ML13317A523
| ML13317A523 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 02/28/1983 |
| From: | Krieger R SOUTHERN CALIFORNIA EDISON CO. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-06-10.A, TASK-07-02, TASK-6-10.A, TASK-7-2, TASK-RR NUDOCS 8303070064 | |
| Download: ML13317A523 (18) | |
Text
Southern California Edison Company P.O. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 February 28, 1983 Director, Office of Nuclear Reactor Regulation Attention:
Mr. D. M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Gentlemen:
Subject:
Docket No. 50-206 Systematic Evaluation Program Topic VI-1O.A (Testing of Reactor Trip System and Engineered Safety Features) and Topic VII-2 (ESF System Control Logic and Design)
San Onofre Nuclear Generating Station Unit 1 Your letter dated November 17, 1981 provided your draft safety evaluation report along with your contractor's draft technical evaluation for SEP Topic VII-2. In your letter you indicated that additional information is necessary to complete the final safety evaluation on this topic. Enclosure 1 to this letter provides the additional information that you have requested for this topic. There were certain areas of your contractor's draft technical evaluation which require clarification. This information is also included in. to this letter is our response to your draft safety evaluation for SEP Topic VI-10.A (Testing of Reactor Trip System and Engineered Safety Features) which was forwarded by letter dated August 28, 1981.
In your evaluation, you noted several instances where the testing and surveillance requirements of the San Onofre Unit No. 1 Technical Specifications diverge from those of the Standard Technical Specifications.
The responses to these discrepancies are included in Enclosure 2. In most instances where these discrepancies occurred, there exists surveillance and testing procedures at San Onofre Unit 1 which satisfy the requirements of the Standard Technical Specifications, but are not included in the San Onofre Unit 1 Technical Specifications. These instances are denoted in Enclosure 2 and the appropriate procedures which satisfy the Standard Technical Specification requirements are referenced. The need to modify San Onofre Unit 1 surveillance and testing requirements for those cases where the San 8303070064 830228 PDR ADOCK 05000206 P
PDR 01O
Mr. D.
February 28, 1983 Onofre Unit 1 requirements diverge from the Standard Technical Specification requirements will be determined during the Integrated Safety Assessment.
Modifications to the San Onofre Unit 1 Technical Specifications to incorporate additional surveillance and testing requirements will also be determined during the Integrated Safety Assessment.
If you have any questions regarding the enclosed, please let me know.
Very truly yours, R. W. Krieger Supervising Engineer San Onofre Unit 1 Licensing Enclosures
ENCLOSURE 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING SEP TOPIC VII-2 ENGINEERED SAFETY FEATURES (ESF)
SYSTEM LOGIC AND DESIGN
VII-2 ENGINEERED SAFETY FEATURES (ESF)
SYSTEM LOGIC AND DESIGN The discussion below provides additional information related to the isolation devices used to isolate the input signal channels on the safeguard load sequencing system, containment spray and the containment isolation system.
Additionally, one point of clarification in regards to your contractor's evaluation of the safety injection system is necessary. Section 3.2, paragraph 5 of the Technical Evaluation (TE) by EG&G currently reads:
"The low pressurizer pressure signals from the three bistable relays feed all four subchannels of the two sequencers."
During the extensive TMI modifications, it was determined that by providing independent low pressurizer pressure signals to each sequencer (i.e., three bistable relays fed into each sequencer) that the safety and reliability of the safety injection system would be enhanced. The bistable relays which feed Sequencer No. 1 are PC 430GX, PC 431EX and PC 432CX. The bistable relays which feed Sequencer No. 2 are PYC 3000A, PYC 3000B and PYC 3000C.
This configuration was implemented during the 1980 refueling outage.
- 1. Section 4.1 of the TE states that there is "Insufficient information to evaluate the isolation between the input signal channels in the SLSS two-out-of-three logic."
-2
Response
As identified in Enclosure 1,Section II.A, paragraph 4 of SCE letter dated August 9, 1979, each SLSS train has redundant logic subchannels for each parameter input to the system. Each subchannel is completely independent and has its own Input Buffer Module (6N213), Logic Sequencer Module (6N297), Relay Driver Module (6N212), Relays and Power Supply Assemblies (8N10 and 8N11).
The Input Buffer Modules (6N213) provide the following:
- a. Accept the input signals from the remote relay contacts.
- b. Provide output signals to the logic circuitry indicative of the status of plant equipment.
- c.
Isolate the SLSS from the voltage transients and filter and filter out noise.
Furthermore, pushbuttons provide the capability to test the buffer modules. A solid state (LED) test lamp extinguishes when contacts are not in their normal state.
- 2. Section 4.2 of the TE states that there is "Insufficient information to evaluate the isolation between the input signal channels in the Foxboro CIS logic."
-3
Response
The containment high pressure input signal to the containment isolation system is generated on two trains from three pressure transmitters (PT 1120A, 1120B and 1120C for Train 1 and PT 1121A,
.1121B and 1121C for Train 2).
The output of each pressure transmitter is sensed by an alarm bistable (PA 1120A, 1120B and 1120C for Train 1 and PA 1121A, 11218 and 1121C for Train 2).
These alarm bistables utilize relay outputs which provide isolation from the input signals and from the separate train. It is our understanding that this configuration is judged acceptable under General Design Criterion 22.
- 3. Your letter which transmitted the draft evaluation for Topic VII-2 (dated November 17, 1981) identified the isolation scheme for the containment spray system to be inadequate.
Response
The draft technical evaluation included in your letter is inconsistent with that position. EG&G has determined in their evaluation that the relay contacts which provide isolation between trains of the CSAS and the non-safety systems is an acceptable isolation scheme, and furthermore, the isolation of power systems by separate buses and by thermal magnetic breakers on the same bus is also acceptable.
MJT:6895
ENCLOSURE 2 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING SEP TOPIC VI-1O.A TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES
VI-10.A TESTING OF REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES The discussion below provides additional information related to the surveillance and testing requirements for San Onofre Unit 1 I. In Section 3.2 of your contractor's draft technical evaluation of this topic it was stated:
"In comparing the instrument channels that can cause a reactor trip at San Onofre Unit 1 with the surveillance requirements that are in the unit technical specifications (Table 4.1.1),
it appears that the following channels are not checked, tested or calibrated:"
A. Manual Reactor Trip
Response
The operability of the reactor trip breakers is verified every refueling outage by Procedure SO1-II-2.4.4, Reactor Trip Test.
Included in this procedure are the requirements to test various manual reactor trip scenarios to verify that the reactor trip breakers receive a trip signal.
-2 B. Startup Rate Excessive
Response
Included in the Reactor Trip Test Procedure (SO1-II-2.4.4) is a procedure to test the High Startup Rate. This procedure entails functionally testing the intermediate range high startup rate alarms and is also conducted every refueling outage.
C. Safety Injection
Response
Reactor Trip Test Procedure (SO1-II-2.4.4) also includes a procedure to functionally test the sequencer subchannel trips for safety injection every refueling outage. This test verifies that the reactor trip breakers receive a trip signal upon initiation of safety injection.
NOTE:
Initiation of safety injection is simulated by decreasing the pressurizer pressure channels and increasing the containment pressure channels to the safety injection setpoint.
-3
.D. Turbine Trip
Response
The On-Line Turbine Trip Test Procedure (SO1-12.3-13) establishes requirements to test (on a monthly basis during Modes 1 and 2) the following turbine trips without actually tripping the turbine:
(1) Overspeed governor oil trip (2) Thrust bearing failure trip (3) Low condenser vacuum trip (4) Low bearing oil pressure trip Additionally, the Reactor Trip Test Procedure (SO1-II-2.4.4) establishes provisions to verify the turbine trip reactor trip scheme every refueling outage. This is accomplished by functionally testing the three turbine trip channels.
The San Onofre Unit 1 Technical Specifications do not establish testing requirements for the above mentioned reactor trip channels; however, as has been noted in each case, there exists procedures which require the appropriate functional testing of these reactor trip channels at frequencies consistent with the Standard Technical Specifications..
-4 II. Also, in Section 3.2 of your contractor's draft technical evaluation the following comments were made:
A.
"Additionally, there are channels called out in the Standard Technical Specifications that San Onofre Unit 1 does not use for a reactor trip signal."
Response:.
At -present,-these additional channels denoted in the Standard Technical Specifications are not required to be part of the Reactor Trip System at San Onofre Unit 1. The need to include additional channels will have to be determined during the Integrated Safety Assessment.
B.
"The San Onofre Unit 1 technical specifications require the channel functional test for loss of flow quarterly, while the Standard Technical Specifications require this.monthly.'.'
Response
Proposed Change No. 5 to the San Onofre Unit 1 technical specifications was submitted to the AEC on December 9, 1970 and was approved and issued October 20, 1971. This change decreased the minimum frequency required for testing the Reactor Coolant Flow scram channels from once every 2 weeks to once every 3 months.
-5 This change is supported by data showing that the performance of the operating instrumentation over a 2-year period had been better than had been assumed in the original basis for the. specification. The need to modify this testing frequency will be determined during the Integrated Safety Assessment.
"...the power range neutron flux channels do not have a channel functional test..."
Response
The Nuclear Instrumentation Power Range Test Procedure (SO1-12.2-3) verifies the operability of the power range nuclear instrumentation system while in Modes 1 and 2. These functional tests are conducted weekly and after maintenance or instrument department calibration of any channel.
This testing frequency is more conservative than the monthly testing requirement established in the Standard Technical Specifications.
NOTE:
The maximum safety system setting for the nuclear overpower trips is.109% of full power (see Unit Technical Specification Table 2-1, Item 3).
-6 D.
"...the intermediate and source range neutron flux channels are not required to be calibrated by technical specifications."
Response
The unit technical specifications do not require these channels to be calibrated; however, the Nuclear Instrumentation System Calibration Procedure.(SO1-II-1.6) establishes provisions for these channels to be calibrated every six months. The Standard Technical Specifications require calibration at least every refueling outage.
E.
"Response time testing to verify that the channel response time does not exceed the required response time is not in the San Onofre Unit 1 technical specifications."
Response
The response time of each individual component of the reactor trip system is not measured and recorded to insure that their components will perform their necessary function within a specified time limit..However, the operability of the components of the reactor trip system is verified every refueling outage by the reactor trip test procedure (SO1-II-2.4.4).
-7 Furthermore, the Control Rod Drop Time Test and L.V.D.T. and Rod Bottom Bistable Calibration Procedure (SO1-II-1.4-7C) establishes provisions to test the rod drop time during every refueling outage.
Also, control rod operability is verified by operation of the control rod drive mechanisms and associated control and indication circuits every 31 days, as specified in Procedure S01-12.3-24.
The verification of control rod operability, the testing of reactor trip system components and the testing of control rod drop time provides adequate assurance that the reactor trip system will perform its necessary safety function. Accordingly, additional logic system response time testing is not necessary. However, the need for additional testing will be determined during the Integrated Safety Assessment.
III. In Section 4.2 of your contractor's draft technical evaluation the following comments were made which elicit response:
A.
"The following surveillance is not done at least as frequently as required for present day licensing:"
- 1. Refueling Water Storage Tank Volume and Concentration.
-8
Response
The volume of the refueling water storage tank is indirectly monitored by the Boric Acid Batching Procedure (SO1-4-14).
This procedure establishes provisions that the level in the tank be maintained at >64% of capacity. This procedure is conducted on a weekly basis and, therefore, the volume of the tank is monitored on a weekly basis.
The boron concentration in the refueling water storage tank is monitored by Station Order SO1-E-2 (Water Chemistry Control).
Section 6.4.6 of this order establishes the testing frequency and concentration limit for boron in the tank. The tests are conducted on a weekly basis.
The Standard Technical Specifications require that the volume and boron concentration for the refueling water storage tank be monitored on a weekly basis. The San Onofre Unit 1 surveillance requirements are consistent with this requirement.
- 2. Refueling Water Storage Tank Temperature.
Response
As noted in your contractor's technical evaluation, the Standard Technical Specifications establish requirements to monitor the refueling water storage tank temperature with the
-9 presumption that the expected ambient temperature will be less than 35 0F.
San Onofre Unit 1 technical specifications do not establish testing requirements as the sustained ambient temperature does not go below 320F.
B.
"The San Onofre Unit 1 technical specifications do not agree with the present Standard Technical Specifications further in that:"
- 1. The Component Cooling Water System is not Under Surveillance
Response
- a. Valve Position - The valve positions on all safety related systems are checked on a monthly basis under Surveillance Instruction S01-12.3-6 (Safety Related Systems Valve Alignment).
- b. Valve Operability - Correct automatic valve operation on a test signal is verified at.least every 18 months during cold shutdown or refueling by Operating Instruction SO1-12.8-17 (Sphere Isolation Valve.Test).
- c. Pump Operability - The pump operability is verified on a monthly basis by the Component Cooling Water Inservice Pump Test (SO1-V-2.14.2).
-10
- 2. The Saltwater Cooling System is not Under Surveillance
Response
- a.
Valve Position -
See response to Item a above.
- b. Valve Operability -
See response to Item b above.
- c.
Pump Operability -
The pump operability is verified on a monthly basis by the Salt Water Cooling Inservice Pump Test (SO1-V-2.14.8).
- 3. The Ultimate Heat Sink is not Under Surveillance
Response
The Salt Water Cooling System is the ultimate heat sink for the San Onofre Unit 1. The response to B.2 above describes the surveillance of this system. In addition, the saltwater
cooling system utilizes the Pacific Ocean as its resource and it is not considered practical to establish surveillance requirements for the Pacific Ocean.
The surveillance and testing intervals specified in this section are consistent with those specified in the Standard Technical Specifications.
MJT:6895