential core damage in addition to the low flow trip.
The safety setpoint for the bus undervoltage trip is, however, omitted in Table 2.1, "Maximum Safety Systems Settings."
This supplement includes the RCP bus undervoltage trip setpoint in Table 2.1, Table 3.5.1-1, and Table 4.1.1 of the Technical Specifications.
The Attachment 2 changes have single change bars to designate those changes made in previous revisions to this proposed change. This supplement adds changes which are indicated with double change bars.
Existing Technical Specifications See Attachment 1.
Proposed Technical Specifications See Attachment 2.
Significant Hazards Consideration Analysis As required by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that a supplmental license amendment to implement revised provisions for this proposed change and operability for SONGS 1 represents no significant hazards consideration. In accordance with the three factor test of 10 CFR 50.92(c),
implementation of the proposed license amendment was analyzed using the following standards and found not to:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The 4 kV Bus-lA supplies power to RCP G-2A and G-2C. The 4 kV Bus-lB is the power source for RCP G-2B. The RCP bus undervoltage trip circuit consists of two detection channels, one on each RCP bus, which opens the respective RCP breakers on detection of bus undervoltage. The RCP circuit breaker trip actuates RCP breaker auxiliary switch contacts which trip the reactor. The undervoltage trip is currently controlled administratively for normal operations.
In loss of flow events caused by a loss of power to the reactor coolant pumps (RCPs), e.g., a loss of bus or manual actions to open an RCP circuit breaker, the low flow trip and the undervoltage trip provide a diverse and redundant reactor protection.
Inclusion of the RCP bus undervoltage trip is consistent with assumptions in UFSAR Section 15.7.1.
03630
-2 Additional changes remove from Table 2.1 and provide to Table 3.5.1-1 clarifications of the function of the Reactor Protection System permissives P-7 and P-8. One of three loops reactor trip logic is enabled by the P-8 permissive above 50% power. The P-7 permissive enables the two of three loops reactor trip logic above 10% power. This provides the necessary protection for a single loop failure for sheared shaft and seized shaft events above 50%
power and for two loop loss of flow events below 50% power.
Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
- 1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?
Response
No.
Currently the undervoltage trip is administratively controlled for normal operations. The undervoltage trip is necessary to satisfy single failure criterion for loss of flow events caused by loss of power to the RCP. The addition of the undervoltage trip safety setpoint in the Table 2.1 ensures that the RCP bus undervoltage trip is available and OPERABLE, consistent with assumptions in the accident analysis in UFSAR 15.7.1.
Therefore, the proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.
- 2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The supplemental change includes the safety system setting of the RCP bus undervoltage trip in the Table 2.1.
This trip is credited to satisfy the single failure criterion in the event of loss of forced coolant flow due to loss of RCP bus.
Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.
0
-3
- 3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response: No.
.The addition of the safety setting of the undervoltage trip in the Table 2.1 provides consistency with assumptions in the accident analysis in UFSAR 15.7.1 and satisfies the single failure criterion for loss of flow events caused by loss of power to the RCPs.
Therefore, the proposed supplemental change will not involve any reduction in a margin of safety.
Safety and Significance Hazards Consideration Determination Based on the Safety Evaluation provided in Amendment Application No. 164 and the information provided above, it is concluded that: (1) the supplemental changes to Proposed Change No. 204 do not involve a significant hazards considerations defined by 10 CFR 50.92; and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
03630 Existing Technical Specifications 2.1
- .AC*: **:
- Limi ng Combination of Power, Prestre, and Temperature APPLICABTI-lj:
Applies to reactor power, system pressure, coolant teerature. and flow during operation of the plant.
OBJECINE:
To maintain the integrity of the reactor coolant system and to prevent the release of excessive amounts of fission product activity to the coolant.
SPECFCAIO: SafetyLimits (1) The reactor coolant system pressure shall not exceed 2735 psig with fuel assemblies in the reactor.
(2) The combination of reactor power and coolant temperature 60 shall not exceed the locus of points established for the I//8.
RCS pressure in Figure 2.1.1. If the actual power and temperature is above the locus of points for the appropriate RCS pressure, the safety limit is exceeded.
Maximum Safety System Settings The maximum safety system trip settings shall be as stated in Table 2.1.
17 BASIS:
Safety Limits 13/as
- 1. Reactor Coolant System Pressure The Reactor Coolant System serves as a barrier which prevents release of radionuclides contained in the reactor coolant to the containment atmosphere.
In addition, the failure of components of the Reactor Coolant System could result in damage to the fuel and pressurization of the containment. A safety limit of 2735 psig (110% of design pressure) has been established which represents the maximum transient pressure allowable in the Reactor Coolant System under the ASME Code,Section VIII.
- 2. Plant Operating Transients In order to prevent any'significant amount of fission products from being released from the fuel to the reactor coolant, it is necessary to prevent clad overheating both during normal operation and while undergoing system transients. Clad overheating and potential failure could occur if the heat transfer mechanism at the clad surface departs from nucleate boiling.
System parameters which affect this departure from nucleate boiling (DNB) have been correlated with experimental data to provide a means of determining the probability of ONB occurrence.
The ratio of the heat flux at which DN is expected to occur SAN ONOFRE - UNIT 1 2-1 Revised:
12/21188
for 4ven set of conditions to theOtual heat flux exper enced at a point is the 0ON ratW and reflects the probability that ONS will actually occur.
It hs been determined that under the most unfavorable conditions of power distribution expected during core lifetime and if a ONE ratio of 1.44 should exist, not more than 7 out of the total of 28,260 fuel rods would be expected to experience ONE.
These conditions correspond to a reactor power of 1251 of rated power.
Thus, with the expected power distribution and peaking factors, no significant release of fission products to the reactor coolant system should occur at ONO ratios greater than 1.30.(1)
The ON ratio, although fundamental. is not an observable variable.
For this reason, limits have been placed on reactor coolant temperature, flow, pressure, and power level, these being the observable process variables related to determination of the OO ratio.
The curves presented in Figure 2.1.1 represent loci of conditions at which a minimum DN8 ratio of 1.30 or greater would occur. (1)(2)(3)
Maximum Safety System Settings
- 1. Pressurizer Hiah Level and High Pressure In the event of loss of load, the temperature and pressure of the Reactor Coolant System would increase since there would be a large and rapid reduction in the heat extracted from the Reactor Coolant System through the steam generators.
The maximum settings of the pressurizer high level trip and the pressurizer high pressure trip are established to maintain the 0MB ratio above 1.30 and to prevent the loss of the cushioning effect of the steam volume in the pressurizer (resulting in a solid hydraulic.
system) during a loss-of-load transient.(3)(4)
In the event that steam/feedflow mismatch trip cannot be credited due to single failure considerations, the 97 pressurizer high level trip is provided.
In order to meet 4/7/as acceptance criteria for the Loss of Main Feedwater and Feedline Break transients, the pressurizer high level trip must be set at 20.8 ft. (50%) or less.
- 2. Variable Low Pressure Loss of Flow and Nuclear Overpower These settings are established to accommodate the most severe transients.upon which the design is based, e.g.,
loss of coolant flow, rod withdrawal at power, control rod ejection, inadvertent boron dilution and large load increase without exceeding the safety limits. The settings have been derived in consideration of instrument SAN ONOFRE - UNIT 1 2-2 Revised: 12/21/88
erro*and response times of all neA ary equipment.
Thus, these settings should prevent the release of any significant quantities of fission products to the coolant as a result of transients.(3)(4)(5)(7) t cl ln order to prevent significant fuel damage in the event of increased peaking factors due to an asymetric power distribution in the core, the nuclear overpower trip setting on all channels is reduced by one percent for each percent that the asymmetry in power distribution exceeds 5%.
This provision should maintain the DNS ratio above a value of 1.30 throughout design transients L1 mentioned above.
The response of the plant to a reduction in coolant flow while the reactor is at substantial power is a corresponding increase in reactor coolant temperature.
If the increase in temperature is large enough, DNB could occur, following loss of flow.
The low flow signal is set high enough to actuate a trip in time to prevent excessively high temperatures and low enough to reflect that a loss of flow conditions exists.
Since coolant loop flow is either full on or full off, any loss of flow would mean a reduction of the initial flow (100%) to zero.(3)(6)
References:
(1) Amendment No. 10 to the Final Engineering Report and Safety Analysis, Section 4, Question 3 (2) Final Engineering Report and Safety Analysis, Paragraph 3.3 (3) Final Engineering Report and Safety Analysis, Paragraph 6.2 (4) Final Engineering Report and Safety Analysis.
Paragraph 10.6 (5) Final Engineering Report and Safety Analysis, Paragraph 9.2 (6) Final Engineering Report and Safety Analysis, Paragraph 10.2 (7) MIS Safety Review Report, April 1988 1.17 12/13/88 SAN ONOFRE -UNIT 1
2-3 Revised: 12121/88
MAII SFl STEM TNCI 12/13/as Three Reactor Coolant Pumes Oneratino
- 1.
Pressurizer 1 20.8 ft. above bottom of pressurizer High Level when steamlfeedflow mismatch trip is.nt credited, or 1 27.3 ft. above bottom of pressurizer when steadfeedflow mismatch trip J1 credited
- 2. Pressurizer 1 2220 psig Pressure: High
- 3. Nuclear Overpower 117
- a. High Setting"*
1 109% of indicated full power ty/a/e
- b. Low Setting
.1 25% of indicated full power
- 4. Variable Low Pressure 1 26.15 (0.894 AT4T avg.) -
14341
- 5. Coolant Flow
> 85% of indicated full loop flow h Credit can be taken for the steam/feedflow mismatch trip when this system is modified such that a single failure will not prevent the system from performing its safety function.
- The nuclear overpower trip is based upon a symmetrical power distribution.
If an asymmetric power distribution greater than 5% should occur, the nuclear overpower trip on all channels shall be reduced one percent for 12
/
each percent above 51.
12/13/a
- May be bypassed at power levels below 10% of full power.
SAN ONOFRE - UNIT 1 2-4 Revised: 12/21/88
TABLE 3.5.1-1 REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 1. Manual Reactor Trip 2
I 2
1, 2 I
2 I
2 31, 4*, 5*
7
- 2. Power Range, Neutron Flux, 4
2 3
I, 2 2#
Overpower Trip
- 3. Power Range, Neutron Flux, 4
1*1 4
1, 2 28#
Dropped Rod Rod Stop
- 4. Intermediate Range, Neutron 2
I 2
1##, 2 3
Flux Un
- 5. Source Range, Neutron Flux A. Startup 2
I**
2 2#1 4
B. Shutdown 2
[**
2 3*, 4*, 5*
7 C. Shutdown 2
0 I
3, 4, and 5 5
- 6. NIS Coincidentor Logic 2
I 2
1, 2 29 3*, 4*, 5*
7
- 7.
Pressurizer Variable 3
2 2
1##61 Low Pressure
- 8. Pressurizer Fixed High 3
2 2
1,2 65 Pressure
- 9.
Pressurizer High Level 3
2 6#
H F W
26 2I
TABLE 3.5.1-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTION UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
- 10. Reactor Coolant Flow A. Single Loop I/loop I/loop in any I/loop in each I
6#
(Above 50% of Full Power) operating loop operating loop B. Two Loops I/loop I/loop in two I/loop in each 1###
6#
(Below 50% of Full Power) operating loops operating loop I. Steam/Feedwater Flow Mismatch 3
2 2
1,2 6#
- 12. Turbine Trip-Low Fluid Oil Pressure 3
2 2
l9l#
6#
(n OD W
0H
TABLE 4.1.1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP ACTUATING DEVICE CHANNEL CHANNEL CHANNEL OPERATIONAL ACTUATION FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST
- 1. Manual Reactor Trip N.A.
N.A.
N.A.
R N.A.
- 2. Power Range, Neutron Flux S
D (2,3)
M N.A.
N.A.
R (3,4)
- 3. Power Range, Neutron Flux, N.A.
N.A.
M N.A.
N.A.
Dropped Rod Rod Stop.
- 4. Intermediate Range, S
R (3,4)
S/U (1),
N.A.
N.A.
Neutron Flux M
- 5. Source Range, Neutron Flux S
R (3)
S/U (1),
N.A.
N.A.
M
- 6. NIS Coincidentor Logic N.A.
N.A.
N.A.
N.A.
M (5)
- 7. Pressurizer Variable Low S
R M
N.A.
N.A.
Pressure
- 8. Pressurizer Pressure S
R M
N.A.
N.A.
E
- 9.
Pressurizer Level S
R M
N.A.
N.A.
- 10. Reactor Coolant Flow S
R Q
N.A.
N.A.
- 11.
Steam/Feedwater Flow S
R M
N.A.
N.A.
OD Mismatch
- 12. Turbine Trip-Low Fluid N.A.
N.A.
N.A.
S/U (1,6)
N.A.
Oil Pressure L.J-L N