ML15238B288
| ML15238B288 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/25/1981 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8103110784 | |
| Download: ML15238B288 (11) | |
Text
FEBRUARY 2 5 IW1 S
NSIC MFai rtil1e eCORB#4 Rdg MPadovan NRC PDR DEisenhut RIngram L PDR RPurpl e IE--3 TERA-3 RTedesco
\\CRS-16 Dockets Nos.LO;Z6 50-270 OEL.D and 50-287 AEOD GLainas Gray File TNovak HOrnstein RReid EBlackwood to Mr. William 0. Parker, Jr.
FRosa Vice President - Steam Production PChek, Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242 10-9
Dear Mr. Parker:
NRC letter dated November 7, 1979 requested that licensees plants (except TMI-1) address the B&W recommendations contained in BAW Report 1564 "Integrated Control "ystem Reliability Analysis".
Your response to this request was provideQto us via letters dated December 21, 1979 and July 23, 1980.
We have reviewed your submittal and have concluded that it does not contain sufficient information for us to determine whether your pro posed actions-are sufficient.
Enclosed' arequestions which we request you review without providing us with a written response.
The questions are being forwarded to allow you a'duate prepara lon time prior to meeting with us at your Charlotte, NC 9eneral offices to discuss our concerns.
We believe that this should provide an opportunity to better understand the details of the ICS"design and toassess your responses by conducting a design review of the ICS system.
We request that ou have necessary drawings, design information, and personnel avatlable so that we may complete this e ort in about three days. We wish to emphasize that no additional analysis or studies should be performud to obtain information that is not presently existing at your offices.
However, B&W attendance at this meeting would be appropriate.
I4e are aware that some of the agenda items have been covered in your IE Bulletin 79-27 program, proposed AtQG pro ram and in other Duke sub
_nittals. -This review is part of NEG-0737 Task Action Plan Itm II.K.2.9. 8103110-..
Wr. Wil iam U. ParKer, Jr.
Please contact the NRC Project Mnager to schedule the meeting at yout offices, preferably early in the month o'f March..
Sincerely, Originai signed by Robert W. Reid Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Meeting Agenda cc w/enclosure:
See next page
- See previous. 318 for concurreIces.
OFFICE) ORB#4:DL ORB#4:D[L3J C-ORB#4:DL
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M'adOVan Ob.MFairtile RReid DATE) 2/20/81 2/ZO/81 2/
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1980-329-824
7-7 DISTRIBUTION:
ORB#4 Rdg RReid Docket File-3 DEisenhut MFairtile NRC PDR RPurple MPadovan L PDR RTedesco RIngram TERA-3 GLainas IE-3 NSIC TNovak ACRS-16 Dockets Nos. 50-269, 50-270 OELD and 50-287 AEOID Gray File H~rnstein EBlackwood Mr. William 0. Parker, Jr.
FRosa Vice President -
Steam Production Duke Power Company PCheck P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 26242
Dear Mr. Parker:
NRC letter dated November 7. 1979 requested that licensees with B&W pi-ants (except TMI-1) address the B&W recommendations contained in, BAW Report 1564 "Integrated ontrol System Reliability Analysis".
Your response to this request was,provided to us via letters dated December 21,11979 and July 23, 1980 We have reviewed your submittal and have concluded that it does not contain sufficient information for us to 4etermine whether your pro posed actions are sufficient.
Enclosed, re questions wq h we request you review without providing us with a written resp onse.
The qestions are being forwarded to allow you adequate preparation time pror to meeting with us at your Charlotte, NC generaI f fices to discuss our concerns.
We wish to emphasize thait no additional analysis or studies should be performed to obtain information that is not presently existing at your offices.
However, B&W attendinc at t1s meeting would be appropriate.
We are aWare that some of the agenda items have been covered in your IE Bulletin 79-27 program, proposed ATOG program and in other Due sub mittals. This review is part of NUR-EG-737 Task Action Plan Item II.K.2.9.
Please contact the NRC Project Manager to schedule the meeting at your offices, preferably early In thie month of March.
Sincerely, Robett W. Reid, Chief Operating Reactors Branch #4 ORB#4:DL Division of Licensing.
\\-adovan/cb
Enclosure:
Questions 2/?o/81 OFFICEO ORB#4:DL1 4C C-R.SB:DSI AD-PS:DSI AD-0R:DL
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t*tREG(,
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON D. C. 20555 February 25, 1981 Dockets Nos. 50-269, 50-270 and 50-287 Mr. William 0. Parker, Jr.
Vice President -
Steam Production Duke Power Company P. 0. Box 2178 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
NRC letter dated November 7, 1979 requested that licensees with B&W plants (except TMI-l) address the B&W recommendations contained in BAW Report 1564 "Integrated Control System Reliability Analysis".
Your response to this request was provided to us via letters dated December 21, 1979 and July 23, 1980.
We have reviewed your submittal and have concluded that it does not contain sufficient information for us to determine whether your pro posed actions are sufficient. Enclosed are questions which we request you review without providing us with a written response.
The questions are being forwarded to allow you adEquate preparation time prior to meeting with us at your Charlotte, NC general cffices to discuss our concerns.
We believe that this should provide an opportunity to better, understand the details of the ICS design and to assess your-respc.nses by conducting a design review of the ICS system.
We request that you have necessary drawings, design information, and personnel available so that we may complete this effort in about three days.
We wish to emphasize that no additional analysis or studies should be performed to obtain information that is not presently existing at your offices.
However, B&W attendance at this niCeting would be appropriate.
We are aware that some of the agenda items have been covered in your IE Bulletin 79-27 program, proposed ATOG program arc indother Dukesub mittals.
This review is part of NUREG-0737 Task Action Plan Item II.K.2.9.
Mr. William 0. Parker, Jr.
-2 Please contact the NRC Project Manager to schedule the meeting at your offices, preferably early in the month of March.
Sincerely, Robert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Meeting Agenda cc w/enclosure:
See next page
Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 2178 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Raleigh, North Carolina 27603 Oconee County Library 501 West Southbroad Street Walhalla, South CaroI na 29691 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Director, Criteria and Standards Division Office of Radiation Programs (ANR-460)
U. S. Environmental Protection Agency Washington, D. C. 20460 U. S. Environmental Protection Agency Recion IV Office ATTN:
EIS COORDINATOR 345 Courtland Street, N.E.
Atlahta, Georgia 30308 Mr. Francis Jape U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 17th Street, N.W.
Washington, D. C. 20036
Enclosure MEETING AGENDA
- 1. Even with the modifications made to improve the rel ability of the power supply of the Oconee ICS/NNI, it appears that there are a number of single failures which can cause a loss of power to the ICS/NNI.
The information submitted does not adequately describe the consequences of a power loss.
The following should be discussed:
a)
Identify the automatic control actions which would occur following loss of the ICS/NNI power supply. The concurrent actions of equipment not a part of the ICS NNI but which receive power from or control signals dependent on the same power supply should also be identified.
The above should consider all plant operaiting modes (i.e., power operation, c
1it cality to powe operation, approach to criticality, hot shutdown cold shutdown). The discussion can be limited to control of plant processes having a significant impact on plant safety (i.e., reactivity; reactor coolant pressure, temperature, flow and inventory; and secondary system pressure, inventory, and flows.
b) Identify the plant transient which would occur following the power loss of (a) above assuming no operator actions.
c) Identify the operator actions and' the times following the power loss at which the actions must take place in order to bring the plant to a safe, stable condition.
d) Identify the information required by the operator for the actions in (c) and confirm that the information would be available with the loss of power.
-2 e) For (a) through (d) above provide a discussion of the consequences resulting when a loss of power supply occurs during the time a control action is actually taking place.
For example, a power loss occurring when pressurizer spray is in operation.
- 2. How does the operator determine specific instruments,.controls, and indications which are not operating following a loss of power supply? Attachment 3 to Duke Power Company letter dated July 23, 1980 infers that the operator must refer to lists contained in emergency procedures. Is this correct? If so, pro vide us with a copy at the discussions.
- 3. Duke Power Company letter dated December 21, 1979, indicated that ICS/NNI loads were not split among separate power supplies since design reviews indicated that splitting the loads would not significantly reduce the probability of a unit trip following loss of power supply.
The benefits of splitting the ICS/NNI loads on plant operations occurring after a trip with loss of a power source should be discussed.
The benefits of splitting the ICS/NNI loads on indications available to the operator following loss of a power source should also be discussed.
- 4. Attachment 3 to Duke Power Company letter dated July 23, 1980, indicates that the Oconee ICS/NNI has been modified to provide indications and controls needed to maintain the plant at hot shutdown independent of ICS/NNI power supplies.
The specific design criteria for these indications and controls should be discussed. Also:
3 a) Are the indications and controls redundant?
b) Do the indications and controls meet all applicable codes and standards for protection? system instrumentation?
c) Is one channel of each indicated parameter recorded for use in trending, instant recall and post-event evaluation?
d Is operator action required to transfer the indications and/or controls from one power supply source to another?
e) Is operator action required to transfer equipment control from one controller to another?
f To what degree are the indications and controls independent of the ICS -
totally independent, supplied by separate power sources, independent transmitters etc.
- 5. Duke Power Company letter dated December 21, 1979, stated that the feasibility of revising the manual transfer scheme to automatically, select each reactor coolant loop's highest flow for input to the ICS was being investigated.
The results of this investigation should be discussed. If an auctioneer is to be used, the consequences of failure of the auctioneer and the means by which the auctioneer's correct operation is monitored should be discussed.
- 6. Other than the reactor coolant flow signal, are there other parameter input signals to the ICS on Oconee which have a high failure rate history or the failure of which would significantly affect control of more than one key plant parameter?, i e, a) Steam Generator Inventory b) Reactivity c) Reactor Coolant Pressure
-4 d) Steam Generator Pressure e) Reactor Coolant Flow f) Reactor Coolant Inventory
- 7. Duke Power Company letter dated December 21, 1979, described several modifications to be made to improve the reliability of the feedwater/
condensate system controls and to improve feedwater pump turbine drive minimum control.
The status of these modifications should be discussed along with the performance criteria used in evaluating the adequacy of the modifications and the ICS/BOP system tuning.
The performance criteria should include performance during plant startups and transition from use of the startup feedwater control valves to use of the main feedwater control valves as well as performance following plant trips.
- 8. Discuss plant operating modes and ICS control features where the operator frequently must intervene and execute manual control.
- 9.
An analysis of the consequences of excessive feedwater caused by a single component failure, e.g., single valve stuck open, controller calling for full feed, etc., should be discussed. The analysis should indicate cooldown rates which occur and the time after failure to reach a limiting condition, e.g., time to fill a steam generator.
Required operator actions and the times after the component failure at which operator actions are required should be included.
The analysis should clearly indicate if and where credit is taken for operation of equipment not designed to protection system standards.
- 10.
An analysis of the consequences of a stuck open turbine bypass valve, including cool down rates, operator actions required and the times at which operator actions are required should be discussed. The analysis should
-5 clearly indicate if and where credit is taken for operation of equipment not designed to protection system standards.
- 11.
Procedures and/or operating practices used at Oconee to detect component failures in the ICS which might go undetected for one plant operating mode (such as relatively stable high power operation) but might lead to a transient when the plant goes to another operating mode (such as a power change or low power operation) should be discussed.
Procedures and/or operating practices to minimize the possibility of undetected failures occurring and continuing to exist until other failures occur should be discussed since this sequence could negate the approach taken in the analysis of BAW-1564.