ML20245E631
| ML20245E631 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1987 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| Shared Package | |
| ML20245E603 | List: |
| References | |
| TASK-AE, TASK-T712 AEOD-T712, NUDOCS 8905020068 | |
| Download: ML20245E631 (38) | |
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4 a
4 AE0D TECHNICAL REVIEW REPORT
- UNITS:
Generic TR REPORT N0.:
AE00/T712 DOCKET NO: N/A DATE:
November 1987 LICENSEE:
N/A EVALUATOR / CONTACT:
T. Wolf
SUBJECT:
UNPLANNED CRITICALITY EVENTS AT U.S. POWER REACTORS SIMILAR TO THAT AT OSKARSHAMN UNIT 3 ON 07/30/87
SUMMARY
On 30 July 1987 Oskarshamn Unit 3, a BWR in Sweden, experienced an unplanned criticality during routine shutdown margin testing. Based on searches of the SCSS and INEL AEOS LER databases and the Abnormal Occurrence reports, several events with characteristics of this event were found to have occurred at U.S. commercial nuclear power plants within the past few years. Of these, an event at Peach Bottom 3 on 17 March 1986 appeared to be most like that of Oskarshamn 3 in that an unplanned flux anomaly occurred as a result of multiple operators being inattentive during plant evolutions whfie a principal safety system was inoperable.
Events at Fenni 2, San Onofre 3, Summer and Vogtle I were similar from the standpoint of flux anomalies as a result of multiple operators being inattentive during plant operation. In these cases, however, no principal reactor control safety systems were inoperable.
As a result of this exercise, best search strategies to be used in the SCSS and INEL databases to find similar events were determined.
Using these search strategies yields listings of 100 to 200 LERs whose abstracts would need to be manually screened to determine if the events had the characteristics of prime interest.
DISCUSSISN Bac kground On 30 July 1987 Oskarshamn Unit 3, a BWR in Sweden, experienced an unplanned criticality during routine shutdown margin testing. This event occurred while the fast acting hydraulic scram system was inoperable, which is a violation of normal procedures. At the time of the event, the control room night shift, consisting of an operator, a physicist, and a shift supervisor, decided to do a standard shutdown margin test, in spite of the non-availability of the fast acting hydraulic scram system. This decision was based, apparently, on the fact that the reactor was at low power levels and the slower acting electric control rod insertion system was available.
In addition, the baron injection system was also available.
The test consisted of the controlled withdrawal of six control rods to determine criticality occurrence. Upon only partial withdrawal of the first
- This document supports ongoing AEOD and NRC activities and does not represent the position or requirements of the responsible NRC program office.
8905020068 871119 PDR orc NEXD 1
4 control rod the core went critical much earlier than had been expected.
The night shif t operators did not immediately nctice the unusual neutron flux rise as it occurred on their instrument panels.
The control logic for the sicwer acting electric scram system initiated action at its icwer range on high neutron flux, blocking further contrcl rod withdrawal and reinserting the ccntrol rod.
The team then reset the electric scram system and continued the test on the secuno control rod, which proceeded uneventfully.
They were proceeding with the test on the third control rod when the day shift took over. The aay shift continued the tests, apparently unaware of the inoperability of the fast acting hydraulic scram system. The th)rd rod test was completed uneventfully and the fourth rod test was initiated. During this test, the test team became aware of the hydraulic scram system inoperability and stopped further testing and l
reported to their supervisors.
j i
Essentially, then, this event had the following major characteristics:
l 1.
Inadvertent criticality during low power testing; j
2.
Procedural violation; 3.
Prime safety system inoperable; l
4.
Multiple personnel failures to recognize an abnormal situation.
An investigation based on computerized data sources was conducted of recent 1
U.S. plant operational experience with the intent of determining if similar events have occurred and, if so, to ccmpare the major event characteristics.
Similar U.S. Operatina Experience i
To determine similar U.S. operating experience, queries were made of the l
databases maintained for AEOD at Oak Ridge (Sequence Coding and Search System - SCSS) and at Idaho (Trends and Patterns Program).
In addition, a review was made of the Abnormal Occurrence (A0) reports submitted to Congress to determine if any similar events had occurred and been so classified.
SCSS ORNL was supplied with a copy of the initial press release of the UsTarshamn 3 event and asked to develop a search strategy and a listing of events having similar characteristics.
(Note: the SCSS database contains information on events from 1980 through early 1997.) One limitation supplied to help bound the search was that the reactor power level should be limited to one percent or below. Some 10 different searches were combined based on different search strategies, e.g., flux anomalies or licensed operator actions linked to the control rod drive system.
This resulted in the computer selec-tion of some 400 events, out of the 25,000 events in the database, as possible candidates. ORNL then manually screened the abstracts of these events and developed a list of some 104 events which had some characteristic of the Oskarshamn 3 event. Abstracts of these 104 events were then supplied to AE0D.
AE0D (TWolf) screened these 104 abstracts and selected only six events as representing those most similar to the Oskarshamn 3 event. These six events i
are briefly summarized in the following discussion. Also included is a listing of the major Oskarshamn 3 event characteristics which were noted in each U.S. event.
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S Browns Ferry 2, BWR, LER 260/84-004, 02/22/84: Automatic scram on high 1
flux at less than 1% power due to ccntinuous withdrawal of a high-worth
{
Caused by inadequate administrate controls of fast
]
period control rods, i
McGuire 2, W PWR, LER 370/85-014, 05/17/85: Manual shutdown following inadvertent criticality. Caused by use of incorrect xenon coefficient in control rod criticality position calculations.
j Oyster Creek, BWR, LER 219/86-032, 12/24/86: Automatic scram on high f
flux at about 1% power due to excessive cold feedwater addition to i
reactor. Caused by a combination of operator inattention and control system design.
San Onofre 3, CE PWR, LER 362/86-006, 04/13/86: Automatic scram from core protection calculator low DNBR and high local power density on unplanned criticality. Caused by multiple operational personnel errors.
)
I Wash. Nuclear 2, BWR, LER 397/85-061, 11/17/85: Automatic scram shortly after criticality on high intermediate range monitor setpoint i
during control rod withdrawal. Caused by inadequate start-up j
procedures below the heating range during highly reactive periods of the fuel cycle.
Perry 1, BWR, LER 440/86-086,11/26/86:
Automatic scram at about 2%
power due to upscale intermediate range monitor trip. Caused by excessive feedwater injection into vessel due to operator error and i
procedural deficiencies.
In summary, while all of the U.S. events found in the SCSS search included j
an unexpected flux change, none of the events had essential reactor protection systems out of service.
INEL A copy of the pertinent information on the Oskershamn 3 event was sent to INEL.
INEL personnel under contract to AE0D/TPAB then screened the databases on reactor scrams, safety system unavailability (NOTE:
, and technical specification violations which they maintain for AE0D.
These databases contain event information from 1984 through the present.) The basic criteria used for this screening was elimination by computer search of j
(1) all events at power levels above 1% and (2) from the scram database, those events which had no rod motion indicated and (3) from the technical specification database, those events involving systems not associated with reactor control. The elimination of events at power levels above 1% reduced the data to about 670 events. Eliminating non-rod motion or non-reactor control system events reduced this number to some 200 events. The events were then manually screened using the abstracts and LERs. with the result being a list of 17 events which appeared to have characteristics similar to the Oskarshamn 3 event. These were subsequently transmitted to AE00 (TWolf) for further review.
Of these 17 events,11 were selected as being most similar. These are sumarized in the following discussion.
3
Browns Ferry 2,-BWR, LER 260/84-004, 02/22/84:
See SCSS discussion.
i Brunswick 1, BWR, LER 325/86-011, 04/06/86: Automatic scram at 1%
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power un IRM upscale trip.
Caused by excessive feedwater addition for control rod pcsition.
Diablo Canyon 1 W PWR, LER 275/84-033, 12/06/84: Automatic scram on SRM at less than 1% power.
Caused by operator error in setting of IRM voltage.
Grand Gulf 1, BWR, LER 416/84-040, 09/05/84: ' Automatic scram at less I
than 1% power on IRM upscale trip. Caused by operator slow to adjust IRM range during start-up.
l Hatch 2, BWR, LER 366/85-036, 11/07/85: Automatic scram at power less than 1%. Caused by excessive cold water injection during vessel level J
adjustment.
McGuire 2, W PWR, LER 370/85-014, 05/17/85: See SCSS discussion.
{
l Oconee 2, B&W PWR, LER 270/85-002, 04/21/85: Automatic scram at power indicated at less than 1%.
Caused by power recorder malfunction, actual power exceeded 4%.
River Bend, BWR, LER 458/86-042, 06/14/86: Automatic scram from 1%
j power on high IRM upscale trip.
Cdused by excessive fetdwater addition-during start-up.
San Onofre 3, CE PWR, LER 362/86-006, 04/13/86:
See SCSS discussion.
Vermont Yankee, BWR, LER 271/86-015, 10/04/86: Automatic scram from 1%
power on IRM high flux due to cold feedwater injection.
Caused by inadequate procedures and training.
Vogtle 1, W PWR, LER 424/87-032, 06/06/87: Automatic scram on SRM from less than 1% power following inadvertent criticality. Caused by errors 6nd inattention by multiple operators (reactor operator trainee, reactor operator, shift supervisor).
Wash. Nuclear 2, BWR, LER 397/85-061,11/17/85: See SCSS discussion.
All of these events were associated with unexpected flux levels. Many.
involved operator inattention problems with the Vogtle event most like that of Oskarshamn. However, no explicit procedures appeared to be violated "and no primary safety systems were inoperable at the time of the events.
Abnormal Occurrences A total of five unplanned criticality events have been classified and reported as Abnormal Occurrences since 1976.
In only one of these five events A0 86-8 at Peach Bottom 3, were any reactor protective systems cut of service.
In the Peach Bottom 3 case, the rod worth minimizer was down due to a ccmputer fault.
Brief descriptions of these five events
- follow, l
4 1
3 A0 86-8 On 17. March 1986 Peach Bottom Unit 3, a BWR, experienced an unplanned criticality during start-up while the rod worth minimizer (RWM) was bypassed due to a computer hardware. fault. As allowed by technical spec 1fications, with the.RWM out of service, a second 1
. licensed reactor operator. was assigned to mor.itor the first licensed 3
reactor operator as the latter withorew control rods in accordance with
.I the sequence prescribed by procedure.. Howevar, the operator withdrew control rod 10-23 out of sequence instead of rod 02-23. The second operatar monitoring the rod withdrawal failed to ' notice the error.
Later, as prescribed in the procedure, both operators. signed off the withdrawal of control rod 10-23, Shortly afterward the reactor attained criticality.. Subsequently, withdrawal of additional control rods was t, locked by.the rod sequence. control system (RSCS) since rod 4
02-23 was not in its prescribed position. The Shift Superintendent and.
Shift' Supervisor then bypasseo the-RSCS rod 02-23 full out logic with a keylock switch without verifying the rod position and conformance to the rod withdrawal sequence as required by the procedure for bypassing RSCS logic. Rod withdrawal and start-up continued with rod 02-23 fully
'l inserted instead of being fully. withdrawn as ' required.
This condition continued until the oncoming shift requested that the
~
RWM be returned to service. This was accomplished and the' operators noted an insert error for rod 02-23. The rod was confirmed to be.out-of position for the sequence. The Shift Supervisor returned the RSCS bypass for control rod 02-23 to norual.
Two control rods were inserted and then the reactor was manually scrammed.
r A0 76-15 On 12 November 1976 Millstone 1, a BWR, experienced an unplanned criticality when a reactor control-rod selection error and movement of control rods were made during shutdown margin testing. At the time the unit was shutdown for refueling and the core had been partially loaded with testing in progress to verify the lccal shutdown margin with the strongest rod fully withdrawn.
Due to-personnel errors and inadequata procedures..an adjacent and errcne' usly selected control o
rod was withdrawn to a predetermined position.
The subsequent movement of the designated high worth rod as part of the planned verification of local shutdown margin resulted in an unplanned criticality and a protective reactor trip.
Following the trip, the same rod selection and sequence of movement was again initiated duplicating the original error and requiring the immediate insertion of-the high worth control rod to avert a second event and reactor trip. All automatic safety features were available and provided automatic control rod insertion on the most sensitive scale for measurement thus terminating the result of the operator error. No loss of. fuel integrity occurred or was-expected.
A0 85-1 On 28 February 1985 Summer, a Westinghouse PWR, experienced a-l premature criticality during re.ctor start-up.
Under the direct supervision of the shift supervisor, control rods were being withdrawn by an operator trainee with no previous reactor or simulator experience-at this facility. The shift supervisor, believing the reactor would go critical at about 168 steps of Bank 0, instructed the trainee to 1
1 I
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J withdraw the bank to 100 steps.
This position should have kept the reactor subcritical, even allowing for a 50 step margin of error in the estim6ted critical condition (ECC) calculations.
This assumption was i
based on erroneous ECC calculations.
The ECC was based on middle of life' rod worth curves instead of beginning of life curves. Also the reactcr was critical for approximately three hours just before the event and the reference critical data used in the ECC was based on this data rather than the data for equilibrium conditions from the previous power history. Also. the trainee was not adequately instructed in the need to anticipate criticality any time rods were being withdrawn or to closely monitor the available instrumentation for indication of criticality. Neither did the shift supervisor provide the necessary 1
attentiveness or monitoring himself, i
i As the Bank D rods passed the 40 step position the reactor went l
critical. Unrecognized by the operator and supervisor, rod withdrawal
'l was continued until a high flux positive rate trip occurred at 76 steps. Two other licensed operators were on duty in the control room j
at the time. The operator at the controls was engaged in other l
start-up activities on another part of the control board. The control room supervisor, a licensed senior operator, was at his assigned 1
station but his view of the instrumentation important to this event was blocked by the shift supervisor and the trainee. All systems perfonned as required.
There were no adverse effects as a result of the event.
A0 85-1 (Update) On 1 July 1985 a reactor operator at fermi 2, a BWR, was withdrawing control rods to achieve criticality of his first attempt ever to bring a commercial power reactor critical.
He pulled 11 rods in Group 3 to the fully withdrawn positic.n (position 48),
i rather thar, position 04 as required by the rod pull sheet.
This i
resulted in the reactor prematurely reaching criticality althcugh this was not fully recognized by the licensee until several days later.
At the time, while the lith rod in Group 3 was being pulled, the short period alarm annunciated five times and the pen for the Channel A 1
source range monitor failed to ink for about three minutes. When the pen began inking again the count rate was increasing. At about t5e i
same time, the rod pull error was recognized and the reactor operator began reinserting the 11 rods. The Nuclear Shift Supervisor (NSS) was called and came out of his office to consult with the reactor operator.
The NSS, who was also responsible for directing his first start-up of a commercial power reactor, reviewed the event with the reactor operator and the Shift Technical Advisor in Training. He determined that the reactor had not gone critical and authorized recommencing of rod pulls.
In making this decision he did not consult with the Shift Reactor Engineer, the Shift Operations Advisor or the Shift Technical Advisor.
However, after the recommencement of rod pulling he did contact the Operations Supervisor at his home and briefly discussed the event with him.
The event did not result in any actual safety consequences.
6
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A0 85-1 (Update _l On 13 April 1986 San Onofre 3, a Combustion Engineering P E, experienced an unplanned criticality while the licensee was performing a recovery from a spurious reactor trip which had occurred the previous day. At the time of the event the reactor was being operated by a reactor operator trainee under the direct supervision of a licensed reactor operator with other licensed personnel present in the control room.
With the core protection calculator (CPC) in service, rod pulling w(as being accomplished based on predicted estimated critical position ECP) calculations. These calculations were incorrect based on xenon tables for cycle 1 operation instead of the correct cycle 2 tables. Thus, when Group 4 rods reached approximately 80 inches withdrawn, the reactor achieved criticality.
This was not immediately recognized by the operator and the supervisor and rod withdrawal continued to approximately 114 inches. At this point the reactor tripped based on signals generated by the CPC on indications of departure from nucleated boiling ratio and local power density trip signals. At this time, the critical condition was recognized.
Contributing to the event was the fact that regulating Groups 1, 2, and 3 were withdrawn to their upper grcup limits but they were not individually aligned to uniform height at the fully withdrawn position, as specified by procedure. The operators' intent was to dress the rods after criticality was achieved I
before out-of-sequence penalty factor signals to thE CPC became ef fective.
Data Ccmparison l
The events found during the INEL search which were not initially found in i
the SCSS search were discussed with INEL and SCSS personnel.
It was detemined from reviews of the search strategies used that the SCSS search strategy had eliminated those events involving source rar.ge monitor (SRM) ena intermediate range monitor (IRM) instrumentation and had eliminated these events that occurred at 1% power.
Adding such events to the SCSS 4
search yielded agreement with the INEL search findings.
A comparison of the Abnormal Occurrences with the SCSS and INEL findings shows that only one of the events, the 1986 event at San Onofre 3, was cornmon. A detailed check of the Abnormal Occurrences with the search strategies used on the SCSS and INEL databases found that:
l The Millstone 1 event in 1976 was not in any of the databases due to its being outside the scope of the years of data in the SCSS and INEL databases.
The events at Femi 2, Peach Bottom 3, and Sumer were in the SCSS and i
INEL databases. However, the actual power levels at which the scrams l
occurred were greater than 1%, the cutoff point for the initial searches of SCSS and INEL databases. A modification of the search l
strategies to include any event in the start-up range up to 15% power would have captured these events.
The principle drawback to such a search is that the number of events which must be further screened would increase between 50 and 100 percent.
7
i 4
. Summary Several events with characteristics of the 30 July 1987 Cskarshamn Unit 3 event were founo to have occurred at U.S. commercial nuclear power plants during searches of the SCSS ana INEL databases and the Abnormal Occurrence reports. Of these, an event at Peach Bottom 3 on 17 March 1986 appeared to be most like that of Oskarshamn 3 in that an unplanned flux anomaly occurred as a result of multiple operators being inattentive during plant evolutions while a principal safety system was inoperable.
Of the other events, those I
at Fermi 2 San Onofre 3, Summer and Vogtle I were similar from the standpoint of flux anomalies as a result of multiple operators being inattentive during plant operation.
In these cases, however, no principal reactor control safety systems were inoperable.
As a result of this exercise, the apparent best search strategies to be used in the SCSS and 1NEL databases to find similar events would include:
Limitation of events based on power levels in the start-up range of from 0 to 15%-
In INEL, eliminate those events not involving rod motion; In SCSS, eliminate all events except those involving flux anomalies, reactor trip or control rod drive systems, nuclear instrumentation, and licensed operator actions linked to control rod movement.
The result of these searches would yield listings of 100 to 200 LERs whose abstracts would need to be manually screened to determine if the events had the characteristics of prime interest.
l 8
9 FODH 1
L o s cus * ** * * * * * * * * ; * * * * * * * *, e *ER S CGS DATA *************************r*******08-25-8 DOCKET YEAR LER NUHRER REVISION DCS NUHBER NSIC KVENT DATE e*******
- 88 1987 saa act**********s*ssee.***xa*****a*************ns***sseva**ss**ss/as/ss spWggspspy a
DOCIET:888 OSKARGRAlti UNIT 3 TYPE:BWR REGION:
8 NSSS:#s ARCHITECTURAL ENGINEER: pass FACILITY OPERATOR: ###ssssssssas###ssWWWWWW SYHBOL: s#W CCitSNTS WATCH 975 - SAIGTY RELATED HYDRAULIC SCRAH SYSTEN 007 OF SERVICE W CONTROL ROOH CREW HOVES CONTROL ROD C.AUSING INADVERTENT CRITIC SCHAM.
AB6 TRACT POWER LEVEL - 000s. THE HYDRAULIC SCRAH SYSTEM BAD NOT YET BEEN RES FOLLOWING ROUTINE MAINTENANCE A SECONDARY ELECTRICALLY OPERATED SCRAM SYSTEH (THE ELECTRICAL CONTROL ROD DRIVE SYSTEM) WAS OPERABLE BUT IS NOT CONSIDERED SAFETY RELATED.
DRIVE SYSTEH HAS A FOUR HINUTE RESPONSE TIHE.THE ELECTRICAL CONT THE CONTROL ROOH HIGHT SHIFT, CONSISTING OF AN OPERATOR, PRYSICIST, AND SHIFT SUPERVISOR, DECIDED TO STICK TO THE STANDARD SHUT DOWN HARGIN TEST SCHEDULE SPITI OF THE UNAVAILABILITY OF THE HYDRAULIC SCRAH SYSTEM ( AND IN IN VIOLATION OF PROCEDURES).
N, POWER LEVELS AT-WRICH THE CONTRob ROD WITHDRAWAL TESTS WERE BEI i
PERFORMED. THE SLOWER ACTING ELECTRIC OQHTROL ROD INGERTICH SYST BE RELIED DPON TO ACT AS A SCRAH SYSTEM.
PARTIAL WITHDRAWAL OF THE 'FIRST I
CONTROL ROD CAUSED PREHATURE CRITICALITY AND AN UNEXPEC7ED SCR ELECTRIC DRIVE SYSTEM BANDLED BY BLOCKING WITHDRAWAL AND RE-lH l
CONTDOL RODc THE ELECTRIC CONTROL ROD DRIVE SYSTEM WAS THEN RESET AND THE 881FT TEAM CONTINUED THE TEST SEQUENCE ON SECONP CONTROL ROD.
CONTROL ROOH BHIFT CAME ON DUTY TH THE MORNING. ANP OILE 00NTINUIN A NEW TEST A THIRD CONTROL ROD, REAL11ED WHAT THE HYDRAULIC BCRAM STETIN WAS
'll0T.AVAILABLE'AND STOPPED THE TESTS.
THE INCIDENT DID NOT GIVE RISE '!V ANY FU L DAMAGE OR RADI0 ACTIVE RELEASES. AND THE SEI HAS SINCE UlVE UTILITY PERHISSION TO RESTART THE REACTOR.
e 9
1 l
FODH 1
LER SCSS DATA 08-c o ce s e * * * * ** * * * * * * * * * *
- e s t. * * * * * * * * * * * * * * * * * * * **** ** * * * * * *
- r s e r
- ==
DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC BNENT DATE ses 1987 saa a
s as ccC** * ** *** * * * * * *
- e s s * *m a s * * * * * * * *** p p p p p p s s p #****************ssass*****/ss/as I
DOCKET:#es 06EARSHAW UNIT 3 TYPE:BWR REGION:
NS86: 8s ARCHITECTURAL ENGINEER: 888s FACILITY OPERATOR: ######ssssssssssssWWWWWW 1
SYMBOL: 889 i
WATCH 975 - SAFETY RELATED HYDRAULIC SCRAM SYSTEM OUT OF SERVICE i
COhTROL ROOH CREW HOVES CONTROL BOD CAUSING INADVERTENT CRITIC 8 CRAM.
STEP LE SLK CAUSE PSYS ISYS CCHP VEND QUAN TR CH DI T FD EFF i
0 A
AT AB XXX 1GT E EK 2
1 8
YC 3
1AT I
UA 4
3 X
1AT I
UA 6
3 RC AA CRA i
1 1AT I
AM 8
1 1AT I
PE T
8 A
VR XX I AC I WN 8
1 1AT I
KE 9
2AT I
UA 10 9
2AT I
UA 11 9
RC AA CRA 2
1 2AT I
AM 12 11 B
RC AA FLA A
i 1PT I
PE 13 B
1 2P7 I
KE 14 XX I XX YC 15 YY HH YC WATCH-LIST CODES FOR THIS LER ARE:
978 POSSIBLE SIGNIFICANT EVENT
STEP:
1 SUBLINK:A CADBE:AY
--UNSPECIFIED HAJHT'MANCE PRIMARY SYSTEH:AB
--CONTROL ROD DRIVE COMPONENT:XXX --ENTIRE SYSTEM EFFECT:KK
--INOPERABLE
-....-STEP:
2 THIS STEP IS DIRECTLY LINKED TO STEP 1
SUBLINK:8 l
EFFECT:YC
--ADDITIONAL INFORMATION l
STEP:
3 CAUSE:SA
--INTRINSIC HUMAN ERROR PRIMARY SYSTEH:PT
--TEST / CALIBRATION ACTIVITY COMPONENT:PLO --LICENSED OPERATOR PERSONNEL EFFECT:UA
--COHNISSION OF UNDESIRED TASE. ANALYSIS. OR STEP 10
STEP:
4 THIS STEP IS DIRECTLY LINKED TO STEF
-3 SUBLINE:I CAUSE:SA
--INTRINSIC HUMAN ERROR PRIMARY SYSTEM PT
--TEST / CALIBRATION ACTIVITY COMPONENT:PUI --UNKNOWN UTILITY PERS000EL EFFECT:UA
--CotMISSION OF UNDESIRED - TASK, ANALYSIS.. OR STEP
STEP 5
THIS STEP IS DIRECTLY. LINKED TO STEP 3
CAUSE RC
--RESULTANT CCHPONENT FAULT PRIMARY SYSTEM:AA
--RIACTOR CORE COMPONENT:CRA --CONTROL BOD ASSEMBLY EFFECT:AH
--WRONG POSITION
-STEP 1 6
THIS 8?EP IS DIRECTLY LINKED TO STEP 5
CAUSE:BC
--RESULTANT COMPONENT FAULT PRIMARY SYSTEM:AA --REACTOR CORE COMPONENT:FLA --FUEL ASSEMBLY QUANTITY A' EFFECT:PE
--FLUX AN0 HALY
STEP:
7 THIS STEP IS DIRECTLY LINKED TO STEP 6
SUBLINK:A i
CAUSE:VR
--HUMAN ERROR l
INITIAL UNIT CONDITIONS: I COLD SHUTDOWN UNIT EFFECT: AC AUTCHATIC SCRAM
......sygp; e
THIS STEP IS DIRECTLY LINKED TO STEP 1 AND STEP-7 CAUSE:RC
--RESULTANT CCHPONENT FAULT PRIMARY SYSTEM:AB --00NTROL ROD DRIVE COMPONENT:CRD --CONTROL ROD DRIVE EFFECT:KE --LONG RESPONSE TIME 4
....--gygp; 9
CAUSE:SA
--INTRINSIC BUHAN ERROR PRIMARY SYSTEM:PT
--TEST / CALIBRATION ACTIVITY COMPONENT:PLO --LICENSED OPERATOR PERSONNEL DIFFERENTIATOR:2 i
EFFECT:UA
--CotetISSION OF UNDESIRED TASK, ANALYSIS, OR STEP l
11
4 b
4 l
GTEP:
10 THIS ETEP IS DIRECTLY LINKED TO STEP 9
I SUBLINE:I CAUSE:SA
--INTEINSIC HUMAN ERROM PRD$ARY SYSTEH:PT
--TEST / CALIBRATION ACTIVITY 00MPONENT:PUX --UNKNOWN UTILITY PERSONNEL DIFFERENTIATOR:2 EFFECT:DA
--CorMISSION OF UNDESIRED TASK. ANALYSIS. OR STEP
~ BTEP:.11 THIS STEP ISrEIRECTLY LINKED TO STEP 9
~
CAUSE:RC --RESULTANT CCHPONENT FAULT i
PRIMARY SYBTEM:AA
--BEACTOR 00RE I
.., COMPONENT:CRA --CONTROL BOD ASSEMBLY l
GOANTITY:2.
.DITTERENTIATORJE!i ---
T EITECT AM --WBONG POSITION
~
TEP Y L$4 RED T BTEP 11'
..:CAU$EtRC '--RESULTANT COMPONENT FAULT PRINARY. SYBTEM*AA
-REACTOR CORE 1
00MPQNINTstLA --FUEL ASSEMBLY QUANTITTiA EFFECT:PE
--FLUX ANOMALY
STEP:
13 THIS STEP IS DIRECTLY LINKED TO STEP 2 AND STEP 12 CAUSE':RC
--RESULTANT COMPONENT FAULT PRIMARY SYSTEM: AB
'; CONTROL ROD DRIVE CONPONEltr CRD --00HTROL BOD DRIVE l
QUINTITY 2-DIFFERINTIATOR:2 EFFECT:EE
--LONG RE8PONSE TIME INITIAL UNIT 00NDITIONS: I COLD SHUTDOWN UNIT EFFECT: XX NO SIGNIFICANT EFFECT RFFECT ON ENVIBotWENT: N NO RELEASE EFFECT ON PERSOIDEL: N NO EXPOSURE 12
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1 I'.'*
1 LER SCSS DATA!
'B-24-67 e.,
e........ ***
eeeeeee....e.....
e.eeeeeeeeeeeeeseeeeeeeeeeeeee
- ~!**E~
YEAR LER NUMBER REVIS!DN DCS NUMBER NSIC E.ENT DATE
. :0 1984 004 0
8403220159 189389 02/22/84 e e e ee e e e eee,o n e s e e,eeee eeeeeeee e e ee e e e e ee e ee es s e e s e eeeeee eeee ee eeeee
- 'ET:26) BROWNS FERRY 2 TYPE:BWR j
l REGICN:
~.
NSSS:GE
]
-:TECTURAL ENG NEER: TVAX j
FAC!LITY GPERATOR: TENNESSEE VALLEV AUTHORITV j
SYMBOL: TVA 1
1 DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50. 73(a)(2) (t v): ESF actuations.
1 REFERENCE LERS:
1 260/79-001 2 260/79-011 ABSTRACT POWER LEVEL - 000%. DURING STARTUP DN UNIT 2. THE REACTOR SCRAMMED DUE TO A HIGH FLUX. SPIKE ON CHANNELS 'C" AND "F" 0F THE INTERMEDIATE RANGE MONITORS (IRM) 0F THE NEUTRON MONITORING SYSTEM.
THE SHORT I
FERIOD (( 5 SECONDS) AND SUBSEQUENT REACTOR SCRAM WERE THE RESULT OF j
THE CONTINUOUS WITHDRAWAL OF A HIGH-WORTH CONTROL ROD.
THE IRMS FUNCTION TO PREVENT FUEL DAMAGE THAT CAN RESULT FROM ABNORMAL TRANSIENTS THAT OCCUR WHILE OPERATING IN THE INTERMEDIATE POWER RAN6E.
1 ADMINISTRATIVE CONTROLS TO PREVENT RECURRENCE OF FAST PERIODS WERE 1hADEQUATE TO ENSURE THAT HIGH-WORTH RODS WERE PROPERLY IDENTIFIED FR;0R TO USE.
PLANT WRITTEN PROCEDURES AND THE NUCLEAR EN61NEER:
- .ALIFICATION PRCSFAM WILL BOTH BE REVISED TO INCLUDE FC
- 94L EMPHAS:S
- '. THE NUCLEAR ENSINEER'S DUTIES AND PROPER CONTROL ROD IE20ENCINS-l 7:~M RESFECT TO 8!GH-wCETH FGDS, e
14
P; " '
1 LER SCSS DATA 08-24-97
... e e e e e e...... e e e e e e e e e e e e e n e e e e e e e e s e i n ge e e e e, e e e e e e e, e e e e e e e e e e e
!'O 1985 014 0
8507120326 195243
-05/17/85
- e e + * *,# # e e e e e e e n es e e ee e e ee e e e e e e s e e e e e e e e ee e e e e e e ee ee e eeeee e e e e eee
- 'ET:370 MCGUIRE 2 TYPE:PWR REGION:
2 NSSS: WE
- -lTECTURAL ENGINEER
- DUKE FACILITY OPERATOR: DU6E POWER CO.
SYMBOL: DPC C 3".M E N T S ETEP 8: MANUAL SHUTDOWN RESULTED FROM REACTOR AT CRITICALITY WITH CONTROL RCDS BELOW INSERTION LIMITS.
WATCH-LIST CODES FOR THIS LER ARE:
233 TRANSIENT PROCEEDS IN A SIGNIFICANT. UNEXPECTED REDORTABILITY CODES FOR THIS LER ARE:
10 10 CFR 50.73(a)(2)(i): Shutdowns or technical specification violations.
ABSTRACT FCWER LEVEL - 000%. CN MAY 17, 1985 AT.0609, THE REACTOR WAS MADE CRITICAL WITH BANK 'C' CONTROL RODS BELOW THE MINIMUM TECH SPEC INSERT!DN LIMITS.
OPERATORS REINSERTED ALL CONTROL-RODS IN ACCORDANCE WITH OPERATING PROCEDURES. SHUTTING DOWN THE REACTOR.
MAIIMUM POWER.
LEVEL WAS ABOUT 10(-8) AMPS ON THE INTERMEDI ATE RANGE NEU?RON DETECTOR m*,
CONTROL ROD BANF "C"
WITHDRAWN 26 STEPS.
THIS EVEN? IS
- .'5S1FIED AS A FROCEDURAL DEFICIENCY BECAUSE THE COEFFIC;INTS IN THE si',;N FREDICT COMPUTER PROGRAMS WERE INC0PRECT.
THIS F RCi: AM WAS USED
- - - ;PERATORS TO PREDICT THE CONTROL ROD POSITION AT WHICH THE REACTOR
=~. D GD CRITICAL.
l l
15
4 8::"
1 LER SCSS CATA 08-:4-17 e e e e.ee e e eee e e eee e e s e se eee e e e e e eee e e e e e e e e e e e s s e +ee e e e e e e ee eeeee e e e.
~:tET YEAR LER NUMBER REVIS!CN DCS. NUMBER NS]C EVENT-LA*E
- "?
1986 032 0
8702020:03 202512
~12/24/86 e ee e eeeeeeeeeeeeeeeeeeeene seeeeeeeeeese e eeeeeeeeeeee**eseseeeeeeeee
1 NSSS:6E
- -ITECTURAL ENGINEER: ENRC FACILITY OFERATOR: JERSEY CENTRAL F0WER & LIGHT CO.
SYMBOL: JCF COMMENTS STEP 2: COMP HTR - REHEATER.
REFORMABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(iv): ESF actuations.
REFERENCE LERS:
1 219/85-006 2 219/05-012 3 219/85-022-ABSTRACT PCWER LEVEL - 001%. ON DECEMBER 24, 1986 AT 0249 HOURS A REACTOR TRIP DCCURRED DUE TO A HIGH NEUTRON FLUX CONDITION CAUSED BY COLD FEEDWATER ADDITION TO THE REACTOR VESSEL.
AT THE T]ME OF THE EVENT REACTOR POWER WAS LESS THAN 1 MEGAWATT THERMAL ON THE INTERMEDI ATE RANGE NEUTRON' MONITORS (EIIS-IG) AND BEING REDUCED TO MAKE REPAIRS TO A STEAM LEAK ON THE TURBINE FIRST STAGE REHEATER (CF1-RHTR).
THE CAUSE OF THE EVENT WAS A COMBINATION OF OPERATOR ERRORS IN THAT PERSONNEL
- NOT MAINTAIN AN ACTIVE AWARENESS OF PLANT SYSTEM ST A?ilS AND
- '. FIGURATION WITH RESPECT TO THE FEEDWATER CONTROL SYSTEM (E!!S-JE),
- R;ECTIVE ACTION CONSISTED OF REVIEWINE THE EVENT WITH EACh GNC;9!NS
' S-;FT FRICR TO ThEIR ASSUMING STARTUP ACTIVITIES.
A MCFE DETA'JLED
- E. !EW CF THE EVENT STRESSING THE NEED TO PAINT AIN AN ACTIVE AWAF.ENE55
- FLANT STATUS AND CONDITIONS WILL EE INCCEPORATED IN C:ERATOR 7:alNING. ONCE THE FINAL TRANSIENT ASSESSMENT. REPORT IS COMFLETE.
S: MILAR OCCURRENCES WERE REPORTED IN LICENSEE EVENT REPORTS 85-06'.
5 5 -12, AND 65-22.
e 16
+
- =
1' LER'SCSS DA1A 08-24-S?'
e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e e s e e e s e e n s e 6,e * *
8605200395 199318
.04/13/86 e e e,e e *
- e n e e e e e e e e s s e e e e eee ee ne n n e ee e e e ee e e e ee e e * *
- e e e e e s s e se e *
- e s ee
- <ET
- 362 SAN ;NOFRE 3 TYPE
- PWR REGION
5 NSSS:CE
"::-ITECTURAL' ENGINEER: BECH FACILITY OFERATOR: SOUTHERN CALIFORNIA EDISON CO.
SYMBOL: SCE COMMENTS STEP 5: CAUSE L1-EXCESSIVELY CONSERVATIVELY PENALTY FACTORS.
REPORTABfLITY CODES FOR THIS LER ARE:
13 10 CFR 50. 73(a) (2) (i v): ESF actuations.
ABSTRACT POWE0 LEVEL - 000%. ON 4/13/86 AT !!!6. WITH UNIT 3 IN MODE 2 AND A RE AChDR STARTUP IN PROGRESS THE REACTOR TRIPPED WHEN A CORE PROTECTION CALCI'L ATOR (CPC) LOW DEPARTURE FROM NUCLEATE BOILING RATIO (DNBR) AND H1 LOCAL POWER DENSITY (LPD) TRIP WAS GENERATED.
MINOR CONTROL' ELEMENY ASSEMBLY (CEA) MISALIGNMENTS NEAR THE ALL RODS OUT-(ARO)-
POSIT 10N DURING THE REACTOR STARTUP RESULTED IN THE TRIP. AT NO TIME DID AN DNBR/LPD CONDITION EXIST-IN THE CORE. AS A RESULT OF AN INACCURATE ESTIMATED CRITICAL POSITION (ECP) THE REACTOR ACHIEVED CRITICA !TY AS CONTROL ELEMENT ASSEMBLY (CEA) REGULATING 6ROUP 4 WAS BEING WITHDRAWN RATHER THAN DURING THE WITHDRAWAL OF REGULATING GROUP UF0h RECDGN!!!NG THAT T AE REACTCR HAD BECOME CRIi!CA. WITH e.
- E;-ULATIND GROL:5 5 AND 6 f:0T YET WITHERAWN, THE OPE,ATO:5 T00k PRCMFT
- FECTIbl A071;N TO MITICATE THE EVENI BY CEA INSE TICN.
DURING THE EE;:0RMANCE OF TWE CORRECTIVE ACTION THE CFC DNBR/LPC TR::
OCCURRED.
!*. ADDITICN 10 AN INACCURATE ELP, AND A LACK OF SUFFICIEN~ DILIGENCE i+ SOME OPERATORS, THE IMPOSIT!DN OF OVERLY CONSERVATIVE :ENALTY F4; TORS 1h CPC SOFTWARE FOR CEA MISALIGNMENTS DURING REACTOR STARTUPS haS DETERMINED TO BE THE ROOT CAUSE OF EVENT.
CORRECTIVE ACTION TO F; EVENT RECURRENCE INCLUDES:
AFFROFRIATE DISCIPLINARY ACTION,-
PROCEDURE MODIFICATION, ISSUANCE OF A SPECIAL ORDER, ENHANCEMENTS TO THE OPERATOR TRAINING PROGRAM, AND A MODIFICATION TO STARTUP INSTRUMENTATION.
e s
I l
{
17
l j
Ff: M
-1 lLER SCSS DATA 08-24 '
- .g 3
- L C D. E T YEAR LER NUMBER. REVISION DCS NUMBER NS]C EsEhi IA'E
]
397 1985.
061-1 8604150207 198633 11/17/85 j
1 l
C;; PET:397 WPPSS 2 TYPE:EWR REGlCN:
5 NSSS:GE A::s! TECT'JRAL ENGINEER BNR0 FACILITY OPERATOR: WASHINGTON PUBLIC F0WER SUFFLY SY37EM SYMBOL: WPP DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a) (2) (iv): ESF actuations.
ABSTRACT l
POWER LEVEL - 000%. ON NOVEMBER 17 1985, DURING A COLD REACTOR STARTUP, A REACTOR-SCRAM OCCURRED.
THE SCRAM WAS CAUSED BY EXCEEDING THE HIGH UPSCALE INTERMEDIATE RANGE MONITOR (IRM) SETPOINT DUR]N6 CONTROL ROD WITHDRAWAL WITH THE REACTOR CRITICAL PRIOR TO REACHING ~THE HEATING RANGE.
THE CAUSE WAS ATTRIBUTED TO PERSONNEL ERROR AS A RESULT OF INADEQUATE PROCEDURES IN THAT THE REACTOR COLD STARTUP PROCEDURE DID NOT ADEQUATELY ADDRESS THE CONTROL OF A REACTOR STARTUP j
BELOW THE HEATlWG RANGE DURING HIGHLY REACTIVE PERIODS OF THE FUEL l
CYCLE.
TWO SEPARATE COMPUTER ANALYSES WERE PERFORMED TO VERIFY REACTIVITY WORTHS OF CONTROL RODS. PRIOR TO THE NEXT STARTUP, THE CONTROL ROD WlTHDRAWAL SHEET WAS ALTERED TO MINIMIZE THE EFFECT.0F THE lhDIVIDUAL ROD NOTCHES INVOLVED IN THIS EVENT AND OPERATORS WERE CAUTIONED TO PAY CLOSE ATTENTION TO THE FIRST FEW 00DS OF ANY ROD w;RTH MINIM!?ER GROUP.
THE REACTOR COLD STARTUP AND CONTROL ROD SEQUENCE DEVELDFMENT FROCEDURES WILL BE PEVISED TO PROVIDE ADDITIONAL
,G.!CANCE AND REQUIREMENTS.
i l
l 18 I
(
LER SCSS DATA SS-24-67 l
.......eee...e,e,e.e......e.....e....e..ee.........ee.....e...ee.e.
q l
U:kET YEAR LER NUMBER ' REVISION DCS NUMBER NSIC
!.ENT DATE
)
440 1986' 086' O
8612230377 202181 11/26/86 I
e e.e..,e e e n.e e e e t,e,e.e e e e n e e e e..e.e +.e s e e e e e.e e s e e e e n e e.e e.e * *
- e e s e i
- ) ET: 440 PERRY 1 TYFE:EWR REGION:
NSSS:SE
- -lTECTURAL ENGINEER: GLET FACILITY OPERATOR: CLEVELAND ELECTRIC ILLUMINATING CC.
SYMBOL: CEI COMMENTS i
j STEP 4: COMP XA - FEEDWATER FLOW CONTROLLER FAIL ALARM. STEP 5: EFF KX -
RAPID CYCLING OF VALVE OPEN AND CLOSED. STEP 20: CAUSE XI - SUSPECTED 3
DAMAGE FROM PREVIOUS TESTS USINS MANUAL VALVE OPERATOR.
]
DEPORTABILITY CODES FOR THIS LER ARE:
l
-l 13 10 CFR 50.73(a)(2)(iv): ESF actuations.
1
' ABSTRACT POWER LEVEL - 000%. ON NOVEMBER 26 AT 0745, A REACTOR SCRAN OCCURRED DUE TO AN UPSCALE TRIP DN THE INTERMEDIATE RANGE NEUTROW MONITORS (IRM).
THE SCRAM WAS A RESULT OF EXCESSIVE FEEDWATER INJECTED INTO THE REACTOR VESSEL WHILE PLANT OPERATORS WERE ADJUSTING THE MANUAL FLOW CONTROLLER OF THE REACTOR FEED PUMP TURBINE (RFPT) A IN.
PREPARATION TO TRANSFER FEEDWATER SUPPLY OVER FROM THE MOTOR-DRIVEN FEED PUMP. THE CAUSES OF THE EVENT WERE PERSONNEL ERROR AND AN l
INSTRUCTION DEFICIENCY.
THE OPERATOR AT THE CONTROLS FAILED TO FE;CGN!!E THE MAGNITUDE OF INCREASED FEEDWATER FLOW AND THE INCREASE
- NTROLLER.
A CONTR!PUTING lAUSE TO'iHE ERROR WAS A LA:$ :: NECESE'S/
8;E~ AUTIONS IN THE CEEDWATER CCNT;0L SYSTEM OPERAi!NG !s57:UCTICN.
- !E INSTRUCTION FAILED TO CAUTION THE OPERATOR TO ENSU;E 'UFFICIEN' ETAi!LIZATION TIME WAS FROV!DED WHILE FERFORMING CONTRCLLE:
ADJUSTMENTS.
TO PREVENT RECURRENCE RETRAINING WILL BE CChDUCTED FOR A.L ON-SHIFT LICENSED OPERATORS FEGARDING THE OPERATION OF THE TEECWATER CONTROL SYSTEM AND THE SEQUENCE OF EVENTS INVOLVED IN THIS
~
REACTOR SCRAM.
IN ADDITION, THE FEEDWATER CONTROL SYSTEM OPERATING INSTRUCTION HAS BEEN REVISED TO CAUTIUN THE OPERATOR TO ENSURE SUFFICIENT STABILIZATION TIME IS PROVIDED WHILE PERFORMING CONTROLLER ADJUSTMENTS.
e l
19
~C:'
l-LER SCSS DATA 05-04-27
!!5 1986 011 0
8600000000 201476 04/06/06 l
)
- ET 325 BPUNSWlCV !
T YPE : P W5:
REG!CN!
- -;TECTURAL ENGINEER: UECI FACILITY OPERATOR: CAROLINA POWER h LIGHT CD.
SYMBOL: CFL CGMMENTS STEP 8: CAUSE SX - OPERATOR DID NOT HAVE TIME TO ADJUST IRMS.
DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(1v): ESF actuations.
)
J ABSTRACT PCWER LEVEL - 001%. ON 4/6/86. AT 1832, THE UNIT 1 REACTOR AUTOMATICALLY SCRAMMED DUE TO AN UPSCALE TRIP OF INTERMEDI ATE RANGE MCNITORS (IRMS) A AND H.
REACTOR POWER WAS GREATER THAN 1% IN THE STARTUP MODE WITH REACTOR PRESSURE - AT 111 PSIG AND REACTOR LEVEL AT-194 INCHES.
CONTROL ROD MANIPULATIONS TO CONTROL REACTOR C00LDOWN l
RATE WERE IN PROGRESS DURING A REACTOR POWER REDUCTION IN PREPARATION l
FOR ANTICIPATED REPA!RS TO THE HYDR 06EN SEALS OF THE UNIT MAIN
)
GENERATOR.
A SCRAN RECOVERY WAS CONDUCTED AND PLANT PARAMETERS WERE i
MAINTAINED.
THE UPSCALE TRIP OF IRMS A AND H IS ATTRIBUTED TO A NET
~
POSITIVE REACTIVITY INSERTION CAUSING THE MONITORS' TRIP SETPOINTS TO EE REACHED.
THIS OCCURRED WHEN WITHDRAWAL NOTCHING OF CChiROL A00
- 4-23 BECAME ADDITIVE TO THE DECREASING REACTOR MODERATOR TEMPERATURE-4 AN: AN APPARENT PROBLEM WIik THE STARTUP LEVEL CONTROL VA 5E WHICH C'. SED LEVEL TO INCREASE FROM 187 INCHES TO 194 INCHEE.
~HE OPERA'OR a;5 INJEST] GATING THIS LEVEL STEP' CHANGE WHEN THE SCRAM C::URRED.
THE
T-;S EVENT WILL BE COVERED BY PLANT OPERATIONS PERSONNEL IN REAL-TIME i*'INING.
I
?
I I
20 i
__-____-___-___.__A
1 r;:".'
1 LER 5055 DATA 08-04-l' e e e e e s ee n ** ne e ve e e,*e e e e e e,e e e e e e e e e e e e,e e e s e e e e e e e e e e e e s e s e s e * *
- e ee i
D0;a ET YEAR LER NUMBER FEVISIDh DCS NUMBER NSIC EVENT D *E I
275' 1984 033-0 8501110411 192541 12/06/54 eeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeeneeeenseeeeeeeeeeeeeeeeeeeeeee 00;iET:275 D!ABLO CANYON 1 TYPE:PWR REGION:
5 NSSS: WE
' A;;-;TECTURAL ENGINEER FGEC FACILITY OPERATOR: PACIFIC GAS & ELEC7A'C CD.
SYMBOL: PGE i
DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50. 73(ai(2) (t v): ESF actuations.
I ABSTRACT
(
POWER LEVEL - 000%. WHILE IN MODE 2 (STARTUP). THE REACTOR TRIPPED WHEN THE SOURCE RANGE HIGH FLUX TRIP SETPOINT WAS EXCEEDED BEFORE THE j
INTERMEDIATE RANGE ' PERMISSIVE (P-6) ALLOWED BLOCKAGE OF THE TRIP.
ALL J
AIF MATIC EQUIPMENT RESPONDED'AS DESIGNED.
THE P-6 SIGNAL WAS NOT lt'4 TI ATED BECAUSE THE COMPENSATING VOLTAGE FOR THE INTERMEDI ATE RANGE i
DETECTORS HAD NOT YET BEEN ADJUSTED TO CORRECT FOR THE CURRENT CORE BURNUP.
ALL SYSTEMS AND EQUIPMENT AFFECTED BY THE EVENT WERE RETURNED TO NORMAL OPERATION.
THE COMPENSATING VOLTAGE FOR THE INTERMEDIATE RANSE DETECTORS HAS BEEN RESET.
IN ADDITION, THE PERSONNEL INVOLVED WILL REVIEW THIS EVENT.
e i
1 1
21
r
- 1 LER SCSSLDATA-08-24-67 e....e *
- e e e e e
..ev e e e e e e e e e s e e s e e e.e * *
...e s e e e s e e e e e e e,s e e,e e s,e e e e
- E*
YEAR LER NUMBER REVISION C;S NUMBER NSIC EVENT DATE
&;e 1984 040
'O S410160113 191668 09/05/84
- e e s ee** **+e ****e eeee ee eeene ne**e see n s e e es e n s e ee se eneeneeess ee n es ee C::'E*: 416 GRAND GULF 1 TfPE BWR REGION:
2 NSSStGE A::-:*ECTURAL ENGINEER: BECH ACILITY OPERATOR: MISSISSIFFi F0WER & L13HT CO.
SYMBOL: MPL REFORMABILITY CODES FOR THIS LER ARE:
i 13 10 CFR 50.73(a)(2)(a v): ESF actuations.
A5STRACT F0WER LEVEL - 000%. ON 9-5-84 DURING RE ACTOR ST ARTUP, A SCRAM 0: CURRED WHILE CONTROL RODS WERE BEING WITHDRAWN FROM.THE REACTOR.
THE SCRAM WAS A RESULT OF THE 1RM'S'NOT BEING RANGED UP AS THEY RESPONDED TO THE POWER LEVEL INCREASE. FOLLOWING THIS EVENT. DURING i
PLANT SHUTDOWN. THE PPESSURE REFERENCE SETPOINT WAS LOWERED TO A POINT WmlCH CAUSED THE TURBINE BYPASS CONTROL VALVES TO FULLY OPEN.
THE REACTOR WATER LEVEL DECREASED TO THE SCRAM SETPOINT.
IN ORDER TD j
PnEVENT THE OCCURRENCE OF SIMILAR EVENTS, THE PROCEDURE REQUIREMENTS
]
HAVE BEEN REVISED TO ALLOW FOR HIGH SPEED OPERATION 0F AT LEAST ONE j
IRM RECORDER.
ALSO, AN OPERATOR HAS BEEN DESIGNATED TO MONITOR l
IMPORTANT PARAMETERS SUCH AS REACTOR POWER, PRESSURE AND LEVEL.
R I
)
\\
J l
'.l l
1
.i
- I 1
22
-___-____--_m___________m_
~5 :
- J-LER SCSS LATA 09-24-57 e v e e e ee,e e e.e e s s e e e s e e e e ee e e es e,*e es ee e e e e s e e e e s e s e e e n e e ee e ee eeeeee s
- *:a IT YEAR -LER NUMBER-REVISION-DCS NUMBER-NSIC EVENT DATE 1985 036 0
8512160040- 196969 11/07/85
- + e * *
- e se e e e e e e e e e e e e e e e e e e n e s e s e e n e s;* e s e n s e e ee ee e e e e e e ee ee e e ee ee e
- tET: 266 HATCH 2 TYPE:9WR REGION:
2 NSSS:GE A::-:iECTURAL ENGINEER: BESS 2
FACILITY OPERATOR: GEORGIA POWER CD.
SYMBOLt GPC DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(iv): ESF actuations.
ABSTRACT J
POWER LEVEL - 000%. AT 1450 CST ON 11-7-85, WITH THE UNIT IN THE STARTUP MODE, A FULL REACTOR SCRAM OCCURRED.
FOLLOWING AN-1 INVESTIGATION, IT WAS DETERMINED THAT A HIG'iER.THAN NORMAL AMOUNT OF STEAM HAD BEEN USED TO EQUALI2E PRESSUR6 AROUND THE CLOSED 1. 80ARD MSIV'S.
THIS CAUSED THE REACTOR VESSEL' WATER LEVEL TO, LOWER TO 2B INCHES ABOVE INSTRUMENT ZERO. CONSEQUENTLY, DURING THE PROCESS OF RESTORING VESSEL WATER LEVEL, AN INCREASE IN REACTOR POWER RESULTED, CAUSING IRM 2C51-K601D AND 2C51-K601E TO SPIKE TO THEIR HI-HI SETPOINT WHICH ACTUATED THE RPS. THIS EVENT IS THE RESULT OF LICENSED PLANT PERSONNEL ERROR IN THAT THE RESPONSE TO THE INCREASE IN REACTOR POWER WAS NOT DETECTED IN TIME TO PREVENT
- HE ' ACTUATION OF THE RPS WHICH u
SCRAMMED THE REACTOR, e
1 l
l 23
~. * ~ "
LER SCSS CATA 08-24-67
+e
++,,eee,seeeeeeeee.eseeeeesesee eeeeeeeeeeeeeeeee,eeeese,teee+ee
~70 1985 002 0
8506040508 195202 04/21/85 e eeeeseenaeene,seeeeeeeeeene*,eeeseseseseeee,seeeeeeeneteveesseeeee
- ET 270 DCCNEE 2 TYPE:PWR REGION:
2 NSSS:BW 1:*-!TECTUFAL ENGINEER: DtBE FACILliY CPERATOR: Duke POWER CD.
SYMBOL: DPC CCP. MEN T S ETEP 1: COMP MSC - ret:RDER PEN: STEP 2: MODEL NL.
RES DRTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(iv): ESF actuations.
ABSTRACT POWER LEVEL - 000%. ON 4-21-85 AT 0051 HRS. OCONEE 2 TRIPPED DURING ZERO POWER PHYSICS TES11NG (ZPPT). WHEN RPS CHANNELS A AND B SENSED A FLUX LEVEL IN EXCESS OF THE SETPOINT.
A MALFUNCTION IN THE POWER RAWBE RECORDER, WHICH CAUSED A LESS-THAN-ACTUAL INDICATION OF FLUX, LED THE OPERATORS TO EXCEED THE SETPOINT, CAUSING THE TRIP.
A SYSTEM TRANSIENT DlD NOT OCCUR BECAUSE OF THE LOW POWER LEVEL.
THE REACTOR WAS RESTARTED IN ORDER TO CONTINUE 2 PPT.
PRIOR TO PROCEEDINS WITH TESTING, CHECKS WERE MADE ON THE FLUX TRIP SETPOINTS. AND THE
!N'ERMEDIATE AND POWER RANGE RECORDERS: ALL WERE FOUND TO BE OPERATING FFCPERLY.
2 PPT WAS RESUMED AT 0202 HRS.
24
: =
1 LE; SCSS' DATA 08-24-B' e e, e e e s e e *,s e s ee e e e ee ee e e e e e e e e e e,e e e e e e e e e e e e e e e e e e e e e e e e e eee e e ee e,
'::: RET YEAR LER NUMBER AEVISIONE DCS NUMBER NSIC EVENT DATE 455 1986 042 0
8607230237 200288 06/14/86 e.e,+ e e e ee e e e e e e ene e e e e see n ee e e e s e e e s e e e e *
- e s s e e s ee s e e s e e ee ee e n s ee ne 4
- 'ET 458 RIVERBEND 1-TYPE:BWR REGICN A
NS35:GE
- - TECTURAL ENGINEER: EWXI FACIL]i1 0FERATCR:' GULF STATES UTILITIES SfMBOL: GSU COnnENTS STEP 7: CAUSE SD-DESIGN OF FEEDWATER CONTROL SYSTEM MAKES RESPONSE TO j
TR ANS!ENTS DIFFICULT DURING STARTUP.
REFCRT ABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(iv): ESF actuations.
REFERENCE LERS:
1 458/86-041 i
ABSTRACT P0hER LEVEL - 001%. AT 2326 DN 6-14-06 DURING RESTART FROM SCRAM 86-17 (TURBINE HIGH VIBRATION. REF. LER 86-041) THE UNIT TRIPPED ON INTERMEDIATE RAN6E MONITOR (IRM) UPSCALE FROM APPROXIMATELY 1% POWER.
WITH THE RE ACTOR AT APPROXIMATELY 750 PSIG THE UNIT. 0PERATOR (UO)
BESAN ALIGNING STEAM LINC DRAINS PER THE STARTUP PROCEDURE IN USE AT j
TwE TIME.
AFTER OPENING TWO 3 INCH DRAINS, PRESSURE BEGAN TO REDUCE
- IDLY.
THE PRESSURE REDUCTION RESULTED IN A REACTCR HATER LEVEL ihE.L AND POWER OE:RE ASE DUE TO INCREASED VD!DS. THE !R"i WERE DOWN f '.3ED BY THE ATC OPERATOR TO MAINTAIN ONSCALE READINS3.
THE SWELL 4.i'J CAUSED CLOSURE OF ThE ST ARTUP AEGUL ATING VALVE. WH "H ALONG h!TH
- a:J RETURN O!VER$1CN. RESULTED IN WATER LEVEL REDUCT10h.
AS LEVEL E:REASED THE STARTUP REGULATING VALVE BEGAN TO OPEN.
W: TH THE S'a~ TUP REGULATING VALVE NOW FULL) OPEN. AND THE IRMS DCas R ANGED DUE T: THE REDUCED POWER LEVELS. THE SIGNIFICANT MASS FLOW Cr COLD TEE ~ WATER TO THE VESSEL RESULTED IN A FLUX INCREASE TO Tni IRM UPSCALE SETPOINT WHICH CAUSED A REACTOR SCRAM.
THE GENERAL OPERATING PROCEDURE HAS BEEN REVISED TO INCLUDE APPROPRI ATE CAUTIONS IN REGARD TO OPENING STEAM DRAINS.
SPECIFIC GUIDANCE AND TRAINING WAS 61VEN TO ALL CREWS AS TO THE CAUSES OF THE EVENT.
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-YEAR LER NUMBER BEVISION DCS NUMBER NSIC EVENT DATE 27' 1986 015 0
8611120141 20158 9 ~
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- ".: ET:271 VERMONT YANKEE TYPE:BWR REGION:
1 NSSS:SE
- - TECTURAL ENGINEER: EPAS ACILITY OPERATOR: VERMONT Y ANL'EE NUCLEAR POWER CORF.
SYMBOL: VYC DEPORTABILITY CODES FOR THIS LER ARE:
13 10 CFR 50.73(a)(2)(s v): ESF actuations.
ABSTRACT F0WER LEVEL - 001%. AT APPROXIMATELY 1733 HOURS ON 10-4-96 WITH THE REACTOR CRITICAL, THE POWER LEVEL AT INTERMEDI ATE RANGE MONITOR (IRM)
RANGE 1, AND THE MAIN STEAM ISOLATION VALVES (MSIV) CLOSED TO ALLOW REPAIR OF EXCESSIVE CONDENSER AIR IN-LEAKAGE AN AUTOMATIC REACTOR SCRAM OCCURRED.
THE SCRAM WAS INITIATED BY THE IRM'S AS A RESULT OF h!6H. NEUTRON FLUX.
THE HIGH PRESSURE COOLANT INJECTION (HPCI) SYSTEM HAD BEEN STARTED TO LOWER VESSEL PRESSURE.AND WHEN IT WAS SECURED THERE WAS AN 8-!NCH VESSEL WATER LEVEL SHRINK.
THE LEVEL SHRINK-CAUSED THE AUTOMATIC INJECTION OF FEEDWATER.
THE COOLER FEEDWATER CAUSED A FLUX SPIKE WHICH RESULTED IN THE IRM HIGH FLUX SCRAM.
THE CAUSES OF THE EVENT ARE ATTRIBUTED TO PROCEDURAL / TRAINING DEFICIENCIES.
CORRECTIVE ACTIONS ARE TO CONSIST OF REVIEWING C:ERATING PROCEDURES AND THE ASSOCIATED TRAINING WITH CONSIDERATION
- -:,EN TO PROVIDING MORE DETAILS FOR THIS MODE OF DPEFATION.
FUTURE
.iE OF THIS MODE OF OPERATION WILL CONSIDER THE LESSONE LEARNED FROM j
' -:1 EVENT.
THE CONDENSER REF AIR WA5 COMPLETED AND FE A:~CR STAR'.F
""E NCED AT 195 4 HOURS ON 10-5-86.
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2 Cn Parch 18 1906, at 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br />, Unit 1 wa s n. ino.il l y sc ra mrteil from approximately 31 power.
Unit 3 was i n s t u t ap ma.le a t the time of the event.
The event wa s causerl by
<a seri os of pe rsonne l errors involving failure to with.1 raw the correct cont rol r o,1.
failure to aloquately veri fy that the correct w i t hil r a wa l sequence was followed, a nil failure to verify the proper ponition of a control ro<l prior to utilizing the l< oil Se.psence Con t rol Syntem bypass function for that rod.
The re.ictor was m.inually scrammel lue to concerns that the oot-of-sequence ro.).in I ro so t t in.) rmi pattern may not have been bounded by the current relna.1 specirle Rod Drop Accialent (RDA) safety analysis.
S u bs eiluen t analyson showed that the Unit 3 reactor wan not ope r i t eil i n ei con.lition such that RDA conscriuences wou t.1 have exce..de.1 the donign criteria at any timo during this event, and It!)A connorloenetus w..ro.
significantly loss severe than the reload specific Itt)A analysis results.
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- Rod Drop Accident PORC - Plant Operations Review Committec i
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Unit 3 in Startup Mode (further details are d i scu s sed in ' Description of Event' section).
U.e s,c r,i_p t_i,,on _o f t h,c,,F.v e n t :
On March 18, 1986, at 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br />, Unit 3 was manully scrammed.
A description of the events that led to the manual scram follows.
U ni t 3 reactor startup commenced on Ma r c h 17, 1906 at 2218 hours0.0257 days <br />0.616 hours <br />0.00367 weeks <br />8.43949e-4 months <br />.
'~.c RWM was bypassed at the time because it could i.ot be proven to be operable.
Two licensed operators were assigned to monitor control rod withdrawal in accordance with Technical Specification 3.3.B.3.b which states, "h'nenever the reactor is in the startup run modes below 25% rated power the Rod Wor t h Mini mi ze r shall cr he operable or a second licensed operator shall.erify that the cperator at the reactor consolo is f > l l ow i n'3 the control rod program".
Control rods 1 thro
's,)f Group I wers withitrawn prior to change of shift at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />.
At shi f t change, two di f f crent 1ic<ensed operators were assigned Technical Speci ficat ion 3. 3.D. 3.h dut ics.
Rod withdrawal recommenced at Step 6 of Group 1 and cont inued to Step 13 of Group 1.
Step 13 specified that rod 02-23 be f ully wi thdrawn.
However, the reactor operator inadvertently withdrew rod 10-23 (which is located near rod 02-23).
Further, the necond licensed operator mistakenly veri fied that rod 02-23 had been correctly withdrawn.
This withdrawal error occurred at 0128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br /> March 18.
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Rod withdrawal continued, in proper sequence, to Step 6 of Group II.
Step 6 specified that rod 10-23 was to be fully wit 1hlrawn.
At this point the reactor operator lef t the control rod pane l to check recorders that were monitoring various parameters.
When the operator returned to his station, he observed that rod 10-23 was fully withdrawn, and, not realizing that he had previously wi thdrawn rod 10-23 a t 0128 hour0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br /> s, signed off the rod withdrawal sheet for Step 6, Group II.
Rod withdrawal continued until all Group II rods were fully withdrawn.
Group III rod withdrawal was attempted at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br />.
RSCS initiated a rod block at this point due to 02-23 being fully inserted.
The Shift Superintendent and Shift Supervisor placed the RSCS bypass keyswitch to the " full-out" position for rod 02-23 without verifying that rod 02-23 was f ully wi thd rawn.
This action bypassed the rod 02-23 insert error and allowed rod withdrawal to continue.
Group 111 rod withdrawal was completed at 0434 hours0.00502 days <br />0.121 hours <br />7.175926e-4 weeks <br />1.65137e-4 months <br />.
With the exception of the last ro 1 in Group IV
( which was only wi thd raan a s far as position 10), all Group IV rods were folly withdrawn by 0611 hours0.00707 days <br />0.17 hours <br />0.00101 weeks <br />2.324855e-4 months <br />.
No further rod wi thdrawals aere per formed on this shift, which endcol at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.
The Shift Superintendent, Shift Supervisor, and both reactor operators ended their shi f t at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.
At 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br />, the Shi f t Superintendent on day shift requented that t troubleshoot i ng be per f or med on the RWM.
At 0740 hours0.00856 days <br />0.206 hours <br />0.00122 weeks <br />2.8157e-4 months <br /> the RWM was re-initialized and returned to service.
The HWM indicated an insert error for rod 02-23.
The reactor operator observed that rod 02-23 was fully inserted although it s hoo l.1 have been f ully wi th:lrawn, rurther-investi 93 tion r.evi it".1 that the RSCS ke 'ock hypass switch for rod 02-23 wan place 1 in the
" full-&_
position, thereby hypassing the insert error.
At 074;l hours the RSCS bypass switch was returne<l to the normal p,sition.
As expected, a rod withdrawal block o c c u r ri." 1 due to rod 02-23 being out-of-sequence.
A controlled shu tdown wa s begun by inserting rods so that rod 02-23 <:ould be withJrawn in corroct sequence.
liowever, due to concerns that the 02-23 innertion may have adver sely a f f ected the Rod Drop Accident analynes, the reactor was manually scrammed at 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br />.
The ser.im occurred properly, and, at 0903 hours0.0105 days <br />0.251 hours <br />0.00149 weeks <br />3.435915e-4 months <br /> the scram was reset.
The Ells code for the a f fected Reactor Prot ect ion System in JC.
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Cor.se,quences of the Cvent :
The Philadelphia Electric Company Nuclear Fuel Management Section and General Elactric (GE) Company have revieeed this event.
Th e various RDA scenarios associa ted with this event were i.len t i f ied during this review.
The worst case RDA results of the scenarios show a peak enthalpy deposit ion of approximately 120 calories / gram compared to the RDA design criteria of 280 t
calories / gram.
The 280 calories / gram design criteria value bounds all currently licensed fuel hundle exposure values.
It is concluded therefore that, during the period of this event, the Unit 3 reactor operated in a condition such that RDA consequences would not have exceeded the design criteria.
Further, it was determined that the current Unit 3 reload analysis RDA results (peak enthalpy deposition of 215 calories / gram) bound this event.
C a u s e _o f,, t h e, Eve n t. :
This event was caused by a series of personnel errors involving fsilure of the reactor operator to withdraw the correct rod, failure of the second operator to adequately veri fy that the correct withdrawal sequence was being followed, and failure of the Shi f t Supervisor and Shift Superintendent to veri fy the proper position of a cont rol ro.1 prior to utiliz:nq the RSCS bypass function for that r o,1.
Co r r e.c t i v e Ac t,i on s :
The four individuals responsible for this event were disciplined.
To prevent recurrence, the following actions have been developed j
and are in various stages of implementation:
1.
A letter was issued on March 18, lWib from the Supe ri n tende nt - Operations to all licensed personnel outlining the event and describing interim requirements for verification of proper rod positioning.
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2.
A procedural control has been developeil to use the plant process computer to generate a core map at the completion of withdrawal of speci fic rod groups.
The computer-generated map will be compared to a ro.l map which is attached to the operator's rod withdrawal
{
sheet.
The two maps wi11 then be verifie.1 to be identical before proceeding to the next rod g r.wp.
3.
Six procedures involving RWM and RSCS have been revise.1 l
and PORC approved to enhance the proceiluril requirements.
Included in these revisions is the HSCS bypass procedure which has been reviseil to proviile a j
Shift Supervision sign off sheet for proper vorifleation
]
of rod position prior to i mp l e me n t i nij the RSCS bypass l
function.
4.
On March 24, 1986 the Plant Manager issue.I a. letter to j
Shift Supervision to ensure that "best e f f or t s" a re m.nle j
to place the RWM in service prior to commencin-) r e.ic t o r startup.
5.
Henceforth, when rod movements are heing performed with the RWM bypassed, the nocond licensed operator will be dedicated to ensuring that the proper ro.1 sequence is l
being followed.
j 1
6.
Plant staff management mootingn are being held with a11 I
Operations Personnel to discuss the event anel i n li v i dua l responsibi1itics.
Previous Similar 9ecurrens:cs:
Peach Bot tom IIR 3-77-19/lT.
One of the cor rect ive inct ions implemented subsequent to the 1977 event w.s s ver i f ica t ion of t he black-and-whi te pa ttern (all rods through Group IV are full out).
Rod wi thd rawa l on Ma rch 18, 1996 was hal ted before t.he black-an'l-whi t e pa t tern was re.iched.
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4 PHILADELPHIA A ELECTRIC COMPANY j
2301 M ARKET STREET P O DOX 0609 I
PHIL ADELPHI A. PA 19101 i
12151 R 41 4 000 I
/.pril 16, 1986 Docket No. 50-278 l
Document Control Desk i
U.S.
Nuclear Regulatory Commisolon l
Washington, DC 20555 SUBJ ECT:
Li,c e n s o o Ev,o n t,,l}cy,o,r t J
Peach not tom Atornic Power Sta t ion - Uni t 3 i
i This LER concerns to ' withitrawal errors and the subsequent ma nua l scram of Peach !!ot tom Unit 3.
l
Reference:
Docket No. 50-270 f
Report Number:
3-06-09 i
Revision Number:
00 J
Event Dato:
March 18, 1986 Report Date:
April 16, 1906 facility:
Peach llottom Atomic Power Ststton RD 1.
Box 208, Delta, PA 17314
)
This LER is being submitteil pursuant to the requirements of i
i Very troly yours.
l k,u_,
C.
M. I.eltch
/
S upe r i n t e nilen t Nuclear Generition 1)i v i n i ori ec:
Dr. Thomas I;. Murley, Administrator, Ro<jion 1,
I!S NitC T.
P.
Johnson, NRC Resident inspector b
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- ACILIT F Naaf t its DOCatt NutetAtil PAGE G PLANT V0GTLE - UNIT 1 42l 4 i lor l 014 oIsIoloIeI 1
TITLS 64, OPERATOR ERROR LEADS TO A REACTOR TRIP ON SOURCE RANGE HIGH FLUX E viert DATE 15e Lt A NUMeta egi aEpost DAf t ifi OTMt A 8 ACILITits tuvotvtO te, WONTw Dav etam
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On June 6,1987 at 0300 CDT during reactor startup an automatic reactor trip caused by the source range neutron flux protective signal i
occurred.
The reactor startup was being perfonned by a reactor operator trainee under the supervision of a reactor operator (RO).
Normal shutdown conditions were quickly reestablished.
The event was caused by operator error.
The RO and trainee were inattentive during reactor startup and did not recognize when the reactor was criticale Corrective actions included:
additional simulator training for the trainee, R0 and the shift supervisor; the use of an inverse count rate ratio plot for each reactor startup; estimated critical condition calculation review by reactor engineering prior to reactor plant startups; and revision of the " Estimated Critical Condition Calculation" procedure and the " Reactor Startup" procedure.
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REQUIREMENT FOR REPORT l
This report is required per 10 CFR 50.73 (a)(2)(iv) because an i
unplanned automatic actuation of the Reactor Protection System (RPS) occurred.
1 B.
UNIT STATUS AT TIME OF EVENT l
Unit 1 was undergoing reactor startup (Mode 3 to Mode 2) per procedure 12003-1, " Reactor Startup." The reactor coolant system pressure, temperature and boron concentration was approximately 2235 psig, 557 degrees Fahrenheit, and 1120 ppm, respectively.
C.
DESCRIPT!0U OF EVENT l
On June 6,1987, a reactor startu The l
estimated critical position (ECP)p was commenced at 0246 CDT.
calculations for this startup were performed by the reactor operator (RO) trainee and the shift I
technical advisor (STA) per procedure 14940-1, " Estimated Critical Condition Calculation".
The calculated ECP was 45 steps indicated j
on Bank D with a Boron concentration of 1120 ppm.
The startup was performed by a trainee under the supervision of an RO. All five shutdown rod banks had been previously withdrawn from the core.
Reactor startup was commenced by sequential withdrawal of the four control rod banks.
Control rod bank A was fully withdrawn (228 steps); control rod banks B and C were being withdrawn.
Per-procedure, the trainee stopped rod withdrawal every 50 steps to verify proper group alignment and count rate stabilization. When control rod bank C withdrawal was stopped at 70 steps, bank B was at 185 steps.
The count rate meters and recorder indicated the reactor was subcritical.
The trainee resumed withdrawal of control bank C.
At approximately 0258 CDT, a low pressure alarm on the No.
3 accumulator was received.
The shift supervisor directed the trainee to stop rod withdrawal.
After approximately one minute, the shift supervisor decided to continue the startup and assigned another operator to fill the accumulator to clear the alarm. The shift supervisor directed the RO to proceed with the startup and then began making a log entry.
The RO instructed.the trainee to proceed. The trainee initiated withdrawal of control bank C and observed a rapid increase in source range counts.
No attempt was made to reduce startup rate by rod insertion, but rod outward motion was stopped at 101 steps on bank C.
The intermediate range anps increased rapidly to 10E-10 amps, which activates the P-6 pernissive annunciator and lights.
At this time, the R0 also w
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w c m onm noticed the high source range counts, but before any action could be taken the reactor tripped at 0300 CDT.
All equipment functioned as designed.
Plant conditions were stabilized at approximately 0301 CDT.
D.
CAUSE OF EVENT The reactor trip on high neutron flux was caused by operator error.
Both the R0 and the trainee were inattentive to the reactor startup and did not recognize the reactor had gone critical during rod withdrawal of bank C.
The calculated ECP was 45 steps on control bank D (158 steps on bank C).
The actual critical was outside the plus/minus 500 pcm (approximately 50 steps) position range specified in procedure 12003-1, " Reactor Startup," for stopping control bank withdrawal.
The R0 and trainee did not expect criticality since they were not close to the ECP.
A review of previous reactor startups showed, there was a large ECP error, during the May 6,1987 reactor startup.
The RO recognized the error in the ECp and manually inserted control rods to stop the startup.
This error was caused by an inadequacy in the plant procedure 14940-1 which resulted in using the incorrect temperature for boron worth and had not been resolved by the time of the June 6,1987 event. Another contributing factor was that an inverse count rate ratio plot (1/M plot) was not required by procedure and was not being performed.
If this plot had been performed, it would have helped the RO to understand when to expect criticality without relying upon the erroneous ECP.
Additionally, the simulator and plant responses to rod withdrawal are not the same. The simulator is less sensitive than the plant to reactivity changes and there is more noise on the in-plant source range instrumentation.
This makes the simulator startup much easier to monitor than a plant startup.
E.
ANALYSIS OF EVENT The reactor trip system is designed to automatically initiate a reactor trip whenever necessary to prevent fuel damage for an anticipated operational transient.
The source range high neutron flux trip is designed to provide transient termination should a continuous control bank withdrawal occur from a suberitical or low power startup condition, without adversely affecting the core and the reactor coolant system.
In this event the reactivity insertion, as a result of manually withdrawing control rod bank C, was considerably less than the withdrawal accident considered in the FSAR.
Therefore, since all protective equipment functioned as designed, there was no adverse impact upon the plant and the public safety was never affected.
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CORRECTIVE ACTIONS Corrective actions for this event include:
1.
The RO was removed from any licensed duties pending adequate n
retraining.
l I
2.
The trainee,.R0 and the shift supervisor have received additional simulator training on. reactor startup.
l 3.
The reactor startup procedure, 12003-1, was revised to
. l l
require an Inverse Count Rate Ratio plot before a reactor startup.
{
}
4.
The procedure 14940-1, " Estimated Critical Condition Calculation" has been reviewed and revised to ensure boron worth at 557 degrees Fahrenheit.is used.
l a
i 5.
All ECP calculations shall be reviewed by a reactor engineer f
prior to reactor plant startups.
j l
6.
Training was provided to all licensed operators to emphasize that rod insertion is the correct response to any abnormalities that occur during a reactor startup.
l 7.
The simulator reaction response has been changed to make the i
simulator, more sensitive to reactivity changes. An increase l
in the simulator source range instrumentation noise was also added.
These changes will make the simulator startup more similar to the actuation plant startups.
l G.
ADDITIONAL INFORMATION 1.
Previous Similar Events NONE i
l 2.
Energy Industry Identification System Control Rod Drive System - AA Incore/Excore Monitoring System - IG 3.
Failed Components NONE
_.l"-""
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610f gI84Wti$0mpany -
I 333 Pieomont Avenue.
.(9 At:aata Georg-a 30306.
Te.eonone 404 526 6525 Maang Acc ess J
Nst Off4ce Bor 4545 l
At4anta Georg-a 39302 b
j Georgia Power i
L. T. Guews W !! "4'~ # C' I 3
- f Y~
Vb*6pe' tesc eB' SB*e's an0 L.Ce1S
- g SL-2782 0383m X7G317-V310 July 6, 1987
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U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D. C.
20555
-l PLANT V0GTLE - UNIT 1 NRC DOCKET 50-424
)
4 OPERATING LICENSE NPF-68 LICENSEE EVENT REPORT-OPERATOR ERROR LEADS TO A REACTOR TRIP ON SOURCE RANGE HIGH FLUX I
1 i
Gentlemen:
i Pursuant to the requirements of 10 CFR 50.73(a)(2)(iv), Georgia Power i
Company is submitting a Licensee Event Report (LER) concerning an event where an operator error resulted in a reactor trip on source range high
- flux, i
~
i l
Sincerely, pr-m -
L. T. Gutwa PAH/im i'
Enclosure:
LER 50-424/1987-032 c:
(see next page) 4 V
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37 I
~.
y 3.. o k
' Gebya Poiver C' i
O j
..~S. Nuclear Regulatory Commission July 6,'1987 Page Two-c: Georaia Power Comoany Mr. R. E. Conway a
Mr. J. P. 0'Reilly.
Mr. G. Bockhold, Jr.
Mr. J. F. D'Amico Mr. C. H. Hayes GO-NORMS 1'i Southern Comoany Services Mr. R. A. Thomas Mr. J. A. Bailey Shaw. Pittman. Potts & Trowbridae Mr. B. H. Churchill, Attorney-at-Law Troutman. Sanders. Lockerman & Ashmore Mr. A. H. Domby, Attorney-at-Law
'U. S. Nuclear Reaulatory Commission l
Dr. J. N. Grace, Regional Administrator
~
Ms. M. A. Miller, Licensing Project Manager, NRR (2 copies).
Mr. J. F. Rogge, Senior Resident Inspector-Operations, Vogtle Gporaians Aaainst Nuclear Enerav 4
Mr. D. Feig Ms. C. Stangler 4
t 3
4 l
0383m 38-i 1
_____1__________1_
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