ML20055J328
| ML20055J328 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 07/25/1990 |
| From: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20055J326 | List: |
| References | |
| EA-90-100, NUDOCS 9008020100 | |
| Download: ML20055J328 (3) | |
Text
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NOTICE OF V10LAT10N AND PROPOSED IMPOS] TION OF CIV]L PENALTY lilinois Power Company Docket No. 50-461 Clinton Power Station License No. NPF i EA 90-100 During an NRC inspection conducted on April 12-27, 1990, violations'of NRC requirements were identified.
In accordance with the " General Statement of.
l Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1990), the Nuclear Regulatory Conmission-proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act),
42 U.S.C. 2282, and 10 CFR 2.205. The particular violations and associated civil penalty are set forth below:
Technical Specification 3.1.4.1 requires that control rods not be withdrawn in OPERATIONAL CONDITIONS 1 and 2 when thc main turbine bypass valves are not fully closed and ' THERMAL POWER is greater than the low power setpoint.of the rod pattern control system (RPCS). The action statement requires that, with any control rod withdrawal when the main' turbine bypass valves are not fully closed and THERMAL POWER is greater than the low. power setpoint of the RPCS, innediately return the control rod (s) to the position prior to control rod withdrawal.
Technical Specification Surveillance 4.1.4.1' requires that control rod withdrawal be prevented, when the main turbine bypass valves'are not. fully closed and THERMAL POWER is greater than the low power setpoint of the RPCS, by a second licensed operator or.other technically qualified member of the unit technical staff.
A.
Contrary to the above, on April 11, 1990, with the' reactor in'0PERATIONAL
' CONDITION 1 at a THERMAL POWER greater than the low power:setpoint of the RPCS, at least fourteen control rod withdrawals were performed by a reactor-operator with main turbine bypass valves not fully: closed. Upon. discovery.
of the open main turbine bypass valves, the load selector was. raised to-close the main turbine bypass valves rather than immediately returning -
the control rods to the position prior to control rod withdrawal.
B.
Contrary to th'e above, on April 11, 1990, with the reactor'in OPERATIONAL-CONDITION 1 at a THERMAL POWER greater than the-low power.setpoint of.the.
RPCS, at least fourteen control rod withdrawals were not prevented by a second licensed operator or other technically qualified meber of'the unit' I
technical staff when the. main turbine bypass valves were not-fully closed.
' This is a Severity Level 111 problem (Supplement 1).
Civil Penalty - $25,000 (assessed equally between the two violations).
9008020100 900725
- PDR ADOCK 05000461
.Q PDC t
Notice of Violation Pursuant to the provisions of 10 CFR 2.201, Illinois Power Company (Licensee) is hereby required to submit a written statement or explanation to the Director, Of fice of Enforcement, U.S. Nuclear Regulatory Commission, within 30 days of the date of this Notice of Violation and Proposed Imposition of Civil Penalty (Notice). This reply should be clearly marked as a " Reply to a Notice of fiolation" and should include:
(1) admission or denial of the alleged violations (2) the reasons for the violations if admitted; and if denied, the reasons why, (3) the corrective steps that have been taken and the results achieved, (4) the corrective steps that will be taken to avoid further violations, and (5) the date when full compliance will be achieved.
If an adequate reply h not received within the time specified in this Notice, an order may be issued to show cause why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken.
Consideration may be given to extending the response time for good cause shown.
Under the authority of Section 182 of the Act, 42 U.S.C. 2232, this response shall be submitted under oath or affirmation.
Within the same time as provided for the response required above under 10 CFR 2.201, the Licensee may pay the civil penalty by letter addressed to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, with a chect, draft, money order, or electronic transfer payable to the Treasurer of the United States in the amount of the civil penalty proposed above, or may protest imposition of the civil penalty, in whole or in part by a written answer addressed to the Director, Office of Enforn ment, U.S. Nuclear Regulatory Commission.
Should the Lit.ensee fail to answer within the time specified, an order imposing the civil penalty will be issued.
Should the-Licensee elect to file an answer in accordance with 10 CFR 2.205 protesting the civil penalty, in whole or in part, such answer should be clearly marked as an " Answer to A Notice of Violation" and may:
(1) deny the violations listed in this Notice in whole or in part, (2) demonstrate extenuating circumstances (3) show error in this Notice, or (4) show other reasons why the penalty should not be imposed.
In addition to protesting the civil penalty, in whole or in part, such answer may request remission or mitigation of the penalty.
In requesting mitigation of the proposed penalties, the factors addressed i
in Section V.B of 10 CFR Part 2, Appendix C (1990), should be addressed. Any written answer in.secordance with 10 CFR 2.205 should be set forth separately from the statement or explanation in reply pursuant to 10 CFR 2.201, but may incorporate parts of the 10 CFR 2.201 reply by specific reference (e.g., citing page and paragraph lumbers) to avoid repetition.
The attention of the Licensee is directed to the o:her provisions of 10 CFR 2.205, recarding the procedure for imposing civil peralties.
Upon failure to pay any civil penalties due which subsequently has been determined in accordance with the applicable provisions of 10 CFR 2.205, this matter may be referred to the Attorney General, and the penalties, unless compromised, remitted, or mitigated, may be collected by civil action pursuant to Section 23.4c of the Act, 42 U.S.C. 2282c.
l
Notice of Violation The response noted above (Reply to Notice ci Violation, letter with payment of civil penalty and Answer to a Notice of Violation) should be addressed to:
Director, Office of Enforcement, U S. Nuclear Regulatory Commission, ATIN:
Document Control Desk, Washington D.C. 20555 with-a copy to the Regional Administrator, Region 111, U.S. Nuclear Regulatory Commission, 749 Roosevelt Road, Glen Ellyn, IL 60137, and a copy to the NRC Resident inspector at the Clinton Power Station.
FOR.THE NUCLEAR REGULATOPY COMMISSION
[
i A. Bert Davis Regional Administrator Dated at Glen Ellyn, Illinois this 25th day of July 1990 o
i
['.c i
I l.:
l I
U.S. HUCLEAR'REGULA10RY COMMISSION i
REGION 111 L
F Repori No. 50461/90009(DRP) l License No. NPF Docket No. 50-461 Licensee:
lilinois Power Company 500 South 27th Street-Decatur,1]L' 62525 Facility Name: 'Clinton Power Station Inspection At:
Clinton Site, Clinton, Illinois
'1 Inspection conducted:
April 12*27, 1990 Inspectors:
P, G. Brochman 1
R. D. Lanksbury S. P. Ray F. L. Brush J. B. Hiciarn O
2 I
C V4e Approved By:
D..L ksbury,. Chief Date j
ReactorProjectsSection3B i
Inspection Summary t.
Inspection from April 12-27, 1990 (Report No. 50-461/90009(DRP))'
Areas Snspected:
5pecial safety team inspection by the resider.t, Region-))I, and NRit inspectors to review the circumstances surrounding the improper with main turbine bypass valves withdrawal of control rods on April 11,1990, not fully closed.
The safety significance of this event is-derived from the lack j
Results:
of attention to plant' indications during important evolutions by licensed personnel, not believing plant indications, not informing supervisors of.
unexpected plant response, and the scheduling and controlling of complex-The actual physical and safety significant evolutions.during shift turnovers.
safety significance of this event on the reactor core was minor because all of the control rod withdrawals made during this event.were:in accordance'with thel rod pattern and were at a rate which was'more conservative than the rod pattern control system limits. Two apparent violations were identified (failure to follow Technical Specifications 3.1.4.1 and 4.1.4.1 Paragraph 8; f ailure to-follow procedures or develop adequate procedures - Paragraph 8).
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DETAILS
)
1.
Persons Contacted lilinois Power Company (1P)
"J. Perry, Vice President
- J. Cook, Manager, Clinton Power Station
- R. Wyatt, Manager, Quality Assurance
. J. Miller, Manager, Nuclear Station Engineering.
- F. Spangenberg, III, Manager. Licensing and Safety
- R. Morgenstern,. Manager, Scheduling and Outage Management
- J. Palmer, Manager, Nuclear Training
- J. Palchak, Manager, Nuclear Planning and Support-
- D. Morris, Director, Plant Operations
- S. Rasor, Director, Plant Maintenance i
i
- D. Miller, Director, Plant Radiation Protection
- J. Hansker, Director, Planning and Programming
- R. Phares, Director, Licensing
- S. Hall, Director, Nuclear Program Assessment
- K. Baker, Supervisor, I&E Interface Soyland Power t
- J. Greenwood, Manager, Power Supply The inspector also contacted and interviewed other-licensee personnel during the course of this inspection.
- Denotes those present during the exit interview on April 27, 1990.
2.
Purpose (41701, 71715. & 93702) j The purpose of this special team inspection was to review the circumstances surrounding the events on April 11, 1990, when several control rods were withdrawn during a reactor startup while the main turbine bypass valves were not fully closed.
The inspectors interviewed the shift crews, training department personnel and managers involved in this event. - The inspectors developed a chronology of events and analyzed this event for its safety implications.
The j
inspectors provided augmented monitoring of control room activities during the restart of the reactor.
f 3.
Description of the Event The following description of the event and the chronology were developed upon review of operating logs and strip charts, and interviews conducted during the inspection.
Some of the times listed are approximate and are based upon a consensus of individual memories of the event.
All of the times listed are CDT.
2
~
b
s 10, 1990,- at approximately 11:00 p.m., the. midnight shift On April "A" reactor operator (RO) (the R0 assigned to the control room panels co'ntrolling reactor power and monitoring equipment, turbine / generator controls, feedwater controls, etc.) assumed the' watch.
The A" R0 had noted during the shift turnover that generator load was at 105-110 MWe, reactor power was at 18-19%, and 'he electrohydraulic control (EHC) system generator load set was at 100 MWe, He also reviewed the plant l
l procedcres, in place at that time, in preparation for continuing the After completing the procedure review, the "A" R0 power ascension.
noted'that reactor power had decreased to approximately 16% due to xenon building in (increasing).
He then commenced withdrawing control rods with the line assistant shift supervisor (LASS) acting as the independent verifier.
During the interviews, the inspectors uniformly were told that the purpose of the independent verifier was to ensure that the correct control rod was selected from the rod withdrawal sequence and moved to-the correct location.
On April 11, 1990, at approximately 1:30 a.m., the "A" RO stopped-withdrawing control rods with reactor power at 23-24%.
During this time i
At this period he did not make any adjustments to generator load set.
point the low power set point.(LPSP) (approximately 22%) of the rod pattern control system (RPCS) had just been-reached and~ there were a number of surveillances that needed to be completed,before the power ascension could be continued.
A repair to a feedwater pump mechanical overspeed relay also needed to be completed and tested prior to increasing power.
At approximately 6:30 a.m., the "A" R0 completed testing of the feedwater Fif teen minutes later, at 6:45 a.m.,
pump mechanical overspeed relay.
the shif t supervisor (SS) instructed the LASS to recommence the power ower, prior to the end of the ascension with the goal of reaching 35% p'A" RO to commence withdrawing shift.
The LASS in turn instructed the control rods, in order to bring the-reactor up _to approximately 35% '
The "A" RO complied and commenced control rod withdrawal with a i
power.
day shif t staff assistant shift supervisor (SASS), who had come in early, acting as the independent verifier.
At this same time, day shift operators were arriving and began the i
process of shift turnover (which included panel walkdowns, a' shift briefing, log reviews, etc.).
Due to this, the normal senior operator oversight of this evolution did not occur.
In addition, there were also more people in the control room than normal. At around 7:20 a.m.,
the normal day. shif t SASS needed to get his turnover from the SASS who The LASS had the day
- shift shift was acting as the independent verifier.
technical advisor (STA) relieve the SASS as the independent verifier. -At about 7:25 a.m., the "A" RO noted that generator load was about 200 MWe, At He was also cognizant that generator load set was still at 300 MWe.
this same time the LASS shift turnover had just been completed and the Prior to leaving, the midnight shift LASS-j midnight shift LASS left.
reminded the day shift LASS about surveillances that were due at 25%
power and also informed him that they had received permission to continue the power ascension to 35% power.
The day shift LASS started reviewing l-3
t the power ascension procedure being.used from the point where they L
were at to see what needed to be completed.next.
He did not go back l
and review the portion that had already been done.
The LASS was engaged in making plans for.the work to be accomplished on dayshift.
Because of this he did not adequately monitor overall unit operation by observing plant indicators in order to detect any unusual or abnormal-trends.
At approximately 7:28 a.m., the rc. actor reathed 25% power.
Since-permission had already been given to continue the power ascension-At to 35% power, the "A" RO continued withdrawing control rods.
about 7:30 a.m., the day shift SS relieved the midnight shift 55.
Eight ininutes later at 7:38 a.m., generator load reached the load set and This went unnoticed the first main turbine bypass valve began to open.
by anyone, including the LASS who was responsible for monitoring overall plant response to the ongoing evolution, and the "A" R0 continued to withdraw control rods.
At about 7:40 a.m. the STA ncted.that generator load had not been increasing even though control rods were being withdrawn j
and reactor power was increasing.
He questioned the "A" RO about this and was told it wasedue to the computer display control system (DCS) not-updating properly.
Even though the STA was not satisfied with this answer.
he did not pursue it any further.
At 7:44 a.m., the first bypass valve j
reached the full open position and the second bypass valve began to open as control rods continued to be withdrawn.
Three minutes later at 7:47 a.m.,'the last control rod withdrawal was made with reactor power reaching approximately 35%.
The STA again questioned the "A" RO about-the fact that generator load had not increased.
The "A R0 attributed the problem to lockup of the DCS and/or xenon buildup.. During control rod withdrawals, earlier in the shift, the "A" RO had noted that generator load increases were not as large as expected.
This fact was attributed to the fact that the control rods being withdrawn were peripheral control rods with little rod worth and due to. xenon building in.
During the time.
that the bypass valves were open a total of 14-17 single notch, gang mode,.
control rod withdrawals were made.
With control tod withdrawals stopped, the STA left the contro1' area.
The "A" RO obtained'a process computer rintout-printout (OD-3) to verify reactor power level.
The 00-3 p'A" R0 waited indicated that reactor. power was approximately 34%.
The approximately three minutes and then obtained a~second 00-3 printout.
This printout' indicated that reactor power was approximately 38%.
The "A" R0 did not check what generator load was and had not looked at that parameter since 7:25 a.m.
At 7:51 a.m., the midnight shift "A" R0 commenced turnover to the day shift "A" RO.
As part of the turnover, the "A" R0s walked down all; the panels in'the control room.
This included the panels containing the indications that two bypass valves were open. They stopped at those panels and discussed items of interest but neither noted that the bypass valves were open.
The "A" R0s continued the' turnover process until 8:30 a.m. When the midnight shift "A" RO was officially relieved by.
the-day shif t "A" RO.
At about 8:00 a.m., the day shift LASS assigned the extra day shift R0 (referred to as "B prime" RO) the task of performing the portions l
of the power ascension procedure necessary to lineup balance-of plant l
4
(BOP) equipment.
The "B" and "B' prime" R0s arc assigned to the control room panels controlling balance-of plant syst ms.
Sometime between 8:00 a.m. and 8:30 a.m., the "B prime" R0 did note that there were bypass valves open, but apparently did not recognize the significance of this fact.
At approximately 8:33 a.m., the open bypass valves were discovered' by the two "B" R0s while performing panel walkdowns as part of their shift turnover. They immediately brought it to the attention of the' midnight shif t "A" RO who was still in the area.
The operatort recognized that the generator load set was too low.
After a brief discussion as to whether any control rod withdrawals had been made with the bypass valves open-(the midnight shif t "A" R0 did not remember making any), the midnight shif t "A" RO increased generator load set, in order to close the bypass-valves.
At
'8:35 a.m., the generator load set was increased to approximately 520 We.
By 8:36 a.m., the bypass valves were fully closed and generator load increased to approximately 300 We.
j The SS directed that the high speed computer monitoring system GETARS =
be checked to determine if any control rod withdrawals had been made with the bypass, valves open.
By 10:00 a.m., the GETARS records had been obtained and they indicated that 14-17 control rod withdrawals had occurred
,l with bypass valves open, The exact number of rod withdrawals could not be 1
determined because the first bypass valve started to open sometime during a series of three rod pulls.
The Operations department management made j
notifications of the initial f acts to-the NRC residents and senior plant management.
Operations department conducted a critique at 4:00 p.m. that day, to review-the event.
Senior licensee management was informed of the initial facts of the event; howe'ver, the results of the critique and the significance of the issues involved in this_ event (reactivity control, performance of licensed operators, and conduct of control room evolutions) were not communicated to senior management until they were expressed by NRC Region 111 management at a previously scheduled meeting the1 next day i
with~1icensee corporate management.
After reviewing this information, licensee management decided to shutdown the unit and implement the corrective actions discussed in Paragraph 6.
The shutdown was commenced i
at 7:34 p.m. on April 13, 1990, by inserting control rods. The generator was off-line at 9:42 p.m. that day.
The unit was subcritical at 6:40 a.m.
on April 14 and reached cold shutdown by 5:45 a.m. on April 15, 1990.
1 4.
Sequence of Events The following is a brief-sequence' of events:
April 10, 1990 11:00 p.m.
"A" reactor operator (RO) took the shift with the generator at 105-110 We, reactor power at 18-19% power, and load set at 300 We, Commenced withdrawing control rods.- Line assistant shif t supervisor (LASS) acting as independent verifier.
5
7 I;tus April-11, 1990 1:30 a.m.
"A" RO stopped withdttwing control rods with reactor
]
power at 23-24% and just above the low power set point i
(LPSP) (approximately 22% thermal power) to perform surveillances.
Generator load was 125 We.
l 6:30 a.m.
"A" RO completed testing of a feedwater pump sechanical overspeed relay that had been repaired.
6: 45 a.m.
Shift supervisor (SS.) directed the LASS who' instructed the "A" RO to commence withdrawal of control rods to i
increase power, with the goal of reaching 35% power prior-to.the end of the shift.
Control room turnovers had begun for some positions.
Staff assistant shift supervisor' l
(SASS) acting as independent verifier.
{
7:20 a.m.
Day shift shift technical advisor (STA) relieved SASS
)
as the independent verifier.
=!
7:25 a.m.
"A" RO noted that generator load was 200 We.
" A'i RO-i also was cognizant that load set was at 300 We, Midnight l
LASS officially relieved.
7:28 a.m.
Achieved 25% power.
Continued withdrawing control rods to 35% power.
7:30 a.m.
Midnight shift SS relieved by day shift SS.
7:38 a.m.
First main turbine bypass valve began to open.
This fact was not observed by the "A" RO or the'STA.
Control rod withdrawals continued.
7:40 a.m.
STA questioned "A" R0 about generator load not u
increasing.
"A" R0 attributed this to display control i
i system locking up (not responding).
7:44 a.m.
First bypass valve was fully open.
7:47 a.m.
Power reached 35%.
Control t'od withdrawals ended.
STA again questioned "A" RO on why generator load has-not increased.
"A" RO 6 gain attributed this to display control system locking up or a xenon building in.
STA 1
t left the control area.
"A" RO recognized generator load' at 200 We, "A" RO obtained a computer printout'of thermal power level (00-3) which indicated it to be 34%.
7:50 a.m.
"A" R0 obtained a second OD-L which indicated reactor power to be 38%.
6 i
-l l
f 7:51 a.m.
"A" RO commenced turnover to day shif t "A" RO, including panel walkdowns.
8:00 a.m.
Extra day shif t "B prime" R0 assigned to perform required balance-of-plant equipment lineups by LASS.
7:51 to Both "A" R0s stopped in front of panel with oypass valve 8:30 a.m.
position indication and looked at and discussed items i
of interest.
Bypass valves were not recognized as being open.
8:00 to
- B prime" RO noted bypass valves open but did not recognize 8:30 a.m.
significance of this fact.
8:30 a.m.
Midnight shif t "A" RO relieved by day shif t "A" RO.
8:33 a.m.
Bypass valves are recognized as being open by "B" R0s performing panel walkdowns as part of their shif t turnover.
8:35 a.m.
Ge'nerator load set is raised to 520 We by midnight shif t "A" RO.
8:36 a.m.
Bypass valves all fully closed.
Generator load has increased to 300 MWe.
SS directed that GETARS records be checked-to determine if control rods were withdrawn while the bypass valves were open.
10:00 a.m.
GETARS records indicated that control rods were withdrawn with bypass valves open.
Initial notification of event was made by operations department to station management and NRC.
4:00 p.m.
Critique conducted on the event by operations department management, t
April 12. 1990 1:00 to Licensee station and corporate management met with 4:00 p.m.
NRC Region III managers for a routine meeting, at l
which the NRC expressed concern over the event on April 11.
April 13, 1990 7:34 p.m.
Licensee management directed that the unit be taken to cold shutdown (Operational Condition 4).
9:42 p.m.
Generator was taken off-line.
April 14, 1990 6:40 a.m.
Reactor was taken suberitical.
7 i
April 15, 1990 5: 45 a.m.
Unit reached cold shutdown.
5.
Evaluation and Analysis Operations Procedure CPS No. 3004.01, " Turbine Startup and Generator Synchronization," was an integrated operating piocedure which provided overall direction to take the unit from 8% power with the turbine shutdown ard steam flowing through the bypass valves to 35% power with the generator synchronized and the bypass valves closed.
This procedure was being utilized by the operators at the time of the event.
Paragraph 6.8, Limitations, stated, "In accordance with Tech Spec 4.1.4.1, when above 20% thermal power (Low Power Setpoint of the RPCS), and when the main turbine bypass valves are not fully closed, control rod withdrawal shall be prevented.
This require;nent shall be verified by a second licensed operator or other technically qualified member of the i
unit technical staff and documented by shift Supervisor log entry."
Paragraph 8.2.5, dealt with Generator Synchronization.
Paragraph 8.2.5.5 synchronized the generator and Paragraph 8.2.5.6 directed that generater load be increased by selecting INCREASE on the Load Selector until the bypass valves are shut and Load Selector is approximately 150 MWe above s
generator load.
Immediately before Paragraph 8.2.5.7 was a caution which stated, " NOTE, Load Selector should be maintained approximately 200 MWe above generator load for further load increaser."
The "A" RO was withdrawing control rods and failed to raise the load selector as required by Procedure 3004.01;. consequently, as reactor power and reactor pressure increased, the main turbine bypass valves began to open.
This ft.t was not noted by either the "A" RO or the STA who was performing the independent verification of control rod movements.
There were multiple indications, within easy view of the "A" R0 and the STA, which indicated that the bypass valves were beginning to open.
However, there were no annunciators which alarm when the bypass valves were open. With the bypass valves opening, generator load did not increase.
The STA did observe that generator load was not increasing.
Twice he questioned the "A" R0 about this probl e.
The "A" R0 did not believe his indications and attributed this to the possibility that the display control system (DCS) had locked up (was not updating).or that because xenon was building in (increasing) the relative worth of control l
rods was reduced; consequently, generator load might not increase.
The "A" RO continued to withdraw control rods.
1 Administrative Procedure CPS No. 1001.05, " Authorities and Responsibilities of Reactor Operators for Safe Operation and Shutdown," Paragraph 8.1.3.1, i
I states "The line Assistant Shif t Supervisor assigned to the Main Control Room Area shall perform the following duties:
a.) Monitor i
/
overall Unit operations for adherence to CPS Technical Specifications.
... f.) Monitor CRTs, indicators, annunciators, and recorders in order to detect unusual or abnormc1 trends and initiate appropriate, timely action to correct or mitigate the situation...." The LASS failed to 8
s
,-a s-.-.-.--
..,_-....-- ---+.
n.~.....-
4 adequately follow the oversight requirements of this procedure because he was planning the work for the dayshif t.
Indications of reactor power, penerator load and bypass valve position were readily available to him and i
located only a few feet in front of his desk.
Administrative Procedure CPS No. 1401.01, " Conduct of Operations,"
Paragraph 8.5.4.. stated, " Indications are provided to monitor plant J
-' be believed, enless verified faulty by two alternate parameters ar:1 independentmeanswhenpossible,orthroughmaintenancetroubleshooting."
The
'A" RD 1.siled to follow this procedure and did not believe his indications.
The STA still had concerns with the fact that generator load was not going up; however, he did not raise these concerns with the LASS (control room supervisor (SRO)).
Nor did the "A" RO inform the LASS of the abnb.nal plant response.
J As a result of these actions, at approximately 7:38 a.m., bypass valves began to open.
During the next 9 minutes 14-17 control rod withdrawals were performed.
Several different control rods were withdrawn, in the gang mode, using single notch withdrawal.
By 7:47 a.m., reactor power had reached 35% and control rod withdrawals were stopped, pending startup Bypass valves remained open for an of balance-of-plant eq(uipment. total of 58 minutes open).
additional 49 minutes The main turbine control valves opened in response to increased reactor pressure, as reactor power was increased; this permitted the control valves to pass more steam and increase generator load.
If reactor pressure was increased and generator load was limited, then the reactor pressure controller would cause the main turbine bypass valves to open.
draulically operated bypass valves.
Each has a capacity There are six hk of rated steam flow. When the control rod withdrawal equivalent to 6 were stopped, reactor power was approximately 35%.
Nominally, 26% of the steam flow (reactor power) was going to the main turbine; the other 9%
was going through the turbine bypass valves.
The reactor did not sense anything different than if all 35% cf the steam flow had been going to I
the turbine.
However, this condhion was significant becau:e the rod pattern control system (RPCS) utilized main turbine first stage impulse pressure as an input.
First stage impulse pressure was utilized as a linear method of measuring equivalent reactor power.
Consequently, with some of the steam flow diverted froni the main turbine, the first stage i
The R)CS impulse pressure indicated a lower than actual reactor power.
used this equivalent reactor power to limit the number of notches a l
l control rod could be withdrawn continuously.
Technical Specification 3.1.4.1 required that control rods shall not be j
withdrawn in OPERATIONAL CONDITIONS I and 2, when the main turbine bypass l
valves wer? not fully closed and THERMAL POWER was greater than the low power setpoint of the RPCS. With any control rod withdrawal when the
(
main turbine bypass valves were not fully closed and THERMAL POWER was greater than the low power setpoint of the RPCS, IMMEDIATELY (emphasis added) return the control rod (s) to the position prior to control rod withdr#wal.
Technical Specification Surveillance 4.1.4.1 required that control rod withdrawal shall be prevented when the main turbine bypass valves were not fully closed and THERMAL POWER was greater than the low
4 power setpoint of the RPCS, by a second licensed operstor or other technically qualified member of the unit technical staff.
The basis for Technical Specifications 3.1.4.1 and 4.1.4.1 stated that the rod withdrawal limiter system input power signal originated from the main turbine first stage impulse pressure. When operated with the steam bypass valves open, this signal indicated a core power level which was less than the true core power.
Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential existed for nonconservative control rod withdrawals.
Therefore, when operating at a sufficiently high power level, there was a small probability of violating fuel Safety Limits during a licensing basis rod withdrawal error transient.
To ensure that fuel Safety Limits were not violated, this specification prohibited control rod withdrawal when a biased power signal existed and core power exceeded the specified level.
The actual safety consequences to the reactor core were mitigated by the fact that all of the control rod withdrawals made by the "A" R0 during this event were.in the proper sequence and were single notch withdrawals; vice, the four notch withdrawal limit which the RPCS would impose from 25-35% power.
Consequently, a second independent error, plus a much higher reactor power level would have been necessary to challenge the fuel safety limits.
I At approxirnately 8:20 a.m., the midnight shif t "A" R0 and the day shif t "A" RO began their turnover.
Both R0s stated that they had stopped at the panels which contained at least six separate indications that bypass valves were open and discussed _the status of the equipment contained on those panels without recognizing that the bypass valves were open. The "A" Ros completed their turnover, at 8:25 a.m.,
The "B" R0s were performing their turnover and at 8:35 a.m. the day shift "B" R0 discovered the bypass valves open.
After a brief discussion, the midnight shift I
"A" RO then raised the load set, which caused the generator load to increase and the bypass valves to close.
At no time did the operators j
refer to the the Technical Specifications to verify that all required actions had been taken.
The action statement of Technical Specification 3.1.4.1 required that with any control rod withdrawal, above the applicable power levels and any bypass valves open, IMMEDIATELY (emphasis added) return the control rod (s) to their prior position.
Typically licensees may comply with a Technical Specification action statement or change plant conditions such that the Technical Specification is no longer applicable.
When the R0s discovered the bypass valves open, they raised the load set rather than l
inserting control rods.
During the interviews, the operators stated that when they discovered the bypass valves open they did not believe that control rod withdrawals had caused the bypass valves to come open, but that some unrelated problem with the bypass valves had occurred.
The operators remembered the procedure requirement to keep load set 200 MWe I
above generator load and with no nexus to the control rod withdrawals which had happened some time earlier, they raised the load set.
- However, that is not what the Technical Specification action statement required.
10 A
l During the interviews, all the control room personnel stated they were f amiliar with the prohibition against withdrawing control rods, with the bypass valves open; however, they were not familiar with the requirements of the immediately effective action statement to insert They did state that if they were withdrawing control rods control rods.
and saw a bypass valve come open that they would immediately insert the control rods.
The inspectors discussed the intent of the action statement with the NRR staff.
In this event, raising the lud set (so that the RPCS would more accurately know the reactor power) was not technically incorrect or Nevertheless, the action statement was based upon the supposition unsafe.
that a rod withdrawal error had occurred and that fuel safety limits had been, or would shortly be, challenged; consequently, to minimite any fuel damage the control rods should be inserted to reduce the power density and enthalpy of the fuel as soon as possible.
Procedure 1401.01, Paragraph 8.5.4.4 stated, "Other permanently installed indicators (pages, meters, recorders, etc.) that are removed from service or operating in a degraded or out of calibration status, should be identified with a yellow caution tag." There were no caution tags on the load set meter or load selector pushbuttons.
During the interviews with the control room crews and the training department, several related facts emerged.
The modeling of the simulator with regard to the difference between the load set and generator load was not identical to the actual plant.
As originally designed the difference between the load and load set was less than 50 We; however, the meters in the plant did not respond that way.
Per Procedure CPS No. 1401.01 Paragraph 8.5.4.4,acautiontagwaspreviouslyplacedonthemetersIn i
the control room to indicate that bypass valves would open if load set was not more than 150 We above generator load.
A modification was proposed.
to correct this problem; however, the modification review committee did not approve the change as being cost effective and a decision was made to change the procedure to reflect the greater difference between load and load set by incorporating a cautionary note; which then allowed the The training department was not informed of caution tag to be removed.
the decision of the modification review committee and consequently did not realize that the simulator did not accurately model the plant.
Statements were made to the inspectors that some control room personnel had felt they were under pressure to get to 35% power and that they did not want to continue with the rod withdrawals during the turnover process; however, when directed to proceed they did so. The oncoming SS was concerned with the continuing rod withdrawals and discussed this concern with the LASS.
Before a decision to step the rod withdrawals was made, 35% power was reached.
A review of integrated operating procedure 3005.01, " Unit Power Changes,"
which was used to take the unit from 35% to 100% power, indicated that 11
i the procedure did not contain any references to raising the load set, as reactor power was increased.
l i
6.
Licensee's Corrective Actions In response to the event, the licensee implemented the following l
corrective actions:
f Each shift crew was briefed on the event and its causes before a.
they came on shift.
Each shift crew received retraining on reactivity management j
b.
procedures, the importance of procedure compliance, and the importance of close monitoring of equipment configuration and The thorough equipment status checks during shif t turnovers.
crews were also required to perform exercises on the plant simulator to assure that they fully understood the proper checks and procedures Each crew to follow during reactivity changes and plant startup.
was required to complete a written examination covering these topics.
The Plant Manager and Vice President met personally with each shift c.
crew to discuss the causes of the event and to reinforce the lessons presented in the training described above.
The Vice President met with.the Clinton Power Station (CPS)
I d.
managers and directors to discuss with them the importance of prompt recognition and response to potential problems, and prompt reporting of these problems up the chain-of command, reporting directly to An experienced senior-level individual theVicePresidentwasassigned-tomonItorOperationsdepartment e.
performance to provide additional assurance that significant problems would be promptly recognized and responded to.
Eight director-level individuals were assigned to monitor power f.
ascension.
These individuals mor.itored each shift's performance until the plant attained full power during the startup subsequent to the event, Experienced personnel, reporting directly to the Vice President, g.
from outside Illinois Power (IP) (one from Stone and Webster and two from Toledo Edison), were retained to independently review the cause of the event and the licensee's corrective actions, The licensee changed its policy so that no significant evolutions b.
would occur during shif t turnover.
If necessary the shift turnover wouldbedelayedtoallowcontinuationofacritIcalevolutionthat could not be stopped at that time.
The plant would be placed in a stable condition prior to turnover.
l 12
. ~
i.
The power ascension procedure 3004.01 was changed to require the operator to increase generator load set to a pre determined level in a plateau method as reactor power is increased.
The pre determined level was specified at appropriate parts of the procedure.
And it was raised to its maximum value at the end of 3004.01.
j.
A label was added by the generator load set meter on the standby information panel and by the generator load set increase and decrease push buttons to annotate that the generator load set meter read 300 S'e high.
k.
The generator load set meter in the simulator was adjusted to reflect the actual control room condition of reading 100 %'e high.
1.
The training department completed a review of disapproved modifications, to ensure that they did not result in procedure changes which would affect the simulator modeling, Integrated operating procedures were changed.'o prohibit the LASS or m.
This allowed them the STA f rom f unctioning as independent verifier, to step back and view the whole plant.
7.
Observation of Restart At the direction of Region !!! management, inspectors were assigned to augmented coverage of prestartup, startup, and power ascension activities, to approximately 3% power.
Areas that were observed were as follows:
a.
Required systems were operable, b.
Plant operating staff was ready, Minimum number of problems / alarms in the control room, c.
d.
Control room procedures were current, e.
Adequate preshift briefings occurred, f.
Inter and intra shift communications were adequate, g.
Operator responsiveness to alarms and control room indications were correct, and h.
Log keeping practices were acceptable.
The inspectors observed control room activities from April 22 26, 1990.
The overall perception was that all operating personnel performed in an extremely conservative and safe manner.
Ongoing evolutions were stopped well before shift turnover would begin. The preshif t briefings were thorough and detailed. The turnovers and walkdowns were also very thorough and detailed.
During the startup some of the operators were observed to be very nervous, but they became more confident 13
-l
1 lhere was no sense of urgency to accomplish j
as the startup progressed, too many tasks, but a recognition of the need for controlled progress.
Operator performance of control rod withdrawals and monitoring of the Control of personnel and plant's response were observed to be very good.
restriction of work activities during the startup were good.
A sign was 4
Control rod posted limiting control room access during the startup.
movements were monitored for compliance with Technical Specifications There were no discrepancies in this and the planned rod pull sheets.
The inspectors did note that some of the LAS5$ would focus too closely on the 680 (reactor control) panel activities and not step back area.
and look at the whole plant, but this was isolated to one or two individuals and they were improving as the startup progressed.
B.
Conclusions The inspectors have identified the following root causes for this event:
The "A" RO f ailed to follow procedure to raise load set at he was a.
increasing power, by withdrawing control reds.
The "A" RO and the STA did r.ot monitor typass valve position b.
af ter the withdrawal of control rods.
The "A" R0 did not believe his indications when told twice c.
by the STA that generator load was not increasing.
1 The STA and the "A" R0 did not identify the problem with generator d.
load to the LASS.
The STA did not pursue his concerns with the LASS after the answer e.
he was given by the "A" R0 did not seem right, The LASS failed to adequately monitor overall unit operations by f.,
observing plant indicators.
The inspectors have identified the fn110 wing contributing causes for this event:
A poor human factors approach in the integrated operating procedure, regarding raising the load set continuously, as directed by a note, g.
versus discrete procedural steps and raising load set to discrete plateaus.
Pressure on the midnight shift personnel to get to 35% power.
b.
Performing complex evolutions during turnover.
I The "A" R0 had only done this portion of a startup once before; and i.
had only done a total of three startups.- This type of activity was typically not practiced on the simulator.
However, this operating i
crew had been given practice on the simulator, within a month of 14
)
]
this event, on reactor startups; but due to time constraints certain f
portions of the startup process were skipped.
j.
No annunciator alarm for open bypass valves existed.
k.
The wording of Technical Specification 3.1.4.1 and L'
Surveillance 4.1.4.1 was cumbersome and awkward. The wording was copied into the operating procedures.
1.
The simulator did not model plant performance on the response of load set versus generator load accurately.
Training was not informed of rejection of a modification and the decision to Consequently, l
utilize a procedure change to correct a problem.
the simulator was not updated from the original design basis.
The lack of understanding by the STA and the "A" RO that a xenon m.
transient would not have a significant affer:t on reactor power over i
a 10 minute period.
j The actual physical safety significance of this event on the reactor core was minor because all of the control rod withdrawals made during this event were in accordance with the rod pattern and were at a rate (single notch withdrawal vs. four notch continuous withdrawa)) which was more conservative than the rod pattern control system limits.
The safety significance of this event is derived from the lack of attent'on to plant indications during important evolutions by licensed persontel, not believing plant indications, not informing supervisors of unexpected plant response, and the scheduling and controlling of complex and safety significant evolutions during shift turnovers.
This event and two other recent events involving operations department personnel (the cracking of condenser water boxes and the de-energization of the Division Il NSPS bus) taken together were not indicative of All three of these events had common aspects adequate performance. adherence to procedures, communication between operating involving:
crew members, turnover of information between operating crews, and management oversight and direction of operating crews.
Technical Specification 5.1.4.1 required that control rods shall not be withdrawn in OPERATIONAL CONDITIONS 1 and 2, when the main turbine bypass valves are not fally closed and THERMAL POWER is greater than.-
The the low power setpoint t.f the rod pattern control system (RPCS).
action statement required that with any control rod withdrawal when the main turbine bypass valves are not fully closed and THERMAL POWER is greater than the low power setpoint of the RPCS, immediately return the control rod (s) to the position prior to control rod withdrawal.
Technical Specification Surveillance 4.1.4.1 required that control rod withdrawal shall be prevented when the main turbine bypass valves are not fully closed and THERMAL POWER is greater than the low power setpoint of 15 u
]
I e
the RPCS, by a second licensed operator or other technically qualified member of the unit technical staff.
With the reactor in OPERATIONAL CONDIT10N 1 at a thermal power above the the low power setpoint of the RPCS,14-17 control rod withdrawals were performed by a reactor operator, with main turbine bypass valves not l
Upon discovery of the open main turbine bypass valves, the fully closed.
load selector was raised to close the main turbine bypass valves; vice, inserting control rods.
The withdrawal of control rods with the bypass valves open is an example of an apparent violation of Technical 3.1.4.1 (461/90009-01a(DRP)).
The failure of a second Specification technically qualified member of the unit technical staf f to prevent control rod withdrawals, with main turbine bypass valves not fully closed.
i is an example of an apparent violation of Technical, Specification 4.1.4.1 (461/90009-01b(DRP)),
10 CFR Part 50, Appendix B, Criterion V, required that activities affecting quality shall be prescribed by documented' instructions or procedures, of a type appropriate to the circumstances and shall be accomplished in accordance with those instructions.
Clinton Power Station Operations Procedure CPS No. 3004.01, " Turbine Startup and Generator Sy'nchronization," Revision 11, Paragraph 6.8,
" Limitations," stated, In accordance with Tech Spec 4.1.4.1, when above 20% thermal power (Low Power Setpoint of the RPCS), and when the main turbine bypass valves are not fully closed, control rod withdrawal shall This requirement shall be verified by a second licensed be prevented.
operator or other technically qualified member of the unit technical staff and documented by shift supervisor log entry." Paragraph 8.2,5.6 contained a note which stated, " NOTE, Losd Selector.should be maintained approximately 200 MWe above generator load for further load increases."
Clinton Power Station Administrative Procedure CPS No. 1401.01, " Conduct of Operations," Revision 2, Paragraph 8.5.4.2 stated, " Indications are provided to monitor plant parameters and shall be believed, unless verified faulty by two alternate indep'endent means when possible, or through maintenance trouble shooting.
Paragraph 8.5.4.4 stated, "Other permanently installed indicators (gages, meters, recorders, etc.) that are removed from service or operating in a degraded or out of calibration status, should be identified with a yellow caution tag."
Clinton Power Station Operations Procedure CPS No. 3005.01, " Unit Power Changes," Revision 12, provided detailed instructions for performing the operations required to effect unit power changes between 35% and 100% power.
Clinton Power Station Administrative Procedure CPS No. 1001.05,
" Authorities and Responsibilities of Reactor Operators for Safe Operation and Shutdown," Revision 6, Paragraph 8.1.3.1, states "The line Assistant Shift Supervisor assigntd to the Main Control Room Area shall perform the following duties:
a) Monitor overall Unit operations for adherence to CPS Technical Specifications.
... f) Monitor CRT's, indicators, 16 l
a
annunciators, and recorders, in order to detect unusual or abnormal trends, and initiate appropriate, timely action to correct or mitigate the situation."
The following are all examples of an apparent violation of 10 CFR Part 50, Appendix B, Criterion V:
The "A" RO f ailed to raise the load set while increasing reactor a.
power to keep the load set 200 Mde above3004.01 (461/90009-02a(DR by Procedure b.
The "A" RO did not prevent the withdrawal of control rods with the main turbine bypass valves not fully closed and reactor power above the LPSP as required by Procedure 3004.01 (461/90009-02b(DRP)).
The STA did not verify that control-rod withdrawals were not c.
permitted with the main turbine bypass valves not fully closed and reactor power above the LPSP, as required by Procedure 3004.01 (461/90009-02c(ORP)).
d.
The "A" RO did not believe his indications when he was informed that generator load (MWe) was not increasing.
He did not verify through other available indications that generator load was responding correctly, as required by Procedure 1401.01 (461/90009-02d(DRP)).
No caution tag was placed on the generator load set meter, e.
located on the standby information panel, to indicate that it was in a degraded condition, as required by Procedure 1401.01 (461/90009-02e(DRP)).
f.
Procedure 3005.01 did not contain any instructions for raising the load selector, even though this action was required to raise unit power from 35% to 100% (461/90009-02f(DRP)).
The LASS failed to adequately monitor overall unit operation by not g.
observing plant indicators and therefore did not detect that control rods were being withdrawn with the bypass valves open, that i
generator load set was not being increased, and that reactor power was increasing without a corresponding increase in generator load.
Two apparent violations were identified.
9.
Recommendations The inspectors have made the following recommendations for the licensee's and the NRC management's review:
The licensee should evaluate the installation of an annunciator for 1
a.
open bypass valves when reactor power is above the RPCS low power setpoint.
i 17 l
NRR and the licensee or the BWR-6 owners group should review b.
the wording of Technical Specifications 3.1.4.1 and 4.1.4.1 and determine if more easily understood wording is possible.
If this is possible the inspectors recommend that the Technical Specifications for all BWR-6 reactors be amended, Licensee management needs to continue reinforcing its philosophy of operation and the.need for communication and procedural compliance.
c.
10.
Exit Interview (30703)
The inspectors met with the licensee representatives denoted in Paragraph I at the conclusion of the inspection on April 27, 1990.
The inspectors j
summarized the purpose and scope of the inspection and the findings.
The inspectors also discussed the likely informational content.of the inspection report, with regard to documents or processes reviewed by The licensee did not identify the inspectors during the inspection.
any such documents or processes as proprietary.
C i
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