ML19308A727
| ML19308A727 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Rancho Seco  |
| Issue date: | 09/10/1979 |
| From: | Fairtile M Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7912100554 | |
| Download: ML19308A727 (20) | |
Text
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7 0/v9 MEETIts sui! MARY DISTRIBUTION
_0_RB#4 Mr. Lowell E. Roe, Toledo Edison Company cc:
Mr. William 0. Parker, Jr., Duke Power Company Mr. J. J. Mattimoe, Sacramento Municipal Utility District Mr. William Cavanaugh, III, Arkansas Power & Light Company i
50-312 13, Docket File 346 NRC PDR 269, 270, 287 R. Reid L PCR V. Noonan ORB #4 Pdg P. Check NRR Rdg G. Lainas H. Denton G. Knighton E. G. Case Project Manager - MFairtile OELD I
D. Eisenhut OI&E-(3)
R. Vollrer R. Ingram W. Russell R. Frale B. Grimes Program. y, ACRS (16)
Support Branch T. J. Carter TERA A. Schwencer J. R. Buchanan D. Ziemann Meeting Summary File T. Ippolito NRC Participants W. Gammill L. Shao GVissing RLTedesco J. R. Miller DGarner Dross R. Capra JBeraroya WFKane P. Kreutzer PFCollins ELJordan BABoger PENorian
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September 10, 1979 Dockets Nos. 50-346, 269, 270, 287, 312 & 313 FACILITIES: DAVIS-BES'E 1, OCONEE 1, 2 & 3, RANCHO SECO AND ARKANSAS 1 LICENSEES:
TOLEDO EDISON COMPANY, DUKE POWER, SACRAMENTO MUNICIPAL UTILITY DISTRICT AND ARKANSAS POWER AND LIGHT COMPANY
SUBJECT:
SUMMARY
OF APRIL 24, 1979 fiEETING WITH BABC0CK AND WILC0X (B&W)
AND B&W REACTOR OPERATORS REGARDING RESPONSE TO PLANT TRANSIENTS On April 24, 1979, the NRC staff met with representatives of B&W in Bethesda, Maryland to discuss several considerations related to the safety of nuclear plants, designed by B&W, during a feedwater transient.
Representatives of Toledo Edison Company, Duke Power, Sacramento Municipal Utility District ar d Arkansas Power and Light Company also attended.
A list of attendees is attached (Enclocure 1).
The meeting opened with a presentation of the Agenda (Enclosure 2). The following areas were to be discussed:
1.
Types and frequency of challenging transients, 2.
Role of the Integrated Control System (ICS), especially failure modes, 3.
Thermo-hydraulic behavior of primary and secondary systems, 4.
Mitigation of challenging transients; and 5.
Remedial measures.
r 1.
Challenging Transients B&W stated that the number of transients at their plants had a frequency of about two per year per plant which was compared to about one per year per plant for other nuclear steam supply systems.
In total B&W plants had experienced over 100 feedwater (FW) transients.
B&W in answer to a staff question stated that the frequency rate and total number of transients are not excessive.
As a result of the Three Mile Island Unit 2 (TMI) experience, B&W believes they have taken responsible actions.
In support of this, they presented a summary of the B&W McMillan Report to the ACRS on April 16,1979 (Enclosure 3).
B&W stated that all reactor operators would be retrained on the B&W Simulator in Lynchburg, Virginia as soon as possible; so far 35 have been retrained.
. 2.
Integrated Control System (ICS)
B&W gave a slide presentation of the ICS (Enclosure 4).
The ICS precludes reactor operation with an idle loop, (both pumps in a single loop not operating).
If a single main FW pump is lost power level is automatically reduced to 50%. The reactor input signals to the ICS are flux demand and T average, (average temperature between hot and cold leg of a loop).
The only steam generator (SG) input for level is at the 15% reactor power level where SG water level is held constant.
In response to a staff question, B&W said the plant can be operated manually a t any power level. The utilities said it is done and that it is a " routine non-difficult task".
The ICS can control power changes to the plant, on automatic, between 15% and 100% power at about 10% per minute.
Duke and SMUD said they change power closer to 3% per minute.
In the event of a turbine trip the reactor will be run back, by the ICS, to a 15% power level.
Dr. Hanauer asked for the failure rate of the ICS.
B&W said there were 21 power failures of the system. Hanauer asked if there were any failure modes analysis performed.
B&W replied there weren't any as the ICS is a non-safety system.
Hanauer requested reliability data, B&W said none was available.
B&W promised to perform a reliability study and to submit the failure data.
Huauer asked how long it would take to do a failure effects analysis on ICS, B&W said about one to two months. Tedesco stated it wasn't clear as to what safety criteria are used in ICS design; B&W replied that individual system safety aralysis in the FSAR envelope this concern.
D. Ross stated that a failure of the ICS could result in a total loss of FW; B&W did not rebut Ross's concern.
Toledo Edison stated that the Davis-Besse Auxiliary FW system is not controlled by the ICS as in other B&W plants.
Duke stated that Oconee has a separate auxiliary FW header to the SGs independent of ICS.
SMUD stated that tne Rancho Seco auxiliary FW is controlled from the ICS. We requested a list of plants that ICS can or can't induce auxiliary FW system failure in addition to main FW failure, as well as types of transients so induced.
We requested this information to be submitted in one to two months.
3.
System Transient Response Duke Power described a transient where all power, normal and back-up, to the ICS failed.
Both main FW pumps failed, while the emergency FW pump started, the discharge pressure was too low, and an operator was dispatched to speed up the pump, but a SG went dry for 15 minutes and the other SG for one to two minutes. The station staff handled the evert with no fuel or eystem damage.
In response to a question, Duke said they couldn't get the main FW pumps restarted before the SGs went dry. The pilot operated relief valve (PORV) opened during che transient before the reactor tripped.
SMUD stated their
. auxiliary FW system operated everytime it was required. Toledo Edison experienced only one auxiliary FW failure that was due to a design error in the control circuits, since that occasion there has been no problem with the AFW system Denton asked B&W for the reliability criteria for the AFW. They replied that AFW l
comes on within 40 seconds of demand.
Denton asked if a probability of 1 in 20 is reasonable for needing high pressure in jection because AFW fails and PORV sticks open, (no response from B&W recorded).
Hanauer asked Duke if Oconee in its 15 years of operating experience ever turned on HPI because of loss of all FW.
Duke replied there was one event where main FW failed, emergency FW came on and HPI also came on because of primary system overcooling.
4.
Mitigating Actions Denton stated he is looking for means to avoid SG dryout and would like to hear ideas along this line.
B&W asked for time to explore this with the utilities but stated they believe there are such means.
5.
Remedial Measures a.
IE Bulletins All the utilities present stated that their operators are comfortable with the procedural changes in the IE Bulletins 79-05, 05A and 05B.
The utility manage-ments said the bulletins are easily understood and have been implemented.
b.
Accident Analysis Documentation B&W stated a desire to meet with a small group of NRC reviewers to go over the current status of their analyses and discuss what needs to be done.
The staff provided various combinations of equipment failures to be considered including the role of the operator.
Scenarios discussed were stuck open PORV, one HPI and two HPI available, no small break and various delays in AFW coming on. These analyses were promised by B&W by May 8, 1979.
c.
ICS Reliability Analysis B&W promised this analysis in two months.
d.
Auxiliary Feedwater System Reliability As each AFW system is different at each plant, the utilities discussed this problem individually.
Duke considered constant operation of EFW pumps but rejected the idea.
Duke believes best approach is to start all EFW pumps on loss of main feed on any unit.
Duke has the reliability question under study and will develop a schedule for the work.
Duke stated two EFW pumps could supply enough water for all three Oconee Units.
Toledo Edison said Davis-Besse experienced AFW trips when the system first started in operation but that after two years of debugging, they are now confident of system reliability.
Davis-Besse has run a natural circulation test. Toledo Edison stated
. their reliability is on the order of one failure in 100 demands rather than the one in ten flRC mentioned.
SMUD said the plant has experienced no FW failures for 23 demands or tests.
Arkansas Power and Light has had failures in the system but haven't derived any failure data. AP&L is studying means to upgrade reliability of the system especially in regard to passive single failures.
Tedesco asked if any utility is considering auto-trip of reactor from secondary side. Utilities were not prepared to respond to this question.
M hg bY Morton B. Fairtile, Project Manager Operating Reactors Branch #4 Division of Operating Reactors Enclot;res:
1.
List of Attendees 2.
Agenda 3.
Excerpts from B&W Presentation to the ACRS 4.
B&W Slide Presentation cc w/ enclosures:
See attached distribution list
Sheet 1 LIST OF ATTENDEES April 24,1979 MEETING
_ Attendee Organization NRC M. Fairtile R. W. Reid J. Benaroya P. F. Collins B. A. Boger D. H. Beckham N. Zuber L. P. Crocker a
S. Diab G. S. Vissing A. Szukiewicz L. B. Marsh M. M. Mendonca G. C. Lainas D. C. Dilanni C. Grimes F. Shroeder S. H. Ilanauer R. L. Tedesco D. G. Eisenhut D. F. Ross E. G. Case W. F. Kane J. Carter E. L. Jordan Paul E. Noriar.
Chris Nelson J. J. Watt P. S. Check I. Villciva S. Varga H. Denton Z. Rosztoczy J. A. Calvo f
Sheet 2 Organization Attendee TOLEDO EDIS0N F. R. Miller E. C. Novak T. J. Myers BECHTEL F0WER M. Malcom DUKE POWER K. S. Canady W. O. Parker, Jr.
1 P. M. Abraham J. N. Pope Warren H. Owen C. L. Sansbury PACIFIC GAS AND ELECTRIC CO.
- 0. H. Davis B&W D. W. LaBelle D
H. Roy S. D. Carl on H. G. Smith J. A. Castanes
.)tj,H. Taylor l
BAILEY Davey Cannon R. R. Walker D. Conownight SMUD R. J. Rodriguez J. J. Mattimoe CALIFORNIA STATE 0FFICE Elinor Schwartz ARKANSAS POWER & LIGHT Basil Baker David Trimble
4 AGENDA APRIL 24, 1979 Meeting with HRC Staff and Utilities with Operating B&W Plants
Purpose:
To assess the safety of plant response to feedwater transients, especially as they might be precursors of I
TMI-2 event, and to consider remedial measures (including partial or total reduction in power) that may be necessary entil a complete understanding is 1
available.
Main Topics of Discussion V
1.
The Challenge:
Types and arrival rates of challenging transients 2.
The Control Response:_
Role of the ICS; especially failure modes 3.
The System Response:_
Thermo-hydraulic behavior of primary and secondary systems 4.
Hitigating Systems:
What turns on, and how, to mitigate challenging transients?
i
.t 4
4
s 2-s Remedial Measures:
5.
e A.
Reduce probability ICS fixes?
B.
Reduce consequences AFW design?
OTSG inventory?
, Power reduction?
C.
Dispel concerns Better analyses? tests?
-Detailed Agenda 1.
Cnnilenge rate Briefly_ summarize, for class of B&W plants, the arrival rate of challenging transients 2.
Control Response Discuss Identify the role ICS plays in plant response.
failure modes; false signals (like recent Rancho Seco event);
safety role in controlling 0TSG 1evel; a.
When ICS works; b.
ICS Fe51ure Modes c.
ICS as initiator of transients d.
ICS/ operator interface l
3.
System Response Describe typical system response to challenging transients; uhat tells AFW, HPI to come on; what. tells reactor to trip; what is typical subcooling cargin; what is special response to stuck-open PORV; if natural circulation is interrupted, what assurance is there it will be restored.
l
L4 4.
Mitigating Equipment What has been operating experience with needed mitigating equipment, including power, air, AFW, HPI.
What is role of operator?
5.
Remedial Measures a.
Probability How can control be separated from safety so that as a minimum ICS does not contribute to fast and undesirable swings in pressure, level, and temperature.
b.
Consequences Would plants behave better or worse at 50% P?
Can unreliability of mitigating systems be decreased on short-term?
How can adequate subcooling be assured?
How can better operator action be obtained?
hMU m
EXCERPTS FROM B&W'S PRESENTATION TO THE EM APRIL 16, 1979 NEAR-TERM ACTIONS 1.
ISSUE SUPPLEMENTARY ADVISORY REQUESTING REVIEW 0F INDICATIONS OF AN OPENING IN THE REACTOR COOLANT SYSTEM BOUNDARY RESULTING FROM AN OPEN RELIEF VALVE (THIS WEEK) 2.
CONDUCT A SPECIAL TRAINING PROGRAM FOR OPERATORS TO BECOME MORE FAMILIAR WITH THE TMI-II SEQUENCE OF EVENTS (APRIL 9) 3.
INVESTIGATE AND RECOMMEND DESIGN IMPROVEMENTS WHICH DO NOT AFFECT OTHER PLANT SYSTEMS OR REQUIRE EXTENSIVE ANALYSIS (SIX WEE'(S) o M0.,E POSITIVE INDICATION OF PORV POSITION o
INTERLOCK TO ISOLATE CERTAIN CONTAINMENT PENETRATIONS I
ON ECCS ACTUATION l
e SATURATED CONDITION INDICATOR i
I4, ESTABLISH A TECHNICAL REVIEW COMMITTEE TO PROVIDE WITHIN THREE MONTHS:
o
-RECOMMENDATION F0Ps EQUIPMENT IMPROVEMENTS, OPERATOR INTERFACES, RECOVERY REQUIREMENTS AND INCIDENT SUPPORT AN IMPACT ASSESSMENT OF THE TMI-II OCCURRENCE AND POTENTIAL o
RESULTING CHANGES IN REGULATORY REQUIREMENTS ON TECHNICAL ACTIVITIES
EXCERPTS FROM B%i'S PRESENTATION TO THE ACM APRILlE,_lSZ9.
LolGfjLIERM ACIL0E EXAMINE FOR FEASIBILITY AND DESIRABILITY FURTH ITEMS DESIGN MODIFICATIONS BEYOND THAT IMMEDIATELY ADDRESSED, FOR CONSIDERATION ARE:
REACTOR VESSEL FLUID LEVEL INDICATION INSTRUMEN 1.
~
REAUTOR fRIP ON LOSS 0F FE5DW5fER FU!I ~ ~~~ ~
~
~
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2, CONTAINMENT ISOLATION ON ECCS ACTUATION; PENETRATIONS T 3.
ISOLATED, PENETRATIONS TO BE MAINTAINED 4.
RECOVERY FROM AND MITIGATION OF TRANSIENTS ISOLATED AND SHIELDED LONG-TERM DECAY HEAT COOL 5.
6.
REACTOR VESSEL VENTING-MORE FORMALIZED. STRUCTURE FOR COMMUNICATI 7.
AND OUTSIDE SUPPORT-8.
HANDLING'0F HYDROGEN GENERATION
[
4 IIEid II!!E r
o SUB'STANTIALLY REDUCE CHALLENGES TO PORV 2-3 DAYS o
PERFORM CALCULATIONS, WORST-CASE BREAK WITHOUT AFW FOR 30 MIN, 2-3 DAYS DOCUMENT NATURAL CIRCULATION TESTS CONDUCTED e
AT DAVIS BESSE AtlD 0CONEE 2 1/2 WEEKS o
DOCUMENT ALL OCCURRENCES OF NATURAL CIRCULATION WHICH HAPPENED INADVERTENTLY; INCLUDE A DESCRIPTION OF UNEXPECTED BEHAVIOR 2 1/2 WEEKS' o
DOCUMENT NATURAL CIRCULATION ANALYTICAL METHODS 4 WEEKS o
SUMMARIZE AND DOCUMENT SENSITIVITY-IN KEY PARAMETERS (DEFINITION AND AGREEMENT WITli 8 WEEKS l
STAFF IN TWO WEEKS REGARDING SCOPE) o DEFINE AND DOCUMENT THERMAL SH0CK CRITERIA l
FOR OPERATION AT LOW TEMPERATURE WITH HPI PUMPS RUNNING AND NO NATURAL CIRCULATION 2 WEEKS i
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