ML20203C817
| ML20203C817 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 05/19/1986 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18155A267 | List: |
| References | |
| 4.08, NUDOCS 8607210081 | |
| Download: ML20203C817 (106) | |
Text
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PALISADES NUCLEAR PLANT Proc No 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 1 of 10 TITLE: POST TRIP REVIEW REQUIREMENTS l
i I
1.0 PURPOSE This -procedure is to provide a systematic method for diagnosing the cause(s) of a reactor trip, ascertaining the proper functioning of safety related and other important equipment during the trip, deter-mining any detrimental effect on plant equipment caused by the trip, and making the determination that the plant can be restarted safely.
2.0 SCOPE A documented review should help ensure events that have had an impact on the cause of the trip and subsequent equipment responses are iden-tified and thoroughly understood. The review results will permit a determination to be made as to the readiness of the plant to be safely returned to operation. The format of this procedure can also be used to diagnose reactor plant transients other than those involving reactor trips.
3.0 REFERENCES
l 3.1 NODS-A17, " Management Overview" 3.2 KRC Ceneric' Letter 83-28, dated July 8, 1983 3.3 Palisades Administrative Procedure 10.41, " Procedure On Procedures" 3.4 Palisades Administrative Procedure 10.42, " Procedure / Document Matrix" 3.5 Palisades Administrative Procedure 10.46, " Plant Recceds" 3.6 Palisades Administrative Procedure 4.01, " Shift Operations" 3.7 Nuclear Assurance Department Procedure XIX-2, " Plant Safety Engineering Assessment Program" 3.8 Palisades Plant Site Emergency Plan 4.0 RESPONSIBILITIES 4.1 PLANT GENERAL MANAGER The Plant General Manager or the Duty and Call Superintendent is responsible for making the decision to start up the reactor. For the purpose of this procedure, reference to the Plant General Manger includes the Duty and Call Superintendent when applicable.
In addition the Pirnt General Manager shall verify the cause of the trip has been determined and appropriate corrective action taken.
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PALISADES NUCLEAR PLANT Proc No 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 2 of 10 TITLE: POST TRIP REVIEW REQUIREMENTS 4.2 PLANT SAFETY ENGINEERINC (PSE)
PSE shall perform an assessment of every inadvertant plant trip and provide the results of such assessment to the Plant General Manager in accordance with Reference 3.7.
4.3 OPERATIONS SUPERINTENDENT a.
For unplanned reactor trips involving control rod motion when the reactor is not critical or for plant transients other than reactor trips, the Operations Superintendent may direct that the Posc Trip Review Process be performed.
b.
For a Condition II event, as determined by Step 5.3.4, the Operations Superintendent or his designated alternate shall:
1.
Report to the Plant site to assist in the further investigation of the trip and to determine necessary corrective action before restart.
2.
Analyze the event reconstruction and supply information per Step 5.4.2 to the Plant General Manager.
4.4 LEAD SHIFT ENGINEER The Lead Shift Engineer shall review all completed Reactor Trip Reports.
4.5 DUTY SHIFT ENGINEER a.
The Shift Engineer (SE) on duty shall:
1.
Complete offsite personnel and agency notification in accordance with Reference 3.6 and 3.8 as appropriate.
2.
Interview plant personnel involved in the trip.
3.
Collect data for inclusion in trip report.
4.
Make recommendations to Plant General Manager on reactor startup.
5.
Prepare Trip Report.
b.
The Shift Engineer (SE) on duty should notify the Operations Superintendent of:
1.
Any unplanned reactor trip involving control rod motion when the reactor is not critical.
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PALISADES NUCLEAR PLANT Proc N3 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 3 of 10 TITLE: POST TRIP REVIEW REQUIREMENTS 2.
Significant plant transients other than reactor trips (eg, Engineered Safeguards Feature actuations, significant feedwater or turbine transients, significant PCS pressure, level or temperature transients, etc).
3.
Any Condition II event.
4.6 SHIFT SUPERVISOR /0PERATIONS a.
The Shift Supervisor / Operations (SS) shall:
1.
Assist the SE with data cc.llection as Plant conditions allow.
2.
Make recommendation to Plant General Manager on reactor startup.
3.
Ensure Plant shutdown actions are completed after the reactor trip.
4 Maintain the Plant in a safe condition.
b.
The Shift Supervisor / Operations (SS) should notify the Operations Superintendent of:
1.
Any unplanned reactor trip involving control rod motion when the reactor is not critical.
2.
Significant plant transients other than reactor trips (eg, Engineered Safeguards Feature actuations, significant feedwater or turbine transients, significant PCS pressure, level or temperature transients, etc).
3.
Any Condition II event.
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PALISADES NUCLEAR PLANT Proc No 4.08 ADMINISTRATIVE PROCEDURE Revision L Page 4 of 10 TITLE:
POST TRIP REVIEW REQUIREMENTS 5.0 PROCEDURE 5.1 GENERAL INSTRUCTIONS 5.1.1 Post Trio Review Process i
The post-trip review process is a five-step process. The five steps are as fotlows:
Step Responsiblity (1) Data collection Shift Engineer (2) Trip investigation Shift Engineer (3) Restart decision Plant General Manager (4) Trip investigation review Plant Ceneral Manager /PRC (5) Independent assessment /
Plant Safety Engineering j
comparison with past trips 1
for trends.
5.1.'2 Initiation Any unsch3duled reactor trip when the reactor is critical shall require the post-crip review process to be initiated. For unplanned reactor trips involving control rod motion when the reactor is not critical, or for plant transients other than reactor trips (eg, s
Engineered Safeguards Feature actuations, significant feedwater or turbine transients, significant PCS pressure, level or temperature transients, etc), the post trip review process may be initiated as directed by the Operations Superintendent. The post-trip review shall be initiated after plant conditions have stabilized. The Post-Trip Review Report, Attachment 1, guides and documents the post-trip process. The post-trip review shall not distract the Shift Supervisor, Shift Engineer, or operating personnel from their primary responsibility of monitoring plant parameters and maintaining the plant in a safe condition.
5.2 DATA COLLECTION 5.2.1 Hard Copy Information The Shift Engineer is responsible for the collection of required hard copy information. The purpose of this data is to reconstruct the transient from prior to the initiating event until plant parameters have stabilized. Part 2(a) and (b) of the Post-Trip Review Report, j, lists the information that should be collected. Strip chart recordings must accurately reflect real time to have meaningful information.
If this is not the case, the SE should ensure the chart paper is annotated with a time mark,' chart speed (chart speed may change during the transient), and time scale.
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PALISADES NUCLEAR PLANT Proc N3 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 5 of 10 TITLE: POST TRIP ItEVIEW REQUIREMENTS 5.2.2 Operator Statements After-the plant is in a safe, stable condition, the SE should ensure each individual involved in the trip (eg, reactor operator, repair-man / technician, Maintenance Supervisor, etc) provides a statement concerning his/her involvement in the reactor trip. These statements may be obtained in one of the following ways:
Written operator stacaments (use Attachment 2).
a.
b.
SS/SE interviews with personnel involved in the reactor trip.
i c.
Critique with all involved personnel.
If either of the last two techniques are used, the information should be recorded to ensure future availability of the information. The operator statements should be restricted to facts concerning the event, and the facts should be stated chronologically, if possible.
The statement should include the following:
a.
Plant conditions prior to the trip (for maintenance personnel, this will include the status of maintenance or testing).
b.
First indication that a problem existed.
c.
Individual's actions as a result of the indications.
d.
Subsequent indications and plant response, including manual actions.
e.
Noted equipment malfunctions or inadequacies.
f.
Procedure deficiencies identified during the situation.
g.
Recommendations to prevent recurrence.
The signed written statements or signed transcripts of tape recordings shall be included in the reactor trip data package to assist in the event reconstruction.
5.2.3 Post-Trip Review Repore Data The SE shall complete Part I and Part 2 of the Post-Trip Review Report, documenting the initial plant conditions and the plant response.
Information for the Post-Trip Review Report shall come from a compilation of available data.
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PALISADES NUCLEAR PLANT Proc N3 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 6 of 10 TITLE:
POST TRIP REVIEW REQUIREMENTS 5.2.4 Post-Trip Review Data Package The SE should collect the hard copy information, operator statements, and the Post-Trip Review Report. This wiLL be the post-trip review data package that will be used during the post-trip investigation.
5.3 POST-TRIP INVESTICATION 5.3.1 Event Reconstruction The Shift Engineer is responsible for the initial post-trip investiga-tion. The purpose of this investigation is to determine the cause of the trip, verify proper plant response, and to assess the plant's readiness to return to operation. The SE will reconstruct the tran-sient in Part 3 of the Post-Trip Review Report, Attachment 1, using the collected data. A chronological description of the event will be developed, using all available data. Pertinent alarms, trips, actu-ations, and isolations will be listed or marked on the sequence-of-events or Feedvater Purity Data Logger printout. Pertinent plant parameters should be incorporated into the chronological list of events during the reconstruction.
5.3.2 Analysis and Evaluation The Shift Supervisor, and the Shift Engineer shall analyze and evaluate the event reconstruction to determine the cause of the trip and the following:
a.
If all major safety-related and other important equipment involved in the trip operated as anticipated or expected.
b.
If the trip / transient caused any detrimental effects on plant equipment.
c.
If it is acceptable to restart the reactor.
i The Shift Supervisor and the Shift Engineer should look beyond the obvious indications to diagnose the cause of the trip and evaluate the plant response. They should review the available information thoroughly, looking for:
a.
abnormal indications or degraded trends in equipment performance, b.
events occurring out of the normal or anticipated sequence, c.
failed or degraded response of equipment to control signals, d.
unusual chemistry results or esdiation readings, and e.
unanticipated alarms.
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PALISADES NUCLEAR PLANT Proc No 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 7 of 10 TITLE:
POST TRIP REVIEW REQUIREMENTS The actual or suspected cause of the trip and any abnormal or degraded indication identified during the transient shall be documented in Parts 4 through 6 of the Post-Trip Review Report.
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5.3.3 Preliminary Safety Assessment A preliminary safety assessment of the trip and subsequent plant response shall be performed by the Shift Supervisor and the Shift Engineer. The maximum and minimum values of selected parameters shall be compared with their established limits in the Technical Specifi-cations. Parts 7 and 8 of the Post-Trip Review Report will document this safety assessment.
5.3.4 Trio classification Based on the results of the analysis and evaluation of the plant trip and subsequent response, the Shift Supervisor and the Shift Engineer j
shall classify the event as one of the following conditions and l
document their classification in Part 9 of Attachment 1:
a.
Condition I The cause of the trip is positively known and has been corrected; all safety-related and other important equipment functioned properly during the trip, or if not, the malfunction has either been corrected l
or redundant equipment is available for reactor start-up.
b.
Condition II 2
1.
The cause of the trip is not positively known.
Or 2.
Safety related equipment or other equipment functioned in an abnormal or degraded manner during the trip and the malfunction has not been corrected.
Or 3.
Safety related equipment or other equipment functioned in an abnormal or degraded manner during the trip and redundant equipment is not available for startup.
If the Shift Supervisor and Shift Engineer cannot agree on classifica-tien of the transient, the event shall be classified as Condition II and referred to the Operations Superintendent for evaluation.
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PALISADES NUCLEAR PLANT Proc N3 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 8 of 10 TITLE:
POST TRIP REVIEW REQUIREMENTS 5.3.5 Notifications once~the reactor trip event is classified, the Shift Supervisor / Shift Engineer shall inform the Plant General Manager of the classification of the event, eneir recommendation for plant startup, and the Plant Safety Engineering investigation findings and recommendations.
If the i
event is classified Condition II, the S3/SE shall also inform the Operations Superintendent.
If necessary,_ emergency plan notifications shall be made as required in References 3.6 and 3.8.
5.4 RESTART DECISION The Plant General Manager has the responsibility and authority to grant permission to commence a reactor startup following a reactor trip. This permission shall be documented by the Shift Supervisor or Shift Engineer in Part 11 of the Post Trip Review Report, Attachment 1.
5.4.1 Condition I Events The Shift Supervisor / Shift Engineer shall inform the Plant General Manager if the unscheduled reactor trip event is classified as Condition I.
Based upon this classification, the SS/SE recommends restart of the reactor.
5.4.2 Condition II Event The Shift Supervisor / Shift Engineer sbell inform the Plant General l
Manager when an event is classified Condition II.
The Operations Superintendent or his designated alternate shall report to the plant site to assist in the further investigation of the trip and to determine necessary corrective action before restart.
The Operations Superintendent will analyze the event reconstruction, emphasizing the root cause of the trip and the resolution of abnormal or degraded indications. He should use available expertise to resolve questions concerning the cause and plant response. Sources of exper-tise that should be considered include nuclear steam supply vendors, vendor engineers, onsite engineering staff and other experienced operations and maintenance personnel. The Operations Superintendent should supply the following information to the Plant Manager and document it in Part 9 of the Post Trip Review Report:
a.
The actual or most probable cause of the trip.
b.
The maintenance and testing necessary before reactor restart including additional measures to verify the most probable cause.
Additional monitoring or trending required during and/or after c.
reactor restart.
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PALISADES NUCLEAR #LANT Proc No 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 9 of 10 TITLE:
POST TRIP REVIEW REQUIREMENTS d.
Necessary briefings to operations and/or maintenance personnel concerning specific equipment indications or possible salfunctions.
e.
The conditions necessary for a reactor restart.
5.4.3 Plant General Manaaer Evaluation and Decision The Plant General Manager should evaluate the recommendation made by the personnel performing the trip investigation. The Plant General Manager should consider convening the PRC to review the trip investi-gation prior to reactor restart for Condition II events. The PRC can advise the Plant Ceneral Manager on the thoroughness, technical accuracy and consistency of the trip investigation.
The Plant General Manager's decision to restart the reactor shall include the following considerations:
a.
The most probable cause of the trip is known and corrected.
b.
Major safety-related and other important equipment functioned properly during the transient, or corrective maintenance and satisfactory testing has been performed or will be completed when plant conditions permit.
c.
The plant response during the event has been analyzed and the plant responded as anticipated, or all abnormalties are understood and corrected as required by Technical Specifications.
d.
If the cause of the trip has not been positively identified, the Plant General Manager shall determine if the cause and the circum-stances surrounding the cause have been analyzed adequately. He shall take adequate measures to prevent repetitive challenges to safety systems during future power operations.
5.5 INVESTICATION REVIEW The Lead Shift Engineer, Operations Superintendent, Plant General Manager and PRC Chariman shall document their reviews in Part 12 of the Post-Trip Review Report.
5.5.1 Condition I Events Condition I event trip reports shall be reviewed by the Lead Shift Engineer, Operations Superintendent, and Plant General Panager within two weeks of the trip. The report should be reviewed by the PRC during the next regularly scheduled PRC meeting or routed for review.
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A PALISADES NUCLEAR PLANT Proc Na 4.08 ADMINISTRATIVE PROCEDURE Revision 1 Page 10 of 10 TITLE: POST TRIP REVIEW REQUIREMENTS 5.5.2 Condition II Events i
If diTected by the Plant General Manager, a Condition II event shall be reviewed by the PRC before a reactor restart is commenced. Other-wise, a Condition II event will receive a PRC review during the next i
regularly schedultl PRC meeting or routed for review.
5.6 PLANT SAFETY ENGINEERING ASSESSMENT PROGRAM Plant Safety Engineering will perform an independent assessment of each unplanned reactor trip as addressed in Reference 3.7.
The Lead Assessor shall provide a specific Plant Safety Engineering conclusion related to acceptability of Plant Startup or continued operation. This conclusion shall be documented in Part 10 of the Post-Trip Review Report.
6.0 ATTACHMENTS AND RECORDS 6.1 ATTACHMENTS 6.1.1, " Post-Trip Review Report" 6.1.2, " Plant Personnel Statements" 6.2 RECORDS The post-trip review data packages shall be retained for the life of the plant. This will allow the data package to be available for comparison with future trips.
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Pese N3 4.08 4
Revision 1
)
2 Page 1 of 14 POST-TRIP REVIEW REPORT Date of Occurrences f//9/d Time of Occurrence:
)y/[
PART 1: INITIAL CONDITIONS
)7 a.
Reactor Power b.
Reactor Coolant Pumps Operating (circle)
B C
c.
Main Feed Pump (s) Operating (circle) d.
Status of Control Stations (circle) 1.
Turbine Control Man derAut)
@ Amp Out 2.
A Feedpump Control Man Auto Ek &
h B Feedpump Control Man 3.
A Main Feedwater Reg Valve Man B Main Feedwater Reg Valve Man 4.
Turbine Bypass Valve (**CV-0511)
Man 5.
Pressurizer Level Control Channel A/B Man 6.
Pressurizer Pressure Control Channel A/B Man e.
PORV Block Valves (circle) OPEN f.
Off-normal Status (of any trains / portions of safety systems)
DETAILS s
1.
Tri t.p byf,2<<J 2.
Safety Injection Actuation System
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Proc NS 4.08 Revision 1 Page.2 of 14 POST-TRIP REVIEW REPORT DETAILS 1.
Containment Cooling & Spray h - *tbtf,1 a 4 a u%. /
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Auxiliary Feedwater r70 a $ #,er,
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Testing /Surveillances in progress (include test i and tesc status):
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Proc N3 4.03 Revision L Page 3 of 14 POST-TRIP REVIEW REPORT PART 2: PLANT RESPONSF.
Obtain a copy of the applicable parameter plots given below:
a.
Check If Included Parameter 1/
- 1. Rx Power t/
- 2. Pzr Level V
- 3. Par Pressure V
- 4. Tave V
- 5. Tc WR
/ 6. SC Levels b.
Obtain a Printout from:
/
- 1. Tennecomp Data Logger
- 2. Event Recorder V
- 3. Critical Function Monitor: Obtain hard copies of pertinent parameters to include, at a minimum, S/G "A" pressure (Point ID **PT-07518) and S/G "B" pressure (Point ID **PT-07528) c.
Obtain copies of pertinent logs from past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Chemistry, Control Room, Auxiliary Operator, etc).
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Safety System Actuation and Performance 1.
Reactor Protection System Type of Trip:
Niyl kreuur/Eu [reirure. Time of Trip:
/ Y lf.
2.
Safety Injection Actuation System Actuation Time
[
l Actuation Signal:
System Response ado 285-0014b-93-89
Pete No 4.08 Attcchsent 1 R visien i Page 4 of 14 POST-TRIP REVIEW REPORT 3.
Containment Isolation Actuation Time l\\)/f _:
Actuation Signal System Response:
h 4.
PZR Code Safety Valves Lifted YES Did they resent properly?
YES NO How Confirmed?
S/k frsstures esstelsA
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5.
Main Steam Relief Valves Lif ted Sakty L.{t set /cinti Nu jfressuer IO W /k A Did they resent properly?
t NO How confirmed?
I)i su al ad h Mt tem,m tw se ressn-se i
e.
Control System Actions 1.
Turbine Trip Trip Time 19 : /b Trip Signal fS Tril
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2.
PCS Pressure Control h
L d PORV lift?
YES Was PZR heater response normal?
h NO Was PZR spray response normal? co-resY 4/ het jo [//p YES h
c los e]
ormal?'[C4/ ba) ban sc4ure YES ses,tefontrol /p $e( A tr /
Was PZR level was l
Nou f.
Manual Actions Were any control star. ions taken from Auto to Manual?
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Prac No 4.08
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Revision 1 l
Page 5 of 14 I
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EVENT DESCRIPTION:
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Pr:c N3 4.08 Attcchment 2 Revision 1 Page 1 of 1 PLANT PERSONNEL STATEMENT (Include the Plant conditions prior to the trip, your indications that a probles existed, your action as a result of those indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies. Also, include any information you consider important to review this unscheduled reactor trip. Use additional sheets as necessary).
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Pete N3 4.08 Revision 1 Page 1 of 1 PLANT PERSONNEL STATEMENT (Include the Plant conditione prior to the trip, your indications that a probles amisted, your action as a result of those. indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies. Also, include any information you consider important to review this unscheduled reactor trip. Use additional sheets as necessary).
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Prce ND 4.08 Attcchment 2 Revision 1 Page 1 of 1 PLANT PERSONNEL STATEMENT (Include the Plant conditions prior to the trip, your indications that a
problem existed, your action as a result of those indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies.
- Also, include any information you consider important to review this unscheduled reactor trip.
Use additional sheets as necessary).
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/ Dice Time ado 285-0014c-93-89
Pres Na 4.08 Attcchment 2 Revision 1 Page 1 of 1 PLANT PERSONNEL STATEMENT (Include the Plant conditions prior to the trip, your indications that a probles existed, your action as a result of those indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies. Also, include any information you consider important to review this unscheduled reactor trip. Use additional sheets as necessary).
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Prce N2 4.08 Attcchment 2
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Revision 1 Page 1 of 1 i
i PLANT PERSONNEL STATEMENT (Include the Plant conditions prior to the trip, your indications that a probles existed, your action as a result of those indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies.
- Also, include any information you consider important to review this unscheduled l
reactor trip. Use additional sheets as necessary).
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Sig6ature Date Time ado 285-0014c-93-89
Prac N2 4.03 Revision 1 Page 1 of 1 PLANT PERSONNEL STATEMENT (Include the Plant conditions prior to the trip, your indications that a
problem existed, your action as a result of those indications, noted equipment malfunctions or inadequacies and any identified procedure deficiencies. Also, include any information you consider important to review this unscheduled reactor trip. Use additional sheets as necessary).
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Prac Na 4.08 Attechnent 1 Revision 1 Page 7 of 14 POST-TRIP REVIEW REPORT PART 4: PROBABLE CAUSE OF TRIP
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j ots PART $
UNEXPECTED ASPECTS OF TRANSIENT BEHAVIOR (if event compared with previous similar
-transient, note the transient with which compared)
Compared With Previous trip on: 1/M/O/0'f k
'Date Tirae befe U tr e.
he U n es pecie Aspec ts b frdwc,*es. T I
l
[
O t) W ado 285-0014b-93-89
6 tot No 4.08 Revision 1 Page 8 of 14 POST-TRIP REVIEW REPORT j
i PART 6: IDENTIFICATION OF SYSTEMS WITH INADEQllATE PERFORMANCE System / Component Description of Problem L U ~ O T//
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l Parts 3 thru 6 reviewed by:
pu +</u m Shift Supervisor
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Signature Date Time j
Shift Engineer
/ J/////( / 2 2 3 0 Sign'ature Date Time ado 285-0014b-93-89
Prce Na 4.08 Revision 1 Page 9 of 14 POST' TRIP REVIEW REPORT PART 7: TRANSIENT DATA FOR PERTINENT PLANT PARAMETERS Maximum Minimum a.
PCS Pressure Ch 1 2250 2 13 r1 Ch 1 R 40 2
/6 IfC b.
PCS Average Ch 1 SN 2 379 Ch 1 833 2 8'37 Temperature c.
S/C Press S/C A lO 15 8 to i9 S/C A 4 s.O 8 TSY d.
S/C Level S/C A CS 8
70 S/C A ll 8
l1 I
PART 8: PRELIMINARY SAFETY ASSESSMENT I
a.
PCS Pressure Remained Above Setpoint for -
Automatic SI Actuation.
NO b.
PCS Pressure Remained Below Setpoint for PZR Code Safety Valve Actuation.
h NO c.
PCS Temperature Decreased Less Than 100* F/hr h
l (Tech Spec).
YES d.
Reactor Coolant Was Contained Within The PCS and Quench Tank.
NO e.
Indicated PZR Level Remained On Scale.
NO f.
Indicated SC Level Remained On Scale (vide range and narrow range - specify below if either went off scale.)
NO Comunents:
PCs
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1 Pece Na 4.08
^
Revision 1 Page 10 of 14 POST-TRIP REVIEW REPORT PART 9 Pl. ANT EVALUATION AND RECOMMENDATION Classify trip as Condition I or Condition II according to guidelines in a.
procedure.
The event on f//f/ff at 19 : d is a condition I
event Date Time I,II Signature indicates agreement with c ndition k"'l'. A ~oenm 7
I <~Misil 3930 Shift Supervisor Date Time 3
I I//f/#I J-O O SKift Engineer Date Time l
Comments and recommendations:
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/
l b.
Operations Superintendent's Analysis (for Condition II events only) 1.
Actual or most probable cause of trip 60l-fnw / Cc,a sala
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r t'2,' O n$ tL Frinar, power ru nlu was C-c u w n--
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..i
Prec N3 4.06 Revision 1 Page 11 of 14 POST-TRIP REVIEW REPORT 2.
Maintenance and testing required before reactor restart.
h one 3.
Additional monitoring or trending required during and/cr efter reactor restart.
% etLa +tstiue Ars tw t/m e re lie L;111s ci 8 HC fru u 7
f l
ses de1m a Y crr's r to eu.ceed.*sa f e., ss/i er
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- A N f_ m*
r 4.
Necessary briefings to operations and/or maintenance personnel concerning specific equipment indications or possible malfunctions.
Naus S*
CALs,e't C <ss 1. +, a -be Lc tu i s, e it s
v w/ueeaY laek ta orr Us. a l J au w fe.as}e o' / heV" is i
v rr r n ae s-/
6~ Gr 1
e 5.
Conditions necessary for a reactor restart (other than listed above) ktew
/vr D4cm b [m-6 J / tdt a.p
/
Operations Superintendent ado 285-0014b-93-89
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Free N3 4.08 tavision 1 POST-TRIP REVIEW REPORT PART lit PERMISSIQW TO START UP Plant Ceneral Manager notified and permission granted to start up the reactor, Y%an g> >.-
Y_O IM SQl(
Shift Engineer ir Shift Supervisor Dacd Time DAN kb A
A
~nilo Comments:
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Proc Na 4.08 Revision 1 Page 14 of 14 POST-TRIP REVIEW REPORT PART 12: INVESTICATION REVIEW d n tr u e y a A R P P
review conducted:
igen-einp-rgro dM:ir:it M WdO ost-Tetp Report W^'""-
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b - $4.4 hwlk - wb J
l I
Y!JO
, meeting number:
PRC review of event on:
Minutes of the meeting (s) ari attached
/
PRC Chairman Date Coment s:
l i
ado 285-0014b-93-89
? roc No 'O.4L Aetacr. ment 5 Revision 7 CHANCE NUMBER M'(b -~@I
/
!!M?ORARY CHANCE TO A PROCEOURE
&M4-
// //v '//,f'J-Procedure Type fs N,sn + Procedure No el-/
Rev No/3 ate O/Getty /ts,./96uuy crowsWk Procedure Title
//enufw d Dk TITE' Paragraph No Attachment
.(
ATMS No di 07/#-(2*pf& /3 y -/f Reasan foe chis change 1 15
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Recommend change be made permanent? h No) If NO, expiration date No);
Does the Procedure require PRC review per Administrative Procedure 10.42? g %
The above meets the cetterta on the reverse side
/ J~)b2411
/
Initiator Date Approved Approved PRC Membe(
Date SRO-PRC Mehber
/Dat e /
Change to become permanent? Yes No PRC Review Meeting Number
/
Department Head Date ado 582-0003f-93-154
APPENDIX C PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO THE Mt.Y 19, 1986 REACTOR TRIP l
1 4
4
. =,. _.
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-1 COMPONENT:
Turbine Generator Electro-Hydraulic Control (EHC) System Power Supplies PROBLEM DESCRIPTION:
On May 19, 1986 reactor / turbine trip occurred. Cause of the reactor trip was high pressurizer pressure, although the initiating event was inadvertent closure of the turbine control valves.
1 After the reactor / turbine trip the +15 volt power supplies in the turbine control panel were found to be tripped off, (both the primary and secondary j
power supplies). The turbine control system is dependent on the +15 vole power supplies for operation, therefore with the loss of both +15 volt supplies the control valves are expected to fail in the closed position i
as they did.
NOTE:
The pcwer supply system is designed to be redundant. The primary power supply is powered from the lighting panel 120 volt 60Hz. The secondary is powered from the permanent magnet generator at 400Hz.
4 4
HISTORY:
]
The original power supplies for the turbine control panel were replaced l
in 1981. They were changed because they were a high maintenance item, and Westinghouse recommended replacement.
1 MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-1 Ther'e is no history of any problems with the new existing power supplies since their installation in 1981.
SIGNIFICANCE TO OPERATIONS:
Failure of both power supplies (ie primary and secondary) can directly cause s loss of turbine control, and indirectly cause a loss of control for turbine-generator auxiliary systems and instability in the secondary i
plant presenting a challenge to the reactor protection systems.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
The turbine EHC system is a " fail safe" system. If either the electric or hydraulic power is lost, the turbine is shutdown by spring closure of the turbine steam admission valves. This is the desired action for a reactor trip or for any accident sequence.
Sudden closure of these valves during power operation however, imposes a transient on the plant and requires a reactor trip, a transfer of AC power to the off-site supply, initiation of auxiliary feedwater and a turbine-generator trip. During the 5/19/86 trip, all of the above safety features functioned correctly.
1 MIO686-0566A-MA05-MD02
d PALISADES PLANT EQUIPMENT PROBLEMS i
RELATED TO MAY 19 1986 REACTOR TRIP C-1 STATUS / ACTIONS:
I i
Status J
1]
It has been determined that the loss of the +15. volt power j
supplies was due to a voltage transient (spike). The voltage f
spike into the power supplies caused them to trip on an over-1 voltage condition. The spike was caused by the unplugging of
.t cooling fans located in the turbine control panel, (a PM was i
being performed which requires the cleaning of the fan filters).
During the investigation of the power supply failures the I&C f
Department was able to duplicate the conditions of up to and i
proceeding the turbine trip. This was done by duplicating the i
j voltage spike (unplugging fans) and observing the power supplies tripping on over-voltage. This provides a high confidence level in identifying the cause of the failure.
1 J
2]
During the investigation it was found that the 200 volt power j
supply was not functioning properly due to capacitors failing.
i a
3]
A group from CPCo Lab Services performed an independent investigation, and was able to record voltage spikes on the system while the 1
{
fans were being unplugged.
I i
l 4]
A Westinghouse employee also performed an investigation of the power supplies and concurred that unplugging cooling fans affected i
the power supplies.
MIO686-0566A-MA05-MD02 i
1
... ~~
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-1 Actions 1]
Wiring from the circuit breakers to the power supplies inside of the EHC cabinet has been rerouted through conduit to provide physical separation.
2]
The ventilation fans have been rewired through a switch (formerly plugged in) and powered from a supply external to the EHC cabinet.
3}
Voltage suppression filters have been installed on the AC input side of the power supplies.
4]
The 200 volt D-C power supplies are being repaired (faulty capacitors.
5]
Investigate why the other power supplies (+48, -15 Volt) were not affected by the voltage transient.
6]
Determine if a form of voltage filtration is available and would be effective in preventing a similar reoccurrence.
7]
Investigate the options available to p'rovide an audible alarm in the control room on the loss of a power supply.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-1 8]
Prior to start-up simulate the turbine running to insure no logic cards were damaged during the voltage transients.
RESOLUTION (REFERENCE OBSERVATION NO TGS-16):
Prior to Plant Start-up:
l f
1]
Repair the 200 volt DC, 60 Hz power supply.
I i
2]
Remove the fans from the primary 60 Hz supply and provide power from an external source.
Ensure fan source does not induce transients into power supply voltage sources.
3]
Install transient suppressors on power supply input connections.
4]
Install circuit breaker lock on the AC power source to the primary power supplies at the AC panel to prevent inadvertent opening.
)
5]
Verify proper operation of the -15 VDC supplies; check voltage regulation and ripple.
i 6]
Verify proper operation of the +15 VDC supplies; check voltage regulation and ripple.
1 MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-1 7]
Modify the outside air cooling duct to prevent moisture incursion and resultant corrosion in EHC cabinet.
Before the End of REFOUT '87:
1]
Install an audible alarm in the control room for turbine panel trouble (
Reference:
Observation TGS-11).
2]
Evaluate the effects of aging on the power supplies; replace if necessary.
As Part of the 5-Year Plan:
1]
Evaluate the addition of backup power supplies on each power source to provide double redundancy.
4 i
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-2 COMPONENT:
Turbine Lube Oil Lift Pumps PROBLEM DESCRIPTION:
Following turbine-generator trip and subsequent coastdown on May 19, 1986, the generator bearing lif t pump (P-28), the LP turbine lif t pumps (P-124A, B, P-125A, B) and the exciter house fan did not automatically start at 600 RPM per design. The turbine was allowed to coast to less than 300 RPM when the above equipment was started manually.
HISTORY:
The original design of the turbine-generator had a lift pump only for the two generator bearings. Historically this pump did not supply. adequate pressure to lift the generator shaft. This problem was repaired during the 1985/86 maintenance outage when a leakage path was identified and repaired by installing a gasket. Since startup, this system / equipment has worked very well.
LP turbine lift pumps were installed as a Facility Change in 1981 and were subsequently upgraded per a Specification Change in 1983. The reliability of these pumps has been questionable at best. During post-maintenance testing in 1984 and 1986 these pumps did not provide MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-2 adequate pressure to lift the shaft. On both occasions the pump and piping assemblies were disassembled.
In 1984 no problems could be identified, and, in fact, the pumps started providing adequate pressure later on for no apparent reason.
In 1986 the system o-rings were found to be very brittle and broken. Also, some pump casing bolts had failed.
New pumps and o-rings were installed. They have worked as intended since that time.
l On a previous turbine trip the pumps did not come on, I&C investigated l
and repaired the 14X/SD switch. The switch was verified to work properly i
)
on subsequent plant startup.
4
-l i
It should be noted that according to Westinghouse, operation of the lift pumps during coast down and turning gear operation is not necessary to protect bearings and journals f rom damage although they do provide extra protection that increases bearing and journal life.
SIGNIFICANCE TO OPERATIONS:
l Failure to start during a plant trip results in the need for an operator i
j to start the pump manually.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
The turbine lube oil lif t pumps are provided to ensure establishment of a bearing oil film prior to placing the turning gear in operation during l
MIO686-0566A-MA05-MD02
t l
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-2 startup. The pumps are not needed during a plant trip, nor during any accident sequence. The automatic start of the lift pumps at 600 rpm i
during shutdown is provided only as insurance against bearing rubs at very low shaft speed, such as that experienced during turning gear operation.
2 STATUS / ACTIONS:
Status 1]
Electrical drawings have been reviewed with I&C and it has been
{
determined that the problem was the turbine speed pickup (14X/SD) which is supposed to close at 600 RPM and thus auto start the bearing lift pumps and exciter house fan. A post maintenance test is currently being written to assure operability. A Jumper, Link and Bypass has been initiated to keep the turbine on turning gear during cooldown and repair of the EHC cabinet power supplies.
1 2]
Investigations by I&C have shown the reason that the 14X/SD speed pickup did not energize was because this contact is powered i
from the EHC cabinet. The power supplies to this cabinet were lost, which initiated the plant trip. This subsequently prohibited I
the auto start of the bearing lift pumps.
I 3]
Post-maintenance testing of these pumps is performed at the i
completion of each refueling outage. The testing involves the
,l MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-2 verification of the performance of the mechanical function only.
Testing'is not performed to verify the auto start feature is operable. The test of the auto start feature at 600 RPM will be added to the post-maintenance test.
4]
Westinghouse has performed an independant check and has concurred that the cause of the auto start failure was the loss of the EHC power supplies. Westinghouse is also investigating lift pump reliability at other plants.
Actions 1]
Develop tests which check the auto start feature of the lift pumps.
2]
Monitor lift pump pressures during shutdown and note any changes or problems.
3]
Remove and inspect lift pump asamblies at next refueling outage.
4]
Evaluate replacing the generator lift pump cuno filter with a i
cartridge type filter.
(The LP turbine lift pumps already have cartridge filters.)
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-2 5]
Develop Preventive Maintenance for the lift system. Consider such areas as:
a]
Replace LP lift pump suction filters each refueling, b]
Replace LP lift pump discharge filters each refueling, c]
Replace generator bearing lift pump cuno filter with a cartridge type filter.
Replace filter each refueling outage.
i d]
Investigate possibility of an on-line test of 14X/SD and implement if possible.
RESOLUTION (REFERENCE OBSERVATION No TGS-10):
Prior to Plant Startup:
1]
Test and ensure the lift pump starts at 600 rpm on a turbine trip.
As Part of the 5-Year Plan:
1]
Evaluate the reliability of the lift system and replace as required.
MIO686-0566A-MA05-MD02 i
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PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-3 COMPONENT:
CV-1059, Pressurizer Spray Valve PROBLEM DESCRIPTION:
1] Valve did not fully close following plant trip.
2] History of packing leaks.
3] Attempt to rebuild past outage.
HISTORY:
Document review of the work history of CV-1059 and the other spray valve, CV-1057, reveals a chronic problem with packing leaks. There is no I
history to suggest that operability has been unreliable.
During Refout '85, CV-1057's bonnet and operator were removed to facilitate a complete repacking effort. The internals were cleaned and packing replaced.
Attempts were also made to clean and repack CV-1059 using the same methods, however, the degraded condition of the operator yoke and yoke MIO686-0566A-MA05-MD02
i 1
PALISADES PLANT EQUIPMENT PROBLEMS
^
RELATED TO MAY 19 1986 REACTOR TRIP C-3 4
j nut prevented disassembly of the valve. A new yoke nut was not available and therefore, we could not remove the old one because it was in such 4
poor condition.
Three rings of packing were installed on top of the lantern ring, i
I 4
i, Due to the history of chronic packing leaks, the position indicator switch mounting bracket was severely corroded. This condition was identified during the maintenance performed in the refueling outage.
The bracket could not be replaced at that time because the operator is 4
required to be removed l
SIGNIFICANCE TO OPERATIONS:
1 The component is important to reliable plant operation.
Its failure can i
{
result in a high probability for challenging safety systems by axcessive cool down rate and can result in operator distraction, a plant trip or consequences unfavorable to plant operations during a trip.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITICNS:
i The pressurizer spray valves ara modulated to maintain primary coolant l
system (PCS) pressure during normal operation. The valves would only 1
rarely be fully shut. The valves not only provide spray flow for pressure j
control but are also required to maintain the pressurizer boric acid 4
j concentration.
I I
M10686-0566A-MA05-MD02
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-3 During those accident sequences where PCS pressure increases, (continuous rod withdrawal, rod ejection, turbine trip without reactor trip) these valves would open and assist the safety valves in limiting PCS pressure.
No credit is take'n for this action in our safety analysis. During those accident sequences where PCS pressure decreases, (loss of coolant, steam
+
line break, steam generator tube rupture) the spray valves would close because of decreasing pressure.
A large LOCA or MSLB, however, not only quickly empties the pressurizer overriding any effect of spray flow, but requires tripping of the primary coolant pumps which terminates spray flow by removing the driving head.
1 A steam generator tube rupture requires rapid reduction of PCS pressure in order to reduce primary to secondary flow rate. The pressurizer spray valves would be opened by the operator.
In those accident sequences where off-site power is assumed to be lost, spray valves have no effect due to lack of PCP head.
STATUS / ACTIONS:
Status 1]
Initial inspection of the spray valve in the "As Tripped" condition indicated that the valve was at least 95% closed.
M10686-0566A-MA05-MD02
PALISADES PLANT EOUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP i
C-3 2]
The close micro switch mounting bracket was corroded and has slipped slightly which contributed to the lack of close indication 1
in the control room.
3]
The control air pressure was checked by Instrument and Control and while it is calibrated to 3 PSI (+/-l PSI) per the calibration control (Spec) sheet, in its "As Tripped" condition the control i
j air pressure was observed at 5 PSI using an installed and uncalibrated i
gauge. This is within the "As Found" acceptance criteria indicated l
on the calibration spec sheet.
4]
Per the calibration specification sheet which is used to set up and calibrate the valve controller, an "As Found" control air pressure of 3 PSI (+/- 3 PSI) is within the acceptance criteria.
It should be noted that the relationship of control air pressure to valve position is linear and that for each 3 PSI of control air, the valve position changes 25%. This implies that given the i
expected drift, the valve should not be expected to be positively closed or open, even with a full signal from the control room. or from the pressure transducer.
5]
Packing leaks had contributed to the corrosion of the position switch bracket.
1 i
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-3 j
Actions 1
1]
CV-1059 has been disassembled and rebuilt to eliminate packing leakage.
2)
A new position switch mounting bracket has been fabricated and installed.
3]
An inspection of CV-1057 Pressurizer Spray Valve, has revealed this valve to be in excellent condition.
4]
A hot post-maintenance test has been developed and approved by operations to verify success of the repair effort.
5].
Evaluate calibration specification setpoints and. acceptance criteria for the Pressurizer Spray Valves.
6]
Include PZR spray valves -3 candidates for live loaded or EPRI type packing.
RESOLUTION (REFERENCE OBSERVATIOh" No. PCS-01):
Prior to Plant Startup:
1.
Valve should be disassembled and inspected or physical obstructions that would prevent valve from going fully closed.
D-PAL-84-212 identified foreign material and a cocked seat ring for CV-1057.
M10686-0566A-MA05-MD02
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PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-3 r
2.
Inspect and repack CV-1057 in accordance with established Maintenance Procedure (PCS-M-8).
3.
Both CV-1057 and 1059 have had considerable packing leaks that 4
could have sprayed the valve auxiliaries (ie, switches, stem, actuator, etc.) with borated water. Valve auxiliaries should be thoroughly inspected with those demonstrating significant degradation (corrosion) replaced.
4 Upon return to hot shutdown conditions, CV-1057 should be repetitively stroked and packing re-adjusted. Also, open/close indication should be verified with each stroke.
5.
Calibrate positioners and E/P's for both CV-1057 and CV-1059.
6.
Prepare, review and perform a post-maintenance test.
Before the End of REFOUT '87:
1.
Establish PACS for packing adjustments during any future outages to prevent or minimize packing leakage.
(Current practice indicates no maintenance on valves until leakage occurs).
o As Part of 5 YEAR PLAN:
1.
Replace valves. Spare parts for existing valves are becoming difficult to procure.
MIO686-0566A-MA05-MD02
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PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-h COMPONENT:
Turbine Bypass Valve CV-0511 PROBLEM DESCRIPTION:
During the 5/19/86 trip sequence the turbine bypass valve failed to open in automatic or in manual.
HISTORY:
Past hist-~y of CV-0511 indicates repeated failure to operate properly, and has been a long standing operator concern.
Valve replaced in 1979 Valve would not open - May 1985 - Broken diaphragm. Replaced diaphragm.
Valve would not open - Nov 1985 - Booster not supplying enough air - to the actuator, rebuilt booster.
Valve would not close - March 1986 - Bushing gauled. Rebuilt valve and valve packing ring. Remachined bushing to original tolerance. Rotated valve from unsupported 45* to vertical position (SC-86-086).
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY L9 1986 REACTOR TRIP Ch Replaced valve spring with larger size.
This outage indicates that the larger spring installed was significantly larger then specified.
SIGNIFICANCE TO OPERATIONS:
Failure to operate hinders plant response on a trip or transient.
The valve is utilized during start-up and shutdown.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
The turbine bypass valve (TBV), CV-0511, provides two functions during accident sequences:
First, it provides a method of remov1ng decay heat so as to avoid repeated cycling of steam generator safety valves. Repeated cycling of a safety valve, over a long period, could lead to its sticking open. Redundancy for this function is provided by any of the four atmospheric dump valves.
Probabilistic risk assessment (PRA) models show that manual or automatic I
opening of one or more of these five valves will avoid excessive cycling of a safety valve.
j MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-4 Second, the bypass valve provides a " closed cycle" cooling path in the event of a tube rupture in a non-isolatable (on the steam side) steam generator.
In that event its use would allow cooldown without significant release to the atmosphere. Redundancy is provided by any of several manually opened paths to the condenser.
It should be noted that the TBV provides no function during accident sequences where off-site power is assumed lost (no condenser vacuum) or where the main steam isolation valve closss.
STATUS / ACTIONS:
Status Conducted a test to verify proper air pressure to the actuator.
Test results identified that the replacement spring was sized significantly larger than specified.
Specified size was 1700 #/ inch. Testing revealed replacement spring rate was 3200 #/ inch.
Westinghouse completed an evaluation of the current valve and operator.
Test results indicate the operator is sized correctly. To resolve current problems the actuator spring should be returned to the original size of 400#/ inch and a procedure developed for setting the bypass orifices. The valve is marginal in the center operating range due to stability of the controller. This is the range where the valve operates most frequently.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP Ch Identified a replacement valve with better characteristics than the current valve. Most improved characteristic is the balanced plug design vs the unbalance plug of the current valve. The new valve does not have the high pressure drop across the plug which was the cause for the original valves poor performance. A Masoneilan valve has been procured and hydro tested.
Installation will begin this week.
In addition to plug differences, a vendor performance evaluation was made based on NPRDS data base. Research of NPRDS indicates a high failure rate in industry among Copes-Vulcan control valves.
Actions 1] Complete installation of the new valve.
2] Develop overall test program which should indicate as a minimum.
a]
Post-Maintenance conditions.
Verify correct operation with no steam pressure. Compare with calculated air pressure.
Verify correct operation with full steam pressure. Compare with calculated air pressure.
MIO686-0566A-MA05-MD02
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PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP.
C-4 b]
Periodic testing.
Valve operation verified during each startup, shutdown and by periodic testing.
c]
Local monitoring that can be made by maintenance personnel.
d)
Post-maintenance testing with Operations personnel stationed in control room and at valve to confirm proper maintenance.
e]
Develop PM's and spare parts inventory to ensure valve can be maintained in optimum operating conditions.
RESOLUTION (REFERENCE OBSERVATION No MSS-01):
Prior to Plant Startup:
I 1]
Repair or replace.
2]
Check, adjust and insure operability.
31 Evaluate testing in conjunction with monthly turbine valve testing.
(Add to Operation Weekly Schedule)
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-5 COMPONENT:
Atmospheric Steam Dump Valve - CV-0779 PROBLEM DESCRIPTION:
CV-0779 failed to open on signal when turbine tripped 5/19/86.
HISTORICAL DATA:
Twenty-nine (29) MOs/W0s have been initiated on this valve since 1976.
The problems experienced from most common to least common are; packing, operating air discrepancies, diaphragm, seat leakage.
SIGNIFICANCE TO OPERATIONS:
Operability of the valve is important to reliable plant operation.
Its failure poses a hindrance to response during a plant transient.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS.
The atmospheric dump valves (ADV) CV-0779, CO, 61 and 82 provide a method of removing decay heat. During accident sequences, any of the four ADV's (or the turbine bypass valve) opening vill prevent continued cycling of a steam generator safety valve and thereby greatly reduce the probability of a stuck open safety valve. Redundancy is provided by the installation of four ADV's and one TBV.
MIO686-0566A-MA05-MD02 l
PALISADES PLANT EQUIPMENT PROBLEMS.
RELATED TO MAY 19 1986 REACTOR TRIP c-5 The ADV's are also used to cooldown, however,' numerous manually opened paths are available to dump steam for cooldown purposes.
STATUS / ACTIONS:
Status 1]
It was discovered through electrical trouble shooting that a polarity problem existed at the remote shutdown panel C-33, which prohibited operation of CV-0779 from that control panel.
2]
Ruptured diaphragas, showed evidence of prolonged leakage of the glycol damping fluid into diaphragm housing. This was discovered upon disassembly and inspection of the valve.
31 Vendor recommends change out of the diaphragms every three to five years.
4]
Diagnostic testing determined that the loss of glycol due to diaphragm leakage was root cause for the valve failure.
Actions 1] Ensure all valves test satisfactorily from remote shutdown panel C-33.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS
^
RELATED TO MAY 19.1986 REACTOR TRIP C-5 2] Diaphragms on all four dump valves are being replaced.
3] Post-maintenance testing at operating conditions is being developed.
4] Periodic maintenance procedures are being developed to change out diaphragms each four years.
5] A Deviation Report has been initiated to investigate root cause and implications of polarity problem on hand switch in remote shutdown panel.
6] Improved packing is being evaluated for the valves.
7] Testing Post maintenance testing will include " snoop" testing all air fittings for leaks, stroke testing valve to ensure 100% travel, verify limit switches are properly set, and inspect for glycol leakage.
Operability testing will be performed with full DP across the seats.
Valve stroke time will be checked and compared with the other 3 valves; packing, air, and glycol will be checked for leakage; valve leak through will be checked af ter reseating.
MIO686-0566A-MA05-MD02
l 5
PALISADES PLANT EQUIPMENT PROBLEMS-RELATED TO MAY 19 1986 REACTOR TRIP C-5 Periodic testing during the monthly turbine valve testing is being suggested.
If no problems occur after 6 months, then the frequency will be extended to quarterly.
Several logistics problems are involved with this testing; releasing steam to the environment requires notification of state and local governments. Hccith Physics have to be notified because of the affect on radiation monitors. This will be a small price to pay for the increase in valve reliability we expect to gain.
8] Preventative Maintenance A preventative maintenance program for these. valves has been established.
1]
Change diaphragms on valve operator every four (4) years.
2]
Visually inspect for Glycol leakage on the operator and charge as necessary every three (3) months.
3]
Inspect the air lines and valve operator casing for air leakage every three (3) months.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-5 RESOLUTION.(REFERENCE OBSERVATION No MSS-03):
Prior to Plant Startup:
1.
Change out diaphragms on all valves.
2.
Check for glycol leaks on all valves.
3.
Adjust packing on all valves.
4.
Stroke and test all valves.
Before the End of REFOUT '87:
1.
Write PACS to replace diaphragms every 4 years or so.
2.
Add to forced outage shutdowns to adjust packing during startup (PAC) i
)
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS.
RELATED TO MAY 19 1986 REACTOR TRIP c-6 COMPONENT:
Charging Pump P-55A PROBLEM DESCRIPTION:
At the time of the plant trip, P-55A was inoperable due to a cracked cylinder block. The Operations department had been instructed that although the pump was inoperable, it could be run in an emergency condition, ie., severe transient. Following the plant trip, P-55A was started numerous times. Each time, the pump started, ran for a short duration and then tripped.
HISTORY:
The charging pumps have a significant number of work orders on file.
Main problems include:
1]
Many work orders are related to packing life which averages 4 to 6 weeks.
2]
Low lube oil pump start problems have occurred on P-SSA and /or P-55B with unsuccessful adjustments of the lube oil relief.
3]
P-55A was out of service due to a cracked block.
MIO686-0566A-MA05-MD02
A PALISADES PLANT EQUIPMENT PROBLDiS.
RELATED TO MAY 19 L986 REACTOR TRIP C-6 4]
Several problems related to suction and discharge accumulators.
5}
Various pumps have had valve and valve spring failures.
SIGNIFICANCE TO OPERATIONS:
P-55A is safety related and Technical Specification related. Failure or degradation can result in a forced outage. Pump failure poses a hindrance to response on a plant transient. The charging pumps provide two (2) functions;
- 1) they maintain or restore PCS inventory during operation and following trips or transients and 2) they inject boric acid for reactivity control. Three pumps are provided.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
The accident sequences where the charging pumps are most important are those where the PCS remains at, or can return to, high pressure (ie.,
steam line break, small break LOCA or loss of feedwater). The capability to injeet cc,centrated boric acid into a fu?ly pressurized PCS would be important in casa, of ATWS.
4 In accident sequer.ces which depressurize the PCS, the HPSI or LPSI pumps provide the necessary water and beric acid.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUlPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-6 STATUS / ACTIONS:
Status The problem is still under investigation. The following is a summary of information as of 6/24/86.
1]
P-55A did start. Generally trips occurred after approximately 30 seconds of run time. The 30 seconds run times correspond with trips associated with low suction pressure and low lube oil pressure.
2]
All three charging pumps receive water from the same source.
This reduces the likelihood of a suction pressure trip since neither P-SSB or C tripped.
3]
Based on recently history, the low lube oil pressure is the likely cause of the trips. However, a new pump was installed with no success.
4]
A review of the monthly pump surveillance test results is underway.
Preliminary data indicates that pump sp ted and lube oil pressure do not consistently correspond.
5]
The lube oil filter was replaced on 3/16/86.
It is periodically replaced every 3 months.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-6 6]
A similar problem occurred on P-55B approximately one year ago.
Replacement of the oil pump corrected the problem. This incident was documented on D-PAL-85-052.
7]
Electricians checked P-55A's breaker and found everything satisfactory.
It is highly unlikely that the cracked cylinder block was the cause of the pump trips. P-55A was run for four days after the crack was discovered without any incident.
8]
Pump testing continues with focus on line by line checks of the pumps lube oil system.
9J During testing, concerns related to the fluids drives ability to engage properly were verified. Through testing, this has been shown not to be the cause of the pump trip' problem, but a separate issue. During starting, the motor starts, however, there may be up to a 10 second delay before the fluid drive begins to drive the pump. Evaluations are underway which point to a possible internal seal problem with the drive.
10] Feview by the verdor indicates that pump design cf the lube oil I
system may not allow reliable pump starts after pump has been idle for a period of time due to oil draining from the system.
Pumps may require the addition of an electric auxiliary pump to
{
maintain prire, i
MIO686-0566A-MA05-MD02
1 PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP c-6 11] Trouble shooting for the starting problems included I&C analyzing instruments initiating-trip signals. No problem was found.
12] Plant electrical analyzed trip circuit. It was confirmed that low lube oil pressure was tripping the pump. Also, no problem in circuit function was found. Vendor inspected pump. Recommended that the oil pump be replaced which could be a source of low lube oil problems. No other problems noted by vendor. Visually saw that fluid drive was not engaging pump. This is a cause of the 5 second and 15 second trips.
13] Maximum pressure that is being seen is 8 psi which is not high enough to defeat trip. This leads to either a defective lube oil pump, air in the oil lines or the pump shaft not rotating fast enough. The last can be discounted because pump was operating at max RPM at max flow. A strobe will be hooked up on pump shaft to confirm eccentric shaft RPM.
l Actions 1]
Continue testing of P-55A.
l 2]
Rebuild fluid drive if internal inspection indicated need.
Develop specification for a replacement drive.
MIO686-0566A-MA05-MD02 T&=
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PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP c-6 3]
Evaluate need of an auxiliary lube oil system.
4]
Evaluate multiple starts on motor for possible degradation.
5]
Resolve problems as identified.
6]
Activities are underway to solve the cracked block which include analysis of the structural integrity, development of weld procedures and procurement of replacement and spare blocks.
7]
Identify and develop necessary procedure periodic tests and PMs to maintain quality status of the pump.
RESOLUTION (REFERENCE OBSERVATION No CVC-15):
Prior to Plant Startup:
1]
Insure P-55A operable either with repaired block or justification for operability with leak and leakage collection system installed.
2]
Adjust the time delay in the oil pressure switch or repair lube oil system.
Before the end of REFOUT '87:
1]
Replace P-55A block.
MIO686-0566A-MA05-MD02
i PALISADES PLANT. EQUIPMENT PROBLEMS
.r RELATED TO MAY 19 1986 REACTOR TRIP l
c-6 i
I As part of the 5-Year Plan:
.i 1)
Evaluate packing life and oil leakage and make recommendations.
I s
i MIO686-0566A-MA05-MD02
PALISADES PLANT. EQUIPMENT PROBLEMS RELATED TO MAY 19.1986 REACTOR TRIP C-7 COMPONENT:
Intermediate back-pressure regulator; CV-2012.
PROBLEM DESCRIPTION:
Regulator failed closed during the 5/19/86 trip requiring operator action to transfer to redundant controller.
HISTORY:
1]
Due to let down orifice and trim valve irregularities in operation the pressure control valves and control circuitry has required frequent tuning and adjustment. This, coupled with the fact that the positioner on CV-2012 is obsolete and aged, led to erratic and improper operation of the intermediate pressure control valve CV-2012 at the time of the plant trip.
i 2]
While this component appears to be a "High-Maintenance" item, di3cus:ior. with optratore irdicated that this is a recent problem and not a chronic one.
f 3]
Due to the problems with let :!own flow control, the control loop
~
for the intermediate pressure control has not been allowed to operate per design, and trouble shooting has been generally inde te rminate.
MIO686-0566A-MA05-MD02 i
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-7 SIGNIFICANCE TO OPERATIONS:
The component is important to reliable plant operation and its failure can result in operator distractions during plant trips or transients.
Its failure can also initiate failure of RV-2006.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
i No impact on accident sequences, i
STATUS / ACTIONS:
i Status 1]
The valve positioners for both intermediate pressure control valves are old and obsolete. Over the life of the plant, they have been adjusted and calibrated frequently.
i 2]
Frequent adjustment and calibration has been required in a large part, due to the improper operation of the letdown orifice and trim valves.
3]
Evaluation of the electrical control circuit is not complete as yet, largely due to the several variables in the system induced by other than design flow through the orifices.
MIO686-0566A-MA05-MD02
I
^
PALISADES PLANT-EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP
{
C-T Actions i
i i
1]
The letdown orifice and trim valves are being replaced and calibrated to ensure design flow, to eliminate the variables in the system which affect the operation and adjustment of the intermediate pressure controllers.
2]
New positioners are being installed on both intermediate pressure control valves, CV-2012 and CV-2122, to eliminate the aging and obsolescence problems associated with these controllers.
3]
After the let down system is restored to design, and the system calibrated, testing and evaluation of the control circuitry (loop) which controls CV-2012, and CV-2122, will be performed.
4]
Based on the control circuit evaluation, modification or repair will be implemented to ensure the total intermediate pressure i
control and letdown systems operate as designed.
5]
Letdown stop and bypass control valves are being rebuilt to correct seat leakage.
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO.MAY 19 1986 REACTOR TRIP C-7 RESOLUTION:
Prior to Plant Startup.
i 1]
Change out positioners on both CV-2012 and CV-2122; intermediate pressure control valves.
2]
Perform dynamic testing and adjust positioners.
i l
l i
MIO686-0566A-MA05-MD02
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-8 COMPONENT:
Reheat stop valve CV-0544 PROBLEM DESCRIPTION:
Valve position indication did not show valve to be closed on trip.
HISTORY:
Due to the environment that the valve ILait switches are in, they have historically been susceptible to sticking, thus giving false indications to control room and instrumentation.
SIGNIFICANCE TO OPERATIONS:
Lack of reliable limit switch position indication for reheat / intercept valve position poses an operator distraction during startups/ shutdowns and plant transients.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
Little or none.
l MIO686-0566A-MA05-MD02
PALISADES PLANT ECUIPMENT PROBLEMS RELATED TO'hA" 19 1986 REACTOR TRIP c-8 STATUS / ACTIONS:
Status 1]
Limit switches are not reliable without periodic maintenance, given the environment they are expected to operate in.
Actions 1]
All 16 turbine valves will be inspected and repaired by electrical department to ensure operability and reliability.
RESOLUTION (REFERENCE OBSERVATION No TGS-12):
Prior to Plant Startup i
1]
Replace existing limit switches with more reliable switches if feasible.
2]
Ensure all limit switches for indication on all turbine valves are properly set and tested.
BEFORE END OF REFOUT 87:
(
1]
Replace limit switches with more reliable switches if not completed j
prior to startup from current outage.
MIO686-0566A-MA05-MD02 l
i l
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-9 COMPONENT:
Control Rod #34.
PROBLEM DESCRIPTION:
Rod bottom light failed to light on the control panel following the recteor trip of 5/19/86.
HISTORY:
On occasion, rod bottom lights do not illuminate, requiring thr. >perators to use backup means of verifying rod insertion. Some of the indicator 1
problems are related to seal leakage which results in boric acid contamination of the switches.
SIGNIFICANCE TO OPERATIONS:
Poses a hindrance to response on a plant transient; requires additional verification of control rod insertion.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
Requires additional operator verification of control rod insertion on a reactor trip; no ot*er effect on any other sequence.
STATUS AS OF JUNE 18, 1986:
Control rod drive package #34 has been replaced. The defective drive package was repaired by replacing bad wires from the terminal blocks to the synchro and limit switch connections.
i CNTRL ROD 34/MD01 1
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP c-9 i
RESOLUTION (REFERENCE OBSERVATION No CRD-05)
Prior to Plant Startup:
1.
Ensure operability of CRD #34 primary position indication and limit circuitry (including rod bottom light).
i 4
=
l CNTRL ROD 34/MD01 2
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-10 COMPONENT:
Condensate Recirculation Valve - CV-0730 PROBLEM DESCRIPTION:
On April 9, 1986, a work request was written on CV-0730 which stated "CV-0730 appears to drift open when at 100% power".
The work request was sent to the I&C department because the description of " drift" indicated the controls operating the valve were malfunctioning. I&C attempted to troubleshoot the problem, however, operations would not allow them to stroke the valve because it would cause a plant trip. Instrument air to the valves flow controller was isolated to prevent the spring-to-close valve from opening again.
It was understood that if the plant were to trip, operator action would be required to open the valve to prevent the condensate pumps from deadheading. On May 19, 1986, operator action was required to open CV-0730 following the plant trip.
HISTORY:
After reviewing maintenance history, three areas of concern were identified.
1] The valve positioner had been rebuilt or replaced three' times in the last four years. The causes of failure were indicated as vibration and water in the air supply.
MIO686-0566A-MA05-MD02
1 PALISADES PLANT EOUIPMENT PROBLEMS.
i RELATED TO MAY 19 1986 REACTOR 1 RIP C-10 2] The valve flow controller (FC-0730) was rebuilt once over the same four year period because of vibration.
3] The control valve and actuator had not been rebuilt in the seven years.
SIGNIFICANCE TO OPERATIONS:
Loss of the feedpump or condensate pump recirculation capability presents a significant chance that a pump will be lost and/or that immediate operator action will be required during a durate or plant trip in order to assure: recirculation flow exists.
Failure of the valve can also result in a plant trip.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
CV-0730 is of significance only in those sequences where:
1]
Secondary PCS cooling is required (ie not large LOCA) and, 2]
AFW is not available (Condensate System is one possible backup) and, 3]
Off-site power is available and MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-10 4]
Main feedwater piping is intact.
In order to stop condensate pump recirculation flow, CV-0730 would.have to fail closed in such a manner that it could not be opened by operator actions.
(Failure of instrument and/or air would be two such failure modes.)
STATUS / ACTIONS:
Status 1]
Calibration checks of FC-0730 and CV-0730 positioner were found to be satisfactory. Troubleshooting testing of the valve, with no dP across it, revealed a slow, jerky motion. Based on this testing and the fact that the valve and actuator had not been rebuilt in seven years, a work request was initiated for their rebuild.
2]
Initial installation testing of CV-0730 indicated that 6000 gpm of condensate could pass through the valve on its recirculation path back to the condenser. Upon disassembly of the drag-type valve, the disk stack was found to be 60 to 70% plugged. This plugging could reduce the anticipated recirculation flow to 1800 gpm or less than 10" of the condensate pumps total capacity.
Also discovered after valve disassembly was a bent plug stem and MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-10 mechanical binding wear patterns. These factors lead us to believe that the valve did not " drift open" but it never closed once the flow controller told it to.
Operations suspected CV-0730 was open when condensate pump amps were increasing higher than normal as reactor power increased. Upon inspection, CV-0730 was found to be open. To close the valve an operator failed the flow controller high and the valve positioner, which also points to a mechanical binding problem.
j 3]
Valve body has significant erosion in the seat area.
4]
Currently evaluating the horizontal orientation of the valve.
l Valve parts indicate wear patterns which may have cause possible binding of the valve.
1 5]
In addition to the control valve rebuild, the system was analyzed as to its possible affects on the flow controller actually sensing low flow which conses CV-0730 to open.
iI The orifice on the outlet of the air ejector inner-nf ter condenser was suspected of being eroded, however, inspection showed the orifice was intact. Plans are underway to inspect the gland steam condenser for possible tube fouling and/or plugging.
i MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS.
RELATED TO MAY 19 1986 REACTOR TRIP C-10 Actions 1]
Rebuild CV-0730 and V0P-0730 Once the valve was completely disassembled, body erosion was discovered in the seat land area which requires repair prior to rebuilding the valve. The valve will be cut out during the week of June 23 and require two weeks to weld repair and remachine.
The erosion found on the valve body does not directly correlate with the valve's mechanical binding.
2]
Inspect PO-0789 Inspection revealed no erosion. Orifice has been reinstalled.
3]
Inspect gland steam condenser Inspection plans are being developed.
4]
Evaluate replacement valve Five different vendors have been contacted. As of 6/23, no valves have been found which are immediately available and meet design criteria.
MIO686-0566A-MA05-MD02
^
PALISADES PLANT EOUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-10 5]
Evaluate replacement flow controller and positioner Vendor recommendations to arrive June 25.
6]
Reorient the valve to a vertical position to allow better performance and maintenance. Reorientation will be developed as a part of the 5-Year Plan. A new style valve will also be considered during the design phase of the project.
7]
Develop Preventative Maintenance to maintain:
al Clean hotwell each refueling outage to reduce possibility of plugging CV-0730 disk stack.
b]
Disassemble and inspect CV-0730 each refueling outage.
Extend rebuild period to 5 years, if confidence is gained that disk stack plugging is slow process.
c]
Calibration of FC-0730 and CV-0730 already in place.
8]
Post maintenance testing should include:
a]
Verify correct operation of CV-0730 M10686-0566A-MA05-MD02
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-10 b]
Verify correct valve positioning with simulated flow, c]
Verify correct valve positioning with actual flow.
RESOLUTION (REFERENCE OBSERVATION No CDS-01)
Prior to Plant Start-up:
1]
Properly test and adjust CV-0730 controls during start-up.
2]
Identify and complete required valve maintenance, i
i 3]
Determine correct controller operation and instruct Operations of proper operation (CV-0730).
4]
Evaluate whether valve configuration (ie, laying on side) is proper for this valve and take appropriate action.
5]
Inspect the orifice at discharge of the air ejector (RO-0789).
6]
Open and inspect the gland steam condenser for possible tube blockage in order to assure correct functioning of the differential pressure indicator.
7]
Rebuild CV-0730 and V0P-0730.
3 MIO686-0566A-MA05-MD02
1
)
l PALISADES PLANT EQUIPMENT PROBLEMS l
RELATED TO MAY 19 1986 REACTOR TRIP C-11 COMPONENT:
Air Ejector Condenser Steam Pressure Control Valve PCV-0633 PROBLEM DESCRIPTION:
During the 5/19/86 trip RV-0633 lifted apparently due to PCV-0633 not l
l responding properly to the increased steam pressure.
HISTORY:
Relief Valve (RV-0633) was worked on during the 1985/86 refueling outage because of seat leakage, the valve was subsequently worked on during the startup at which time it was removed, repaired and ratested.
SIGNIFICANCE TO OPERATIONS:
l The pressure control valve is important for proper air ejector operation.
i Failure poses a hindrance,co response on a plant transient.
IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
Degraded control has no effect on accident sequences.
l 1
(
l MIO686-0566A-MA05-MD02 t
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO M Y 19 1986 REACTOR TRIP C-11 STATUS AS OF JUNE 18, 1986:
k'ork Order has been initiated to troubleshoot, disassemble, inspect and repair both PCV-0633, and RV-0633.
RESOLUTION (REFERENCE OBSERVATION No AES-01)
Prior to Plant Startup:
1]
Troubleshoot and repair PCV-0633 as necessary and test.
2}
Retest RV-0633 for proper setpoint pressure.
MIO686-0566A-MA05-MD02
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-12 i
COMPONENT:
Turbine stop valves CV-0571 and CV-0575 PROBLEM DESCRIPTION:
May not have closed until 3 (0571) and 25 seconds (0575)
HISTORY:
Due to the envirotaent that the valve limit switches are in, they have historically been susceptible to sticking, thus giving false indications to control room and instrumentation.
SIGNIFICANCE TO OPERATIONS:
Operability of the turbine stop valves is important to reliable plant operation.
Failure of a turbine stop valve poses a hindrance to response during a plant transient and presents a high potential for operator distraction following a plant trip.
IMPACT OF FROBLEM UNDER PO5TULATED ACCIDENT CONDITIONS:
Turbine admission valve closure affects no accident sequence except turbine overspeed. Redundancy is provided by pairing main and reheat admission valves, and by the main steam isolation valves.
MIO686-0566A-MA05-MD02
- - - - - - ~.
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-12 i
The turbine admission valves, (stop, control, reheat stop, and intercept valves) provide a means to control steam flow to the turbine during operation and a means to stop steam flow during a turbine trip. Complete failure to stop steam flow to the turbine following a generator trip would lead to turbine overspeed. Failure to stop steam flow can only occur if both valves in one of eight specific pairs (ie #2 stop and #2 control valve or #3 intercept and #3 reheat stop) fail to close at the same time. Monthly testing of each of these valves assures high reliability.
STATUS /ACTIbNS:
Status 1]
Limit switches are not reliable without periodic maintenance, given the environment they are expected to operate in.
Actions 1]
All 16 turbine valves will be inspected and repaired by electrical department to ensure operability and reliability.
I MIO686-0566A-MA05-MD02
PALISADES PLANT. EQUIPMENT PROBLEMS
)
2 RELATED TO MAY 19 1986 REACTOR TRIP C-12 RESOLUTION (REFERENCE OBSERVATION No TGS-13).
Prior to Plant Startup Inspect, test and repair as necessary. This includes complete verification of closing times and limit switch and events recorder j
circuitry.
l t
MIO686-0566A-MA05-MD02
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-13 COMPONENT:
Plant Data Logger / Events Recorder PROBLEM DESCRIPTION:
4 Thirteen (13) data logger system (DLS) points did not respond when comparable event recorder pens operated during the 5/19/86 trip, i
HISTORY:
Equipment failures in the data logger system (Tennecomp Feedwater Purity Datalogger) are not well documented. The assigned technician usually checks the system on a daily basis to ensure operability. Environmental conditions (ie heat) in the station power room where the computer and Field Remote Station No 1 are located is not conducive to good computer operation. The equipment is unique and obsolete. Over the last few years, equipment reliability has not been good. An engineering service request has been generated to relocate the control room printer to the station puver room to eliminate misoperation by Operations Department j
personnel. As this system monitors several inputs to the events recorder system, loss of event recorder power can result in loss of signals to the data logger.
i 1
4 SIGNIFICANCE TO OPERATIONS:
Failure or degradation can result in difficulty in reconstructing events I
following a transient and inhibit determining the actual cause of an event similar to that of the 5/19/86 trip.
PLT DATA LCR/ EVENT RCRDR/MD01 1
^
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-13 IMPACT OF PROBLEM UNDER POSTULATED ACCIDENT CONDITIONS:
Failure or degradation can result in difficulty in reconstructing events I
following a transient and inhibit determining the actual cause of an event similar to that of the 5/19/86 trip.
STATUS AS OF JUNE 18, 1986:
A memo detailing the thirteen discrepancies was forwarded to the System Engineer on June 2, 1986. Repairs are in progress.
EESOLUTION (REFERENCE OBSERVATION No DTA-01 AND No DTA-02)
Prior to Plant Startup:
1.
Repair inputs 2.
Evaluate Engineering Service Request and relocate data logger printer to station power room if action is warranted.
3.
Improve documentation of maintenance activities on data logger systems by implementing open work orders on each data-logger I
system so that equipment histories may be maintained.
Before the End of REFOUT '87:
1.
Plans for the replacement of the plant sequence-of-events monitor and data-logging system should be completed. This should include evaluation of the present inputs versus required inputs in light of the recent addition of the Critical i
Functions Monitor.
4 PLT DATA LGR/ EVENT RCRDR/MD01 2
PALISADES PLANT EQUIPMENT PROBLEMS RELATED TO MAY 19 1986 REACTOR TRIP C-13 As Part of the 5-Year Plan:
1.
Replacement of the Plant Information Processor (PIP) Secondary Position Indication (SPI) and Plant Datalogger (Tennecomp) is Presently in the 5-Year Plan (as well as other plant computer systems).
s 4
s i
PLT DATA LGR/ EVENT RCRDR/MD01 3
8 APPENDIX D PALISADES PIANT REVIEW COMMITTEE MINUTES OF SPECIAL MEETING MAY 20, 1986
Palisades Nuclear Plant PLANT REVIEW COMMITTEE MEETING Special Meeting - May 20, 1986 No 86-037 3
i Members Present i
Chairman / Scribe..... RAFenech, Technical Engineer
)l RMRice, Plant Operations Manager RMBrzezinski, I&C Engineering & Maintenance j
Manager j
SGhidotti, Shift Supervisor i
JGlewis, Technical Director i
Guests.......... MDFerens, Senior Engineer j
RSWesterhof, Senior Engineer RAVincent, Safety Engineering Administrator REMcCaleb, QA Director GFord, Senior Engineer I
KHDunham, Reactor Operator JFFord, Reactor Operator j
JLHanson, Shift Engineer j
TJPalaisano, Plant Projects Superintendent JDAlderink, Mechanical Engineering & Maintenance i
Superintendent i
j The number of members present constituted a QUORUM.
4 l
A special meeting of the Palisades Plant Review Committee was held on Tuesday, t
May 20, 1986. The purpose of the meeting was to review an Event Report titled i
" Notification of Unusual Event Due to Reactor Trip on Pressurizer High Pressure" l
and the Post Trip Review of the 5/19/86 reactor trip.
I.
The following items were reviewed and recommended for approval at the i
meeting. All were found not to constitute an unreviewed safety i
question. Unless otherwise indicated, the vote was unanimous.
j E-PAL-86-057 NOTIFICATION OF UNUSUAL EVENT DUE TO REACTOR TRIP
{
ON PRESSURIZER HIGH PRESSURE I
Reactor trip occurred at 1416, May 19, 1986 due to l
EHC power supply failure which resulted in turbine 1
valve closure and a reactor trip on pressuriser high pressure.
Initial assessment of incident was that trip was caused by loss of load; however, on further investigation, was determined that trip initiator j
was high pressure in PCS which necessitates declaring Unusual Event. Notification made @
l 1745.
f MM0586-0034A-LIO1 i
_. _. _. -. - _, - - _ _ _. ~. _ _ _ _. - _.. _ _ _ _.... _ _ _ _ _ _ _,. _. _ _. _
SPECIAL PRC MEETING MINUTES Page 2 No 86-037, May 20, 1986 The root cause of the problem was ti.ought to be failure of both EHC power supplies. The power i
supplies were tripped when a technician unplugged a cabinet fan and induced an "RF" signal into the circuitry.- Additional testing was proposed to try to duplicate the failure before making repairs.
Post Trip Review 5/19/86 Reactor Trip The committee reviewed the trip report and interviewed the operators on shift at the time of the trip. A concern was raised regarding the number of pieces of equipment that failed to operate properly (see i
Post Trip Review). After discussing the safety significence of the failed equipment, the committee madi the following recommendations to the Plant Manager:
1.
.ue to the increased potential to challenge the code safeties, either the atmospheric dump l
valve or the Turbine Bypass valve must be returned to service before coming on-line.
2.
Due to the extent of equipment that did not i
operate properly following the reactor trip, PRC recommends the Plant Manager give the highest priority to the definition of equipment malfunctions, prioritization and repair.
l Recoimeended priorities are:
f 1.
Turbins Bypass Valve, CV-0511 1
2.
Atmospheric Dump Valve, CV-0779 3.
Charging Pump, P-55A 4.
Letdown pressure control system 5.
Condensate recirculation valve, CV-0730 6.
Turbine lift pumps.
1 M
Prepared (L
I,17-Flo Reviewed
(
PRC Secretary Date Plant Technical
' Da t'e Director Reviewed k.
8-2 / Jf.
Approved PRC Chairman Date VPlant General
'Date Manager MM0586-0034A-LIO1
APPENDIX E CHRONOLOGY OF MAY 19, 1986 REACTOR TRIP 1
i e
e 1
CHRONOLOGY i
i A chronology of the significant events after 1400 on May 19, 1986 hours0.023 days <br />0.552 hours <br />0.00328 weeks <br />7.55673e-4 months <br /> follows below. Entries are identified as follows:
SSL = Shift Supervisor Log; RXL = Reactor Log; CRL = Control Room Log; SEL Shift Engineer Log; DTL = Plant Data Logger (Military time including seconds).
DTL 1406:50.514 HPSI Pump P-66A Stopped SSL 1413 Main Turbine valves went closed j
DTL 1415:13.966 Main Turbine Speed Low i
(Authors Note: This has been diagnosed as a false signal due to the AEH power supply failure which powers the j
turbine speed sensing circuitry)
DTL 1415.06 Turbine Speed 1796.67 RPM i
1 DTL 1415:21 Pressurizer Level:
56.54%; Pressure 2026.33 psia Loop 1 Temperatures:
Inlet: 540.05'F; Outlet: 580.60*F i
Loop 2 Temperatures:
Inlet: 538.15'F;6Outlet: 588.01*F i
Steam Generator Steam Flow: 10.07 x 10 pph Steam Generator Pressures: A: 732.56 psia, B:726.70 psia
{
Steam Generator Levels: A: 67.65%, B:69.84%
DTL 1415:26 Pressurizer Level: 56.70%; Pressure: 2028.29 psia Loop 1 Temperatures: Inlet: 540.07'F; Outlet: 588.71*F Loop 2 Temperatures: Inlet: 538.' 19'F; 6Outlet: 588.06*F Steam Generator Steam. Flow: 9.92 x 10 pph i
Steam Generator Pressure: A: 738.08 psia; B: 730.08 psia Steam Generator Levels: A: 67.67%; B: 68.70%
I DTL 1415:51.266 Steam Generator E-50A Pressure High I
Steam Generator E-50B Pressure High
{
j DTL 1415:56 Pressurizer Level: 61.85% Pressure: 2097.43 psia Loop 1 Temperatures: Inlet: 544.39'F; Outlet: 590.39'F Loop 2 Temperatures: Inlet: 541.23*F;6Outlet: 590.13*F i
Steam Generator Steam Flow:
7.08 x 10 pph Steam Generator Pressure: At 840.54 psia; B: 838.19 psia l
Steam Generator Levels: A: 45.61% B: 47.94%
4 DTL 1415.06 Turbine Speed 1796.67 RPM' DTL 1415.11 Turbine Speed 1160.35 RPM i
(Author's Note: This is an incorrect value due to the i
failed +15 VDC EHC power supply) 4 f
DTL 1415:58.066 Steam Generator E-50A Level Low Steam Generator E-50B Level Low l
SSL 1416 Rx & T/G Trip CRL 1416 Turbine Trip; Rx Trip (Loss of Load);
Loss EH PWR Supply-Power; All Governor Valves closed RXL 1416 Rx & Turb Trip due to loss of EH pressure SEL 1416 Rx Trip - normal plant response, declared 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> non-emergency due to RPS actuation.
DTL 1416:01.264 Pressurizer T-72 level high DTL 1416:06 Pressurizer Pressure: 2245.49 psia (maximum)
DTL 1416:09.312 Letdown orifice stop valve CV-2004 opened DTL 1416:09.594 Letdown orifice stop valve CV-2005 opened DTL 1416:10.710 Pressurizer T-72 Chanel-D Pressure High DTL 1416:10.742 Pressurizer T-72 Chanel-B Pressure High i
DTL 1416:10.806 Reactor Control-Rod Drive Clutch-A Power Relay K-1 De-Energ
.806 Reactor Controi-Red Drive Clutch-B Power Relay K-4 De-Energ
.814 Pressurizer T-72 Channel-A Pressure High
.828 Reactor Control-Rod Drive Clutch-B Power Relay K-3 De-Energ
.828 Reactor Tripped DTL 1416.11 Stm. Gen. Press.(max): A: 1025 psia; B: 1019 psia i
DTL 1416:11.132 Pressurizer T-72 Chanel-C Pressure High DTL 1416:11.222 Bus 1C Fast Transfer Operated DTL 1416:11.280 Rehest Stop Valve CV-0534 Closed DTL 1416:11.282 Reheat Stop Valve CV-0548 Closed 3
I DTL 1416:11.320 Reheat Stop Valve CV-0537 Closed l
DTL 1416:11.366 High Pressure Turbine Stop Valve CV-0569 Closed High Pressure Turbine Stop Valve CV-0573 Closed DTL 1416:14.522 High Pressure Turbine Stop Valve CV-0571 Closed j
DTL 1416:16.436 Auxiliary Feedwater Pump P-8A Drive Motor Started i
DTL 1416:28.490 Charging Pump P-55B Started DTL 1416:36.418 High Pressure Turbine Stop Valve CV-0575 Closed DTL 1416:50.686 Charging Pump P-55A Started CRL 1417 Follow-Up E0P 1 Immediate Action DTL 1417:21.620 Charging Pump P-55A Stopped
. RP0686-0083A-MD02
~
d DTL 1417:57.410 Charging Pump P-55A Started 1418:28.776 Charging Pump P-SSA Stopped 1418:32.074 Charging Pump P-55A Started 1419:02.618 Charging Pump P-55A Stopped 1419:07.290 Charging' Pump P-SSA Started 1419:38.712 Charging Pump P-55A Stopped 1419:45.672 Charging Pump P-55A Started 1420:16.556 Charging Pump P-55A Stopped 1423:23.344 Charging Pump P-55A Started 1423:28.222 Charging Pump P-55A Stopped CRL 1420 Stopped P-2B, P-10B SSL 1430 Determined Rx Trip due to loss of load - declared 4 s
i hour non-emergency report.
CRL 1435 Stopped P-10-A SEL 1450 Notified State of Unit Trip - they requested they be notified when we start back up.
SEL 1453 Notified County of Trip SEL 1458 Notified NRC of 4 He non-emergency report for RPS actuation.
SSL 1458 Completed all notification's with exce ption of Public l
Affairs Director.
SSL 1500 Aborted M0-22 and cancelled performance of MO-16.
I SSL 1501 Issued to CO-1 GOP's #10 + 8 3
SEL 1503 Notified CPCo/ Power Control of 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> non-emergency report. Reactor trip appears to have been to a EHC l
system failure, Trip was on loss of load.
SSL 1505 Had A0 open CV-0730 SSL 1506 Problem appears to be with EHC cabinet power supplies.
j RXL 1511 Borating PCS to Hot S/D +50 SSL 1612 Released I&C to work EHC-23 EHC Power TGS 246055100 SEL 1616 Notified NRC that trip was due to PZR High Pressure; not loss of load but the PZR High Pressure was due to the loss of load.
SEL "C" Shift Rx in HSD i
SEL 1745 Declared unusual event due to Rx Trip on High PZR Pressure and terminated.
r 4 RP0686-0083A-FD02
- - ~ - - - -
- - - ~ ~ - - - --
r SSL 1750 Completed checklist 6.1 - Sat -
SEL 1759 Completed Notification of Unusual Event and termination notifications.
CRL 1856 Closed MSIV's - Bypasses open RXL 1938 Rx Reset for RPS CL 36 2230 Completed Post Trip Review Report SSL 2232 Completed CL35 NI's - SAT -
RXL 2300 Completed MO-22 (HPSI's)
RXL 2301 Completed CL36 SSL 2318 Completed CL36 RPS - SAT-May 20, 1986 SSL 0015 D&C approved Rx startup.
RXL 0254 Reset Reactor CRL 0254 Reset Reactor SEL 0255 Rx in hot standby RIL 0300 PCS Leak Rate Total = 0.5 GPM Unidentified = (no value)
RXL 0303 S/D Group A Withdrawn SSL 0303 Withdrawing S/D Rods Rx in Hot Standby RXL 0312 S/D Group B Withdrawn, Withdrawing Reg. Rods RXL 0445 PCS Boron @ 874 PPM SSL 0445 NRC Plant Status Check RXL 0751 Rx Administrative 1y critical Cat 0751 Reactor is critical SEL 0751 Rx is critical CRL 0835 MSIV's open, MSIV Bypasses closed CRL 0853 Chemistry reports we're on a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Chemistry hold @
<2% pwr.
RXL 1045 PCS Boron @ 871 ppm SSL 1101 Released for repair CV-0779 Atmospheric Dump Valve W.O.#24605524.
CRL 1200 Comp GCL 8-1 (Plant Shutdown to Hot Standby, MSR Valving etc)
CRL 1345 Rolling main turb up for work by I&C, appears to be a problem with indicated speed not matching up with
" reference speed" on EHC panel.
N1630 Completed PRC Meeting #86-37 Review of E-PAL-86-057,
" Notification of Unusual Event due to Reactor Trip on Pressurizer High Pressure." RP0686-0083A-MD02 l
CRL 1800 Turbine rolling @ 1200 RPM for I&C - Primary Power Supply is removed for test.
SSL 2326 Unisolated CV-0511. Did not stroke.
RKL 2342 PCS Boration for Xenon / Rod withdrawal May 21, 1986 RXL 0400 PCS Leak Rate +0.298 GPM total, + 0.282 GPM Unid.
SSL 0610 NRC Plant Status Check SSL "B" Shift Rx in Hot Standby, Chemistry hold on sulfites.
(
SSL 0855 Pre-planned and discussed possibility of Turbine Trip 1
and autostart of D/G while werking on EHC power supplies.
SSL 1015 Isolated CV-2012 for I&C work.
CRL 1320 Opened Steam Dump #0779 for one minute total time (1 I
min).
CRL 1406 Steam Dump 0779 open for 46 seconds.
3 SSL 1500 Completed testing CV-0779, the valve does operate but does so quite slowly.
It takes 15 to 25 seconds to i
begin to open, then opens very quickly. It takes %45 secs to begin to close and then closes very quickly.
]
However, the HIC 0781B at C-33 does not operate either CV-0779 or 0781 in manual. Note 0779 was tested with steam.
SSL "C" Shift Rx critical @ 1%.
1200 RPM on Turbine 1
N1800 Received verbal notification from the US NRC, Region i
III, that a Confirmatory Action Letter would be issued, requiring the plant to take the facility to cold shutdown and not restart the facility until a thorough investigation into the causes and e
implications of the trip and a thorough investigation of plant safety systems and balance of plant systems impor tant to safety with regard to operability and required maintenance is completed.
SSL 2000 Operational Test of CV-0511: Stroked CV-0511 satisfactorily when inlet steam supply closed.
Cracked open inlet steam supply and CV-0511 would not j
open. Closed inlet steam supply and CV-0511 stroked
)
OK.
Put CV-0511 at 10% open and cracked open inlet steam. Steam flowed for a few seconds and then i
5-RP0686-0083A-MD02
l-1 CV-0511 drifted and stuck shut. Shut inlet steam supply and CV-0511 went back to 10% open. Opened CV-0511 to 50% open and cracked in steam and valve drifted and stuck shut. Shut inlet steam supply and l
CV-0511 went to 50% open. Opened CV-0511 to 100% and
(
cracked in steam. Valve stayed open. Shut CV-0511 to 80% open and it stayed at 80%. Shut CV-0511'to 50%
open. Valve drifted manually to 60% open and then quickly drifted and stayed shut. Left CV-0511 inlet steam isolated.
%2030 Following verbal exchanges with the USNRC, Region III, regarding operating time for testing of balance of plant systems at temperature and pressure required to effectively test those systems, the order was received to immediately proceed to cold shutdown condition.
SSL 2045 Per RMRice: Make preparations and bring plant to cold shutdown in a controlled and orderly manner. Testing on CV-0779 in progress will stabilize plant.
SSL 2046 Shutting down"B" MFP CRL 2050 Clear up on CV-0779 for DWilliams. Comp. @ 2058 SSL 2110 Shutting down "A" MFP SSL 2152 Completed testing CV-0779 & CV-0781. All controls left as is before test! g.
Unisolated CV-0779 and CV-0781. Verified all atmospheric dump air isolations open. Verified CV-0779 and CV-0781 unisolated.
(Stroked CV-0779 and CV-0781 Sat.. from C33 and Col.
Some discrepancy was found at C33 but all conditions returned to normal). CV-0781 stroked slow and lost N1/2 gpm of glycol. Sys. Engineer is aware of all problems.
SSL 2153 Order man. T/G S/D CRL 2154 Turbine tripped manually SSL 2158 Man T/G S/D 1-1 D/G auto started' properly. All conditions normal.
SSL 2207 Bypassed and shut MSIV's SSL 2209 Rx Suberitical
(
RXL 2221 All Reg Rods Inserted l
l RXL 2225 All part length rods inserted
.RXL 2226 Commenced borating + Cold S/D concentration l
l,
RP0686-0083A-MD02 l
L
6 May 22, 1986 RXL 0102 Manually Tripped Reactor, plant in Hot S/D 4
SEL 0123 Tave <525'F, proceeding to cold S/D per D&C direction.
I CRL 0432 Turbine on turning gear SEL 0707 Notified State and county of unit S/D and steam dump ops.
i RXL 1605 LTOP operable (amber light) @ 300*F I
RXL 2159 PCS in cold S/D l
J i
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9 a
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RP0686-0083A-MD02
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APPENDIX F SYSTEMS IMPORTANT TO RELIABLE PLANT OPERATION 2
i 1
'I 4
I i
1-(
Systems Important to Reliable Plant Operation Rev 3 June 4, 1986 I
SYSTEMS USED TO CONTROL PLANT SYSTEMS A.
Electrical Power 1.
AC power 2.
DC power 3.
Diesel generator 4.
Instrument & preferred AC B.
Instrument Air C.
HP Air, Safeguards & Turbine D.
RPS E.
Rod Control F.
Rod Position Indication G.
Nuclear Instrumentation H.
Control Room Instrumentation & Alarms I.
Communications II SYSTEMS WHOSE FAILURE COULD CAUSE CORE DAMAGE A.
PCS B.
Main Steam System III SYSTEMS USED TO COOL THE CORE A.
Condensate storage tank l
2.
Steam supply 3.
Nitrogen system B.
Normal Feedwater 1.
Main feedwater piping and valves 2.
Condensate system 3.
Condenser NOTE: Systems are listed only once, even if they fall under more than one category.
SYST IMP / PLT OP/MD01
4 2
~
C.
ECCS 1.
Safety injection tanks 2.
Low pressure safety injection 3.
High pressure safety injection 4.
Component cooling water 6.
Main Steam System 1.
Code safety valves 2.
Atmospheric dump valves 3.
MSIV's & bypasses-4.
Turbine bypass valve IV SYSTEMS WHICH LIMIT FISSION PRODUCT RELEASE A.
Containment Airlocks B.
Containment Isolation Valves C.
Containment Spray System D.
Containment Air Coolers E.
Hydrogen Recombiners F.
Iodine Removal System V
SYSTEMS WHICH PROTECT SYSTEMS LISTED ABOVE A.
Fire System i
VI SYSTEMS WHICH REPEATEDLY CHALLENGE SYSTEMS LISTED ABOVE A.
Main Feed Pumps B.
Feedwater Regulating System C.
Main Turbine Controls D.
Main Generator Controls VII MISCELLANEOUS SYSTEM A.
NSSS Sampling B.
Control Room HVAC C.
Emergency Lighting NOTE: Systems are listed only once, even if they fall under more than one category.
SYST IMP / PLT OP/.E 01
.