ML13319A212
| ML13319A212 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/21/1979 |
| From: | Baskin K Southern California Edison Co |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| TAC-44652 NUDOCS 7911260341 | |
| Download: ML13319A212 (14) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE ROSEMEAD, CALIFORNIA 91770 K. P. BASKIN TELEPHONE
- MANAGER, GENERATION ENGINEERING November 21, 1979 213-572-1401 Director, Office of Nuclear Reactor Regulation Attention: Mr. D. G. Eisenhut, Acting Director Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:
Subject:
Docket No. 50-206 Additional Information in Support of Responses to NRC Requirements Related to the Three Mile Island Accident San Onofre Nuclear Generating Station Unit 1 By letter dated October 17, 1979, we submitted the report entitled "Responses to NRC Requirements Established Tb Date Following the Three Mile Island Accident, San Onofre Nuclear Generating Station, Unit 1, October, 1979."
The report contains our commitments to each of the actions in NUREG-0578,
"'OMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," as modified or supplemented by the NRC letter dated September 13, 1979.
By letter dated October 30, 1979, you provided additional clarification of the NRC requirements contained in NUREG-0578 and your September 13, 1979 letter, and delineated those NRC requirements which require prior NRC review and approval and those for which post implementation NRC review is acceptable. In addition, you outlined action items necessary to allow you to complete your review concerning the acceptability of our responses for the NRC short-term requirements established following the Three Mile Island Accident.
The action items are as follows:
- 1.
Submit a revised schedule for implementation of all items which do not meet the established NRC implementation schedules, and for those items which can not be met by January 1, 1980, submit a description of the degree of compliance expected on that date and a detailed justification for the delay within fifteen days from receipt of your October 30, 1979
- letter,
- 2. Submit a detailed description of our methods which are not in complete agreement with the NRC requirements along with justification for any differences within fifteen days of receipt of your October 30, 1979
- letter, O
1
.7 911260~7
-2
- 3.
Submit design details for those items requiring prior NRC approval in a timely manner so that NRC approval and our implementation can be completed by the required date,
- 4.
Document our method of implementation for those items which do not require prior NRC approval by the required date.
Specific responses addressing each of the above action items are provided below (referenced sections correspond to the enumeration contained in our aforementioned October 17, 1979 report):
- 1.
As requested in your October 30, 1979 letter, we have reviewed the implementation schedules contained in our aforementioned October 17, 1979 report with the purpose of improving the schedules to meet those established by the NRC.
Based on this review and discussions with members of the NRC staff, we have improved our implementation schedules as discussed below:
- a. For those items required by the NRC to be implemented by January 1, 1980, a special station outage has been scheduled to commence on or about January 1, 1980 to complete the modifications associated with the backup nitrogen pneumatic supply to the PORV block valves (Section 2.1.1),
the PORV, block valve and safety valve position indication (Section 2.1.3.a),
the diverse containment isolation signal and prevention of automatic reset of containment isolation valves (Section 2.1.4), the controls grade automatic initiation of the Auxiliary Feedwater System (Section 2.1.7.a), and the controls grade auxiliary feedwater flow indication (Section 2.1.7.b).
The precise shutdown date is predicated on the timely completion of the engineering, procurement and preoutage construction activities associated with each of the above modifications. As discussed with members of the NRC staff, it is anticipating that the current status of these activities for all of the above modifications will support a station outage on or about January 1, 1980 to complete the modifications, except those associated with the controls grade automatic initiation of the Auxiliary Feedwater System.
Based on the current status and projection of activities associated with modifications to the Auxiliary Feedwater System, it may not be possible to complete those modifications during the special January 1980 outage in which case the modifications would be completed during the refueling outage currently scheduled for March-April 1980.
The delay is associated with the required sequence of engineering and design activities necessary to implement the design details for the modifications to the Auxiliary Feedwater System which are provided in Enclosure 1 to this letter.
During the interim period until the modifications are completed, the operator stationed at the manual Auxiliary Feedwater System isolation
- valves, as required by IE Bulletin No.79-06A, will promptly initiate auxiliary feedwater flow when necessary.
As described in to this letter, controls grade auxiliary feedwater flow
-3 indication will be provided at this location and in the control room.
It is anticipated that the current status of the activities for this modification will support a station outage on January 1980 to complete the modifications.
Following implementation of the design details described in to this letter, we will terminate continuous stationing of an individual to promptly initiate auxiliary feedwater.
We will periodically advise you concerning our progress relative to completing the above modifications.
As additional information becomes available, we will advise you of the exact date when the station will be shut down.
In addition to the above modifications, the leak rate testing of systems which would carry radioactive fluids outside containment described in our aforementioned October 17, 1979 report (Section 2.1.6.a) will be corpleted by January 1, 1980.
The procedures prepared to perform the leak rate testing will also identify manual isolation valves which would be closed prior to use of the systems following a TMI type accident.
This will limit the potential flow of radioactive fluids within these systems.
As required, the results of the leak rate testing will be reported by January 1, 1980.
All other items required by the NRC to be implemented by January 1, 1980 will be so completed as identified in our aforementioned October 17, 1979 report.
- b.
For those items required by the NRC to be implemented by January 1, 1981, plant modifications associated with some of the items have been proposed to be deferred pending completion of the integrated assessment in connection with the Systematic Evaluation Program (SEP).
The remainder will be completed independent of the SEP.
Those items for which plant modifications have been proposed to be deferred to SEP include new instrumentation to detect inadequate core cooling (Section 2.1.3.b),
combustible gas control (Section 2.1.5.a),
plant shielding for personnel and equipment (Section 2.1.6.b),
safety grade automatic initiation of the Auxiliary Feedwater System (Section 2.1.7.a),
safety grade auxiliary feedwater flow indication (Section 2.1.7.b), improved post accident sampling system (Section 2.1.8.a),
improved post accident onsite radiological spectrum and chemical analysis capability (Section 2.1.8.a), improved post accident onsite stack effluent radio-iodine analysis capability (Section 2.1.8.b),
and reactor coolant system high point venting (Section 3.2).
Although the actual extent of the modifications for each of these items has not yet been established, our judgment is that modifications may be extensive and that each of the items will be affected by potential modifications identified by SEP topic reviews of station design and operation. Accordingly, the basis for deferral of each of these items as described in our aforementioned October 17, 1979 report remain applicable.
However, we intend to proceed with the evaluations required by the identified items on the NRC required
-4 schedules and re-evaluate our assessment following completion of the evaluations to determine if any of the modifications can be completed outside the SEP.
We will advise you of the results of our re-evaluation in a timely manner subsequent to January 1, 1980.
Those items which will be completed independent of the SEP include extended range noble gas effluent monitor (Section 2.1.8.b),
high range in-containment radiation monitors (Section 2.1.8.b),
onsite technical support center data display and transmit system (Section 2.2.2.b), containment pressure indication (Section 3.1.1), contain ment hydrogen concentration indication (Section 3.1.2),
and containment water level indication (Section 3.1.3).
Sufficient information is not yet available to accurately project if these modifications can be completed by January 1, 1981.
As the design details for these modifications become available, they will be evaluated with the purpose of establishing a reliable implementa tion schedule. We will advise you of the implementation schedules and the basis therefore if the modifications can not be completed by January 1, 1981.
- 2. As described in our aforementioned October 17, 1979 report, some of our methods of implementation are not in complete agreement with the NRC requirements.
For each such case, a description of our proposed methods along with justification for the differences is provided below:
- a.
The backup nitrogen pneumatic supply to the PORV block valves (Section 2.1.1), the PORV, block valve and safety valve position indication (Section 2.1.3.a), and the diverse containment isolation signal and prevention of automatic reset of containment isolation valves (Section 2.1.4) to be installed during the special outage commencing on or about January 1, 1980 may not be seismically or environmentally qualified.
If qualified equipment cannot be installed, a qualification program will be implemented to qualify the equipment consistent with the seismic criteria for the existing component or system to which it is a part and with the environment in which it is expected to operate.
It is anticipated that the qualification program will be completed and appropriate action taken by January 1, 1981.
We will advise you if we are unable to qualify the equipment by January 1, 1981.
- b.
The power operated relief valves (PORV's) and their associated block valves will not be supplied from different emergency power buses (Section 2.1.1).
At San Onofre Unit 1, two parallel PORV paths are provided for overpressure relief capability. The control power to the PORV and its associated block valve in one path is fed from the same vital bus, and the control power to the opposite PORV and its associated block valve is fed from a separate vital bus.
The PORV's and block valves are air operated, and in the event of loss of power and/or loss of instrument air, the PORV's fail closed and the associated block valves fail open.
Based on a failute modes and effects analysis, the current method of supplying power to the PORV's and their associated block valves (i.e., "train aligned") provides higher operational reliability relative to single failure protection and redundancy than the NRC require-
-5 ment.
In addition, reactor overpressurization protection safety considerations would be degraded if the NRC requirement is implemented.
Therefore, the NRC requirement will not be implemented.
- c.
A controls grade primary coolant saturation recorder in conjunction with a saturation temperature/pressure curve (Section 2.1.3.b) will be utilized rather than safety grade, redundant primary coolant saturation meters required by the NRC.
As clarified by your October 30, 1979 letter, an acceptable alternative method of complying with the stated NRC requirement involves the use of a highly reliable single channel environmentally qualified, and testable system plus a backup procedure for use of steam tables. to this letter provides the information required by your October 30, 1979 letter necessary to demonstrate the reliability of the controls grade primary coolant saturation recorder.
- 3. Of those items requiring prior NRC review and approval, design details are only provided at this time for the item required to be implemented by January 1, 1980 (i.e, Section 2.1.7.a, Controls Grade Automatic Initiation of Auxiliary Feedwater System). to this letter describes our method of providing controls grade automatic initiation of the Auxiliary Feedwater System.
Although not required as part of this response, Enclosure 1 also describes our method of providing controls grade auxiliary feedwater flow indication to the steam generators (Section 2.1.7.b) for completeness.
The remaining items requiring prior NRC review and approval are to be implemented by January 1, 1981. The schedule for submitting the design details for these items is as follows:
- a. The functional design of any new instrumentation for inadequate core cooling (Section 2.1.3.b) will be submitted by January 1, 1980.
The design details for the new instrumentation will be submitted in a timely manner subsequent to that date when available.
- b.
The design details for the high range radiation monitors (Section 2.1.8.b) will be submitted in a timely manner subsequent to January 1, 1980 when available.
- c. The functional design for reactor coolant system high point venting (Section 3.2) will be submitted by January 1, 1980.
The design details for the venting system will be submitted in a timely manner subsequent to that date when available.
- 4. Our method of implementation for those items which do not require prior NRC review and approval will be documented and maintained on file available for NRC inspection.
-6 During a meeting in Bethesda, Maryland on October 24, 1979 with members of your staff, two such items were discussed:
(1) shift technical advisor (Section 2.2.1.b),
and (2) onsite technical support center (Section 2.2.2.b).
For completeness, our method of implementation for these items as discussed at the October 24, 1979 meeting, as modified following additional considerations, are provided below:
- a.
Commencing on January 1,
- 1980, the position of Shift Technical Advisor (STA) will be filled in the following manner:
o During all off-normal hours (i.e., grave and swing shifts, weekends and holidays),
San Onofre Units 2 and 3 operations personnel with previous experience at San Onofre Unit 1 will function as the STA.
In addition, a Station Engineer with a bachelor's degree knowledgeable in the basic engineering and science subjects will be on-call and available at the station within one hour after reactor trips.
o During normal working hours or as otherwise required, either a Station Engineer with a bachelor's degree knowledgeable in the basic engineering and science subjects or station personnel holding a senior reactor operating license on San Onofre Unit 1 will function as the STA.
o Station personnel functioning as STA's or providing on-call assistance will have experience in station design and response.
Commencing on January 1, 1981, the position of STA will be filled in the following manner:
o The STA will have a bachelor's degree in a scientific or engineering discipline or will have equivalent education and experience.
The STA will have received specific training in the response and analysis of the station for transients and accidents and in station design and layout, including capabilities of instrumentation and controls in the control room.
o Once San Onofre Units 2 and 3 become operational, the STA will be assigned to function at these units as well and will have specific training in the station design and response of the units.
- b. The Onsite Technical Support Center (OTSC) is currently established as described in our aforementioned October 17, 1979 report.
However, we cannot provide a direct display of plant parameters by January 1, 1980 since San Onofre Unit 1 does not have a plant process computer and thus cannot utilize a teletype to provide data acquisition in the OTSC.
0
-7 However, the OTSC is imediately adjacent to the Control Room and contains a viewing window through which some of the plant parameters and system's status can be directly observed.
In addition, this close proximity allows ready access to technical information in the form of recording charts, incore thermocouple maps and incore flux data.
Procedures will be prepared to allow key OTSC personnel access to the control room for the purpose of obtaining technical information.
If you have any questions, or desire additional information, please contact me.
Very truly yours, Enclosures (2)
ENCLDSURE 1 CONTROLS GRADE AUTOMATIC INITIATION OF AUXILIARY FEEDWATER SYSTEM AND AUXILIARY FEEIWATER FLOW INDICATION Introduction The existing manually controlled Auxiliary Feedwater (AFW)
System will be modified to achieve automatic initiation of the system along with the remote control capability and AFW flow indication in the main control room (MCR) to each of the steam generators.
Flow Path The AFW flow path that will be modified will consist of electric driven AFW pump supplying flow through a newly added motor operated valve (MOV) installed in the path connecting the AFW pump header to the west main feedwater line which ultimately supplies each of the steam generators.
Flow to the steam generators is controlled by the main feedwater line by-pass valves CV-142, CV 143 and CV-144 which can be remotely throttled from the MCR.
The attached Figure 1 depicts the flow path to be utilized. An appropriate flow indication system will be added downstream of each of the CV's 142, 143 and 144 to indicate remotely in MCR the flow to the steam generators.
Automatic Initiation Logic The AFW system will have the capability to be initiated automatically or remote manually from the MCR.
Automatic initiation will be based on a low steam generator level signal processed from each of the new level transmitters (LT 450X, LT 451X and LT 452X) associated with each of the steam generators.
In order to avoid any spurious and unwanted actuations, the automatic initiation logic will be based on two out of three (2/3) low level signal logic. Remote manual initiation will be accomplished by manual actuation of a control switch in the MCR.
Appropriate indication and annunciation will be provided in the MCR when automatic or manual initiation of the AFW system occurs.
In addition, remote control capability to operate components of the system from the MCR will be provided.
The attached Figure 2 depicts the system operation.
Automatic Initiation Control Panel and Logic Rack The steam generator level transmitters will supply a 4-20 MA signal to the logic rack installed in MCR area, where a low level signal and a 2/3 logic will be developed. A control board (panel) in the MCR will provide monitoring and manual control capability of the system.
If the 2/3 logic is initiated, the operator will be informed through the status indication on the control board; however, no action will be required by the operator to initiate flow since the system will be automatically initiated.
Since auxiliary feedwater flow to the steam generators will be controlled by throttling of main feedwater by-pass control valves CV-142, CV-143 and CV-144, annunciation and
-2 status indication will be provided alerting the operator that "Auto System is Initiated" and that he should control the flow to the steam generators through these valves.
The logic is designed such that once the Automatic Initiation Signal is initiated, it will go through and latch the output which must be manually reset if individual component control is desired.
In case the Automatic System does not initiate actuation of the components, system level manual initiation will achieve this function also. At present, valves CV-142, CV-143 and CV-144 are closed on SIS. However, the control circuit for each of these valves will be modified to override SIS if the Auxiliary Feedwater Automatic Signal is initiated during or after the occurrence of SIS.
This implies that the control valves will always be available to the operator whenever this system is desired.
If the auxiliary feedwater system was initiated prior to the occurrence of SIS sinultaneously with LOP, the system will be tripped so that the emergency diesel generators are placed on line on a dead bus, then SIS related loads are automatically initiated and after 21 seconds of time delay (adjustable), the auxiliary feedwater loads will be placed on the diesel generator and operation resuned automatically. However in case of SIS with no LOP the AFW system will not be disturbed.
The logic and the monitoring panel along with the individual conponents will be powered from the emergency diesel generators.
Flow Indication Since AFW flow to the steam generators will be controlled by manual throttling of CV-142, CV-143 and CV-144, an appropriate flow measurement system will be provided downstream of these valves so that flow through each of the lines to the steam generators will be indicated in the MCR.
This AFW flow indication will provided with diverse back up steam generator level indicators in the MCR.
The existing manual capability of the AFW System will be retained and in addition new flow measurement devices (FC 2002A, FC 2002B and FC 2002C) will be installed with a flow indication locally and in the MCR so that this system is also available should the automatic initiation of AFW not be operable. By appropriate communication between the MCR operator and an operator at the AFW isolation valves, the manual valves can be controlled to obtain flow to the steam generators through this system also. Figure 1 also depicts this system.
Periodic Tlesting The system will be designed to be testable.
In order to avoid unnecessary component operation during periodic testing, the operator will be able to disable the initiation logic by means of "Auto Enable/Disable" control switch.
Whenever the system is disabled, annunciation will be provided indicating this condition. However, if during testing the Automatic System is initiated, the operator will still be alarmed of such event so that he can either place the "Enable/Disable" control switch in the "Enable" mode and control the AFW flow through the control valves or he will be directly able to operate individual components (namely the motor driven pump and cross connect MOV) from the MCR and again control the AFW flow through the control valves.
ENCLOSURE 2 Information Required On The Subcooling Meter Display Information Displayed (T-Tsat, Tsat, Press, etc.)
Hot Leg Temp.,
Pressurizer Pressure, T-Tsat Display Type (Analog, Digital, CRT)
Analog Continuous or on Demand Continuous Single or Redundant Display Single Location of Display Control Room (J-console)
Alarms (include setpoints)
T-Tsat < 50 F Overall uncertainty (OF, PSI)
Accuracy of T-Tsat
+ 1.3% of span Range of Display Hot Leg Temp.
450 to 700'F Pressurizer Pressure 400 to 2400 psi T-Tsat
-100 to +100 0F Qualifications (seismic, environmental, IEEE323)
See Note 1 Calculator Type (process computer, dedicated digital or Dedicated analog analog calc.)
calculator If process computer is used specify availability.
N.A.
(% of time)
Single or redundant calculators Single Selection Logic (highest T., lowest press)
Manual -
any sensor Qualifications (seismic, environmental, IEEE323)
See Note 1 Calculational Technique (Steam Tables, Functional Calculator duplicates Fit, ranges)
Steam table Tsat curve within +0.5% over range 450 to 7000F
-2 Input Temperature (RTD's or T/C's)
RTD's meperature (number of sensors and locations)
Three, one per loop on hot leg Range of temperature sensors 32-10000 F Uncertainty of temperature sensors (OF at 1 )
Accuracy is +.25% of span Qualifications (seismic, environmental, IEEE323)
Qualified commensurate with equipment to which it is attached Pressure (specify instrument used)
Foxboro Ell Series Pressure (number of sensors and locations)
One, pressurizer Range of pressure sensors 0 to 3000 psig Uncertainty of pressure sensors (PSI at 1 )
Accuracy is 1/2% of span Qualifications (seismic, environmental, IEEE323)
Qualified commensurate with equipment to which it is attached Backup Capability Availability of Temp. & Press.
Incore thermocouples, three cold leg RTD's, three Taug indicators and four pressurizer pressure indicators other than those used for the saturation recorder are available Availability of Steam Tables etc.
Yes, control room Training of Operators Yes Procedures Yes
NOTE 1 Qualifications of the Display and Calculator These items are located in the control room and are not subject to adverse environmental conditions.
The display recorder is a Foxboro Spec 200 Series 220S similar to fully qualified safety grade units utilized on San Onofre Units 2 and 3.
The calculator utilizes high quality, reliable components (e.g.
+ 1% resistors).
All components are installed in safety related consoles and instrument racks which meet the original seismic design criteria for the plant.