ML18087A789
| ML18087A789 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/21/1983 |
| From: | Fairtile M Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8303290676 | |
| Download: ML18087A789 (25) | |
Text
e e
UNITED STATES NUCLEAR R~GULATORV COMMfSSION WASHINGTON, D. C. 20555 MINUTES OF MEETING BETWffN B&W PLANT OPERATORS' REGULATORY RESPONSE GROUP ( RRG) AND NRC SALEM GENERIC. IMPLICATIO_NS TASK FOR.CE The meeting opened at 9:00 a.m. on March 8, 1983 in Bethesda, Maryland.
at the NRC offices.
The B&W RRG was invited to the meeting to discu~s the impact of the Salem 1 events of February 2_2 and 25, 1983 on the safe.
operation of B,&W* reactors.
The. 1 ist of attendees is Enclosure-1.
R. Mattson, Chairman of the Task Force described the importance of. tQe Salem ATWS event, the establishment of the NRC Task Force, and the.
- Apri1'18 report' to be issued to the Commission.
He pointed out th.ree.
general areas of NRC interest:
- 1.
Shaul d additional prompt actions be taken?
- 2.
How well are NRC and industry doing in learning management lessons?
- 3.
How should the Salem event affect the ATWS ru~emaking_?
Mattsori referred to the agenda of nine questions as a way of structuring the meeting.
The RRG.. preferred to follow its own agenda (Enclosure 2) which included responses to the NRC questions *.
Di~k Crouse, of TECo.and the Chairman of th~ RRG, initiated the Owners' presentation.
He sta.ted that the Salem event did not affect safe....
operation of the B&W plants, principally due to the electrical di*versity in the B&W trip sys tern.
Howard Stevens of B&W presented details of the Reactor Trip System as shown in slides (_Enclosure 2).
There are two B&W designs; one i.s used
.at the Oconee Units, TMI, CR-3, AN0-1 and Rancho Seco (Fig. 2.1); the
- other design is used at Davis-Be~se (Fig. 2.2) ~nd in Dnits not yet 1
licensed for operatio~.
In a discussion of the B&W RPS design, several key points were emphasized.*
The reactor protection system (RPS) is a four-chanriel redundant system.
The control. rod drive control system (CRDCS) has redundant power supplies.
_For the Oc6nee design the A and B channel outputs of the RPS feed into.
the undervoltage relays of the AC trip breakers on the CRDCS.
The C.
and.D outputs feed into trip breakers on the DC power supply side of the safety rod drives and turn off silicon controlled rectifiers (SCRs) in circuits of the regulating rods.
In the 11 Davis-Besse 11 design all 4 channels feed circuit breakers and selected combinations of two _of the four breakers interrupt power to*
all control rods.
The C and D channels also turn off silicqn controlled rectifiers in circuits for a11 rods.
The regulating rods alone in the 110conee 11 design are sufficient to take the reactor subcritical.
All breakers a~d silicon controlled.
( 83032906 76 830321
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' PDR
3/8/83 rectifiers (SCRs) are tested monthly, with plant in Modes 1 or 2, Certain relays in the 110conee 11 desi:gn, feeding the SCRs, part'of the reoulating rod groups, are not "safety related", as there are sufficient "safety grade" relays in the safety rod group.
Similarly,.
the circuits which interrupt rod control power through the SCRs on the Davis-Besse design are not "safety grade.
The regulating rod system, per* Mr. Crouse,. is not a part of the "safety system but* i.s provided for power generation control; however, it does provide
- "electrical diversity".in. terms of safe operation.
The RPS system breaker design specs did not contain a specific reliability criterion but did have a specific requirement for a 80 millisecond operating time.
All RPS components including breakers and breaker internal components are covered by the same safety-related maintenance* procedures.
The staff pointed out that at Salem a maintenance worker did not treat a breaker as "safety related" which resulted in an improper maintenance procedure being used, apparently resulting in breaker failure.
The operating personnel from several of the B&W plants discussed procedures* at their plants to demonstrate*
that the Salem lack of maintenance couldn't happen.
One such procedure is attachment
- 3.
The plant computers are relied on to read out status of RPS and CRDCS
. breakers after a trip.*. Plants have many indications on control panel of success of trip when signalled.
There are apparently no indic~tors
.in the B&vJ plants to tell what signal was effective in causing a t_rip,.
but some plants have "first out" panels.
The. staff asked if emergency procedures call for manual trip when there is. indication of an auto-trip.
The owners present gave varying answers some do, some don't.
The ATOG procedures will call, 'fi0r a manual trip after verification that an auto trip did.not occur~_: The staff asked what sacrifice would be made if manual trip were called for immediately upon indication of an auto *trip signal; licensees *res.ponded that sequence information might be lost and that there may be. too many trips if the manual trip was initiated when one bistable alarmed.
The RRG agreed that the proper place to pursue this question was between the staff and the ATOG procedures group.
H. Stevens said there is a "trip confirm" indication given on control panel in two places and in the plant computer.
The indication is not all "safety reJated 11 and that control panel could give false information.
However, licensees pointed out that there is redundant indication.
Plant computer is:on separate power supply, not on NNI power.
Protection sys:tem power supply, which is safety related, provides power to "rod.
bottom" 1 i ghts.
The Reactor Trip Breaker for B&H plants is shown on Fig. 2.3.. It contains
- a. shun.t trip dev-ice that is not part of the protection system and activates only for overvoltage protection of CRDGS.
Binding of trip shaft or in mechanism of undervoltage trip device (GE Model AK-2A-2S) have been sources of breaker problems at B&W plants.
The staff asked for a sample of the GE undervoltage trip device used in the breaker; the owners group agreed to supply one.
- e.
3/8/83 Bob Gill presented Operating History, Testing and Maintenance.
B&l~
was the source of the information shown on this slide.
Of 469 scram
, demand situations there have been no.failures of any breakers.
The staff asked how reliable this information is. Jim Taylor responded that the only breaker fa i 1 ures known to B&~/ occurred during s urvei 11 ance testing.
Out of about 5000 surveillance tests only 19 failures are reported in LERs.
Of the 19 failures, 14 separate breakers were involved.
In 1982 on three separate breakers there were five surveillance failures at AN0-1 and Rancho Seco.
Since there was increased surveillance in 1982 the failure-per-demand in 1982 may not be different from 1980 or 1981.
A 11failure 11 includes sluggish operation.
Licensees donl't believe increased number of failures in 1982 were due to any causes different from previous failures.
The staff asked what the B&W owners are doing to follow failures of safety related components.
Licensees responded that with this type of 11ow failure rate history it is difficult to detect relevant failure trends with a small data base so they tend to rely in INPO and other large systems of data.
Ths staff asked if similar equipment (breakers) in other applications in plants have any failure history.
The licensees could not, at this meeting, provide any useful information.
Licensees suggested we query breaker manufacturers for fail~re data.
Jim Taylor pointed out that information is disseminated to owners by B&W on the basis of perceived safety significance.
The staff asked what actions weTe being taken by Arkansas and SMUD in response to 1982 failures.
SMUD was unprepared to respond; Arkansas personnel described their studies and in light of the Salem event, wiJl look at what additional activities are needed, if any.
Gill ~ointed out that breaker maintenance is performed at regular frequncies, apparently a different situation than at Salem.
The actual breaker overhaul program at B&W plants was presented in a detailed slide showing 13 checks and procedures to be followed.
RRG said breakers are tested after overhaul and before startup.
W. Johnson of.FPC presented Management GontPols.
The first slide was 11 Post Trip Review"; it describes a management approach to a restart decision after a trip.
The staff provided an overview of the Salem event, pointing out that Salem management had missed the Arns event on February 22 and allowed restart.
Each licensee indicated that prior to restart after a trip event, the cause of the trip must be understood.
Some licensees have independent reviews by two groups and they must agree on the cause of the trip prior to restart. It was pointed out by the RRG that the computer-based "events recorder" at Sal em may have not recorded the events in chronological order.
(NRC staff subsequently reconfirmed its under-standing of the Salem events recorder.)
Some B&W plants have a post-trip procedure (check list).
Duke supplied one for Oconee - Attachment 4.
Duke uses a specially designated performance en!!J'ineer to study trips before restart.
The plant manager or designee makes restart decision at all B&W plants.
The staff pressed the B&W Owners for thoughts on how they would assure they aren't missing an ATI1S event just as the Salem station personnel had missed it. It was clear that a trip perceived by the utility as a "normal trip" might not receive an in-depth review before restart.
The staff and owners discussed the value of event recorders for short and long-term post-event analysis.
3/8/83 /
/
In the Manage nt Control presentation the owners also described how
~*
information co cerning operating events reaches operating pl ants.
Al$O B&W said it int ds to provide a list to each utility of equipment~
related informat1 n previously supplied to that utility, includinlwgeneric
- material that B&W etermines is needed by each utility.
The utiJities will check the lis for holes in their existing information.
~
l' The staff explored wi h the RRG how management philosophy i0mparted to operators; are they co*cerned about continued power produc)lion, or will they trip the plant if' afety is in doubt?
There were si;./ong RRG management views that th. highest priority was to be pfd to safety.
The RRG presented equipmen classification and implem1ntation information.
The Staff expressed concern ver operational QA at ~rating plants and..
asked RRG if there is a simpl r means to solve cla),;s'ification problems other than recilassifying all.s fety-related compoQients.
The SMUD repre-sentative described that uti l ity\\s new computer p~sed equ~1pment list, saying it had been preempted by e~eir response fo the EQ'rTulemaking and would help the classification prob~\\m.
/
Ted Enos described th~ response of s\\w plan;glto ATWS events like Salem, only at fUll power.
He said that all\\B&W llicensees have ATWS related
. procedures; however, ATOG program wn1ll 'l>'miefrove the ATl~S procedur.e*s.
ATOG is being implemented at AN0-1 and later ~n the other B&W plants.
-'/ \\
Enos discussed the time an operator h~ aft.er failure of auto-trip to initiate manual scram so as to avoid/fexce~d\\~g service level C.
There are four rod insert possibilities orl receipt\\of auto-trip signal:
\\
/
- l. If all rods insert - no prob~m
\\
- 2.
If only safety rods insert,1E no problem
- 3.
If only regulating rods ipsert - no short-term problem.
A curve sho~ing the case of only a portion o/the regulating rods 'ripping indicates that while system pressure win;::ach about 2800 psi it ~*11 recede to about 2550 psi, in about one mi2~te.
- 4.
If no rods insert, ttie pressure rises as shown on a secbnd graph,
- The calculation. assumeJ/loss of main feed, 95% MTC, PO~Y opens.
The peak pressure reaches/3,450.
Service level c.can be av~ided if scram occurs within about140 seconds.
Auxiliary feed responds\\utomatically within 35 seconds1t"Grning primary pressure downward.
R. Crouse summ~rf~ed the RRG conclusion that there is no im ediate concern regarding saf. operation of the B&W plants because of the Sa em event.
A written r,.
- ort is to be submitted by the B&W RRG by March 1 its basis or that conclusion.and providing the material prese meeting.
Enclosures:
- l.
L~st of Attendees
- 2.
Slides cc w/enclosures:. See next page Morton B. Fairtile, Prdject Manag Operating Reactors Branch #4 Division of licensing
.I'
- e.
3/8/83 In the Management Control presentation the owners a 1 so described how information concerning operating events reaches operating plants. Also B&W said it intends to pro vi de a 1 i st to each utility* of equipment related infonnation previously supplied to that utility, including generic material that B&W determines is needed by each utility. The utilities will check the list for holes in their existing information.
The staff explored with the RRG how management philosophy is imparted to operators; are they concerned about continued power production, or will they trip the plant if safety is in doubt? There were strong RRG management views that the highest priority was to be paid to safety.
The RRG presented equipment classification and implementation information.
The staff expressed concern over operational QA at operating plants and asked RRG if there is a simpler means to solve classification problems other than recJassifying all safety-related components.
The SMUD repre-sentative described that utility's new computer based equ~pment list, saying it had been preempted by their response to the.EQrrulemaking.and would help the classification problem.
Ted Enos described the response of B&W plants to ATWS events like Salem, only.at full power.
He said that all B&W licensees have ATWS related procedures; however, ATOG program wli1l l improve the ATWS procedune*s.
ATOG is being implemented at AN0-1 and later on the other B&W plants *...
~
Enos discussed the time *an operator has after failure of auto-trip to initiate manual scram so as to avoid exceeding service level C.
There are four rod insert possibilities on receipt of auto-trip signal:
- 1. If all rods insert - no problem
- 2.
If only safety rods insert - no problem
- 3. If only regulating rods insert - no short-term problem. A curve sho\\'iing the case of only a portion of the regulating rods tripping indicates that while system pressure will reach about 2800 psi it will recede to about 2550 psi, in about one minute.
- 4. If no rods insert, the pressure rises as shown on a second graph.
The calculation assumes loss of main feed, ~5% MT£, PORV opens~ The peak pressure reaches 3,450.
Service level.'Ci:can be avoided if scram occurs within about 40 seconds. Auxiljary feed responds automatically withi-n 35 seconds. turning pr.imary -pressure downward.
R. Crouse summarized the RRG conclusion that there is 'no immediate concern regarding safe operation of the B&W plants because of the Salem event.
A written report is to be submitted by the B&W RRG by March 11 summarizing its bas is for that conclusion and providing the material presented in th.is meeting.
Morton B. Fairtile, Prdject Manager O~e~a~ i ng Reactors Branch #4 A'ft:eriCiees***..
res*:*****See**ne 4:**page*****......................................................
NRC FORM 318 00-80) NRCM 0240 OFFICIAL RECORD COPY
/
USGPO: 1961-335-960
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MEETING
SUMMARY
DISTRIBUTio'N Licensee:
DPC, GPUN, FPC, SMUD~ AP&L,.TECo
- Copies also sent to. those people on service (cc) list. for subj~ct plant(s).
ORB#4 Rdg
- Project ManageF MFairtile, ADe A'gaziq, Mco*nner,.GVissing, SMiner, JVan Vliet JStol'z GLainas Liceniing As~istant-Ringram OELD JHeltemes, AEOD
- . NRC Meeting
Participants:
JPartlow Wlanning PShemanski GHolahan ERossi HL i WKennedy
. l~Hodges LC rocker HSil ver Sis rae 1 RKendall
- RMattson Task. Force and Oversight Group NRR Division Directors (5)
HDeriton ECase RMinogue RDeYoung VStello Paul F. Collins, l<JMC Inc.
Suite 1111 1747 Pennsylvania Avenue Washington, DC 20006
Enclosure. 1
. B&\\IJ RRG/NRC 11 SALEM 11 EVENT MEETING -
MARCH 8, 1983 - ATTENDANCE LIST
I ----- ---
.e REPORT ON B&W PLANT TRIP BREAKERS MARCH 8, 1983 UTILITIES WITH B&W DESIGNED OPERATING NSS:
- ARKANSAS POWER & LIGHT
- DUKE POWER COMPANY I FLORIDA POWER CORPORATION
~ GPU NUCLEAR CORPORATION I SACRAMENTO MUNICIPAL UTILITY DISTRICT I TOLEDO EDISON COMPANY
I.
\\
~
III.
IV.
- v.
VI.
. VI I I B&W OWNERS GROUP MEETING WITH NRC STAFF SALEM GENERIC IMPLICATIONS MARCH 8~ 1983 AGENDA INTRODUCTION REACTOR TRIP SYSTEM DESIGN BREAKER OPERATING HISTORY, TESTING, & MAINTENANCE ATWS CONSIDERATIONS 1MANAGEMENT CONTROLS CONCLUSIONS QUESTIONS
/
I
. ~
B&W OWNERS RRG ACTIONS RELATED TO THE SALEM TRIP BREAKER EVENT I
COMPARED BREAKER AND TRIP SYSTE~ CONFIGURATION AT B&W PLANTS~.
To* SALEM I
REVIEWED BREAKER OPERATING HISTORY I
REVI~WED BREAKER TESTING. AND MAINTENANCE PRACTICES I
REVIEWED POST-TRIP/PRESTART ASSESSMENT PRACTICES I
- REVIEWEb *APPROPRIATE MANAGEMENT CONTROLS I
REVIEWED ATWS CONSIDERATIONS I
PREPARED
SUMMARY
REPORT
B&W OWNERS RRG FINDINGS
SUMMARY
- 1.
BASED ON EVALUATIONS OF THE IMPLICATIONS OF THE SAL~Mi EVENT, THERE IS NO CONCERN REGARDING CONTINUED SAFE OPERATION OF THE B&W PLANTS.
2, TRIP SYSTEM COMPARISON SHOWS BOTH DIFFERENCES AND SIMILARITIES GE VS. WESTINGHOUSE BREAKERS UV F~ATURE IS SIMILAR FAILURE TO TRIP ALL RODS IS NOT LIKELY DUE TO ELECTRICAL DIVERSITY
- 3.
ALL FAILURE$ TO OPEN HAVE BEEN DURING SURVEILLANCE TESTS.
FROM. 1979 TO 1981 FAILURE DECREASED STEADILY, AN INCREASE IN SURVEILLANCE FAILURES IN 1982 AT two PLANTS REQUIRE FURTHER REVIEW 4,
To DATE, NO SPECIFIC DEFICIENCLES HAVE BEEN IDENTIFIED WITH REGARD TO MANAGEMENT CONTROL OF QUALITY ASSURANCE AND MAINTENANCE PROCEDURES BUT REVIEW IN ~HESE AREAS WILL CONTINUE AND CHANGES WILL BE MADE AS APPROPRIATE.
Reactor Trip System Oconee, TMI, CR-3, AN0-1, Rancho Seco CRDCS 31 AC 311 AC MAIM SECONDARY BUS BUS i J AC L l rl!!_K TRIP
~
1 BREAKERS I
I
. I
- l..;._ - - - - - - - - --GJ DC POWER
-c--* SUPPLIES.,_ __ D_
DC TRIP BREAKERS SAFETY RODS GROUPS 1-4 Hill---,
. QJ
~~
- o UP TD 12 RODS PER GROUP 11.AC VITAL BUS A RPS 11 AC. 16 AC VITAL VITAL IUS I BUS C 1.1 AC VITAL BUS D TRIP MODULES MANUAL TRIP
~ - - ~ ---*i-- - ~
I 1 I I c
~
Ill A Ill CONTROL --1 1
. POWER A
B SCR's REGULATING RODS
- GROUPS 5-8 ' AUX c
D D
~
~ SEC Ir CONTROL POWER UP TO 12 RODS PER GROUP
[c] -~Cornpllter Readout FIGURE 2.1
CRDCS Reactor Trip System DB-1 3' AC 31 AC MAlll SECONDARY POWEi.
POWEi l J.
AC L l
~-.
TRU*
~-1 I c I mumJ D !
kfu-<..
TB-~
L_________ -----0 16 AC VITAL BUS A TRIP MODULES 2/4 RPS 11 AC 11 AC VITAL VITAL BUS I BUS C 0-11 AC VITAL BUS D MANUAL TRIP
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MAii.. cp
. CONTROL POWER A
c
- .********.**~****.*
SEC
,....---+-.....;...;;.~
. CONTROL SCR'.s
- ALL REGULATING ' SAFETY ROD GROUPS 1.2,3,4,5,6,7,8 'AUX UP TO 12 RODS PER GROUP FIGURE 2.2 POWER D
I STORED ENERGY DEVICE UMDUYOL UGE TRIP DEVICE Reactor Trip Breaker TRIP TRIP SHAFT POWEi SHUNT TRIP*
DEVICE CRDCS.
OYER VOLTAGE PROTECTION t--~ CRD BREAKER*. MOUNTED MAM.UAL TRIP DEVICE
....;....----.1LJ...!~--t RE ACTOR 1---*
115 VAC VITAL BUS I
TRIP I
MODULE CONSOLE MOUNTED MANUAL TRIP SWITCH FIGURE 2.3
. OPERATING HISTORY AND TESTING CUMULATIVE FOR ALL OPERATING B&W PLANTS MORE THAN 5000 TESTS CONDUCTED WEEKLY TRIP TESTS OF ONE BREAKER PRE~TARTUP TESTS POST MAINTENANCE TESTS 19 FAILURES TO-OPEN, ALL DURING SURV~ILLANtE TESTING MANY OF THESE WERE "SLUGGISH" OPERATION SEVERAL WERE THE SAME BREAKER DURING A SHORT.
PERIOD O~ TIME (RETESTl~G) 469 DEMAND SCRAM SITUATIONS No FAILURES DURING-DEMAND SCRAM SITUATIONS
. ~ 5 l.n*
~ 4
- II
~.3 a:
2 1.6' 1.6' a:
CD 0
1973 Number of Breaker Failures-to-Open*
5 5
4 0
D 2
1974 1975 1l7i 19'7 7 1,978 1979 1910 19 81 1982 YEARS
1-I i
I I
BREAKER PREVENTIVE MAINTENANCE
- 1.
BREAKERS MAINTAINED IN ACCORDANCE WITH VENDOR
- INSTRUCTIONS
- 2.
VENDOR MAINTENANCE;l~STRUCTION MANUAL
- 3.
AUGMENTED VENDOR MAINTENA~tE INST~UCTION ISSUED BY B&W IN MARCH 1979 ISSUED BY GE IN APRIL 1979 APPROVED PROCEDURES PREPARED BY EACH PLANT FOR USE BY MAINTENANCE PERSONNEL 4,
ESTABLISHED OVERHAUL FREQUENCY ~ VARIES FROM EVERY 6 MONTHS. TO EACH* FUEL CYCLE J
BREAKER OVERHAUL FEATURES I
CHECK PICKUP VOLTAGE ON UNDERVOLTAGE COIL.
I CHECK DROPOUT VOLTAGE. ON UNDERVOLTAGE COIL I
CHECK ALIGNMENT OF RELAY I
CHECK DISTANCE OF PLUNGER TO PADDLE ON SHAFT I
CHECK TORQUE TO TURN SHAFT
. I CHECK CALIBRATION SPRING I
CLEAN OR REPLACE CONTACTS I
CHECK CONTACT SPRING TENSION I
CHECK ARMATURE TRAVEL I
CHECK SHUNT TRIP DEVICE I
CHECK*ELECTRICAL AND MECHANICAL CONNECTIONS FOR TIGHTNESS I
CLEAN AND LUBRICA~E I
BREAK DOWN BREAKER AND INSPECT FOR DAMAGE
.2 0.
177-FA.LOP'\\..' A'n.~ With Partial
. Ro~d Drop -. System Pr.es sure L.O 6Q B 0.
100
. TIME, 5EC Otvf?
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en
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.e Figure 2-41.
205 LOFW ATWS 4000 *r------------------------------~---,
3800.
3600 3400 3200 3000 2800
. 2600 2400 2200 r*
- <I 205 LOFI
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177 LOFI ATIS
.. 2000..__ _
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to
' 20 30 '
40 50.
' 60 70 80 90
. I Time (sec) lht~
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Babcock & Wilcox I
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TAP mrP INITlAL EVALUATfON DETAf LED EVALUATfON
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SUMMARY
RESTART.
T J4 p ~ (R_ #-/JS l ctv' 'f. A5Sri>5 '1B-tv T Pt/...o(F~ft ~
- ~*
~
VENDORS B&W
.. INDUSTRY
- UTILITY
- EVALUATJON
- 1. IMPLEMENT...,.*. *.***.
I EQU lPMENT RELATED* I*NrORMATI'ON *FLOW
e*
EQUIPMENT CLASSIFICATION AND IMPLEMENTATION PROV I DEs*- PRo'PER CLASS I F1I CAT I ON OF SAFETY RELATED EQUIPMENT AND PROCEDURES.
PROVIDES QUALITY ASSURA~CE AND QUALITY CONTROL REVIEWS TO CONF1RM PROPER PROCEDURE IMPLEMENTATION,
.. i.
'I'*"'
[....
CON CL US IONS
- l~ *'THERE IS'NO coNCERN~REGARDING CONTINU~D SAFE OPERATION OF THE B&W PLANTS.
2, DI VERSE METHOD.OF° INTERRUPT I NG POWER TO THE REGULATING AND SAFETY RODS RESULTS IN LESS POTENTIAL FOR FAILURE
.. OF ALL RODS TO SCRAM, 3 'I BREAKERS ARE GE MODEL AK-2A I THE USE OF THE I UNDERVOLTAGE FEATURE IS THE SAME AS T~E w BREAKER.
I i
4,I UV RELAYS AND ABILITY TO OPEN BREAKER 'ARE QUALIFIED AND MEET APPROPRIATE SAFETY GRADE.REQUIREMENTS,
- 5.
NINETEEN FAILURES HAVE OCCURRED DURING OVER 5000 SURVEILLANCE TEST~.
6,. No FAILURES HAVE O'cCURREDDURING 469 ACTUAL DEMAND
- SCRAMS, 7,
BREAKERS ARE MAINTAINED IN GENERAL AGREEMENT WITH VENDOR INSTRUCTIONS, 8, *OPERATOR HAS INFORMATION.AND IS TRAINED TO RESPOND TO SITUATION OF FAILURE OF BR~AKER TO OPEN, 9~
POST-TRIP REVIEWS. ARE CONDUCTEb PRIOR TO STARTUP,
- 10.
No SIGNIFICANt DEFICIENCES UNCOVERED TO DAlE IN MANAGEMENT CONTROL
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