ML19225C461
| ML19225C461 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 06/25/1979 |
| From: | Head J SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| FOIA-79-98 NUDOCS 7907300382 | |
| Download: ML19225C461 (12) | |
Text
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Southem Califamia Edison Company g5 s' O. som S CO 22A4 wALMuf grove AVCHut ROSCM C AD. C AWCmMI A 9a770 Testa =o=C J. T. M CA D, J 88 I
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.u U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region V Suite 202, Walnut Creek Pla a 1990 North California Boulevard Walnut Creek, California 94596 Attention:
Mr.
R.
H. Engelken, Director dmo. 50=20 c'~~Sha Onofre Unit 1
References:
(1) SCE (J.
H.
Drake) letter to NRC (R. H. Engelken) dated May 3,
- 1979, Docka_t No. 50-206 (2) SCE (J.
T.
Head, Jr.) letter to NRC (R. H. Engelken) dated May 23, 1979, Cocket No. 50-206 References (1) and (2) provided our responses to IE Bulletin 79-06A and Revision 1 thereto which requested information concerning the recent Three Mile Island Incident.
Reference (1) addressed IE Bulletin Items 1-12 while Reference (2) addressed IE Bulletin Item 13 and provided schedular information for cc=pleting the outstanding action items identified in Reference (1).
The purpose of this letter is to provide additional information to supplement our responses submitted by References (1) and (2).
The additional information is contained in the enclosure to this letter entitled, " Supplemental Information to May 3, 1979 and May 23, 1979 Responses to Bulletins79-06A and 79-06A, Revision 1, San Onofre Nuclear Generating Station, Unit 1, Docket No. 50-206."
To fac_.litate your review of the enclosure the supplemental information has been numbered to correspcnd to the IE Bulletin action items and our associated responses contained in References (1) and (2).
Should you have any questions or require additional information, do not hesitate to contact me.
Sincerely A7C ne I u
,p /
4JJ U4'
+
Enclosure Office of Inspection and Enfjreemen O /RO W W ~77p cc:
- Director, Division of Reactor Operations Inspection 9 073 0037 >
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SUPPLEdnTAL INFOMATION TO KAY 3,1979 AND MAY 23, 1979 RESPONSES TO BULLETINS 79-06 A AND 79-06A, REVISION 1 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 DOCKdT NO. 50-206 The supple = ental infor:ation below is nu=bered to correspond to the Bulletin action items.
Ites 2 A thorough review of all transient and accident conditions based on insight gained fres THI-2 is currently in progress. The scope of this review will be sufficiently broad such that subsequent procedure revisions will (a) assure that action steps specifically warn of potential for voiding with a description of all instrumentation which sight provide indication of potential or actual voiding, (b) specifically address operator actions, based on operational modes and instru=ent indication discussed above, for terzinating conditions tending to lead to void formation and (c) provide operators with guidarce for enhancing core cooling given the unexpected conditions of actual voiding in the pri=ary systes.
As stated in our May 23, 1979 letter to the NRC, we have scheduled co:pletion dates of August 15, 1979 for the review and September 15, 1979 for revision of operating instructions based on appraisal of the review.
A summary of the results of the review and subsequent procedure revisions will be suDmitted within two weeks after co=pletion of the review.
Cur review thus far has resulted in issuance of st lon me=orandums and precedure revisions which provide operators with instructions and infor:ation with respect to identification and control of syste:
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e, eh-voiding, enhancing core cooling, and =aintaining natural circulation capability.
We have also identified instrumentation available in the control room which might be utili=ed in void recognition.
The best means currently available to the Jperator to determine if voiding has occurred is a comparison of reactor coolant system pressure with reactor coolant bot leg temperature to verify that a margin to saturation conditions exists. As noted in our May 3 response, a curve has been provided to the operators with instructions as to its use amd signific ance.
Other instru=entation available in the control room which might provide the operators with indication of core voiding during natural circulation include:
(1) Incore ther=ocouples and (2) Scurce range nuclear instru=entation Instru=entation available in the control room to verify natural circulation is being =aintained includes:
(1) Reactor coolant hot to cold legg1T cn all three loops.
(2)
Cold leg reactor coolant te=peratures on all three loops.
(3) Hot leg and cold leg average reactor coolant te=peratures on all three loops.
(4)
Incore ther=occuples.
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In addition, several other indirect indicators of natural circulation would be available in the form of instru=entation and equipment which is indicative of heat removal from the secondary side of the steam generators (i.e., steam generator level and pressure, steam du=p, etc.).
Our continuing review of transient and accident conditions =ay identify further instrumentation which might provide indication of systes voiding and/or natural circulation. As noted above, the review results will be sw=sarised a.td any additional operator aids identified will be reflected in our review description.
In the interi=, the above infor=ation regarding instrumentation available to detect core voiding and verify natural circulation will be reviewed by station operators by June 30, 1979.
Documentation of the review will be =aintained in plant records.
Item 4 Cur operating procedures require manual initiation of containment isolation based on the existence of conditions described therein which indicate that a loss of coolant has occurred. In addition, manual containment isolation is required if the 2 psig automatic isolation setpoint is reached coincident with safety injection initiation to prevent certain valves from reopening if pressure subsequently drops below 2 psig.
Further= ore, we will revise operating procedures as necessary to require manual actuation of all automatic containment isolation valves in the event of automatic initiacion of safety injection by July 15, 1979 We are currently evaluating a plant design enange which would (1) provide for auto =atic contain=ent 5
044
4-1 solation upon receipt of a SIS actuation signal or 2 psig containment pressure, and, (2) require =anual reset of containment isolation. This change would eliminate the necessity for manual operator actions as described above.
The contain=ent isolation signal (CIS) isolates all but the following containment penetrations:
(1) Component cooling water supply and return lines.
(2) Main steam and feedwater lines.
(3) Instrument air supply lines.
(4) RCS Letdcwn line.
(5) Reactor Coolant Pump Seal Return line.
(6) Lines associated with engineered safety features.
(7) Lines which are normally closed during operating modes requiring containment integrity and remain closed following an accident.
The ccmpenent cooling water system is completely closed within containment and thus was not provided with isolation valves.
The turbine stop valves serve as isolation valves for the main steam lines. These valves close automatically on turbine trip.
The main feedwater lines are isolated via actuation of the Safety Injection system when the feedwater pumps switch to safety injection service.
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- ' The instrument air supply line is auto =atically isolated when supply pressure falls below 60 psig. Since isolation is achieved at a pressure which is above the contain=ent design pressure.
(49.4,sig), isolation of this line is provided if the air supply pressure will not assure positive inflow to containment.
The RCS letdown line downstream of the letdown orifices is not designed to acco==odate full RCS pressure.
Auto =atic actuation of the centain=ent isolation valves, either during routine CIS testing or inadvertently, would cause lif ting of the relief valve which is provided to prevent overpressurization of this line. Therefore, the isblation valves on this line are closel by re=ote =anual actuation when required for centain=ent isolation as specified in our operating procedures.
Is addition, the RCS letdown orifice isolation valves are autc=atically closed upon receipt of a safety injection signal. These valves would isolate the RCS letdown line.
The RCP seal return line is not designed to accoc=odate full RCS pressure.
For the sa=e reasons discussed above for RCS letdown, this line is also provided with re=ote =anual actuation isolation valves which are closed per operating procedure as required.
None of the lines needed for RCP auxiliaries are autc=atically isolated on CIS or SlS. Therefore, auto =atic contain=ent isolation does not prevent RCP cperation, b$
On.6
. Item 7.b IE Sulletin 79-06A, Revision 1, Item 7.a requested that operators be instructed not to override automatic actions of engineered safety features, unless continued opera
- ion of engineered safety features would result in unsafe plant conditions.
Item 7.b provided specific instructions for application of the Ites 7,a request for automatic safety injection.
Since San Onofre does not have the typical HP/LP safety injection syste=, our May 3 response to ites 7.b required modificati'an to the specific requests to comply with the intent of the criteria therein.
HPI at San Onofre is provided by the charging pu=ps while LPI at San Onofre is provided by the Safety Injection pu=ps in combination with the
=ain Feedwater Pu=ps.
Due to the high flow rate of the LPI system, it is possible to pu=p the design basis quantity of water to the RCS in less than 20 minutes.
Our response to Item 7.b(1) is intended to preclude termination of safety injection (h*P and LP) after a LOCA until align =ent of the safety injection recirculation system is necessary, regardless of how much time is required to reach this point.
For a s=all break LOCA, this time could greatly exceed 20 =inutes while for a large break LOCA, this ti=e could be a sinimum of about 8 minutes.
In this case, the inventory of water in the Refueling Vater Storage Iank has been established to ensure that (1) the design basis quantity of water has been pu= ped to the RCS h? t JJ 0$,
7 (i.e., witnin approximately 8 minutes), and, (2) sufficient inventory re ains to continue Centainment Spray for an additional mini =um of 20
=inutes. The plant safety analysis was perfor=ed based on these assumptions.
Our response to Item 7.b(2) is intended to prevent termination of safety injection (EP and LP) when actuation is not a result of a LOCA for at least 20 minutes aa1 until all hot and cold leg te=peratures are and can be maintained at least 500F subcooled unless the pressure / te=perature criteria for the pressure vessel are reached. These pressure / temperature criteria are contained in our Technical Specification Figure 3.1.3b.
Our May 3 response indicated that operating Instruction S-3-5.5 provides direction for post-safety injection operator actions. This procedure was revised as noted in the respcase and included in the operator training progras per Bulletin requirements.
A thorough review of all e:ergency operating procedures is being conducted to verify that no specific instructions conflict with the Bulletin require =ents.
This review will ba completed and any identified conflicts will be resolved oy June 30, 1979 Any procedures which are modified as a result of this reviee will be incorporated in the operator training program.
Our review thus far has identified one conflict regarcing steam generator tube failure. This accident is a special case LOCA which does not resulc in utilization of the safety injection recirculation syste:.
In this case, the criteria of 7.b(2) as stated above will be applied until plant cooldown is acco:plished and decay heat is being re=oved through the residual heat removal system.
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-o_
Item S As described in our May 3 response, existing procedures which govern the align =ent, alignment requirements and manipulation of safety-related valves during maintenance, testing and plant and system start-up are currently being reviewed. This review is scheduled to be co=pleted by July 31, 1979. The objective of the review is to determine whether or not the procedures adequately ensure that such valves are returned to their correct positions following necessary manipulations. A summary of the results of the review and any procedural revisions necessary will be submitted within two weeks after completion of the review.
As discussed above, the review of procedures is limited to ensuring prcper safety-related valve positions following necessary manipulations.
During all other operational modes, routine scheduled supervisory periodic surveillance to ensure proper safety-related valve positiens is not cons 13ered necessary for the following reasons:
1.
Position indication for automatic and remote manually operated valves is located in the control room. Should the valve change position, this change of state would be observed in the control room.
2.
Manually operated valves important to the proper operation of systems are chained and locked in position, requiring a key and positive intentional actica to ccange the valve position (e.g.,
locking the valve handwheel or stem prevents the valve from
" vibrating" closed er open).
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3 There are surveillance requirements incorporated into the Technical Specifications which require equip =ent or syste= tests wnich verify proper valve align =ent and/or operability (e.g.,
monthly hot functional testing of the Safety Injection Syste=,
monthly containment isolation valve actuation tests, diesel generator testing, etc).
4.
Should a valve not be in the proper position, there are process indications indicative of i= proper align =ent (e.g., loss of flow, pressure increase or decrease, level changes, etc.).
5 operating personnel perform routine plant inspections which may detect i= proper valve position (e.g., checking for spills fron open valves, disecarges from relief valves, etc.).
6.
Valves are located in controlled plant areas (both security and radiation) which require preauthori:ation to enter; thereoy limiting access to the valves.
Ite: 9 Cur May 3 response specifically addresses valves returning automatically to original position on resetting of CIS. Ihe first paragraph of our response states that Operating Instruction S-3.5.5 nas been revised to require the operator to manually activate the containment isolation
~
signal when he deter =ines that a loss of coolant has occurred.
This action is required prior to resetting safety injection and will cloce the reopening of these valves even if the CIS resets due to contain=ent pressure f alling below tce 2 psig setpoint.
b5 000
. The design char.ge we are evaluating (s4. response to Ites 4 of 'chis letter) would prevent auto =atic reset of the CIS, thereby avoiding the requirement for manual operator action.
Item 10 The review end modification of procedures described in our Kay 3,1979 letter has been extended to include safety-related systems. The review and modification of procedures is schedul9d to be completed by July 31, 1979 A su==ary of the results of the review and the actions taken will be submitted within two weeks after completion of the review.
Item 10.b Our Kay 23 letter listed a scheduled completion date of July 31, 1979 for this ites.
A sc==ary of the results of our review and the actions taken will be provided within two weeks after ccepletion of the review.
Item 10.c The Superintendent authorizes re= oval of syste=s frem service and return to service. Under emergency conditions, this responsibility may be assu=ed by the Watch Engineer (shift supervisor).
Relay of system status infor=ation from one shift to the next is carried out by operating personnel as required by operating procedures.
Each operator finishing his shift discusses all activity accccplished under his cognizance with his counterpart for the oneccing shift (i.e.,
Watch Engineer with Watch Engineer, Control Operator with Control Operator, etc.).
These discussicas place e=phasis on the status of all C5l
.11 activities in progress at shift change. In addition, the oncoming snift is required to review the Control Operators Log Book for preceeding shifts and to so indicate this review by initialling the book.
Item 11 Station procedures S-E-203 and S-VIII-1.4 have been revised to include an explicit equirecent for notification of the NRC withi one hour of the time the reactor is not in a controlled or expected condition of operation.
Ites 12 Our review and subsequent revision to cperating procedures to provide appropriate acticn fer hydrogen control will be expedited to meet a target co:pletion date of July 31, 1979.
In the interi=, cperaters have been instructed by a May 3,1979 memorandus of operating modes fer dealing with hydrogen gas.
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