ML20153D535
| ML20153D535 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/19/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20153D522 | List: |
| References | |
| TAC-60462, NUDOCS 8602240173 | |
| Download: ML20153D535 (139) | |
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y . i OESCRIPTION AND, RESOLUTION OF ISSUES REGARDING THE DECEMBER 26, 1.985 REACTOR TRIP l Prepared by the Staff of the SACRAMENTO MUNICIPAL UTILITY DISTRICT February 19, 1986 I I 9602240173 B40219 . PDR ADOCK 05000312 S PDR-a
- r
INDEX PAGE NO.- I. INTRODUCTION
- 1. Overview of Event 1
- 2. SMUD Response to the Transient 2
- 3. NRC Incident Investigation Team (IIT) 3
- 4. Transient Analysis Organization 3
- 5. B&W Owners Group Regulatory Response Group (RRG) S
- 6. Region V Confirmatory Action Letter S II. SEQUENCE OF EVENTS 5 III. ACTION LIST 6
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IV. QUARANTINED EQUIPMENT LIST 6
- V. RESOLUTION OF ACTION LIST
- 1. Post Trip Report 7
- 1. Main Steam Line Analysis 7 l 2. Minimum Pressurizer Level 9
- 3. Pressurizer Heater Operation 10
- 4. Control Room Instruments Affected by i Loss of ICS DC Power 10 f
Investigation of Report of Smoke } 5. Prior to Reactor Trip 11'
- 6. Primary to Secondary Leak Investigation 12
- 7. Operations / Security Interface During Transients 12 l
l 8. Main Steam Line Failure Logic (MSLFL) 12 l
- 9. SPDS vs. Strip Chart Recorder OTSG Operate Level 13
- 10. Transient Compared to USAR Design Basis 14 4 . . , & e , .- , , , <y y- -- -
----s-.-- - --
. . ~ ;m. _ _ - ,- --
lINDEX (CONTINbED)' . PAGE NO. . 1 l V. 2. Human Factors Analysis 15 4
- 3. Determine Cause/ Corrective Action for ICS- .l Power Failure -
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'1. .ICS AC Power Sources 15
~2.. ICS Equipment Investigations 16
! ;3. Procedure for Loss of ICS 17 L 4.. Modifications to the ICS' , 17 4.- Makeup Pump (P-236) Failure 20 ) 5. Damage to Radiation Monitor R-15001 21
- 6. RCS Overcooling 22 1
- 7. Health Ph'ysics and Control of-Contaminated j- Air and Water
- 1. Secondary Plant' Steam Relief to Atmosphere 23
- 2. Secondary Plant Condenser Air Ejector 23 l 3. Spilled Makeup Tank Water in Auxiliary Bldg. 24 l- 4. Flooding of Waste Gas Heater 24 i
l' 8. Emergency Plan - i
- 1. Implementation 25 i 2. Technical Support Center Fire Sprinkler System 25 4
- 9. Training
- 1. Sequence of Events Training 26 i
- 2. Plant Modifications 26 i 3. Emergency Operating Procedure Changes 26
- 4. Loss of. ICS Power Casualty Procedure 27 i
f 5. General Procedure Changes 28 1 .
- 6. Emergency Entry Into Areas of Unknown
- i Radioactive Conditions 28 a
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INDEX (CONTINUED) PAGE-NO. - V. 9. 7. Manual Valve Operation 28 .
- 8. . Command and Control Training ' 29 -
- 9. Control Room Simulator ~29
- 10. Operational Review 30 ;
- 11. System and Component Response
- 1. Auxiliary Steam Relief Valve PSV-36012A 32-
-2. Difficulties in Manual Operation of AFW Valves FV-20527/8 33
- 3. Difficulty in Manual Operation of AFW 4
Valve FWS-063 34
- 4. Main Steam Relief Valve PSV-20544 34
- 5. RCP' Seal Injection Flow Interruptions 34 6.. Feedwater Heater Shell-Side Relief Valves Opening 35
- 7. HPI/RCS Injection Valve SFV-23811 Position Indication 36 i
- 8. Preventive Maintenance (PM) Program '
for Manual Isolation Valves 36
- 12. Quarantined Equipment List 37 VI. PLANT MODIFICATIONS RESULTING FROM EVENT
- 1. Loss of ICS DC Control Power 38 j 2. Consequences of Loss of ICS DC Control Power 39 I 3. EFIC Implementation 40
- 4. Other Modifications 41 4
t 1 i
INDEX (CONTINUED) PAGE NO. VII. 400T CAUSE(S) 41
- 1. Loss of ICS Pow'er 42
- 2. Rapid Cooldown of NSS 43
- 3. Damaged Makeup Pump 44
- 4. Health Physics Procedure Implementation 44
- 5. Emergency Plan Communications 45 VIII. CONCLUSIONS Safety Significance
- 1. 45
- 2. Appropriateness of Findings 46
- 3. Status of Resolution of Recommendations, Repairs,.and Modifications 46
- 4. Effectiveness of Modifications and Training 46
- 5. Suitability for Restart 47 i
IX. ATTACHMENTS
- 1. SMUD Sequence of Events .
- 2. GAC 85--1001R2
- 3. Action List
- 4. Region V Confirmatory Action Letter
- 5. Post Trip Parametric Data
- 6. Human Factors Items
- 7. Control Room Equipment Affected by Loss of ILS DC Power
- 8. ICS Power Supplies and Distribution
- 9. AFW Control Independent of ICS
- 10. ADV Control Independent of ICS
- 11. TBV Control Independent of ICS i
- i. !
i 1 l
-0ESCRIPTION AND RESOLUTION OF-ISSUES REGARDING ?
[ .THE CECEMBER 26 1985 REACTOR TRIP' i 4 s I. INTRODUCTION
- 1. Overview of Event
! Early in the morning of December 26, 1985, while operating at a steady 75%FP power, the reactor tripped on high pressure as a result i of loss of main feedwater caused by-an unexpected loss of DC power [ within the plant's Integrated Control System (ICS). The loss of ICS
- power caused the main feedwater pumps to reduce to minimum speed, causing Main Feedwater pressure'to decrease which autostarted both AFW Pumps and permitted AFW control valves to operate. Concurrently,
! all other control devices receiving inputs from the ICS received a { "zero" volts signal which is interpreted as a "50% demand" command
- since the' ICS operates ' on a -10 to +10 ydc range corresponding to 0 t
to 100% demand. This resulted in several steam valves repositioning
- to the corresponding 50% open positions, and likewise, the Auxiliary 1
Feedwater. Control Valves opening to a mid-position. The effect upon the Nuclear Steam Supply system (NSS) was to provide a significant j rate of heat removal via both ' steam discharge and the addition of ' l cold Auxiliary Feedwater to the Once Through Steam Generators j (OTSGs). These effects combined to_ cause a rapid reduction in the i tempe'rature and pressure in the Reactor Coolant System (RCS). The i pressure reduction reached the point where the engineered _ safeguards equipment, the Safety Features Actuaticn System (SFAS), actuated bringing into service the High Pressure Injection (HPI) System (which i had previously been operating on operator demand), the standby { emergency Diesel Generators, the AFW bypass flow control valves, and 1 solation of the Reactor Building as it would in the eventLof a Loss '
- of Coolant accident. The high rate of reactor coolant injection from
- HPI was effective in establishing the reactor coolant inventory j needed at the reduced temperature caused by the overcooling. A detailed " Sequence of Events" is provided as Attachment 1.
3 During the event, all safeguards and related equipment performed as i designed, although post-event analyses has identified enhancements to l non-safeguards related equipment and procedures. The event did progress sufficiently to constitute an " Overcooling". event as a result of. delays and complications in closing the open steam and j feedwater control. valves. Reactor Coolant Rumps (RCP's) remained j online through the event. 1 i The operating crew on duty had been augmented by an extra Senior
, Reactor Operator who was also a qualified Shif t Supervisor. . Early in
, , the event he'left the Control Room (this was appropriate, as he was i
;. 'not assigned a specific duty station) and assisted-in the work of j post-trip recovery and equipment operation. Upon returning to the j . :
f
._ . _ . , _ - _ . _ _ _ _.. _ . _ L _ . _ _ . .-
I . -- cl. (Continued)' . , ' Control Room. .after nearly a half hour of vigorous exertion, he
- became " lightheaded" and had to-lie down. 'This was diagnosed as 4
- hyperventilation and followup examinations have not identified any reason for concern as to the individual's fitness for duty.
-During post-trip recovery activities, the, suction to the Makeup Pump was inadvertently isolated. This led to the rapid destruction of the
' pump seals 'and the release of. approximately twelve-hundred gallons of reactor coolant makeup water within the Auxiliary Building. . The associated release of radioactive gases contained in the coolant
- constituted the primary source of radioactivity released during the event. An Unusual Event (as established by the facility Emergency
- Plan) was declared as a . result of the SFAS. actuation, which initiated the notification of the various emergency response agencies. ;
'This event was defined by SMUD management to'be unique and to require i '
special study, analysis, and troubleshooting prior to authorizing a plant restart. This report details these studies and analyses, > I develops the lessons learned, provides the root causes, and describes the implementation of actions and changes to preclude recurrence of <
; this event, while enhancing the ability to mitigate future transients i at Rancho Seco.
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j 2. SMUD Response to the Transient t { In addition to collecting'the various data available following an . event 'of this type, the Plant Manager inusediately in'stituted a deliberate and detailed process to govern the investigations and troubleshooting necessary to ' establish the root cause(s) of the - event. This program was intended to systematically prepare the i facility for return to power operation. Details of this program were . l defined and published as shown in Attachment 2. In anticipation of industry interest in an overcooling event such as j the one experienced, particularly at those facilities using an ICS, i
- j. the District invited the Institute of Nuclear Power Operations '
i (INPO), the Electric Power Research Institute (EPRI/NSAC), and the l B4N Transient Assessment Team to assist with investigating the event and reporting it to the utility industry. The INPO team developed a Significant Event Report (SER #3-86.) which was issued to the Industry on January 2, 1986. EPRI/NSAC assisted in the evaluation of the Pressurized Thermal Shock
.(PTS) related questions and analysis as they pertain to the reactor i~ vessel and potential effects upon its service life. These results 4
are discussed in a following section, which confirms no Reactor Coolant. System component degradation attributable to the transient. i j l : iP
t I.: 2. (Continued) . The.B&W TAP Team assisted in compiling the sequence of events and in !. defining issues and areas which needed further investigation. The team played an important role in developing the troubleshooting plans , .for the ICS and.the Maintenance Instructions (NIs) which implemented 3 those plans. They also assisted in evaluating the procedural
~a dequacy and operator response to the event. !
- 3. NRC Incident Investigation Team (IIT)
Late on December 30, 1985, the District was notified that the NRC had ' elected to upgrade their involvement'in the event investigations by replacing the augmented Region V investigative effort with an IIT charged with detarmining the:
- 1. Sequence of Events
, 2. Troubleshooting Action Plans to Determine Root Cause, and
- 3. Major Issues surrounding the event.
l- Interviews were held "on the Record" with the operating personnel involved in the event, its mitigation, and the District's 1
- investigative efforts.
l An early effort established by SMUD, with concurrence by the NRC, was ) to develop the " Quarantined Equipment List", of that equipment whose i failure influenced the event or operator efforts to stabilize the ! plant. l Details of the requirements for the process of troubleshooting and l reporting of findings, relevant to the Quarantined Equipment, were ! developed and published in revisions to the District's investigation' ! efforts. These~were established by Attachment 2. Once the IIT had completed their interviews,'and. agreed with the , content of the developed Troubleshooting Action Plans, they left the
- l site to write their report ~and await the results of the :
L troubleshooting efforts. The Rancho Seco staff provided these l results, as they became available, directly to the IIT'and the NRC l Site Resident Inspector.
- 4. Transient Analysis Organization ;
Attachment 2 established the organization and process for insuring i that all available' and potentially useful material necessary to gain j a full understanding of the events of December 26th was obtained and utilized. It established an Action List which collected the i developing issues and oversaw their. resolution. ' l The implemented troubleshooting program identified four major allestones needed to resolve an issue. These are as follows: 1 L. -.___.--_.,, .. . . , - , _ . - . . _ _ _ __ ..,__-~ _ .._ - _ _ . _ - -.- _ _. -- --,-. ,.- _ -
. . I. 4. 1.~ Troubleshooting Action Plan' s Following a description of the question, issue, or problem being. investigated, a summary of information supporting the probable cause is developed. Included is a review of the components - maintenance,' surveillance testing, and modification history.
. From this body of information,'the potential root cause(s) are developed and an outline of the troubleshooting plans to prove / disprove each is presented.
- 2. Engineering Report
.This is a report of the results of the troubleshooting efforts and provides the conclusions and justification of the identified root cause.
- 3. Repair Action Plan Once concurrence is obtained that the root cause(s) have been i identified, then those steps required to repair, change, or
- modify the apparatus or procedure for return to service are developed. This step is imposed to insure that repairs / changes are ' properly coordinated and that troubleshooting is complete and sufficient to allow repair.
f l 4. Action Item Closure Report . } This report consolidates.all of the developed information from the above efforts and completes the explanations and j justification of root cause(s) of the item. It also provides j for recommendations which will be useful in the development of ! lessons learned and programmatic improvements to guide in achieving excellence in operations and management while l precluding recurrence. I ! Note, root cause, as used in this section is defined as the direct
- cause of the failure, as1 function, or discrepancy. It is not i
necessarily the programmatic or underlying cause that allowed the-failure, malfunction, or discrepancy to occur. That " root cause" is i
- determined by the Rancho Seco Incident Analysis Group. Section VII l of this report discusses the Root Causes developed by the IAG.
l For those items placed upon the SMUD NRC/IIT Quarantined Equipment l List, or if so designated on the Action List (Attachment 3). the 4 entire four phase program described above was implemented. Several items were added to the Action List for tracking and management which
- did not require the full program. In those cases, a Closure Report alone was required and used to document the scope of the analysis, or '
investigation, and to report the conclusions which resulted from j analysis of the. item. I l
- i l
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p x ? I.- [5. JBW Owners Gr:up Regulatcry Rssp:nso Gr:up (RRG( Sinc'e the ICS Is generic to the ~BW designed NSS System, the NRC addressed its questions, which were of potentially generic interest
- to -all BW plants, . to the RRG. This 'roup g met and/or communicated with the NRC,' and \in general, discus'sed issues which are beyond the scope of this report.
, . . i
- 4
- 6. Region V Confirmatory Action Letter On December 26, 1985, the.NRC Region V Administrator sent to'the District a Confirmatory Action. Letter requiring that Root Cause Analysis of the event be done, and-that the NRC be briefed on the L root cause and provide the justification as to. why the facility is ready to resume power operations. This same letter established the I
hold ,of repair work on equipment which malfunctioned pending , j evaluation by SMUD and NRC inspection teams.- See Attachment 4. NRC Region Y has been briefed on the results of the Root Cause i i investigations and the justification for resumption of power ! operation. '
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II. SEQUENCE OF EVENTS j i! At the time of the event, the Ball y 855-Process Computer was out of' service with problems in it's disk drive control circuits. As a' result, j it was unavailable for data collection and this . source of sequence or events and plant process data was'thwrefore unavailable. The IDADS l (Interim Data Aquisition Display System) rcomputer system was available and performed well, providing a comprehensive view of the sequence of events ' and plant conditions through the event. One consideration was that the analog data points that IDADS obtains.from the Bailey 855 were not viable
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- i during this event, although much of the'information was available.from
[ charts and recorders. j j Extensive interviews with the operators supplemented their usual reports r. ;
- prepared following a reactor trip. While these reports do not have the % U j high resolution time discrimination of computer data, they aid in gaining
- a full understanding of the actions leading to identifiable events and may '
i provide the reasons for actions taken. i Attachment 1 is the Sequence of Events as developed by the. Rancho Seco ! staff. The Sequence of Events issued by the NRC/IIT followed that i developed by SMUD staff, but does ipelude several additional observations l and comments from the perspective'of the NRC/IIT. ,J } g g The TNPO SER includes a summary of the Sequence of Events which did not [1 benefit from the longer preparation and study period available to the I i Rasho Seco staff. This was appropriate given the early publication date.
- offthe SER.
! Attachment 5 is the set of curves developed from the post-trip data base. These are useful for gaining a full understandirig of the plant parameters surrounding the Sequence of-Events. ( l -- 5-
- A yn -ewe-e o,w-e,-,, ,,,-e vw yy , m, .nmme, -w,- w-- . .,w-x,-,Enw-., ,-wn---re jw h w a m m. u, ,,,,- w -r-ww-,,,-,,,4-e
<u ms w + , e es o m e. --~-e x vev,,,,eww----.-
. m III. -ACTION LIST InLthe very active period following the event, during which the issues were being identified and troubleshooting plans developed, daily meetings:
'of.those persons' responsible for resolving.the issues were held.- The Action List served as the means of tracking the progress on a specific
, . issue while providing a forum to consider the significance of the issues and need to expand the scope, or to address additional issues and concerns.
As a living document, the Action List went through numerous revisions. As a prerequisite to startup, or return to power operations, those items identified as 'SU", i.e. "Startup Required",- have been deemed to require
" closing". Attachment ~4 provides the Action List, current as of the date of issue-of this report.
A discussion of each startup required item, or section on the Action List, is included in a following section of this report. Those discussions summarize the troubleshooting done, the identification of root causes, and the basis for determination of the effect upon return to power operations. A copy of the Action List is provided as Attachment 3. IV. QUARANTINED EQUIPMENT LIST The Region V Confirmatory Action Letter, Attachment 5, established the requirement to "... hold in abeyance any repair work planned on equipment that malfunctioned during.the incident ...". In' response tu this requirement administrative controls were immediately placed upon maintenance and operation of that equipment which might in any way have been associated with the event. These broad boundaries were refined by the SMUD NRC/IIT to apply to the fol. lowing equipment:
- 1. Integrated Control System (ICS)
- 2. Power to the ICS DC Power Supplies and Monitor
- 3. Main Steam Pressure Relief Valve, PSV-20544
- 4. Auxiliary Feedwater Flow Control Valves, FV-20527 and FV-20528.
- 5. Auxiliary Feedwater Manual Isolation Valves, FWS-063 and FWS-064.
-For each of these items, except PSV-20544, detailed troubleshooting plans were developed and presented to the NRC/IIT. Discussion with the Rancho Seco staff resulted in concurrence to issue those Troubleshooting Action Plans for implementation via a detailed Maintenance Instruction (NI).
These mis were provided to the NRC Resident-Inspector to allow for NRC monitoring and observation of the troubleshooting activities. l-
- t IV. QUARANTINED EQUIPMENT LIST (Continued) ,
i. In the case of PSV-20544 engineering evaluation concluded that the valve had operated properly. Since the study was not complete prior to the IIT leaving the site, it was agreed that the valve would stay on the list pending submittal of a " Closure Report" justifying its removal from the list. This was subsequently done. i The effect of hav:ng an item on the Quarantined Equipment List was that i it's operation was restricted to safety needs or activities specified under the direction of the approved Troubleshooting Action Plan. No
- servicing, clearances, configuration changes, surveillance or other work was allowed. Work requests which might effect this equipment were held by scheduling. The equipment itself was identified by appropriate barriers.
- and signs.
As of February 14, 1986, all plant equipment had been removed from I quarantine, with three components removed and identified for subsequent analysis by an independent. laboratory. These' are the ICS Power Supply Nonitor and the ICS Power Switches "S1 and S2". Replacement components meeting operational requirements have been installed in place of these
- removed items.
- V. RESOLUTION OF ACTION LIST This section will review each of the fourteen sections of the Action List. It will discuss the troubleshooting done, identify root causes found, establish the status of the item, and will evaluate its effect upon the return to power operation.
l
- 1. Post Trip Report (AP.28)
! There are numerous individual items identified in this section. Each
- will become an element of the final Trip Report, a document which
! compiles the technical basis for the event and establishes the ) historical record for its understanding.. It is not intended to replicate the Trip Report herein; however, the items are briefly discussed below.
- 1. Main Steam Line Analysis k .
j During the transient, OTSG A was overfilled with auxiliary i feedwater. Approximately 11,000 gallons spilled into the "A" main steam line. Because of the large difference.between the i temperature of the steam line and auxiliary feedwater, there was a concern that high thers'1. stresses may have occurred. In addition, consideration was that the flooding may,have caused
" water hammer". Water haimaer noises were heard in the Turbine Building sometime after the flooding had occurred..
To resolve the thermal stress concern, a stress evaluation of the main steam line was performed using an augmented Class 2/3 l , l
V.. 1. .1. (Centinued)
' fatigue. analysis. This evaluation considered.the loads imposed by thermal stratification caused by cold water in bottom half of the steam.line and thermal ~ gradient stresses caused by the l thernal transient. Fatigue usage factors were calculated for' I two cycles of this event, plus operating basis earthquake (08E),
design basis earthqvake.(08E) and .1000 cycles of pressure / I temperature loads. The results of the stress evaluation i indicate that the fatigue usage factors for the most critical j ~ componer.ts in the steam line are all below 0.3. compared to an allowable limit of 1.0. i To resolve the water hammer concerns, walkdowns of the "A" main t i steam line and the "A" main steam line bypass to'the condenser i were performed. A walkdown similar to those assembled for
- previous I&E Bulletin 79-14 walkdowns, packages consisting of i check lists and pipe support' drawings were. assembled for each walkdown. Using the pipe support drawing as a reference, a check of the support configuration was performed. This as found
- ' configuration was then compared to the as found configuration of the I&E Bulletin 79-14 walkdowns. No visible evidence of-water i hammer was found on the "A" main steam line or the "A" main l steamline bypass to the condenser as a result of the conf.iguration comparison.
I i One anomaly was noted where the bypass line penetrates the , Auxiliary Building wall. At this Jocation, the sheet metal ' flashing covering the wall penetration was partially pulled from. ! the wall. Since no apparent damage due to water hammer was , identified, the pulled flashing may be due to thermal expansion j of the bypass line. To ensure that water hammer did not occur in the area of the flashing, a detail examination of the welds and piping of the bypass line is being performed. This examination includes approximately fifteen feet of pipe in each direction from the penetration and encompasses several changes t in piping direction. Any postulated water hammer in this area
- would have been limited to this run of pipe.
l The results of the stress evaluations'and the walkdowns indicate ' that the "A" main steam line and the "A" main steam bypass line to the condenser did not experience unacceptable stresses or l water hammer. The follow-up verification described above for the bypass line will be completed prior to resumption of power ! operations.
. Evaluation of a separate potential effect of water in the steam main, that water may have been injected into the AFW Pump Steam ,
Turbine was evaluated. The result was that no water was injected. This was due to the fact that less than 13,000 gallons of water was spilled into the A-Steam line, and over i 19,000 gallons is required to reach the line to the AFW steam ' turbine. The turbine continued to operate well throughout the event. 4 5 e g a ,w -. - -., l- em - ,, ,-,c- , - , ,.s., ,,. -u--,, , ,- - - , + - - n L ,. m ,,-
__ . _ . _ ~ - _ _ __ . . . _
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t V. 1. 2. Minimum Pressurizer Level l Following a typical reactor trip, the RCS average temperature drops from 582*F.to 545'F. This reduction .in temperature
- corresponds to an increase in density of the cooling water which 1 . requires additional volume to keep the RCS flow path solid _with subcooled water. The- required mass of water is drawn from the Pressurizer until the level is reduced below normal,-and then
- the Pressurizer Level Control Valve opens to provide makeup
] water from the Makeup /HPI System. If the operator determ;nes the rate of Pressurizer level l
- i
- decrease is greater than desired, other HPI Injection nozzles l are opened, providing additional flow paths into the RCS. I Additionally , _ HPI pumps are started to further augment the l supply of RCS Makeup. ;
l During the December 26 event, all of the preceeding occurred, as : the operator observed the effects of overcooling being caused by the large quantities of Auxiliary Feedwater being provided to the steam generators. Normal post-trip reduction in Pressurizer Level was exceeded, and the level decreased to below the lower leve.1 taps on the Pressurizer. Although the level instruments
- indicate zero level, approximately 260 cubic feet of water remains in the pressurizer, including that in the surge line i which connects the Pressurizer to the RCS piping. To obtain a ,
full understanding of the actual RCS volume during the period ' when the Pressurizer was indicating offscale low, mass balance calculations were performed.- These calculations included the , effects of Makeup /HPI flow, letdown flow, RCP seal injection
. flow, and the temperature and pressure within the RCS. The
- results of this calculation show that for a few minutes the RCS may have withdrawn _the water from the pressurizer and surge line which would indicate that steam voids may have formed for several minutes during the overcooling.
, The degree of subcooling is -important, for if subcooling is lost l then the situation is changed as the Emergency Operating i Procedures treat the loss of subcooling prior to any overcooling or other symptoms. During the December 26 event, approximately 80*F of subcooling existed during the period of jnterest, and this margin was being increased through the duration of the- i event. This fact alone demonstrates that the core was properly and effectively cooled during the event. If at any time during this event the RCPs would have been lost 1 or shut off, the steam volds in the, system would not have been larae enough to have impacted the natural circulation cooling ; process which would have been' instituted. Previous operator ; i training would have been appropriate to provide core cooling and l re-establish pressurizer control via;the injection of HPI water ' and heat removal via the OTSGs. t
- g --
l l l
. -, --,3--.--a__ +, _: + , . . - . - .- - , , , . . - - - . - - .,, .
e V. 1. 2. (Continued) In swumary, the fact that pressurizer level indicated offscale
' low, for approximately five minutes during the event, did not
, ~cause the operator difficulty or confound the recovery and ! control efforts which were in progress. f 3. Pressurizer Heater Operation i As a result of the overcooling, water in the RCS contracted causing water from the Pressurizer to be drawn into the RCS. At an indicated level of 40 inches. a lovel switch interrupts the supply of. electricsi power to.the Pressurizer Heaters. This is i necessary, because the heater elements would be damaged if l
- operated while not submerged in water. Since it was observed j that Pressurizer Level went well below the heaters, it was- i prudent to verify that the protection device had functioned' l properly.
4 No direct recording or monitoring of this protectiva device exists. Thus the condition of the heater elements themselves was the only way to verify functioning of the protective interlock. Measurement of the individual heater element resistances showed no significant change from their historical valu'es; thus, it was concluded that the interlock functioned as intended and that the heaters remain serviceable. -Note that following the event, the plant was stabilized and no problems were experienced controlling RCS pressure using the Pressurizer Heaters.
- 4. Control Room Instruments Affected By Loss of ICS DC Power
. A complete list of the Control Room instrumentation affected by ICS DC Power is provided as Attachment 7. This list was !
i' developed during the troubleshooting effort following a comprehensive series of tests and observations. A drawing study provided a similar list. It had nat been generally known that the Auxiliary Steam Reducing Station received its demand f rom a controller powered from the ICS. Since auxiliary steam is needed following a reactor trip for pleat services, such as Condenser Air Ejectors and Turbine shift seals, it was determined to retain this , feature and manually throttle the supply when desired. A review
- " of the-effect this station had upon the overcooling showed that it was small. if any.
l The Auxiliary Steam System normally obtains-itsisteam from the main steam system following a turbine trip. In this case,'with the demand set to 50%, excessive steam flow occurred which caused one of the two Auxiliary Steam relief valves to open , until the main steam pressure had decreased and/or the: load on Auxiliary Steam, (condenser hogging ejectors, turbine seals, l 10 -
F , V. 1. 4. (Continued) . __ etc.) increased. This control station is not being modified or-
- repowered pending a full study. In any case, the impact on RCS 4- cooldown is limited.
- Other than losing ICS control of the normal process devices'(AFW
- Valves, Atmospheric Dump Valves (ADVs), Turbine Bypass Valves (T8Vs), Main Feedwater (MFW) Pumps, Feedwater (FW) Values, etc.), it was'also noted that the " Generator Frequency Error Indicator" and " Main Feedwater Loop Flow Recorders" also provide erroneous indications. These are acceptable conditions since
- a. The Reactor / Turbine trip terminates electrical generation, hence Frequency Error is meaningless.
- b. Loss of ICS Power causes the Main FW Pumps to go to minimum speed, hence, Main Feedwater flow to the OTSGs has l stopped. Alternate Main and Startup Feedwater flowmeters
- are located on the same panel which are not affected by the
. ICS, or its power supplies. In addition, the flow indicators move to the 50% flow position and any operator action based on this excessive indicated post-trip flow would be to trip MFW pumps, an appropriate action required
- by the " overcooling" Emergency Operating Procedure. Plant i training materials are being revised to include this
! information. l S. Investigation of Report of Smoke Prior to Reactor Trip. This issue concerns Operators' reports of the smell of smoke in j the controlled area shortly before the loss of ICS power. The
; purpose of the investigation _was to clarify the Operators' 1
observations and to determine the relationship between those observations and the loss of ICS Power, if any. Although the scope of this investigation did not determine the . source of the smelled smoke, it.is concluded that this was not a factor contributing to the loss of ICS Power. Room 211, where the smoke was sensed, is isolated from any significant electrical power equipment. Two walls-and,a corridor separate the electrical equipment rooms (om 214 and Rs. 216)-from the. ventilation equipment. - Furtnermore, the electrical equipment i rooms have separate heating and ventilation systems from the ( radwaste areas, which includes Room 211, Smoke sensed in Room 1 211 would not have originated in the electrical equipment rooms. The Operatnrs said the smell was like that from wood or paper burning, but that it did.not smell like either electrical-L equipment or insulation burning. They also noted that the smell l~ went away as soon as they exited Room 211~. Investigation of the l room ano adjacent areas showed no evidence of smoke or fire. l Thus, the report of smelled smoke appears to be unrelated to the loss of the ICS power.
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/ -
_ . _ . _ _ . ~ l 6. Primary to Secondary Leak Investigation Prior to the' event, the Condenser Air Ejector radiation monitor l was indicating approximately 400 cpm of activity.- This is a low value, however, it is above background and may be indicative of
- a very small primary to secondary leak.
)
During the event; review of the main steam line radiation 1 monitors did not indicate any changes in activity which would be expected from a significant primary to secondary leak. Following the event it was deemed prudent to monitor the OTSGs, l while in wet layup .to evaluate and quantify identifiable i leakage. These results were inconclusive as some minute traces i of primary radioisotopes were found, so a more sensitive ' p technique was applied. Helium leak detection did identify two very small leaks lin the 8-0TSG. Investigations', using eddy current techniques, are being used to characterize the leaks and to provide confidence that any incipient leaks in either OTSG are found. The size and number of leaks identified are commensurate with the amount of activity observed in the secondary plant prior to December 26.
- 7. Operations / Security Interface During Transients !
. During the event, one of the operators lost his security badge. He notified Security and they escorted him until a replacement badge was issued. During the event the security area access card reading time at - the Control Room door began to impact the timely dispatch, and reporting back, of operators as they established manual control : i of the various equipment outside the Control Room. The Shift , Supervisor requested Security post _a guard at the door, thus l allowing it to remain open for the duration of the' event. A ' security officer was provided. I These hinderences did not significantly affect the outcome of I the event. Yet it'is apparent that improvements could be made to ensure that Security is a part of event response and that effects of anticipatible occurrences,111ke misplaced badges, could be minimized.by appropriate plans and procedures. Enhancements in the Rancho Seco Security Plan and Procedures are being developed which will improve the support provided by each organization to insure optimum response to plant emergencies and transients.
- 8. Main Steam Line Failure f.ogic (MSLFL)
Post trip review of- the IDADS alarm printout showed that OTSG - pressure was approximately 370 psig when MSLFL alarmed as actuated. The setpoint was 435 psig. In the ensuing analysis
, =.
V. 1.- 8. (Continued) . .
.and troubleshooting, it was found that the MSLFL is monitored by l IDADS once per minute. During the period of interest, OTSG i pressure was decreasing approximately 70 psig per minute. The l
' concluston is that the MSLFL operated'as intended. .The IDADS
' sample frequency for the parameter is going' to be increased to improve its resolution.
Separate analysis determined that a higher MSLFL setpoint is desirable to insure that the condensate pumps will not'begin feeding a low pressure OTSG. . Engineering has developed,'and Operations implemented, a new setpoint of 575 psig which will accomplish this purpose. It is important.to note that MSLFL operates independently of the ICS and overrides any ICS demand signal, to the Main and Startup Feedwater Flow Control Valves, directly causeing thet to close. MSLFL is independent of ICS and ICS power.
- 9. SPOS vs Strip Chart Recorder OTSG Operate Level Several steps in the plant Emergency Operating Procedures (EOPs) have the operator taking specific actions based upon indicated
, OTSG Operate Level. An instrument of significant benefit to the operator, and utilized in this event, is the SPOS (Safety Parameter Display System). It is capable of an integrated .
- ., presentation of several parameters simultaneously and allows the I
operator to observe trends in pressure and temperature, and anticipate developing conditions. l As a part of this SPOS display, a numerical value representing l OTSG Operate Level is provided. Several of the operators reported that these values did not exceed 97% and 93% on the A-0TSG and 8-0TSG respectively yet other Control Room o indications were that Operate Levels were in excess of 100%. , The SPDS uses the same uncompensated level source data as do the
- other Control Room instruments. There is a difference in the way temperature compensation occurs, the SPOS calculating it "
{ internelly based upon the OTSG pressure and a steam tables algorithm. Simulations over the range of interest, using the algorithm showed good co relation with the other Control Room 4 instruments, although the SPUS Operate Level tends to be 1 to 2% lower than comparable strip chart recorders and the computer. It was concluded that, while the SPOS was likely indicating less f than 100% Operate Level, when tne level was actually higher, the maximum -2% error would not adversely isipact operator actions
-directed by.the E0Ps since a'ction-is required upon exceeding
-95%. The operator observations of SPOS Operate Level, during i the event, were not benchmarked against other indications of level, with the result that no discrepancy could be verified. '
This study suggests t, hat even with actual levels in the OTSGs i i l' l I s
6 4 LV.- .l. 9. (Conti,nued) greater.than 100%, the SPDS will' indicate a value just under 100%. As discussed _above, the operators: Were aware that OTSG-Operate level was greater than 100%, .hence the observation-
- investigated here.- The consequence of the slightly low reading does not adversely impact the proper ano prompt implementation of the E0Ps or mitigating action.
- 10. Transient Campared to USAR Design Basis Loss of ICS. Power-is not a specific event considered within the plant design basis as reported in the facility Updated Safety-Analysis Report (USAR). - The plant design basis considers " worst case". scenarios.for non-safety equipment; Should any of the plant safety parameters tue exceeded, the - ,
Reactor Protection System shuts down the reactor independently 4 of the ICS. l Should the RCS pressure boundary integrity be challenged. .or the containment pressure be excessive, the SFAS/ECCS (Emergency Core Cooling System) operates independently of the ICS to -insure core integrity and isolation from the environment. For comparison, the analyzed event Main Steam Line Failure, begins with a :large uncontrolled steam release followed by a l rapid cooldown. The pertinent parameters and results-are: USAR Chapter 14 , Main Steam Line Failure Analysis , Assumotions and Results 12/26/85 Event 1% defective fuel rods. .<<1% defective fuel rods. Complete severance of 36' Nain 6 code safety valves open Steam pipe. Duration of for <1 minute. 2 ADVs and release, 4.5 hours. 4'TEls 50% open for-<10 , minutes. Steam line break'between Reactor- Open valves located in the Building and a turbine stop same portion of the main valve. -steam system. t 490*F reached in 43 seconds. ~ 490*F reached in 6 to 7 ,- minutes, i
- . 10 gpm primary to secondary <0.001 gpm - no measureable ',
, leakage. leakage. !
Site area dose thyroid 12,700 Site area thyroid dose 0 mRes. Whole body 46 mRes. mRes. Whole body 0.02 mrem.
.14 - -=y-' + 1 r--- --
e ae---* vw w -, +4 *,e +wa-+- =we.~wn- e w--- e -m =-v---e+e- <------ +* - - - - ' + - -
~
V. 1. 10. (Cont 1nued) A consequence of such a rapid cooldown is the likelihood of exceeding the Technical Specification limit restricting normal heatups/cooldowns~to 100*F/hr. That specification was exceeded during the December 26 event. A full discussion of the significance of having exceeded this limit, and a related phenomenon, Pressurized Thermal Shock (PTS) is included in a following section. To summarize here, there was not excessive
" usage". imposed on the RCS or the Reactor Vessel by this transient, and the Technical Specification limiting Nil-Ductility Temperature (NDT), was not challenged. The NSSS was not adversely challenged by the December 26 transient.
In conclusion, this event was within the bounds of the analyzed. event, and its environmental impact was a' tiny fraction of that resulting from the analyzed event.
- 2. Human Factors Analysis The District's Human Factors program was applied to this event for the purpose of collecting the Operators' input and observations of the effectiveness of integrating the Control Room, plant equipment, procedures, and operator training and performance.
This process identified 26 items which were worthy of study. Of these,12 had been previously identified-in the Control Room Design review process, and the other 14 were on items beyo'nd the scope of that review. A complete listing of the 26 items is provided as A.ttachment 6. Disposition of the individual
. recommendations is being handled by assigning appropr.iate priority and resources to each. Those which justify immediate resolution have bee.n scheduled for resolution prior to startup.
- 3. Determine Cause/ Corrective Action. for ICS Power Failure
- 1. ICS AC Power Sources Early studies of the potential causes of the loss of ICS DC power clearly eliminated the independent and redundant supply sources of 120v. AC, single phase, 60 Hertz, power as being involved. These studies looked at the performance of other equipment powered by these sources and how that equipment performed during the event. In addition, the Automatic Bus Transfer switch did not shift, meaning that its normal positioning to the vital bus had not been changed; hence, there had been no interruption in the nornal supply from the vital bus. Attachment 8 schematically shows the ICS Power Supplies and Distribution. .
i
- -r , ,.. -- . _ _ , , , , . , . ,- - - . - , - , , ,..&. ,
p V. 3. 2. ICS. Equipment Investiga'tions Post event monitoring of the ICS, and the equipment it operates, shows no problems which would have-created the need for tripping switches S1 and S2, through.which AC power-is fed to the ICS DC i- Power Supplies. .This focused the development of Troubleshooting plans-into the operation of the " Power Supply Monitor", the 2: device which initiates the trip feature of the S1 and S2 switches. A detailed " Troubleshooting l Action Plan" was prepared which systematically. investigated potential root causes without-adversely effecting the "as-found" conditions, or ability to. redirect the investigations should new potential. root causes or + i -considerations be identified. Implementation of this - troubleshooting plan.was accomplished. following NRC-IIT concurrence, by use of very detailed Maintenance-Instructions , (Mis). . Early findings determined that the Power Supply Monitor trip setpoint was " drifting." This led to revision of the. I implementing mis to better understand the phenomenon. This
. understanding was greatly aided by the receipt of a new Power Supply Monitor (PSM) which was setup in the laboratory where a full set of performance data could be 'obtained for comparison with the insitu "as-found" data available from the original Power Supply Monitor.
Significant to the event was the finding that small amounts of a resistance, as little as one ohm, in series with the " sensed" voltage into the PSM could cause less than optimum performance of the device, i.e., " drifting," and could cause tripping, 4 similar to that observed on the insitu PSM. A full investigation'into this effect and it's likely impact on an operating ICS, has been completed. JThe finding of the resistance at the input of the PSM affects performance, and led-to troubleshoctingithe voltages and resistances within the ICS: DC distribution cabling. The result was identification of severai poorly made factory cable lug crimps on the +24 VOC supply to the PSM bus. Goe was actually so loose that it fell off of the wire when disconnected from the bus. An additional I problem was traced to the PSM +24 VDC circuits, but this was not a contributing root cause as its effect would have been "no-trip, or delayed trip." A third finding was that the S1 and - 2 S2 " time delay before tripping" was approximately one-third the intended half second, making the Si and S2 switches much more
- sensitive to transient voltage conditions.
I
- 16.-
t
- -- - ~ . - w , -.
l i l l Y. 3. 2. (Continued) l The root'cause for the loss of ICS event has' been identified as ! the loose power connection on the bus ICS Cabinet 1 supplying ' power to the PSM. .The factory bus wiring'has subsequently been
- replaced with current standard wiring. . Investigations.into the-
-PSM and it's sensitivity to _ input resistance are continuing,
. although this will be pursued as a 8&W Owners Group issue.
-The. inservice PSM and the Sl/S2 switches are being sent to an independent laboratory _for a full analysis.
- 3. Procedure for Loss of ICS l l
At the time of the event, a Casualty Procedure to respond to y Loss of ICS events did not exist. Procedures for repowering the power supplies were available, and the Emergency Operating Procedures (EOPs) were in' place and sufficient to mitigate the consequences of a Loss of ICS. A review was accomplished for the purpose of:
- a. Establish the " Failed" condition of equipment controlled by the ICS. .
b.- The desirability of " repowering" the ICS prior to post trip stabilizing of the NSS.
- c. The appropriateness of the E0Ps for mitigating and controlling Loss of ICS events. -
From this review, it is apparent that changes to the E0Ps which provide milestones, e.g., temperatures or pressures, to assure timely operator actions, would go far to preventing recurrence of consequences, such as overcooling, from happening. There is no need for a specific " Loss of ICS" E0P. Infonnation obtained upon effects of repowering the ICS, which will aid the operator
, .in the post-event troubleshooting and return to normal control, has been developed into an appropriate Loss of ICS Power Casualty Procedure.
Modifications to the ICS
- 4. ,
Modifications have been developed which will enhance plant reliability in two ways by:
. Making ICS DC Power more reliable.
. Enhancing ability to mitigate Loss of ICS Power transients from Control Room.
' .1 Within the ICS itself..the rewiring will ensure that system components receive full electrical power, and the PSM is.
being separately wired directly to the. ICS Power Auctioneer
-Panel 124 VOC buses it monitors. Failures as occurred on December 26 will be eliminated.
L-
/ _
.V.- 3 '. 4. l .- (Continued). '. .
New " human engineered" labels are~being'appliea to switches
.Sl and-S2 to aid the operator in recognizing a tripped condition and in repowering the ICS, should it become
~
necessary. While'similar. switches are used elsewhere in
. the plant. . they are uniquely position' indicated and controller under. separate procedures.
An indicator light-is being'added to the Control Room operator's panel which will show when control is shifted , away from the ICS, i.e., it illuminates upon Loss of ICS DC
- Control - Powe r.
Study showed.that there are hundreds.of years of successful operating history on this type.of.PSM, and that this one is the only one to have tripped an ICS as a result of a-failure. . In addition, this failure was external to the PSM
- itself. For these reasons, the consideration of a two-out-of-three PSM arrangesent has been abandoned.
Should the. ICS Power be' again lost, for whatever reason,
'the modifications discussed below, in conjunction with the E0Ps and Casualty Procedures, are sufficient to mitigate i the consequences with no operator action outside the Control Room.
- 2. External to the ICS, modifications are being installed which'will enable the Control Room operator to operate important valves independent of.the ICS. This is being done by installing a relay which ICS AC Power energizes through PSM contacts. Upon loss of ICS power, the relay :
connects a secure battery backed power source into the
- circuit controlling the necessary. valves. The majority of these circuits were installed as a response to Appendix R
" Fire Protection" analyses and provide a high quality.
separate, and diverse remote control without adversely effecting the Appendix R capability. The specific modifications are:
- 1. AFW Flow Control Valves ,
l A new Hand / Auto controller,-indepeadent of the ICS, is I provided for each Aux ~FW Valve. Located in_the l Control Room, it is adjacent to the existing AFW Fluw i indicators. New OTSG Startup Level. indicators which i are also being added. With the new Hand / Auto station 4 in Auto, ICS controls the valve, except on loss.of ICS Power, when the new H/A station automatically controls , the valve to the pre-established initial flow rate set , f at a nominal 280 gpm per OTSG. Placing the H/A I station in Hand allows manual control of the flow, independent of the ICS. I
,- 4 __ . - . . . ., .-_-a.-
V. . 3. 4. 2. 1. (Continued) .
.Still available are the existing controls which will cause the valve to go full open (on operator demand),
-provide ICS normal controls (when ICS is powered), or use of the local' manual operator. Remaining also is
.the parallel and independent safety grade bypass valve operated by SFAS.and controllable from the Control Room.
i Attachment 9 provides a' schematic view of the modified n AFW Control Valve, independent of the ICS.
- 2. Atmospheric 1 Dump Valves (ADVs)
There are three of these valves on each of the two
- - main steam lines. Normally, only one is.in service.
This modification operates all inservice valves in
~
parallel . It does so by interposing a selector switch in the Appendix R provided control scheme which is enabled when aLloss of ICS Power.is sensed. As a result, the valves stay in their normally closed positions, although the operator in the Control Room 4 can select back to ICS at any time. If the ICS is not powered, then the valve will open to its 50% ~ i position. By this technique, steam pressure _can be- } cycled about a nominal value. Similar and separate controls are provided at the shutdown panel, external to the Control Room, in addition to local controls, which' allow adjusting the ADVs to intermediate
- positions at the operator's discretion. . Attachment 13 provides a schematic view of the modified control circuits, independent of the ICS. Other methods of controlling these valves remain as before this modification.
i
- 3. Turbine Bypass Valves (T8Vs)
There are two of these valves on each main steam line, , and normally, all four are in service. This modification operates all four valves in parallel. It i does so by interposing a sel?ctor switch in the Appendix R provided control scheme which is enabled
~
when a loss of ICS Power is sensed. As a result, the i valves stay in their normally closed positions, althougn the operator in the Control-Room can select back tc ICS at any tiae. If the ICS is not powered, then the valves will open to their 50% position; By . this technique,' steam pressure can.be cycled about a nominal value. Separate controls are provided at the shutdown panel, external to the Control Room, which allow similar operation at the operator's discretion.
- 19 1 ,
*~
. -i- . , _ _ - - - . - - -, ,
_ ~. - _ _ _ _ - - . V. 3. 4. 2. 3. ~(Continued) , Attachment.11 provides a schematic view of the modified control circuits, independent of the IC3. Other methods of controlling these valves remain as before this modification.
- 4. Summary of ICS Modifications These modifications will provide the operator, from .i within the Control Room, assurance.that upon loss of j- ICS power the steam control valves-will not fail.open, that adequate auxiliary feedwater will be provided to '
the OTSGs, and that if desired, or necessary. . controls ' independent of the-ICS are available and sufficient t for the timely mitigation of a transient. The. likelihood of inadvertently losing ICS DC Power has been reduced, and the ability to react to a loss of ICS Power enhanced. This results in fewer. challenges to the plant safety.. systems and improved l reliability.
- 4. Makeup Pump (P-236) Failure
- 1. Cause The rapid overcooling early in the event caused the RCS pressure to decrease below the 1600 psig setpoint of the SFAS, which.
initiated HPI trains A and 8. .Since there was no loss of AC power to the- Safeguards power buses, the Makeup pump .(which' is the nornal source of high pressure reactor coolant for RCS , volume. control and RC Pump seals) remained in service. uAs a feature of the SFAS, the Reactor Building isolates on the assumption that a LOCA has occurred. Associated valving changes which occur are the opening of the suction valves from the BWST and closing the suction from the. Makeup ~ Tank. This has the t effect of shifting the MU Pump suction from the NU Tank to the BWST in parallel with the A-HPI pump. Concurrently, the normal
" mini-flow".from each of the HPI/MU Pumps is isolated from the Nakeup Tank making full-pump capacity available for delivery to the RCS.
.1 Once RCS requirements for HPI/MU had been met, the -
t operators began to reestablish normal Makeup configuration. First, mini-flow was re-established to preclude ~ damage to any pump as the pump flows were being
- throttled. This. led to a rapid filling of the Makeup Tank, the receiver of the mini-flow, which caused the operator to then close the isolation valve from the Borated Water . i Storage Tank'(BWST).
- This was done to cause the NU pump to !
i draw from the NU. Tank instead of the BWST. -Quitted was the I SFAS valve which had earlier automatically isolated the '
. . .. _ ._ . _ . _ _ _ _ _ _ _ = . _
~
V. 4. 1. .1 .(Continued) . ! 1 Makeup Tank suction. . The result was failure of the pump- 1 l approximately;three minutes later due to lack of water '
- . flow. This in turn, led to approximately.l.200 gallons of
-NUsTank water being drained-onto the pump room floor when the operator opened the MU.. Tank suction. valve. Reactor
. Coolant Pump seal requirements were met from the available, and separate, HPI pump except for a 75 second period during L
switching of HPI . gung)s following -the Nakeup Pump failure. The design of.the RCP seals is such as to successfully tolerate flow interruptions-of'this nature.
.2 Repairs and Modifications On a short term basis, the pump failure does not impact the capability to resume power operation. - Technical
, Specifications require two HPI pumps to be available, which there are, and either of them can provide the Makeup requirements. The Makeup Pump, in actuality, 1s'an
" installed spare."
- The repair / replacement of the pump is on a longer schedule, as is consideration of alarms and/or protective interlocks
'which would preclude the likelihood of recurrence. A study was performed into the basis for the SFAS closure of the MU Tank suction / isolation valve. The study, reconfirmed the.
j. isolation as necessary to prevent. the gas u' inding of-the , MU/HPI pumps in a LOCA when the BWST level decreases. Options for providing the desired pump protection are still being developed. i i 5. Damage to Radiation Monitor R-15001 i Following the SFAS by some eighty minutes was the receipt of a smoke alarm in the zone containing the radiation monitor which serves the Reactor Building during normal operation. Investigation by operators determined the source of the smoke to be the sample pump for radiation monitor R-lS001. It was promptly shut 'down and the smoke soon cleared.--Subsequent investigation found overheated sample pump seals caused by SFAS isolating the suction /discitarge flow paths. This is a condition allowed by the existing. design. Engineering is studying this radiation monitor for desirable or necessery changes, modifications, or possibly replacement. This study will include L design arrangements to prevent this type of damage. This is not a startup requirement. i l 8
V. 6. RCS Overcooling As a. consequence of difficulties Jxperienced in closing the Auxiliary , Feedwater control valves, (which were demanded to 50% open on loss of f ICS), and the resulting filling of the OTSGs, the RCS was-rapidly cooled to below its normal post-trip temperature of 545'F. The Rancho' Seco Reactor Coolant System was designed to accommodate 240 normal cooldowns-at 100 degrees F per--hour. There have been several transients during which this cooldown rate was exceeded and therefore the' Babcock & Wilcox Company was asked to determine thef cumulative fatigue usage factor. This evaluation concluded that the allowable number of remaining cooldowns, at 100 degrees per hour,
- should be reduced from 240 to 235. .Since only 31' cycles of this transient have been used to date, the reduction to 235 allowable.
cycles is expected to have no adverse impact on the current design life of 40 years. ! Confusion over the interpretation of the meaning of 100 degree F per 4 hour cooldown has prompted the District to request.the calculation of' , a new pressure / temperature limit for use in the. Rancho Seco Technical. ' Specifications. This curve will be based on step decreases in i~ temperature making it unnecessary to interpret the linear cooldown
< limit. In addition, a temperature will be'specified, above which, any transient rate can be considered acceptable.
4 Very early in the event, the operators recognized the symptoms of. overcooling and initiated actions necessary to restore normal- , post-trip conditions. The delay in terminating Auxiliary Feedwater ' flow to the steam generators took the RCS into the Pressurized < i Thermal Shock" (PTS) region, identified on a operational curve showing desired operating parameters, in the cooldown procedure. This means that the pressure and temperature parameters of the - 4 transient were outside the bounds of the pressure and temperature I parameters used in the pressurized thermal shock study in 1982. Other parameters of the RCS were well within the bounds of those .
!- parameters (e.g. end of life material properties, etc.) used in that study, BAW 1751.
The evaluation performed by.the B&W Company showed that,for the entire transient duration, the allowable pressure was above 2750 psig, the design pressure of the RCS. During the cooltbwn trarsient, c the pressure never exceeded 1700 psig. This demonstrates that the reactor vessel was not subjected to pressurized tnermal shotk as a result of this transient. 4 i s
. , - - - - , - , e ~ -- .# , ., - ; a .&,, . , . , . ,m
i ( 7 V. 6. (Continued) An. entirely. independent analysis of'the Reactor Vessel beltline was-done by the Electric Power Research Institute (EPRI) using the draft ASME Section XI Appendix XX " Evaluation of Unanticipated Operational Transients." That analysis demonstrated' adequate structural. integrity for the Rancho Seco vessel as long as RCS pressure did not exceed the design pressure of 2500 psig, nor T -RTNDTS.less C than , 55'F. These requirements were met, the~ minimum T -RTNOTS C being l = 169'F, and the maximum pressure was less than 1700 psig.
- 7. Health Physics and Control-of Contaminated Air and Water.
Sources of potential release and/or exposure.to Radioactive materials
- during this event were as follows
- Secondary Plant Steam Relief to. Atmosphere
- Secondary Plant Condenser Air Ejector
- Spilled Makeup Tank water in Auxiliary Building
*' Flooding of Waste Gas Header with Makeup Tank /8WST_ water, i
Eaca of-these potential sources is discussed in turn.
.1 Secondary Plant Steam Relief to Atmosphere Prior to -the event the quantities of radioactivity in the Feedwater and Steam were just at or below detectable levels. As a quantifiable pre-existing primary to' secondary leak did not exist, the steam release doses.were determined by post-trip radioanalysis, which was based upon Cesium being the major contributor. This Cesium was residual in the secondary plant as a result of previous primary to secondary leaks. The secondary release averaged over a one hour period (actually the reliefs were open for less than ten minutes) which resulted in a source at.the site boundary of 0.85 MPCs basqd upor.' insoluble Cesium as the major contributor. There was no radiciodine in this release.
i .2 Secondary Plant Condenser Air Ejector r Any radioactive gases are collected -from the secondary plant. in , the condenser, and are discharged to_ atmosphere through the l Auxiliary-Building filtration equipment and monitors. Post event analysis of the charcoal cartridges, in service durina the event, did not show any radioiodine. As these monitor both the Air Ejector and the Auxiliary Building Exhaust, it.is concluded that no radiolodine was released and, therefore, there was no
- thyroid dose.
4 i
~ ,- --_e-__ - _m- < ._- --m.-- .- ~ e- e- , - - - i.- -- -, -,- - - - - .- ,%.-. - - -.---=,y
~ . ._. __ _ _ . _ .
V. 7. .3' Spilled Makeup Tank Water in Auxiliary Building .
~ , . When the Makeup Pump failed, and the Makeup Tank Discharge Valve !
was subsequently opened, approximately 1,200 gallons.of Makeup Tank water spilled onto the floor of.the Makeup Pump Room at the
-20 foot level of the Auxiliary Building. This was a.
coa 61 nation of filtered, cold reactor coolant, and.BWST water. Radioactive Noble gases contained in the coolant were released into the Auxiliary.8uilding/ Exhaust system where they are monitored before being discharged to atmosphere. Total release of noble gases was determined to have resulted in a maximum 0.93
- MPC at the site boundary.- The whole body dose to a person at
~ the highest downwind sector at.the site boundary for the four hour length of the event was determined to be less than 0.02 mrem.
Approximately twenty minutes into the event, an Auxiliary Operator and an Equipment ~ Attendant entered the Makeup Pump Room 4 to access damage and isolate the Makeup Pump. -Total time in the room was about six minutes each. Due to the perceived need to promptly isolate the Makeup Pump, and since Self Contained Breathing Apparatus was not readily available, the entry was ; made without respiratory protection. Subsequent whole body ' l counting identified 1.0% and 0.3% Maximum Permissible Body , Burdens for Silver-110M, while they received external exposures i amounting to 20 and 10 mrem respectively. In addi. tion, one ; individual had contamination to shoes, socks, and trousers, and the other had contamination on his shoes ~and hand only.
~
These releases and exposures were undesirable side effects of this event which did not significantly impact the health and safety of the public or the plant staff. Subsequent training, ,
- l. procedure changes, and equipment allocations have been l l implemented to preclude recurrence.
, .4 Flooding of Waste Gas Header When HPI/MU Pump " mini-flow" was re-established, as a first. step in returning to normal lineups following the SFAS initiation, a consequence was the direction of as much as 300 gpm of-BWST water into the Makeup Tank. As a result, the tank began filling which caused the operator to secure the A-HPI pump and cicsc ! SFV-25003 from the BWST assuming.that the Makeup Pump would then draw from the Makeup Tank. This action precipitated- the damage to the Makeup Pump. Since the Makeup Tank discharge valve to the. Makeup Pump was closed, the tank soon overfilled and discharged through its relief valve into the Flash Tank and through its vent into the Waste Gas Surge Tank and into the Waste Gas compressors. This entire system is designed to accommodate the accumulation of water and as a result no i .
i
- I
! . T i- . . . -
A V. 7. 4. (Centinu:;d) .
]
significant damage was done. After the event, the water was. j drained and a complete Helium leak test of the system and-its - l components and piping completed. A single'small leak, external' I to one__of the Waste Gas Compressors,.was repaired by disassembly i and reassembly. Only normal consumable: parts were used. No damage was-found. The leak.likely preceded the event. No unusual contribution to radioactive material releake resulted ' from over-filling of.-the. Makeup Tank, and recent procedural
~
changes are expected to preclude recurrence. I
- 8. Emergency Plan +
l
.1 Implement'ation i The initiation of SFAS, at a little over three minutes.into the event, was noted by the Shift Supervisor / Emergency Coordinator as requiring declaration of an Unusual Event per.the facility Emergency Plan. Fifteen minutes 1s-Allowed for the formal notifications to be initiated and this requirement was met.
Follow-up notifications were not performed with the result that the Counties expressed concern that there was "too much unco'nfirmed information to feel comfortable, yet too little information to make a decision." Analysis of the implementation of the Emergency Plan in this' instance'has shown several enhancements which will aid the operators when confronted with complex and rapidly evolving conditions. The Emergency Plan supporting agencies and equipment were judged sufficient-to provide timely communications, actions, and emergency support.
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Improvements in documentation were identified which will enhance the quality and timeliness of communications as required by the plan. The District is responding to the issues surrounding this activation
- of the Emergency Plan with additional training for-the operators i
on Command and Control, in addition to retraining on the , Emergency Plan, its recent improvements, and its effective implementation. l l . 2 Technical Support Center Fire Pre-Action Sprinkler.Sjstea l , A separate observation during this event was that, upon a ' Reactor / Turbine Trip, when'the source of inplant power transfers l from the Auxiliary to Startup Transformers, the motontary depowering of the controls to the fire pre-action sprinkler system for the Technical- Support Center (TSC) occurs. This..
- trips the sprinkler water supply valve and causes its pilot actuator to vent water. This water then spilled onto the TSC floor. - Operations responded by resetting the sprinkler controls and placed plastic sheeting over equipment preventing any damage., Recent modifications have provided a suitable hard pipe -
drain to the control valve pilot while the electrical design is being studied to provide a method of precluding valve actuation when power transfers occur.
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V. 9. : Training The large body of information developed dring the. study and
' troubleshooting done following the' December 26 event, provided the opportunity and need to communicate the salient points to the plant staff. . This was done by developing specific training lessons for ,
, implementation by the training department. The major issues and -- topics taught by that training are described below. i
.1 Sequence of Events Training -
The objective was to ensure that all plant operations personnel have the perspective of the cause of the event and the transient; which ensued. The lessons were given in a two-part format. The first was a detailed review of the actual sequence _of events and the occurrences which took place. The second lesson involved , detailed discussions of the events and actions, with emphasis on i the difficulties and problems which the operators faced while
; mitigating the event. This training was completed by mid-February.
4
.2 Plant Modifications The' modifications covered are those to the following components and systems: ADVs, TBVs, AFW flow control valves and ICS'& NNI
+ power supplies. The basic objectives for the training are: L - Describe the purpose of each modification.
. Describe the controls, interlocks, and operation of the modified system or component.
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Describe identified failure modes and. conditions. Understand related procedure changes and requirements.
.3 Emergency Operating Procedure Changes The plant E0Ps are " Symptom Based" procedures based upon the B&W Owners' Group developed ATOG (Abnormal Transient. Operating Guidelines). Primary focus of this approach is to address the symptoms of abnorinal plant response typiff ed by " overcooling,"
"undercooling," or " loss-of-subcooling." The symptoms of these-situations are readily observable by the operator on available plant instrumentation. By freeing the operator from the
-constraints imposed by " event" procedures (e.g., must first identify which of the numerous ways " overcooling" can be' caused, j Loss of ICS, Smill Break LOCA, Main. Steam Line Failure, MFP Overspeed, etc.,'and then select appropriate response procedure), timely response and mitigation is accomplished.by focusing activities to restore stable plant conditions which
, -insures that the core is receiving proper cooling. l
1 JV . .9 .3 ' (Continued) . . l The "cause* of.the " event" is then sought and corrected ' following the. stabilization of the core cooling requirements, ) separate from the " Emergency" actions. l .
. I During.the December 26 event the E0Ps were utilized and applied to mitigate the event. ' Overcooling was not terminated in a ,
timely manner due to ' lack of clear criteria for taking the action specified. to "~... trip appropriate pumps."~ This i, condition has been rectified by developing " milestones" for the operator which will be effective in insuring that actions.will be timely in accomplishing.the-EOP's pcrpose. As.an example, on
, overcooling the procedure causes isolation of.the services and loads upon the OTSGs. The-operators were proceeding to do the necessary valve manipulations when difficulties and delays were i experienced. The' revised E0Ps clearly state that if isolation '
is not effective and OTSG Operate Level exceeds 951, or RCS Temperature is below 525'F or Pressurizer Level is less than.10 inches, then trip the pumps still supplying flow to the OTSGs." These changes are within the bases of ATOG and are procedural
- enhancements which will improve their implementation.
The basic objectives for the training will be:
. Description of changes.
. Basis for the specific changes.
1
. Effects en plant control.
. Dynamics of Response (Simulator Training) l
.4 Loss of ICS Power Casualty Procedure Such a procedure did not exist at the time of this event. A procedure has.been wirtten and addresses those actions necessary to restore ICS power. Significantly..this is not intended to be e attempted until the E0Ps have stabilized plant conditions and then repowering will occur.
It is not appropriate to install a Loss of ICS Power procedure as an " Event" procedure for the reason that the fault may be within the ICS and it may not be available for repewering. The approach being taught is including the'use of the new modifications being installed which provide power and controls (separate and independent from the ICS) 'in the Control Room which will insure that the demands from the ICS (during repowering) will not cause a subsequent transient. 9 27 - p>w - w.-yye-w wn + ,ym-r .yw -p e.-
i -
- V. .9 .4 I (Continued)- .
The. basic objectives for the training will.be:
. Review the power supply system.
. -Describe the' purpose of the procedure. !
- Describe the basis for actions.
4
.5- General Procedure Changes Various Operating Frocedures,-Casualty Procedures, and Administrative Procedures are being changed as a result of the lessons _ learned from the transient. The training conducted on these changes will follow the following basic objectives:
i . Describe the purpcse of;the procedure change.
. Describe the change.
. Describe the basis for the steps.
.6 Emergency. Entry Into Areas of Unknown Radioactive Conditions Following damage to the makeup pump during the event, operators entered the pump room to isolate the pump and assess damage.
Evaluation of this entry identified the need for several cheliges , 'in procedure, in addition to restatement of manacement policy. Training will address the changes made in the requirements and , roles of health physics.and operators, availability and use of . . protective equipment, and the procedures which govern work in. . areas of unknown radiological conditions. In addition, a health
- physics technician has been dedicated to supporting operations.. '
.7 Manual Valve Operation This lecture will specifically cover the operation of the ADVs :
TBVs, and AFW Valves. It will focus on management's policy that
" valve wrenches" or " cheaters".will .ot.be used on devic.es which include " mechanical advantage," such as linkages, screw Jacks, and levers. The training will include:
l
. Description of basic components of the _vnivt controls.
- Describe basic operation of the valves, both with air and -
manually.
- Describe how to manually operate _the valves.
. Plant tour, operating the subject valves, as plant conditions allow.
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V.- 9.- .8. ' Command'an'd Control Training - Changes are being made to Administrative Procedures to: revise : the roles of the operators during normal and abnornel plant conditions, but this is only one part of the training. The Operations Department has also provided specific areas' of
. . control, mainly dealing with the Emergency Plan, that are to be '
covered. The. training will- be presented based on the following
-objectives. .
1 Define the line of command in normal and abnormal situations.
. Ensure the_ operators understand the philosophy behind the
. changes to the Administrative _ Procedures concerning their specific roles and duties.
. For the specific topics identified by the Operations Department:
.. Provide philosophy of change in' control duties.
.. Provide specific ~ guidance.
.. Discuss how to implement the guidelines.
The Shift Supervisor is to continue to maintain an overall-perspective of the event, its mitigation, and compliance with administrative process. The Sen.ior Control Room Operator will assume the responsibility to directly oversee the activities of
- the Control Room Operators. This will be accomplished by interacting with the control panel operators while monitoring
, the E0Ps and directing ~the activities:both in and outside of the l Control Room. , The individual assigned as " communicator" for the Emergency'
- Plan,-will not be diverted from that assignment. This is to insure commitments to effectively implement the Emergency Plan are not delayed.
.9 Control Room Simulator Rancho Seco operators train ~and practice E0P implementation on the Control Room simulator operated by the NSSS supplier, 8&W.
This simulator is modeled after Rancho Seco and as such does a good job of providing realistic dynamics _for. operator training. In the. December 26 event, the makeup pump was damaged when the operator forgot to open the makeup tank suction when recovering from SFAS. The simulator is similar in this feature. A specific operator manual action is required to open that valve. It is not automatically interlocked. This feature will be again l emphasized during the simulator training which is occurring j prior to resumption of power operations. t l
- as -
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il V.- 9. .9 (Continued) The simulator training will include the following items-
. Emergency Operating Procedure (EOP) training-including all necessary steps to terminate , overcooling or OTSG-overfill
, 'from any cause, including loss of ICS power.
.- Changes to ADV, TBV, and AFW valve operation folloiwng a
- loss of ICS power.
. Command and control training including implementation of-Emergency Plan when applicable. (Watchstanding Principles)
. Recovery from safety: features actuation i.e. restoring normak makeup and letdown flow.
. Differences between the simulator and the facility (operator traps).
. HPI.and AFW throttling and trip criteria.
i
. Pressurized thermal shock recovery actions..
. ~ Cooldown rate interpretation and tracking.
Conversion from AFW to MFW flow.
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- 10. Operational Review i A detailed review of Operations activities, procedures, procedure -
adherence, training, plus operator comments and observations was done to ensure that lessons were learned and incorporated into plant procedures. A discussion' of the major areas which resulted follows. ~ l The station's Emergency Operating Procedures (EOPs) were developed-and placed in use in accordance with the Abnormal Transient Operating Guidelines (ATOG) as developed by the 8&W Owners' Group. These procedures are written to provide the basis for operator actions- that l are to prevent exceeding Technical Specification-(TS) limits, or 4 other critical limits imposed by manufacturers, such as the reactor vessel interim brittle fracture limits (18F limits). Additionally these E0Ps provide the basis for operator training on response actions required for, and philosophy of approach to, off-normel plant conditions and operations. '
; In this event,.the overcooling was allowed to orogress while the , ! operators focused on closing the Auxiliary ~Feeowater Control Valves,
- . rather than . implementing the procedure step which would have shut down the associated feedwater pumps,~and thereby controlled flow.
This was a conscious decision on the part of the operators, and steps 4 to preclude recurrence are discussed.below.
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. . V. 10. (Continued) . The station's Emergency Plan.(EPs) was developed and placed in use in accordance with agreements with responsible regulatory and , local agencies. These agreements establish guidelines to-ensure, specific L - operator actions, not associated with control of the plant, are taken during identified station emergencies. During the event,:the~ Emergency Plan was.not effectively implemented. The resolution is discussed below. The station's Casualty. Procedures (Cs) are provided to address
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I recovery actions necessary when specific failures or events are known < to have occurred. They are not. implemented until the plant is- ) stabilized by application of the E0Ps. In this instance,. improvements have been identified and implemented. Operator performance.and procedure adequacy were evaluated in the following areas:
- 1. Loss of.ICS Power
- 2. Makeup Tank Overfill
- 3. Radiation Monitor R-15001
, 4. Manual Operation of AFW valves 1 5. Communications-with/from Control Room
- 6. Fire Alarm Support Activities - Technical Support Center
- 7. Makeup Pump Damage s i 8 Loss of SRO (RO)
- 9. Emergency Plan (s)
, 10. OTSG Overfill No attempt was made to prioritize these subevents in relation'to significance of impact or influence on operator actions. Examination of the ten subevents identified the following common items for action. These are summarized below. The Emergency Operating Procedures (EOPs) are adequate and include all necessary steps to terminate overcooling or OTSG overfill from any cause, including loss of ICS power. The E0Ps and ATOG do not contain plant parameter criteria (e.g., RCS temperature, Pressurizer Level, etc.) as to when to take certain ' steps, ~like tripping all feedwater pumps to meet the ATOG intent which is to minimize the overcooling. These criteria are being developed and implemented. i - Casualty procedures provide event oriented guidance to the operators as a followup and recovery following plant stabilization by the symptom oriented E0Ps. The ICS power. recovery and-SFAS actuation recovery procedurcs have been implemented and training will be completed prior to return to power operation. I 4
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u k 4 V. 10. -(Continued)' . The Emergency Plan will.be revised to' provide more guidance in
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responding to multi-event emergencies. As written, the-Emergency Plan adequately addresses single-event emergencies; however,-in situations where-the operator must cope with several competing priorities the guidance to assign priorities and deal -l
,with the events is'being provided.
. The plan is to be streamlined and simplified by reducing branching between procedures and reducing the number of forms or 3 logs.
. Training programs are being reviewed for completeness inlight of . l l the experience gained from this event. Areas being reviewed are related to operator knowledge and skills related to:.
t Integrated system operation Manual operation of manual or power operated valves Fire alarm response requirements . Emergency Radiation Control procedures and techniques Integrated emergency plan response i 11. System and Component Response This section discusses individual itams of equipment, or systems,
! which for one reason or another were deemed to require post-trip review.
.1 Auxiliary Steam Relief Valve PSV-36012A The accelerometer monitor mourted on this valve. indicated on i
IDADS that the valve was open following the. Reactor Trip. Since. this valve constitutes a steam-(heat) load on the OTSGs, and
- thereby contributes to'the overcooling, an analysis of the-system and its operation was conducted. The.results of the several investigations was that upon Loss of ICS DC control
- power, the Auxiliary Steam control station, received a "50% '
Demand" signal. During power operation, this would not be observed as the Auxiliary steam demands are shared between Hot . Reheat steam (from the Moisture Separator Reheaters - MSRs) and i the Auxiliary steam supply, whichever is at the higher-j pressure. Following the trip, Auxiliary Steam needs decrease j and with the control station set at-50% demand, the excess must
- be relieved, as happened here.
! Its effect upon the overcooling is somewhat mitigated in that i considerable ' pressure reduction occurs in providing steam through the controls to the Auxiliary Steam Header. As the main , steam pressure decayed, in response to the cooldown, the point was reached at which 50% demand was not excess to the loads on the Auxiliary Steam system.- This reduced pressure then allowed PSV-36012A to reseat and minimize the draw on main steam. The 4 Auxiliary Boilers'were soon placed into inservice, further j reducing the demand for steam drawn from the OTSGs for plant use. l - l
N V .' 11. .2 Difficulties in Manual Operation of Auxiliary feedwater Control Valves FV-20527/8- . y a Upon loss of ICS DC Control Power, these two valves received a-7
% "50% Demand" signal. . Response to this demand brought the valves 5 far enough open to effectively allow unthrottled flow of I auxiliary feedwater into ,0TSGs A and B. These air diaphragm actuated valves are provided with side mounted manual' operators capable of closing the valve against both the opening spring in ,
the actuator and the hydrostatic load dut to . throttling water.
-Th'e valve controlling flow to the OTSG-B, FVtfoS28, was thought-closed by the operator dispatched to close-it, although it was still passing considerable water.~ A second operator observed Q s
, thatitwasnotfullyclosedandcompletedclosingitbyhand(
without difficulty. . s i The valve control 11ng flow to OTSG-A, FV-20527, was positioned-in the closed direction by the operator using the manual. % ',' device. He-thought it was still partially open and proceeded-to obtain a valve wrench.to assist in getting the valve fully closed. Unknown to this operator was the fact that flow:had stopped as the valve was fully closed. Use of-the 431ve wrench then caused the two 5/16 inch dowel pins, which position the manual operator to the valve yoke, to be sheared off. .~At this - i point, the valve again responded to the 50% open demand and
" popped"'open. Uncontrolled flow to OTSG-A was re-established
- and the overcooling continued.
l ' Detailed inspections of both manual operators showed signs of i application of excessive manual closure forces, although , j , indications on FV-20528 were minor in comparison aM predated ! the damage done to'the device on FV-20527. Both devices were
- found to be loosely mounted to the valve. yokes,91though this
- has not been identified'as having had a significant effect on their operation. The dowel pins are the important' load carrying and aligning components. Contributing to these problems is the difficulty which exists when trying to determine when the valve is " closed." .There was no clear indication or index to indicate f closed, nor is there a local-indication of flow through the valve.
Repairs to these manual operators will involve installation of a new manual operator on FV-20527, and rebuilding the assembly for FV-20528. Operator training' on valve wrench policy and proper manual operation of.these devices will provide confidence that reoccurrence is.unlikely. Labels to clearly show the valve position are being installed to , provide indication of position inen manually operated. , e i g
. 33.-
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. V. 11. .3 Difficulty in Manual Operation of Auxiliary Feedwater Isolation Valve FWS-063.-
When FV-20527 " popped" open, .the operator attempted to stop auxiliary feedwater flow by closing the. downstream, normally. locked open, manual maintenance. isolation valve, FWS-063. This was: unsuccessful in that the valve was stuck open. 1The need to close this valve was soon eliminated by restoration of ICS power and return of its control functions to the Control Room. . Troubleshooting suosequently identified a-lack of lubrication and rusted yoke nut bearings. Reworking these components restored the valve to an operable condition. As an element of the troubleshooting effort, the similar valve-on'the OTSG-B line, FWS-064, was inspected, as.were all similar l ! valves in service on the Auxiliary Feedwater System. All were j found serviceable with only normal closing torque required to operate through their~ full travel. Section V;11.8 discusses the preventive paintenance program for manual valves. i
.4 Main Stean Relief Valve PSV-20544-
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i Following a Reactor Trip on October 2,-1985, this valve was found to have simmered for some time prior to reseating. Subsequently, it was tested inplace as its lift.setpoint' was , found to have drifted low. The indication of opening.during the December.26 event was via the attached acoustic monitor, which
- showed the valve opening coincident with the Loss of ICS power, 4
which had opened the Atmospheric Dump Valves. Since PSV-20544 ' and an ADV are both on the same main steam header, a-test of the sensitivity of the installed monitor was run. This showed.that the PSV-20544 monitor was extremely sensitive and'would indicate "open" in sympathy with the ADY. Independent analysis of the main _ steam header pressure-versus ' time showed that at the time PSV-20544 first indicated open, steam pressure was dropping from its pre-trip steady-state , value, and at no time had it challenged any of the main steam relief valves. This analysis determined it was appropriate to ! recalibrate the valve monitor and that PSV-20544 !s serviceable for power operation.
,. .5 RCP Seal Injection Flow Interruptions The first observed indication of problems with the Makeup Pump (results of operation with Makeup Tank isolated) was that RCP seal injection flow decreased when the B-HPI pump was shutdown.
b This was a result of all thess HPI/MU Pumps discharging into a i C common header from which RCP seal injection flow is drawn. :( Since A-HPI was already shutdohm, and the Makeup Pump isolated -l from a source of supply, shutting off the B-HPI would deprive ! g *
._-..o,_..--
V.- 11. .5- (Continued) the RCP Seals of' injection water. This anomaly occurred twice as the operators troubleshot the condition. Total time with' reduced injection flow amounted to about 75 seconds. The RCP Seals are designed to operate indefinitely without injection if the RCP is running, and for 90 seconds if .the-RCP is shutdown.
-At this time, one RCP was shutdown and three were running. .This analysis also looked at the. detailed RCP and seal recorder data which confirmed the conclusion that seal performance or
. Integrity were not compromised by the event and that the RCP Seals are suitable for returning to power operation.
.6 Feedwater Heater Shell-Side Relief Valves Opening l On October 2, 1985, the Reactor Tripped from low power, and a overcooling event resulted. The cause was attributed to overlapping setpoints (when tolerances are included) between the 4th Point pegging steam supply pressure and the associated shell side relief valve. _This was resolved by developing and calibrating a new and higher relief valve setpoint.
On December 5,1985, following a Reactor Trip, lit was noted that the 4A Feedwater Heater Relief again opened. Isolation of the
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pegging steam supply prevented recurrence of the previously-experienced overcooling. Detailed troubleshooting determined that the 4A and 48 pegging steam supply valves wou?d overshoot i their target setpoint by 10 to 20% when initidlly c.pened. ' Analysis was done to develop a new and lower setpoint which-i would provide the desired post-trip feedwater heating, yet accommodate the observed overshoot. It was this configuration , which was in effect during the December 26 event. A report:of-i steam venting from the 4th point feedwater heater relief. valve
, caused the operators to again isolate pegging steam early in the event, to preclude its contributing to the overcooling. As a possible " repeat" event, this item was investigated.
Pre-existing work requests to investigate reports of steam leaking.through the pegging steam control valves were
, activated. It was found that one control valve.was seriously
) " steam cut" and required reworking. This confirmed the likely l cause of the relief opening being that the pegging steam control valve ined excassive seat leakage, even when closed. Such a condition.would p ss steam, and heat, into the feedwater heater j even after feedwater flow through the heater has stopped, resulting in a buildup of pressure. This system is now . serviceable'and repeat relief valve opening is not expected. l l 1 ll . l-L - - l l __ . _ . _ . . __
', 3 l
, V. 11. .7' HPI/RCS Injection Valve SFV-23811 Position Indication Following SFAS initiation during the event, the operator-
" balanced" HPI flow between the four injection nozzles. This is accomplished by switches controlling the SFAS' controlled injection .valv_es from the' Control Room.. While adjusting to achieve approximately -100 gpm through each nozzle, the " closed" indicator light-illuminated for valve SFV-23811.
Post-event troubleshooting found the position switch adjusted to actuate at approximately two turns open. This position is commensurate with throttled flow of approximately 100 gpm.- The switch was reset to about one half turn and the valve operation
' checked. .In-the as-found condition, the valve would still operate to the fully closed position. In a similar fashion, the valve operator open" torque switch bypass will remain effective
- to allow the valve to operate through-its full stroke with the new " closed" switch setting. _ The above described discrepancy is independent of a previously 'I resolved condition relating to the flow meters. -Those meters have to be calibrated to compensate for operating pressure-effects. The procedure which calibrates the flow meters, accommodates:this need and insures that the resulting indicated flow will be conservative. As described above, the flow was determined to have been "real," that is, the valve was not
. closed and the meter indicated 100 gpm. This condition was observed only on this valve; the other three performed'as expected.
.8 Preventive Maintenance (PM) Program for Manual Isolation Valves
- Troubleshooting of the Auxiliary Feedwater manual isolation valve, FWS-063, found its bearings dry of lubricant ~ and in a condition suggesting it would benefit from periodic servicing.
Investigations of five similar valves showed them all to be 4 operable, although evidence of recent lubrication was missing. j Reviews of the existing preventive mainteaance program shows that<these valves are not on a PM schedule. Their need for ! servicing being' determined by observation during frequent l , monitoring of' plant equipment by operation staff. ; In, recognition of the desirability of having certain manual j valves readily operable, the Nuclear Operations Manager has
- identified a list of approximately 100 valves which will be l verified" operable prior to resumption of power operation. These '
= manual isolation valves will be characterized by. their purpose j y and need to allow isolation of important active equipment such i 4
as pumps, valves, and heat exchangers. They will be selected to j include'both primary and-secondary plant systems.- Function, not l l
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}. V. 11. .8 (Continued) l class, being the criteria. The program will involve actual stroking the valve, and where necessary, servicing with lubricants, packing, or adjustments. Statistics will be l collected and evaluated to determine a summary status of valves
. in similar service.
Significant changes are underway with respect to the Preventive Maintenance Program at Rancho Seco. Staff is being added for the specific purpose of expanding the scope and quality of the program. Specific procedures detailing the PMs are being , expanded to provide confidence in the operability of the PM'd i l equipment. This expanded PM program will include the above identified valves in addition to those already receiving periodic maintenance, and any which meet the criteria being developed for this program.
- 12. Quarantined Equipment List l
This section of the Action List was used to provide quick reference j to the scope and status of that equipment assigned to the Quarantined ' Equipment List, as discussed elsewhere in this report. I l l i l l l l W!NN5h8%DMgggfQggyl
f VI. PLANT MODIFICATIONS RESULTING FROM EVENT The event of December 26th, was in reality ~so events. They were: A Reactor Trip, caused by a loose connection in the DC distribution wiring within the ICS; and an Overcooling, one cause of which was the inability l i of the operators to effectively manually control the Auxiliary Feedwater l valves which " failed open" upon Loss of ICS DC power. Section V.3.4 of ( this report discussed the ICS related modifications in detail. Review of the recommendations coming from the event shows they can be categorized as above. The following discussion looks at the minimum set of changes necessary to preclude recurrence of the initiating events while l enhancing the operators ability to effectively mitigate the consequences ' of such events.
- 1. Loss of ICS DC Control Power '
I
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- a. Power Supply Monitor The Power Supply Monitor is sensitive to resistance added in series with the voltage it is monitoring. This is a condition peculiar to its design. To date over 160 service years have l been accumulated on this type of device. in Nuclear Power Plants. The situation at Rancho Seco on December 26,1985, is !
the only time the peculiarity is suspected to have caused a Loss j of ICS Power. Having replaced the power distribution wiring the i likelihood of a recurrence is remote. On this basis, there is no need for a modification at this time.
- b. Redundant Power Monitors If installed in parallel with the one in the Rancho Seco ICS, they would have all initiated a trip condition, for the reason that they would have all been monitoring a degraded signal, combined with the peculiarity discussed in VI.1.a above.
The one advantage a "2 out of 3" arrangement does have is that a l random failure of a PSM would not thereby cause a trip, while a l l disadvantage is that they constitute new failure modes not l presently existing. One such issue is " crosstalk," wherein a I low voltage is sensed by one PSM causing it to trip. When I tripped, the PSM current. draw is reduced, which then unloads the ' source allowing voltage to increase. The result is that the PSM l reacts and that a cycling can persist. Still, the history on ' these units is that they are suitably reliable. On this basis, such a modification is not desirable prior to resumption of j power operatione. l d l I i ( MELUNM5I$$$GN$BN$4.2}iV! M W N W N!S M $ 5 5 MON W ' O N
r { VI. .l. c. Time Delay on S1/S2 Trip l Testing showed that the time delay to trip interval was shorter then the maximum 0.5 second. Investigations with the Manufacturer continue to determine the appropriate value. The switches are being replaced with similar new switches to allow for independent accessment of switch characteristics. Adjustment of the delay to achieve the maximum value, is not practical with this type of . switch.
- 2. Consequences of loss of ICS AC or DC Control Power I
- a. Turbine Bypss Valves Modifications are being installed in the control path to these valves which will cause them to " fail closed" upon Loss of ICS 1 DC Control Power. The operator will be provided with a switch in the Control Room to allow cycling the valve as necessary to maintain the plant stable following loss of normal valve control,
- b. Atmospheric Dump Valves Modi.fications are being installed in the control path to these valves which will cause them to " fail closed" upon Loss of ICS DC Control Power. The operator will be provided with a switch in the Control Room to allow cycling the val.ves as necessary to maintain the plant stable following loss of normal valve control,
- c. Auxiliary Feedwater Flow Control Valves
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Control of these valves is being revised in a similar fashion to the TB'Is and ADVs, except that, rather than " fall closed", they will " fall to setpoint." The "setpoint" has been selected to preclude the situation where the valves fall closed when flow is actually required. Operator action will be available to fully close the valve, or throttle to intermediate positions without interaction of the ICS, as deemed necessary. Operator action could of course open the parallel SFAS Auxiliary Feedwater j Valves, but to preclude the requirement for operator action, the l minimum position is preferred. A spectrum of AFW system criteria was reviewed and considered in the selection of this setpoint,
- d. Auxiliary Steam Header Regulator No modifications are necessary as the operator can isolate the main. steam supply independent of the ICS from the Control Room, '
should that be desired. A future modification to power this control independently from the ICS is being considered. The i extent of cooling available through this path is limited in any case and does not require prompt action. l l l
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e - m o i VI. 3. EFIC Implementation By letter dated August 15, 1980, the'NRC identified a situation where failure of power supplies to NNI or ICS could result in ADVs opening
.to 50% open position. The District concurred with this scenario in ,
its October 6,1980 submittal. ,The. Distr.ict proposed to correct this j ADV response as part of its' EFIC ~(Emergency Feedwater Initiation and Control) AFW system upgrade. The design concept was presented to-the ; NRC at a September 4, 1980 meeting. Equipment delivery for EFIC was originally estimated to be in early ! 1982. When actually signed, the contract specified equipment delivery for April 1983. During the initial _ design review process, ,. additional improvements to EFIC were identified. As a result of !- these design changes, the deliver schedule was adjusted to May 1984. NUREG 0737 required AFW automatic initiation and flow ind'ication (II.E.1.2.1 and II.E.1.2.2). -The NRC issued Safety Evaluation I Reports in January and September 1982. In October 1982, the' District indicated that -it would install interim safety grade AFW modifications and that EFIC was separate and beyond the AFW upgrade requirements of'NUREG 0737. The District also submitted a new schedule for EFIC implementation showing completion by Cycle 7. This schedule 'was confirmed by the District in December lod 2. The District informed the NRC in April 1983 that the installation was tied to Control Room Design Review (CRDR) and RG-1.97 modifications. This was based on the need for an EFIC control panel in the Control Room that was compatible with the CRDR effort. Part of EFIC are the associated RG 1.97 instrumentation commitments for Rancho Seco; as a -
,re,sult it was necessary that EFIC be rescheduled for Cycle 8 (1'.e.,
the next scheduled refueling). In late 1983, the District implemented an Integrated Living Schedule to better control resources, scheduling of_ modifications, and enhanced operations at Rancho Seco. Since the AFW requirements of NUREG 0737 were previously c.ompleted. EFIC was considered a plant betterment. Using the Living Schedule, to prioritize the use of' District resources, the District scheduled EFIC to be' installed in two phases-Cycle 8 and Cycle 9. The Living Schedule process determined that other NUREG 0737 modifications.-10CFR50.49 - Environmental Qualification of Electrical Equipment, Appendix R - Fire Protection, Generic Letter 83-28 ATWS, and NUREG 0737 Supplement 1, items receive high pricrity which resulted in heavy consnitment of District resources during.the Cycle 7 outage. e
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' VI. 3. .(Continued)
It became clear in meeting the requirements of NUREG 0737, that the number of modification!, imposed ~1n Rancho Seco would' exceed the electrical capacity of its existing emergency diesel generators. The
' District decided in 1980-dl to' purchase two additional diesel generators to augment the existing system. The District originally
. planned the installation of these new generators during the Cycle-7 refueling outage. This schedule was compatible with the installation of_the majority of the TMI modifications, as well as the implementation of EFIC. The diesels purchased were made by TDI and the District, as well as several other utilities, were forced into a_-
' major TDI generator requalification program as a result of design problems discovered on the Shoreham plant diesels. .This requalification program required both time (several years) and resources-to complete. The current schedule will have the diesels i operational during the Cycle 8 refueling outage. 1 i
Since EFIC, and several other. modifications, were tied to the i installation of the diesels, the District was forced to defer implementation of EFIC. This delay also afforded the District time to take a closer look at EFIC as installed at CR-3 and ANO-1. Because of some initial startup and operational. difficulties at these installations,'the District decided on installing the indication portions 'of EFIC during Cycle 8. This_would allow the operators to- 0' gain familiarity with the system during an' operating cycle. Likewise, the District has been interfacing with the staff of ANO-1 to minimize any operational problems and benefit from the ANO - experience, as the District's EFIC will closely resemble the ANO EFIC. In October 1985, the District committed to accelerate implementation of EFIC. The District outlined the specifics of the EFIC implementation in a letter to the NRC-dated January 17,L1986. This implementation will result in the majority of the EFIC actuation and control functions being . operational at th,e completion of the Cycle 8 refueling outage.
- 4. Other Modifications Many of the event related recommendations deal with potential modifications. These are beyond the requirements for startup and are being evaluated on their merits for subsequent incorporation.
VII. ROOT CAUSE(S) ( The Rancho Seco Incident Analysis Group (IAG) has responsibility to determine the " root cause" of problems or events at the facility. This involves addressing the programmatic causes as well as the direct causes-unique'to specific occurrences. ~ l l l' l~ 9 (thhkh?hh hNkhk M'# b SM Eb.f$$DNUdNDN Ib5Nb$NDbbOE
z VII. -(Continued) The IAG charter causes this ef fort to be separate from other efforts to seek and resolve causes. In the instance of the December 26 event, the t IAG monitored the activities and findings of the Transient Analysis-Organization, but used methods and resources to independently arrive at the " Root Cause(s)" of the event. These root:causes are~ presented below. Although developed independently, their basis is supported by the material presented in this summary report. The December 2f, event was subdivided into five " event themes" for root cause determination. They are: Loss of ICS Power
. Rapid Cooldown
. Makeup Pump j
. Health Physics
. Emergency Plan Analyses of these event themes were then studied for the following i considerations: !
. Procedural Adequacy
. Unique Design Features J' Human Factor Consideration Related Issues -
.. Training Preventive Maintenance
.. Personnel Access
. .. Vendor Technical Analysis The following discussion identifies the Root Causes and the contributory causes, as appropriate.
- 1. Loss of ICS Power Root cause: Manufacturing Error A lug was improperly installed on a factory prepared wire. The resulting connection exhibited variable resistance which was on the input to the Power Supply Monitor. The resulting variable voltage lead to the PSM tripping when the ICS was still being supplied with nominal voltage and power. Corrective action involved installing new wiring and lugs.
Contributory Causes:
- 1. The Power Supply Monitor is sensitive to resistance in series with its voltage input. As little as one ohm was found to cause the trip point to increase. Appror.imately 5 ohms at the failure
- point was sufficient to cause the PSM to trip at its nominal l
operating voltage, 24 VDC. Corrective action. involves wiring the PSM directly to the DC source bus rather than the end of the l distribution bus. l l l 1 WRIMib@$$$GGiMVYbYWMkWN$$hN WA$$$$$5EERS5$MSE* sin *A
, .o VII. 1. 2. The Si and S2 source switches were found to have short built-in time delay characteristics, approximately 0.15 second while the l specification is for 0.5 second. This made them more sensitive I
to short term trip signals generated by the PSM. Corrective action is to obtain new switches.
- 2. Rapid Cooldown of NSS Root Cause: Delay in Implementing Design Changes to Mitigate Effects of Loss of ICS Power The susceptability to this event has been recognized for some time.
In response, a Class I design modification called EFIC has been developed. Revisions in design criteria, delays in reaching scheduled refueling outages, equipment manufacturing problems and delays have compounded to delay its implementation. Interim ], corrective action is to install modifications which will provide the l t Control Room Operator with necessary controls powered independently of the ICS, while aggressive efforts are being made which will install EFIC at the next refueling. Contributory Causes:
- 1. Proc'edures The overcooling procedure did not clearly identify criteria which would cause the operator to take second level actions to terminate the condition when the initial effort proved ineffective.
Corrective action has been taken to incorporata this guidance into the E0Ps. A casualty procedure to address Loss of ICS Power was not provided. A procedure for recovery from SFAS operations was not available. Corrective action is to provide these procedures and the training to effectively implement them. i l 2. Training could have compensated for the lack of the above j features or procedures. Training which had been given was not i able to compensate for those deficiencies. Corrective action; training developed to understand the event, revised policies, and new procedures will increase the awareness and knowledge of the operators. 4s I
^
f VII. 3. Damaged Makeup Pump Root Cause: Procedures i I A procedure specific to restoration of normal equipment lineups following SFAS initiation was not available. Procedural references in the E0Ps for insuring pump suction / discharge paths were not consistently included. Corrective action is to provide the missing procedure and add the appropriate caution steps. l Contributory Cause: Training f l The operator was trained and aware of the consequence of operating this pump without suction. This event demonstrates , that training alone may not always be sufficient to insure requirements will be remembered. Training should not be expected to compensate for lack of appropriate procedures, , Corrective action; this event reinforces previous training with an example of the consequences. Retraining and new procedures are sufficient to preclude reoccurrences.
- 4. Health Physics Procedure Implementation Root Cause: Human Performance The perception of the individuals involved was that they were following appropriate guidelines for the conditions which prevailed and the directions they have been given. Corrective action is to clearly restate Administrative Policy that procedures will be followed and provide training on implementing procedures.
, Contributory Causes:
- 1. Training These individuals did not clearly understand their obligation to follow established procedures and to utilize the protective equipment and other personnel available to them. Corrective action is retraining of all operational personnel to assure that others may not harbor similar concepts and perceptions.
2.
~
Imprecise Definition of HP Responsibilities Authority, Duties Administrative policy is being restated while policy is being changed to assign a HP Technician to Operations for the single purpose of supporting their activities. t . 5
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2 __ ,, 2 L _VII. 5. . Emergency Plan Communications i Root Cause: Training l
' Training-was not sufficielt to. insure the Emergency Plan would be.
properly implemented during a plant transient. Priority of responsibility to that plan was not fully appreciated by some. Corrective actior:.is to provide training on implementing command and control policy and obligation to assigned duties. Contributory-Causes:
- 1. Human Performance The operators were faced with several independent _y
" Emergencies." The overcooling... fire alanns, radiation alarms, damaged equipment, disabled operator, and failed equipment were the major ones. These diverse events challenge the ability to be effective and prioritize. Corrective action is to restate l policy and process in the training program while developing .
training exercises which will better practice complex scenarios. The relationship between the roles of'the Shift-
. Supervisor / Emergency Coordinator and the Senior Reactor Operator are being defined and practices in plant simulator training. I
- 2. Procedures l
Complex branching within the Emergency Plan and its implementing procedures makes smooth implementation difficult. This event-highlighted the situation and led to revisions to remove many of the obstacles. Corrective action, in addition to increased tralning, the necessary forms for an event are being packaged into a folder which will insure all are convenient'and ready to use. i VIII. Conclusions
- 1. Safety Significance The Loss of ICS Power initiated events which led to a Reactor / Turbine trip and a subsequent overcooling of the NSS.. This report analysed the causes of those events and describes changes and modifications whose effe' cts are-to reduce the likelihood of subsequent losses of ICS power while significantly improving plant and operator response to such an occurrence. Neither the facility design bases, typified by the Main Steam Line Failure event, or the NSS components were stressed to the values assumed in the design for nonnel cooldowns.
t Extensive study and analyses have not determined any unreviewed safety questions, while showing that the facility is ready to return
.to power operations with its reliability significantly enhanced.
Challenges to safety systems should be correspondingly reduced. L Qff28iMWN2" "M2MhtMSMUiAAQdEh?N$$iNbMW%%&M;kUW'M51dbk%1!M
~ , -.c VIII. 2. Appropriateness of Findings Study and_ analysis of this event benafited from the systenatic-identification of-issues, logically developed. troubleshooting and careful. investigations which substantiated or rejected the postulated root causes. Independent from the troubleshooting and repair
. activities, the Rancho Seco Incident Analysis Group developed the Root Cause(s) of the events. This summary report relates the
) collected information on each issue which is subsequently reselved in the root cause analysis. The correlation between the findings and , the root causes demonstrates the appropriateness of the methods used ' and validates the results. -This consistency provides confidence that the changes and modifications made will accomplish their-intended - purpose of precluding recurrence while enhancing reliability and safety.
- 3. Status of Resolution of Recommendations, Repairs, and Modifications Each of the several investigations resulted in recommendations and k proposed modifications to the facility. its procedures, or the training program. Those issues, which are necessary for effective plant operation, have been identified as required for startup and will be completed accordingly. Items of. specific interest which will be completed are recapped below:
Repair Hand Operators on AFW Control Valves Institute PM and exercise important Manua'l Isolation Valves Rewire the ICS Cabine' t DC Power Distribution circuits Wire the ICS PSM direct to the supply bus i Provide Controls, independent of the ICS, to cause valves to go
-to desired positions of Loss of ICS Power: ,
AFW Flow Control Valves
- Atmospheric. Dump Valve Turbine 8ypass Valves
- Provide permanent drain for TSC Fire Control Valve Revise E0Ps to provide specific action points Provide Loss of ICS Casualty Procedure Incorporate Lessons Leitrned Procedure Changes Institute formal Command and Control concepts Provide Operator Simulator Training on event and procedure revisions
- 4. Effectiveness of Modifications and Training The described modifications will provide the desired, and necessary, capability to control events such as loss of ICS power, from within the Control Room. This control is achieved independently of the ICS or its sources of power.. With the modified controls, the transient resulting from Loss of'ICS Power will be much less severe, even assuming no operator action.
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VIII. 4. (Continued) Training related to this event, and the proper and effective use of the modifications and revised procedures, will be demonstrated to be effective by observation of the operators at the simulator. -Practice and experience so gained will enhance and reinforce the effectiveness of these changes. The result will be an improved ability to mitigate transients in a timely and effective manner.
- 5. Suitability for Restart This report describes the investigations done, the results determined, and the actions taken to reduce the likelihood and consequences of loss of ICS power. The improvements and modifications made go beyond that single event to provide to the operator additional capability to ensure the timely mitigation of a '
wide variety of events. The applicability of the operators emergency operating procedures has been validated, while the specifics have been improved to enhance confidence and response. Rancho Seco has been physically configured to provide similar control features and capability similar to its sister B&W plants. The achievement of this configuration provides a significant enhancement of safety'and the degree of confidence necessary to support its return to power operation. i l . JA$fdViikfE9hfk$h&/$4AJLMWEU AMhEE5NMS$#i@$$[79MIjd%efdONUihtf42dWM&59
i I e ATTACHMENT 1 SMUD SEQUENCE OF EVENTS - l .
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I 2.-7-T6 ; f' SEQUENCE OF EVENTS KEY M [ j g ~ TIES ~ 04':16:20 Clock time. -The IDADS compute.r time was used as the official clock. Alltimeshavebeenno'rmalized.t$theIDADSclock. r 0 [07:23] Time after initial event in minutes and seconds, f l 04:20:40? Time is estimated to the nearest se::ond. However, the L associated event may have occurred a few seconds earlier or later. 04:16:7 Time is estimated to the nearest minute. The event has occurred sometime during this minute. 04:587- This is the best estimate for the occurrence of.this event. t
. It could have occurred a few minutes earlier or later.
. SOURCE A88REVIATIONS -
IDADS from the Interim Data Acquisition and Display System computer.
.x OPS from operator personal statement or control room logs.
FUI from operator follow up interviews, f CALC from an engineering alculation of pressures or flowrates. SCRTY from security log.
- ECL from Emergency Coordinator's log.
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CHRONOLOGICAL SEQUENCE OF EVENTS 12/26/85
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INITIAL CONDITIONS Unit operating at steady state power of 76%, 710MW(e) Integrated Control System in full automatic Bailey Computer out of service (one of the Control Room's two main computer systems) TRANSIENT INITIATOR TIME SOURCE EVENT / ACTION 04:13:47 IDADS Loss of ICS is caused by the s'multaneous i de-energizing of all redundant ICS DC power supplies. All lights on the ICS , stations go out. All Bailey station demands go to 50%. "ICS OR FAN POWER FAILURE" amiunciator alarms. l SEQUENCE OF EVENTS 04:13:47 IDADS "ICS OR FAN POWER FAlt.URE" annunciator alarms. [00:00] Upon the loss of ICS DC power, all ICS demands went to midscale, corresponding to 0 volts. :The startup .and'asin feedwater valve's closed to 505 because of this decrease in demand signal. The main feedwater pump speeds reduced to hw speed stop, 2500 rps. Oe-energization of the ICS DC power supplies caused the main feedwater block valves to close. Main feedwater flow to the steam generators decreased.
. The loss'of ICS DC power also sent a demand'to the Bailey AFW control valves, ADVs, and T8Vs to open to 50% demand. The main turbine cannot respond to changes in steam header pressure because signa'Is to change governor valve position originate'in the ICS. Total steam flow increased. Steam generator pressures began to decrease.
The reduction in feedwater flow had a greater effect than the extra steam flow. RCS pressure began to rise rapidly as the RCS heated up. The main steam to auxiliary steam pressure reducing station also failed to mid scale. Auxiliary steam pressure began to increase. 04:13:557 OPS Operators notice NFW flow decreasing rapidly.and RCS pressure increasing. Operators manually open one of the pressurizer spray valves in an attmpt to stop the RCS pressure increase. 9
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3-04:14:01 IDADS Main ferdpump discharg) prassure d creases to~ less than 850-psig, automatically starting the electric driven AFW pump, P-319.- At this time, there was no main feed f. low to the steam generators because steam pressure was higher than f feedpump discharge pressure. 04:14:03 IDADS Reactor trip on high RCS pressure. TtIeturbineandgenerator
.[00:16] trips are also initiated by the reactor trip. Operator closes pressurizer spray valve.
OPS Immediately upon reactor trip, transformer yard fire alarms, seismic trouble alarm, and SFP high temperature alarm are received on the main annunciator panels.. A TSC. fire system actuation alarm is received on the IDA05 computer. An auxiliary steam relief valve lifts.. 04:14:04 SCHRT Momentum of RCS carries peak pressure to 2315psig. -AFW flow IDADS begins to both OTSGs. 04:14:06 IDADS AFW dual drive pump, P-318, autostarts on low main feedpump discharge pressure (850psig). RCS hot leg temperature reaches. a peak of 606.5 F. 04:14:077 OPS Operators perform the actions of Emergency Procedure Section E.01. This included reducing letdown flow. Operators then proceed with Emergency Procedures Section.E.02. 04:14:12 IDADS Six OTSG code safety valves are lifting. . 04:14:25 10A05 Operators fully open "A" inject valve for more makeup addition to the RCS due to low pressurizer level in accordance with E.02, Vital System Status Verification. , 04:14:26 IDADS All OTSG code safeties have reseated. 04:14:48 OPS Nakeup Tank level decreasing rapidly due to high rate of makeup to RCS. Operator opened 8WST suction valve on 'A' side (SFV-25003). 04:15:04 IDA05 Operators start "B" HPI pump to increase reactor coolant inventory from BWST - 04:15:18 10AOS No level. remains in OTSGs. AFW is removing heat.. 04:15:307 OPS Operators sent to close AFW flow valves and place covers on MSR relief valves.
. The MSR's ',o into a vacuum following a turbine trip. The '.
relief valves have been.a source of major vacuum leaks in
- the past. There were no* condenser vacuum prob 1, ems during this event. .
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The T8Vs and ADVs could have been shut from the Remote . Shutdown Panel located two floors below the Control Room. However, the operators failed to remember this fact. Ranch Seco does not have Main Steam Isolation Valves (MSIV's). . 04:16:02 IDADS Operator trips both main feedpumps. Two operators verify AFW 16:04 flowrates to each OTSG were greater than 800 gpe. Operators noted that both AFW pumps were running and .that they had no control over main feedwater due to the loss of ICS. 04:16:14 IDADS OTSG levels begin to increase. ] ! 04:167 OPS Operator secures pegging steam from Control Room to ensure that it would not contribute to the cooldown. A Control Room operator heard a steam relief valve blowing on the turbine deck. Pegging steam had caused feedwater heater relief valves to lift in the past.. This had contributed to a recent overcooling event at Rancho Seco. . 04:16:40 IDAOS RCS temperatures decr' ease below expected post trip value (550 F.) RCS pressure is 1670 psig. 04:16:57 ' IDADS RCS pressure has decreased to 1600 psig. Pressurizer level is [03:20] 15 inches. SFAS automatically initiates on low RCS pressure.
"A", "B", "C", "D" HPI injecti6n valves travel to prethrottled position. Selected SFAS equipment, including motor driven AFW pump P-319, trips and block loading of SFAS equipment begins.
AFW SFAS valves travel fu]l open. "A" and "B" DHR/LPI pumps autostart in their recirculation mode. Diesel generators autostart but do not close onto vital busses. There has been no loss of power to the vital 4160 volt busses. SFAS also actuates containment building isolation. The stripping of P-319 from its vital bus causes total AFW flow to decrease by half. 04:16:59 IDAOS 'A". HPI pump autostarts from SFAS signal. 04:17:00 IDADS Pressurizer level goes offscale low. Subcooling margin is 75 F. and increasing. 3-
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04:177 OPS / Opsrators unlocked the ICS cabinets. They found all DC power FUI power supplies de-energized. The dual switches (S1-S2), which provide power to the supplies, were both found in the full down position and were assumed to be closed. The'A8T, which supplies all ICS power (A.C. and DC), was found powered from its normal supply. An operator switches the A87 to its alte'rnate supply, but ICS DC power does not return. The A87 is switched back to its normal supply. 04:17:107 OPS Operator closes AFW SFAS valves which were fully opened by the SFAS actuation. These valves are in parallel with the already open ICS controlled AFW valves. - 04:17:15 IDADS "A" and "8" CR/TSC Essential HVAC units start from the SFAS signal. '~ 04:17:27 IDADS Motor dri /en AFW pump P-319 is block loaded back onto its vital bus and immediately restarts. The dual drive AFW pump-has been running continuously since it started on -low feedpump discharge pressure a few moments after the loss of ICS power. AFW flow to both steam generators increases and is now in excess of.1000gpm to each steam generator. 04:18:58 IDA05 kCS temperature goes below 500F, - One RCP should have been stopped at this time to avoid core lift concerns. An' RCP was tripped at 04:28. 04:19:00 CALC Pressurizer surge line empties. RCS pressure begins steeper [05:13] decline. Subcooling margin drops 8'F to 77'F and then begins i to increase. Water at vessel head is flashing to steam.' l 1 04:19:15 IDADS Operators secured 'A' train of CR/TSC Essential HVAC to reduce the ambient noise level in the Control Room. 04:20:01 IDA05 Auxiliary steam relief valve reseats and does not lift again. Main steam pressure was 550psig at this time. There was enough flow across the pressure reducing valve to reduce the auxiliary steam pressure below the relief valve setpoint. 04:20:20 IDADS Steam generator pressures have decreased to 500psig. Main feedwater flow begins. At this pressure the running condensate pumps began to supply feedwater to the OT38s through the idle Main Feedpumps. This added approximately 1000gpa to the feed rate of each OTSG for a little over two minutes. (.y n. 7 . .
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i-04:207 OPS Op3 rat:r s:nt computer technician to look at ICS power. He confirmed all DC power supplies were de-energized and the A8T had not transferred. 04:21:25 IDADS RCS reaches minimum pressure of 1064psig. Operator is closing ADVs. - l The combination of water flashing in the head and the reduction in steam flow as the operator manually isolates the ADVs has stopped the RCS pressure decrease. HPI flow was now sufficient to keep up with the cooldown rate. 04:21:30 CALC / Pressurizer surge line begins to refill. RCS pressure and (07:43] OPS /' subcooling margin are increasing. IDADS Formation of steam at the head has ceased. HPI has f refilled all steam voids. # 04:22:00 IDADS 4 Violated B&W recommended PTS curve for the reactor vessel. l 04:22:40 OPS / Operator isolating T8Vs. RCS pressure is rapidly increasing. IDADS { 04:22:50 IDADS Steam generator pressures have. decreased to 435psig. Main steam 11n.e failure logic closes the startup and main feed valves. FW flow from the condensate pumps is stopped. 04:23:7 OPS Local isolation of TBVs and ADVs is completed. ADVs were the i first valves closed, followed by the T8Vs.. . 04:23:10 . OPS / 'B' 'AFW control valve partially closed using handwheel. t [09:23] IDA05 ' The operator thought he had completely closed the valve "
- at this point. Feed flow to the '8" 0T56, however, has decreased by about 605. This increased' flow through the "A" AFW valve.
04:25:00 OPS / Operator begins to close "A" AFW control valve with its : IDADS handwheel. { 04:25:30 IDA05 Operator unisolates HPI pump SFAS recirc valves, opening the ! i recire path to the Makeup Tank. 04:26:15 iOA05 CR/TSC Essential HVAC train "B" is secured. 04:26:20 IDAOS Pressurizer level returns on scale. 5-
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04:26i22 FUI/ 'A' AFW valva cicsed. This steps AFW flow to th) *A* OTSG. IDADS
- Operator believes the valve is only 80% closed, but cannot close it any further~by hand. Leaves to locate valve wrench. , ,
' 04:26:47- IDAOS Pressurizer level rising rapidly. Subcooling margin is 170*F.
Operators start to throttle HPI injection valves to minimize
, RC'S repressurization.
04:28:00 IDADS Operators stopped the "C" RCP per core lift requirements. RCS temperature is 410*F. l Flow from four RCP's at low RCS temperatures, may give j excessive lifting force to the core components (fuel assemblies). 04:28:43 IDADS RCS letdown flow is reestablished to help control pressurizer level. Letdown flow is directed to the Makeup Tank.- l Letdown flow can be directed to either the Makeup Tank or the Flash Tank. It is normally lined up to the Makeup Tank. 1 04:28:45 CALC / Makeup Tank level goes offscale high. Makeup Tank' relief valve IDADS lifts and discharges to the Flash Tank. 04:28:59 IDA05 Operators stop 'A' HPI pump. - The RCS pressure has peaked at 1616psig. RCS temperature at this time is 422*F. , 04:29:40 I l OPS / Operator uses valve wrench on 'A' AFW valve. Manual operator [15:53] IDAOS is damaged. Valve reopens. Operator calls Control Room and ( is told to close downstream manual isolation valve FWS-063. AFW flowrate to the 'A' OfSG is greater than 1300gpe, resuming IDADS the rapid,cooldov.. 04:29:45 IDADS "C" and "D" HPI injection valves are closed to reduce the rate l of increasing RCS pressure.
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,, . -- . -o 04:30 OPS *A* sida BWST sucticn valva, SFV-25003, is closed in an ~
attempt to decrease Makeup Tank level. . The operators assumed the running makeup pump would take a suction from the Makeup Tank if the BWST was isolated. They Forgot that the Makeup Tank outlet valve (SFV-23508) closes on SFAS actuation. Their action isolated the' suction'of the Makeup pump, "A" HPI pump, and the~"A" ' Decay Heat pump. (Note that the "A" HPI pump was not running at this time.) Operators shift letdown flow to Flash' Tank. This action was taken to reduce'the inflow to the Makeup Tank. HPI recirculation flow continued to the Makeup Tank. Shift Supervisor declares Unusual Event. A Senior Control l Room Operator begins notification of the state, counties, and the NRC. 04:30:39 CALC Makeup Tank relief valve begins ,to pass water. Flash Tank level begins to increase. The relief valve discharge is routed to the Flash Tank. The Flash Tank pumps start on high level but do not have the capacity to handle the letdown flow and "B".HPI pump recirculation flow. -Flash Tank level began to slowly increase. ' 04:30:40 IDAOS Soth OTSG operate levels off scale high. AFW flow continues to both OTSGs.
"A*: > 1500gpe, "B": 670gpe..
04:33:00 OPS / Started depressurizing RCS to return to condition outside PTS IDA05 region using normal pressurizer spray. 04:33:20 OPS / Operator arrives at '8" AFW valve and finds it partially open. 10A05 He closes it the rest of the way. Feedwater to the '8" OTSG has been stopped. This increases flow to the "A" 0TSG to greater than 1700gpe. 04:33:40 IDA05 "A" 0TSS is full up to the top of the steam shroud and begins [19:53] to spill water into the steam annulus. 04:367 OPS Operator has attempted to close FWS-063 but it will not move, even with a valve wrench. 04:39:00 IDADS RCS'subcooling margin reaches peak of 201*F and begins to decline. RCS temperature was 390*F. RCS pressure was 1430psig. This is approximately a00pst into the PTS region.
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'04:40:00 OPS / _ ICS power is restcred. Op3ratcr clcsed S1-S2 switch 2s 'in ICS
[26:13] 10 ADS cabinet 3. Upon restoration of power, ADV demands went to 100%,-T8V demands went to 0%, "A" AFW valve demand went to 1005, 8 AFW valve demand went to 0%. Operators reduced the 100% demand signals to 0%. The "A" AFW valve responded and l closed. ~ Operators ~ regained control of ADVs T8Vs, and AFW flow control valves. The operators immediately reduced the demand signals to these valves to 05._ Since the ADVs, were held closed by the manual handjacks, they could not respond to their demand signals and did not lift. The "A" AFW control valve was closed by the operator action. All feedwater flow to both OTSGs was then stopped and the RCS began to heat up. The lowest RCS temperature of 386*F was reached and at this time RCS pressure (now 1413psig) is~ being reduced to achieve conditions outside j the PTS region. The RCS had cooled down 196 degrees in 26 minutes. f 04:40:10 IDADS Minimum OTSG pressures reached. ."A": 221psig, '8":202psig. 04:417 OPS / Operator calls control room and informs them that FWS-063 is IDADS- stuck open. Told to disengage the handjack from the '8' AFW
. con l trol valve. Told other operators to unisola,te T8Vs.
l 04:41:10 IDADS "A" 0TSG level goes below steam shroud. . ; Main steam flow is continuing to supply AFW pump P-318. ! An estimated 10,000 gallon's of water spilled over the ' shroud and into the steamlines. l 04:42:42 - IDA05 Shutdown '8' HPI pump. Makeup pump continuing to run. 04:42:56 IDAOS Closed "A" and "3" injection . valves. Pressurizer level is 130 inches and increasing'. All sources of makeup water to the RCS l have been closed except RCP seal injection. l l 04:43:307 OPS Operators noted loss of RCP injection flow. Seal flow slowly decreased as.the '8' HPI pump coasted doun. The operators were puzzled by the fact that seal l flow was being lost with the Makeup Pamp continuing to ! run. They were not yet aware of'the Makeup Pump's isolated suction. , 04:43:54 IDA05 Operator restarts '8' HPI pump to supply RCP seals. Seal ' Flow returns,to normal. . l 04:50:lg IDADS operator stops 'O' HPI pump again.
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, The operators checked the RCP seal injection valves but could not find anything wrong with the valve lineup so
,the pump was shutdoun again.
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. 04:50:30- IDAO'S 0paraters again nstico less of seal flow and restart 888 HPI pump.
04:527 OPS Senior operator collapses in front of control panel. He is moved to an office adjacent to the_ control room. 04:57 OPS Started blowdown of both OTSGs to reduce level. s 04:58 - OPS Both Emergency Diese1' Generators are shutdown. . They had run unloaded since the SFAS initiation. 04:587 OPS Operator in Control Room hears loud noise. The " MAKEUP /NPI [447] PUMP LU8E GIL LOW PRESSURE" annunciator alarms. The operator notes that the Makeup Pump ammeter is reading only a fraction of normal running current. He realizes the pump has been damaged due to lack of suction. J i The pump became separated from the motor when the pump seized and broke the coupling between the speed changeir and pump. 1 05:00:10 IDA05 Operator trips the Makeup Pump. 05:01:22 OPS /. Operators open the Makeup. Tank outlet valve (SFV-23508). 10AOS Water from the tank spills out of the damaged Makeup pump seal's and onto the pump room floor. Approximately 1200.
- galluns is spilled. The operators Ismaediateiy deduce wh sharp drop in Makeup Tank level had occurred and recloss' ythe the I outlet valve.
The operators were aware that' bo'th the Makeup Pump and "A" HPI' pump had their suctions isolated. These pumps share a common suction from the ' Makeup Tank and 8WST. The "A" HPI pump had not been run since its suction was isolated. By taking this step, the operators were attempting to establish a supply path to the "A" HPI Pomp. 05:05 IDA05 Crossed out of B4W recommended PTS region: 3 hour soak is in OPS progress. Ambulance is called for 540. l . . t
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-05:08 OPS Auxiliary Building Gas Radiatien Monitors (R150028 and R15045) I go into alarm.
These' instruments monitor air being exhausted from the
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Auxiliary Building basement. One of these rooms is the Makeup Pump room. The instrument is detecting gas.from the mildly radioactive water spilled in this room. 05:09 IDAOS Both AFW pumps are stopped. 05:12:26 IDADS Both Decay Heat pumps are stopped.
- 12:40 i
The "A" DHR pump had been running without a suction from the 8WST since the closing of the 8WST outlet valve l (SFV-25003). The pump r::n on its own recirculation and always had adequate NPSH. There was no damage to the pump. 05:13 OPS Soth Auxiliary radiation monitor alarms clear. l 05:277 OPS Makeup Pump manual suction and discharge isolation valves are closed. This isolates the pump from the Makeup Tank /8WST suction header. The "A" HPI pump is now available for service, if needed. 05:29:04 IDADS Operators stop lhe "A"'RCP per procedure. 05:33 SCHN1 ' Zone 20 fire alarm trips the Rae Weste Area Exhaust Fan [79:] (A-542A or A-5428). An interlock prevents restart of the fans untti the fire alarm is manually reset.
. The $FAS' initiation has c1'osed the suction and discharge valves to the Reactor Building radiation monitor (R15001A&B). The compressor which normally pulls a sample from the Reactor Building' overheats due to its lack of suction. The smoke from this compressor actuated the smoke detector in this room. The fans are located in an adjacent room within the same' fire zone. If a fire is detected in either room, the fans are tripped. The fans drew air from the -20 and -47 foot areas of the Auxiliary Building. Spent Fuel Building and Chemistry labs. The Makeup Pump room is located in the -20. foot elevation of the Auxiliary Building.
- 05:357 OPS Control Room operators stop R15001A48 com'ressor. p 05:38 SCHRT .
Zone 20 fire alare is reset and one of the exhaust fans is restarted. It runs for less than a minute and the fire alarm trips it once again. 05:407 nps . MainSt[eamlineFailureLogicisinhibited. This action permits the normal feed flow pathway to the steam generators ! to be used. . l g*', . .
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05:48 053 TSC sprinkler system is isolated. O'5:50' SCHRT Zone 20 fire alarm is reset and one of the exhaust fans is restarted. It runs for only a few seconds when the fire alarm trips it again. r 05:517 CALC Flash Tank begins to overflow into Waste Gas System. 05:54 OPS Outside operator reports loss of security badge to Watch Commander. The operator lost his security badge and radiation film badge around the time he was helping the licensed operator manually isolate the ADVs. He was busy assisting in the stabilization of the plant and did not i report the loss until this time. He did not enter any radiologically controlled areas without his film badge.
- 05:59 SCHRT Zone 20 fire alarm is reset and one of the exhaust fans is restarted. It runs for less than a minute and the fire alarm trips it once again.
06:01 .SCHRT Zone 20 fire alarm is reset and one of the exhaust fans is restarted. The smoke cleared from the area and the fan continues to run throughout the remainder of the shift. 06:02 OPS Health Physics sample showed 1.5 MPC's.of Xenon 133 and 135 in
-20 foot level of Auxiliary Su11 ding.
06:04 OPS Operator bypasses Safety Features signals.
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06:11 OPS "ICS OR FAN POWER FAILURE" alarms on main annunciator knel but immediately resets. There is no loss of power. No equipment response is noted. . 06:14 OPS "ICS OR FAN POWER FAILURE" alaries on main annunciator panel. ICS DC Power is lost. Operator immediately resets SI-52 . switches and ICS power is restored. Many ICS demands go to 1005. Operators reduce demands to 05 on ADVs. main and startup feeduster valves, and AFW control valves. Operator response is quick enough to prevent another overcooltag. Feedwater flow from the condensate pumps through the open startup feedwater valves added approximately 750 gallons of water to each steam generator before operators can cumplete the closing of the valves. 06:15 SCRTY Security brings a spare visitor's security badge to the* Control Room. -
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07:02 OPS Site boundary release of: .93 MPC from the Auxiliary Building . stack is calculated. State OES is informed. 07:057 OPS Makeup Tank outlet valve'is opened. . s 3 t 07:15 OPS Started "A" HPI pump. Stopped "B" HPI pump, e t' ' The "A" HPI pump took its suction from the Makeup Tank . l Makeup Tank level began to decrease. This terminated the .1 Flash Tank overflow into the Weste Gas System. -' 08i41 ECL Unusual. Event is terminated. . s
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, AUXILIARY BUILDING EXHAUST AIR UN TS (A-542A/B)
DECEMBER 26, 1985 TRANSIENT ABRIDGED SEQUENCE OF EVENTS L TIME EVENT SOURCE 04:16:57 SFAS initiation (Rad Monitor R 15001 A/B isolated). l 04:43:50 Operator noted loss of RCP seal injection Operator personal flow.(Loss of flow is indicative of Make statement Up Pump P-236 damage. Approximately 450 gallons of contaminated water from breached t Make Up Pump spilled on floor in Auxiliary Building -20 ft. level.) - About this time, an Equipment Attendant located in the Rad Monitor Room smelled smoke. He was dispatched to perform other duties before he could identify the source. 05:03:49 Auxiliary Building stack hi rad alarm Computer printout (caused by transport of airborne activity from Make Up Pump room). 05:10:09 Auxiliary Building stack hi rad alarm Computer printout cleared. , 05:36:04 Auxiliary Building Exhaust Air Handler Computer printout (A-5428) tripped (probably tripped by smoke detector in Fire Alarm Zone 20). NOTE: R 15001 A/B is located in Fire Alarm Zone 20. 05:39:39 Attempted star of Auxiliary Building Computer printout Exhaust Air Handler A (A-542 A) preempted Operator personal by smoke detector. statement. 05:40:28 Attempted restart of Auxiliary Building Computer printout, Exhaust Air Handler A(A-542A) preempted Operator personal by smoke detector. statement. 05:47: R 15001 A/B declared out of service-- Control Room log blower motor overheated. 05:59:07 Attempted restart of Auxiliary building Computer printout, Exhaust Air Handler A(A-542A) preertpted Operator personal by smoke detector. statement. 05:59:50 Attempted restart of Auxiliary Building Computer printout, Exhaust Air Handler A (A-542A) pre- Operator personal empted by smoke dectector., statement. f a Y kN!I M IO h di N d d h A M ikN b b I N 1 @ S $ d N IY M O d 5 D E i d N # N E 6 Mk M M iG eR W 6:MSSMMNWd
AUXILIARY BUILDING EXHAUST AIR UNITS (A-542 A/B) (Continued) TIME EVENT SOURCE 06:01:21 Attempted restart of Auxiliary Building Comouter printout, Exhaust Air Handier A (A 542A) preempted Operator personal by smoke detector. statement. 06:02:15 Started Auxiliary building Exhaust Air Computer printout Handler NOTE: A-542A/B Trip Devices
- 1. Smoke Detectors - automatic
- 2. Ground Fault - automatic
- 3. Control Room Pushbutton - manual The Auxiliary building Air Handlers are not tripped by high radiation levels in the stack.
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I i I. l ATTACHMENT 2 GAC 85-1001R2 M e 9 9 9 9 e
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SACRAMENTO MUNICIPAL UTILITY DISTRICT OFFICE MEMORANDUM To: J. V. McColligan O. D. Whitney DATE: January 4, 1986 R. W. Colombo J. J.. Field GAC 85-1001 Rev. 2 . S. J. Redeker FROM: G. A. Coward susarcT: TRANSIENT ANALYSIS ORGANIZATION, TROUBLESHOOTING, AND EQUIPMENT REPAIR FOLLOWING 12-26-85 TRANSIENT I am again establishing an organization for this analysis and startup l l similar to that used for other recent startups. The organization is ! outlined on the attached chart.
- The troubleshooting and equipment repair guidelines used for the previous trips are to be used for this event and are repeated and )
expanded below. These guidelines apply to those Action List items marked with an asterisk with special requirements added for equip-ment on the Quarantined Equipment. List. Other equipnent may be added to the list by any of you; Dan Whitney will maintain and coor-I
.dinate the master Actic.n and Quarantine List. :
l It is important ' hat t ..; carefully plan and document all troubleshoot-l ing efforts that take place to fix the problems surrounding the event. } An engineer is to be assigned to each specific 7roblem, and will pre- ! pare an action plan for troubleshooting, eva]u;.tfon, and correction } of the problem. The Action List Coordinator will, as a minimum, ap . I prove all Action Plans. A format fo: the Action Plans is included. l Specific equipment involved in the event has been placed on the Quarantined Equipment List which will require additional planning and monitoring during the troubleshooting and repair efforts, in addition to NRC Incident Investigation Team (IIT) review prior to consnencement of any actual work. Information copies of MI's will be given to the IIT prior to beginning any troubleshooting. The troubleshooting and investigative activity shall be preceded by , an event evaluation and analysis to determine probable causes of failure or abnormal operation. The analysis and evaluation shall proceed as follows:
- 1. Collect and analyze known information/ operational data for conditions prior to, during, and after the transient.
- 2. Review maintenance and surveillance / testing history, includ-ing recent modifications to relevant equipment, procedures, and operation.
L . e 6 I - 7 -
7, . o Distribution January 4,1986 l
- 3. Develop a summary of data, including 1 and 2 above, that support any proposed probable cause of failure.or abnormal
. operation. .
4 Based on above items, develop a probable root cause(s) of the problem.
- 5. Develop plans for testing the probable causes (i.e., checks, verifications, inspections, troubleshooting,etc.). In devel-oping inspection and troubleshooting plans, care must be taken -
to insure, when possible, that the less likely causes remain testable. When planning troulbeshooting activity, try to simulate as closely as practical the actual conditions under which the system or component failued to operate properly.
- 6. Document the action above in the Troubleshooting Action Plan.
7 Prepare the MI's to implement the Troubleshooting Action Plan. It is very important that our investigations do not in any way result in the loss of any information due to disturbances of components or systems prior to establishing the cause of the failure. Investiga-tions are to be conducted in a logical, well thought-out, and docu-mented manner. The Maintenance Instruction (MI) shall be used to . implement the Action Plan, and where appropriate, to accomplish the-investigation / repair, with special attention being given to insuring that current drawings and vendor Jnanuals are used. If vendors or vendor representatives are involved, their use must be documented and their role explained. ' Action Plans will be approved by the Action List Coordinator prior to any work being started to insure that the Troubleshooting Action Plans conform to these guidelines, and that the described Maintenance Instructions (MI's) are appropriate and effective in resolving the concern. The Action Plans will include, as appropriate, the follow-ing:
- 1. Asterisked equipment on the Action List is to be treated as QA Class 1 for purposes of documentation, inspection, pur-chasing, and control.
- 2. Troubleshooting and repair activities are to be accomplished on separate Work Requests. Separate Action Plans are to be prepared for troubleshooting and repair for equipment on the Quarantined List.
9 s . k * '
, , ~l . l Distribution January 4,1986 l i l
- 3. Doev:nent on the Work Request all "as found" conditions. !
Visually inspect and document any missing, loose or damaged components,' noting positions, abnormal environmental condi-tions, leaks, cleanliness, fluid conditions, jumpers, abnor-mal circuits, etc. Describe the overall condition or appear-ance. Whenever possible, use photographs to document "as found" conditions, and retain samples of fluids and residues i for analysis. l
- 4. Should a change in scope or direction be warranted, than a revised Action Plan must be processed. No materials er com-ponents are to be shipped off site unless specifically called "
for by the Action Plan.
- 5. It is the Action List " responsible" lead person's responsibility I to assure that the investigative actions are appropriate, suffi-cient, properly defined, documented, and that data is preserved.
This person shall also approve MI's prior to their use and in-sure that a copy is provided to the Action List Coordinator. QC is to be involved in the troubleshooting and repair process. By copy of this memo, I am asking~Andy Schwieger and his Ouality organi-zation to support the entire process. Upon completion of the work described in the Troubleshooting Action
- Plan, an engineering report on the cause(s) is to be prepared. This report is to justify and explain the cause(s), and include a descrip- -
tion of the concern, statement of cause, and justification for this conclusion based upon the findings. A format for this report is attached: An Actior Item Closure Report is to be developed and sub-mitted to the Action List Coordinator following completion of both troubleshooting and repair activities for review and acceptance. A format for this report is attached. In all cases, applicable procedures must be followed. The require-ments of this memorandum must be communicated to all involved per-sonnel to avoid any confusion or misunderstandings during this in-vestigative and repair period. Attachments (5) cc: N. Brock R. Lawrence H. Canter V. Lewis S. Crunk C. Linkhart R. Dieterich B..Rausch J. Eckhardt R. J. Rodriguez ' B. Fraser L. G. Schwieger J. Jewett B. Spencer L. R. Keihnan J. Sul'livan F. Kellie R. Wichert t 4 h , e a
o L _ TROUBLESHOOTING ACTION PLAN (Format, Rev. 2) ACTION LIST ITEM NUMBER ACTION LIST DESCRIPTION i QUARANTINED EQUIPMENT LIST ITEM NUMBER RESPONSIBILITY OF i I PREPARED BY DATE DESCRIPTION OF ISSUE: (Describe purpose, problem, or reason for investigation) SUMARY OF INFORMATION SUPPORTING PROBABLE CAUSE: (Include analysis / review of operational data, procedures, etc.) REVIEW OF MAI'NTENANCE, SURVEILLANCE TESTING AND MODIFICATION HISTORY:
-(0escribe results of review.)
POTENTIAL ROOT CAUSE(5): l (Identify primary and alteinative hypothesis (ses) as appropriate.) OUTLINE OF TROUBLESHOOTING PLAN: (Describe scope, content, and objectives of planned Maintenance Instructions sufficient to resolve potential root causes.) i APPROVED BY DATE Action List Coordinator .SMUD ) i RELEASED FOR IMPLEMENTATION BY DATE Act1on List Coordinator - SMUD
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ENGINEERING REPORT I ON ACTION ITEM l (Format, Rev. 2) i ACTION LIST ITEM NUMBER ACTION LIST DESCRIPTION l l QUARANTINED EQUIPMENT LIST ITEM NUMBER 1
- RESPONSIBILITY OF I
PREPARED BY DATE DESCRIPTION OF ISSUE:
SUMMARY
OF IhFORMATION SUPPORTING PROBABLE CAUSE: (Similar to section in Troubleshooting Action Plan.') REVIEW OF MAINTENANCE, SURVEILLANCE TESTING AND MODIFIC'ATION HISTORY: (Similar to section in Troubleshooting Action Plan.) IDENTIFIED ROOT CAUSE: (Present results of analysis and/or troubleshooting.) ( CONCLUSIONS AND/0R JUSTIFICATION FOR ROOT CAUSE: l l l APPROVED.BY DATE Action List Coordinator - SMUD e L
/ 1 ^ -
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REPAIR ACTION PLAN (Format, Rev. 2) ACTION LIST ITEM NUMBER ACTION LIST DESCRIPTION 1 QUARANTINED EQUIPMENT LIST ITEM NUMBER RESPONSIBILITY OF PREPARED BY DATE REPORT OF "AS FOUND" CONDITIONS FROM TROUBLESHOOTING: (Describe findings and results of troubleshooting efforts to establish root cause(s).) DESCRIPTION OF REQUIRED REPAIRS: (Describe purpose, problem, or reason for repair actions.) l OUTLINE OF REPAIR PLAN: - l I (Des'cribe nature of repairs, procedure changes, modifications, or changes necessary to return item to service.) ' s I i APPROVED BY DATE Action List Coordinator - SMUD l RELEASED FOR IMPLEMENTATION BY ! DATE Action List Coordinator - SMUD 1 I l 1 6 e a e r g - - g -
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ACTION ITEM CLOSURE REPORT . (Format, Rev. 2) ACTIONLISTITEMNUMBER ACTION LIST DESCRIPTION QUARANTINED EQUIPMENT LIST ITEM NUMBER RESPONSIBILITY OF PREPARED BY DATE A. DESCRIPTION OF ISSUE / CONCERN: B. INVESTIGATIONS DONE: } (Attach: 1. Troubleshooting Action Plan.) ( 2. Engineering. Report on Action Item.) ( 3. Repair Action Plan.) C. CONCLUSIONS / EXPLANATION (should include Root Cause): * (Recap of results from "B".) D. SHORT-TERM FOLLOWUP REQUIRED / RECOMMENDATIONS: E. LONG-TERM RECOMMENDATIONS: F. PROGRAPHATIC IMPLICATIONS: REVIEWED* AND ACCEPTED DATE Action List Coordinator - 5 MUD
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a'4 g - otyl h*h TRANSIENT ANALYSIS ORGANIZATION d -
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QM G. Coward If , Plani' Manager 6 - l
**! D. lhitney Ellen Banapian J. Biccolligan Action List Coordinator Assjstant Investigation Coordinator s.4
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l tj Tech.' Support Root ICS RCS operational Emergency IIRC NSE Isto K , Trip Report Cause Investigation overcoollag Review Plan EPRI saar TAP Team h . r,g . u - bT' q . 5 . i [.*, y*y - Ts . .
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ATTACHMENT 3 ACTION LIST l l G 1
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pff LEGEND: SU o Startup R:quir:d ' . bts LT o Long Term SIATUSDATE 02-14-86 If PE = Power Escalation DECEMBER 26, 1985 TRANSIENT i TIME 0800 (35 NA = Not Applicable Page 1 of 19 ? ST - Short Teru - ACTION LIST - i
-[
- bN!h!oSkN/ ENufhebesPe GAC 85-1001 Apply i DESCRIPTION RES?0NSIBILITY . SCHEDULE STATUS WR No./NCR/. COMENTS
@p[ I ETC. ' k9 ! SI 1. Post Trip Report (AP.28)** J. Field ST Compiling TR No. 7S Grant Simons to PRC 2/22 Complete 2/15 [ - Requested by NRC - IIT if Q[ v-available prior to 2/13. k w
- c. a) TAP Team Assistance R. Colombo i
SU Complete i $f.' b) Sequence of Events R. Wichert SU Complete Rev. 3 issued 01-08-86. b?: ; NOM 86-98 Transmitted Rev. 4 to NRC M ' NON 86-99 227-86. Rsv. 4 issued 02-07-86,
- m. NOM 86-112 Dist of Aux Bldg Exhaust k! -
2410-86. lR NOM 86-113 Transmittal of Aux Bldg g,'. Exhaust to NRC 2-10-86 p $ c) Aux FW System Initiation Response P.eport G. Paptzun SU Complete Ahproved 12-29-85. k.. . i d) Post Trip Shutdown Margin G. Keney SU Complete Always > 2 to 5 percen.t
?L shutdown. Approved
, 12-31-85.
I e) Aux FW Flow to OTSG J. Field SU Closure Rpt l Analysis Due 2/17 l f)** Main Steam Line Analysis J. Field SU l
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. 1. Thermal Calc Received 1/27 By Bechtel
~
g 2. . Stress Calc Received 1/27 Bf Bechtel sg . g . l5 3. Closure Report Due 2/17 CIosureReporttocover 4 walkdown
7 ' '
- LEGEND.: SU o Startup R: quired
. LT - Long Term ,
STATUSDATE 02-14-86 J- PE - Power Escalation DECEfEER 26, 1985 TRANSIENT I TIME 0800
,i NA = Not Applicable Page 2 of 19
[} ST - Short Tem - ACTION LIST - i k hf !$ N!h!okkN/N h Nu Ne b es Pe GAC 85-1001 Apply kY 7: :. DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ l COMMENTS ETC.
/$ 1, i $f fil, g)** Minimum Pressurizer J. Field SU Completion RV Head Bubble generated.
ks Level calculation boing B and W' analysis complete W s. Reviewed 1-29-86. kT Ec Due 2/18 NOM 86-88 Closure Report issued T <'
- 245-86.
h[. NOM 86-87 Transmittal to NRC, 2-5-86. we . ,
"ICS Reliability Stu@"
i,($ i
.h) R. Dieterich SU Complete Input to RRG stu@.
i)** Pressurizer Heater J. Field SU Complete NOM 86-39 Troubleshooting plan N Operation *
, approved 01-15 86, NOM 86-33. Heaters are OK.
d g j)** Control Room Instruments N. Brock .SU Complete Testing complet'e.
- w. . which Fail on Loss of Info to Ops due 4t' ICS Power NOM 86-128 Closure Report Issued-
$lR: , 2J13-86 NOM 86-127 Transmittal to NRC a - 2-13-86. -ji2 k)** Investigate report of E. Banaghan SU Complete NOM 86-12 Closure Report approved ZC SM0KE prior to event 01-11-86. Revision due 2/7.
\
iq: 1)** Primary to secondary J. Field SU Closure Rpt. Testing continuing. 4 Leak. Investigation Due 2/25 p, d w.: m) Security Interface B. Spencer /D. Ross ST Closure Due 2/17 Complete actions per GAC 86-016 b ~ l N B l Y i e ,
}
k.Y LEGEND: SU = Startup R:quir:d /g LT = Long Tcra STATUS DATE 02-14-86 ap PE = Power Escalation DECEMBER 26, 1985 TRANSIENT t TIME 0800 hi NA = Not Applicable Page 3 of 19 gif ST - Short Ters - ACTION LIST - j
- n. -
kNNYEhb!iN/Nh*Nuf0ebes Per GAC 85-1001 Apply b ', DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ ' COMENTS y ETC. i y fli m) 1. Review procedures to 50 to Integrate Ops / Security g,p , involve security in Revise AP 506. gg , event. ; A 2. Response to Loss - B$dgesinCASS fy of Badges. 9 . $f ,
- 3. Locked door control S0 to Address g;j policy and procedures. !
9 - t hK > n)** 1. Main Steam Line N. Brock SU Complete NON E6-107 Closure Report issued 6 l Failure Logic 248-86.
, NON 86-114 Distribution to NRC.
@$i m MSLFL worked properly. @ 2. , Change MSFL setpoint V. Lewis Due 3/1/86 SU ECN R-0357 From 435.5 psig to 575
- 5 h .
psig. DER in 3.f.5. & I 7,4p 3. Human Factors Improve- V. Lewis SU Due 2/18 Paint / Tape / Labels to H2YS.
/ ments to MS Isolation 7p E. Controls '
a o)** SPDS vs Strip Charts for J. Field SU Complete NOM 86-92 To Distribution on 02-07-86. '1 OTSG operate level NOM 86-91 To NRC on 02-07 86. p)** Determine if transient was J. Field SU Complete NOM 86-55 Requested by NRC/IIT <
-within USAR Design Basis telecon of 1-17-86. Issued r to NRC 1-24-86.
g NOM 86-81 Distribution to Action List g 0 86.. s
- l 3
- Ml LEGEND: SU = Startup.R:quirrd '
[Ag LT = Long TOrm . STATUS DATE 02-14-86 dE PE = Power Escalation DECEMBER 26, 1985 TRANSIENT TIME 0800 (l I NA = Not Applicable Pdge 4 of 19 g ST = Short Tens - ACTION LIST - i N Ye!h!o!iN/ h r# uk0ebesPe GAC 85-1001 Apply @ DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ I COMMENTS
- g. ETC. .
V F 1, q)** Discussion of no Operator R. Dieterich - SU Complete Requested by NRC/IIT telecon f action fo'r 10 minutes on of 1-31-86. J .' O Loss of ICS NOM 86-83 To NRC 02-04-86. Pq g NOM 86-82 To Distribution. i ' Was Essential HVAC operable J. Field g .' l r) . SU Due 2/21 NRC Region V question of 02-03-86. 1 2, Human Eactors/ Root Cause ,j '- Analysis- - 54 . Ety.1 a )~ Report' S. Crunk , ST Closure Rpt. Held up by ICS gfjf Due 2/15 troubleshooting W f Closure Report forcast 2/24 jyi , Pl. _ b)** Human Factors Review J. Jones ST Complete NOM 86-48 Incorporate into Long Term Human Factors program.
' Issued 1-23-86 to NRC.
b. if; f 3. Determine Cause/ Corrective N. Brock Action for ICS Power failure. {' h]h [8 F : [ a) , Closed Ddleted. See 3c. 4 : 4 - b)*' ICS Power distribution SU Complete NON 86-84 Removed buses from p investigation quarantine 02-04-86. q j
,4N . ?e.
. , i w ,
.h in
g i LEGEND: SU = Startup,R:quir:d ' x LT = Long Tern. STATUS DATE 02-14-86
' PE = Power Escalation DECEMBER 26, 1985 TRANSIENT TIME 0800
.I NA = Not Applicable Page 5 of 19 ST = Short Ters i< - - ACTION LIST - 2 < f T N!h!!!k/h r*Nufbes Per GAC 85-1001 Apply If% DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ ETC. COMENTS 3,_ b)' 1. C Inverter Complete NOM 86-37 Closure Report 01-16-86.. I 2. J Inverter and E and Complete NOM 86-03 i 1 F buses , i .- , f
- 3. Replace 30 amp Closure-44 breaker w/40 amp Report due !
g ' breaker on J Bus. f 2/21 l . i
, c)* ICS Equipment Investigation SU Complete NOM 86 Edgr Report to NRC 02-07-86.
NOM 86-95 Distribution of Engr Re s NOM 86-132 Removed NRC Quarantine. port. g ,
- h. '1. MI 1. Visual Complete MI complete 1-21-86.
Inspections .and i 7] ff Photographs j
] 2. Mt 2. Insitu Veriff- ' Awaiting Micomplete1-21-86.
cation of Power Supply Closure Odtstanding NCR and MIDR. Setpoints l
.- 3.. MI 3. Load Carrying Complete Mkcomplete1-21-86.
Capability of DC ; 1 Power Supplies
- 4. MI 4. Insitu Verifi- Awaiting- MI complete 1-21-86.- MIDR cation of Power Supply Closure open pending closure of
~.,. Monitor Bus Voltage MI.2 MIDR. Identified
.Setpoints , drifting setpoint'. -
5.1 MI 5. Insitu Verifi- Complete 0.'12/0.15 sec vice 0.5 set cation of S1 and S2 Work complete 1-21. i Trip Delay , ,
, c
'h r $y LEGEND: SU Startup Rtquired LT = Long Tcra l STATUS DATE 02-14 hf PE - Power Escalation DECEMER 26, 1985 TRANSIENT i TIME 0800 kt -
NA = Not Applicable Page 6 of 19
, ST = Short Ters - ACTION LIST - -
f hf3YEhNkiN/ hah *Nufbes Per GAC 85-1001 Apply DESCRIPTION RESPONSIBILITY SCHEDULL STATUS WR No./NCR/ COMENTS q . ETC. .
.x . . ,
(- i3 c)
- 5.2 MI 5. ICS Repowering -
Trbishooting NON 86-63 Revised Troubleshooting
'N .
Plan to NRC 01-30-86. NOM 86-72 and NRC on 1-31-86. 7S i 6.1 MI 6. Bench Test SU . Awaiting M(beingrevisedto
}31) c) 4
) -
Power Supply' Monitor Closure include insitu testing.
@ 6.2 MI-6'. Remove and N. Brock ! SU Complete Found defective E replace defective crip/ connection on +24VDC f;.1 i wiring for further i
bds. Probable cause of ICS - FE troubleshooting
- S /S2. trip. -
Af$ . G 7. MI 7. Bench Test to , Awaiting. Teansferred to BWOG
-; , Determine Hand / Auto Closure NRC Concurrence 1-26-86 d
c. Station i
. I g ,
d)* Repair Action Plan for ICS *
, SU Due 2/18 I preparation.
f . @g y e) Loss of ICS Procedure l .
~
.! 1. Equipment Input N. Brock SU Complete Recovery of ICS Power.
M-ATOG review. RJR 86-19, ['M ~ Items 5 and 6. Ep NOM 86-128 Closure Report issued g 2413-86
', 2. Procedure Development B. Spencer SU Due 2/7- PRC Approval 2/5 i -
- 3. AT0G Review of ICS 8. For.d SU Due 2/1
- S9e item 10.a for' action.
e i
Th. LEGEND: SU = Startup R:quir:d 7 LT.= Lcng Tcra STATUS DATE 02-14-86 if: PE = Power Escalation DECEMBER 26, 1985 TRANSIENT TIME 0800 W ' NA = Not Applicable Page 7 of 19 s ST = Short Tem - ACTION LIST -
& l NY$h Sk /h r GuNebes Per GAC 85-1001 Apply & DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ COMMENTS @ ETC.
e 3 'f) Engineering review of DBRs on.all mods to be 2[?2 proposed ICS Mods: implemented requested by [g: NRC-IIT. N1 1. BTU Limits V. Lewis ST Complete Edgineeringmodfollowing iff receipt of B and W Analysis. p f. . NOM 86-108 Closure Report issued 3 2-8-86. r
- 2. RCS Flow Noise V. Lewis ST Complete Eng1'neering mod following p
[% - receipt of B and W Analysis. f .. NOM 86-108 Closure Report issued b-a 248-86. i M' 3. WR Recorder Alarms V. Lewis SU Chaplete ' ECN R-0245 ECNs released 1/27.
$ NOM 86-109 Closure Report issued sq 2-8-86.
Q . 4? ** 4. Modify ADV, TBV V. Lewis SU Complete ECN R-0357 Ddsign Package by 01-27-86. R to fail closed on Sub B R loss of ICS NOM 86-90 DBR Info to NRC 2-7-86. k . HOM 86-129 ClosureRptissued2-13-86. NOM 86-130 Transmittal to NRC 2-13-86. k{;;. ** 5. Provide remote manual V. Lewis SU Complete ECN R-0357 Cdnsider including MFP
$ control of AFW Sub A speed and SU FW Valves.
control valves Design Package by 01-27-86.
; independent of ICS NOM 86-90 DBR Info to NRC 2-7-86.
. NOM 86-129 Closure Rpt. issued 2-13-86
,, - NOM 86-130 Transmittal to NRC 2-13-86.
i 5 J - t 1 f l 4' s .. .
LEGEND: SU = Startup Requir:d
% LT = Long Term STATUS DATE 02-14-86 Lh PE = Power Escalation DECEMBER.26, 1985 TRANSIENT TIME 0800
-l NA-- Not Applicable Page 8 of 19 r ST = Short Tem - ACTION LIST - l
~
I NNh!bN/h! *Nufhebes Per GAC 8S-1001 Apply gj DESCP.IPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ ! COMENTS 4 ' ETC.
$ 3, f)** 6. Evaluate multi-channel V. Lewis ICS/NNI power supply SU Closure Multi-channel monitors not due 2/14 -
appropriates ECN will monitors to trip improve wiring.
~
q N
,' function rd n
** 7.'
kg g. , Evaluate ICS Trouble V. Lewis SU Revising Mge't Review Draft 2/3. Annunciator and NNI gt s / Closure Rpt ' g - g
. . Due 2/14. !
[/ ** 8. Provide protective V. Lewis SU Complete' RdR 86-19, Item 2f. t fusing for AC/DC - Closure Report: w
, supplied loads NON 86-110 Transmitted to FRC 2-8-86.
~
. . NOM 86-111 Distributed Closure Report h.
W 2j8-86. - Y ** 9. Investigate Signal V. Lewis SU Investigating !
.- Converson/NNI -
Report Due !
\ Interface J 2/14 8 %: l r
% ** 10. Investigate ICS/NNI V. Lewis SU closure Change 15 amp trip to 20 amp 3 k Power Supply Load - Due 2/21 1-27-86 meno from Williams.- hR current setpoint Process Standards change @ rdquired. Formal calc 1 - package and ECN due 2/21.
- 11. Remove Aux Steam V. Lewis 3 ST Closure Rpt. Consider >roviding
- 24. VDC l Control from ICS Due 2/21 from anotier source. '
Can Loss of.ICS Power C. Linkhart SU Complete RJR 86-19, Item 7. cause failure to annunc. NOM 86-133 Closure Rpt. issued 2-14-86.
, Ig) Loss of ICS Power j i
.y - '
? !
__ . __b
)5; LEGEND: SU = Startup R:quircd ,
LT o Long Tsra - STATUS DATE 02-14-86 ylR PE = Power Escalation DECEMBER 26, 1985 TRANSIENT TIfE 0800 NA = Not Applicable Page 9 uf 19
; ST = Short Tem - ACTION LIST -
. :
- Tion N!hN S N/ h b Gufhe knes pee GAC 85-1001 Apply DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ C009 TENTS
- j ETC.
; f 35. h) Motor Operated TSV V. Lewis LT Closure RJR 86-19. Item 8.
, isolation valves due 2/21 Living Schedule Item.
J .. ' 4,~ lel Pump (P-236) failure ' e.d 1" a)* Repair / replace M. Price LT Active NOM 86-42 Troubleshooting Plar,
? -
/ released for implementation
{4 1f21-86. L.. b)* Cause M. Price SU Closed NOM 86-f2 Issued 1-24-86.
, , c) 1. Revise SPs for running J. Field . SU Closure -
. , w/o,MU Pump due 2/15 ,
' ,q ~
_ 2. Detemine remaining V. Lewis SU Complete NOM 86-79 C1'osure Report issued
^;
. service life on
- 02-01-86 4.8 years of, e HPI motors. continuous operation remain.,
d)* Effect of SFV-25003 J. Field SU Closure Rpt. KOM 86 T oubleshooting Action Plan O closure Due 2/14 approved 01-10-86. Closeds. 1'. . 02-07-86. e) 1. Consider changing V. Lewis ,ST Complete WR 109713 Long Tera change, requires { suction valve Safety Analysis. i SFV-23508 from NOM 86-80 Issued 02-01-86.
" Modulating" to .
" Seal-In."
i l
LEGEND: SU o Startup RIquircd ! . L 3 *
, LT = Long Tcra STATUS DATE 02-14-86 , - PE - Power Escalation DECEMBER 26, 1985 TRANSIENT TIME 0800 M NA - Not Applicable Page 10 of 19 h
ST = Short Terin - ACTION LIST.-
$ l Y' -
NYNh!!!iN/h r*NNes Per GAC 85-1001. Apply i DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/. COMENTS .
- j. ETC.
k 14. e) 2. Consider Valve V. Lewis ST Complete Idterlock M/U Tank- discharge d interlock modifications valve w/8WST valve per RJR ' 86-19 Item 2d. g '# jpl , NOM 86-80' Issued 02-01-86 p ., , l f) lik Consider reroute of V. Lewis - LT Comple te NON 86-106 Closure Report distributed
*h c. HPI/MU miniflow to <
2-8-86. prevent flooding MU Tank
}"' l
.e g) Reconsider Design Basis R. Dieterich LT Complete NOM 86-51 USAR Design Basis requires SFAS of MU Tank Outlet - SFAS closure. p 'h) Review of IE Info Notice 85-94,' Miniflow on V. Lewis , SU Complete- NOM 86-78 Closure Report issued 04-01-86. Review shows not Safeguards Pumps a plicable. s.
' 5,** Radiatior) Monitor R-15001- l
[ .a) ' Repair M. Price SU Complete WR107890 Mechanical Work Complete-g
. WR107891 (Replaced Seals) d:p Design evaluation b) V. Lewis SU Complete NOM 86-38 Closure Report 01-18-86.
Q f
- 6. RCS Overcooling l
i: a) 1. Tech Spec review R. Colombo SU Complete RWC 85-804 Reportable Cooldown f exceeded Tech Spec by 108*F
~ i J -
M , 70 LEGEND: SU e Startup R: quired g LT = Lcng Tsra STATUS DATE 02-14-86 (p , PE,=. Power Escala. tion DECEMBER 26, 1985 TRANSIENT TIME 0800
;g NA = Not Applicable -
Pdge 11 of 19
'5 ST = Short Tenn - ACTION LIST -
h N YNhoN YN/ h YGu Ne nes Per GAC 85-1001Epply
$ DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/
CO M NTS g ETC. f
- 6 a). 2. Overcooling R. Dieterich SU Due 2/15 RWC 86-55 Requested Official-1 Calculation calculation to be
$ J .
completed by Engineering. i b) Analysis / evaluation
.i 1. B+W Calculations J. Field f SU ,
- 1
- a. Preliminary Complete- NOM 86-14 No adverse effects on NSS.
Ep Evaluation ' k . . - I
- b. Core Lift Report Complete NOM 86-14 0 to 21 percent core lift.
g{.9. : .
, N adverse effects.
J c. Fatigue Report Complete Od-05-86.
- 2. .NSAC Calculations J. Field SU Complete NOM 86-13 Determined no adverse (G
M
. effects.
( i J. Field
- 3. C1,ospre Report SU Complete NOM 86-117 Distributed Closure Rpt.
W 2211-86. l y; NOM 86-118 TEansmitted Closure Rpt. j g ' to NRC 2-11-86. M ! $ 7.** Health Ibsics ]
~
hi a) Health Phsics Aspects / F. Kellie SU Complete NOM 86-43 Riport issued 1-21-86. $ Radiological Evaluation NOM 86-62 Transmitted to NRC 1-28-86 G i NOM 86-96 Distribution of Rev. 1
' Closure Report.
4 NOM 86-97 Transmitted Rev. I to NRC, f?l; 02-07-86..
f LEGEND:. SU = Startup R; quired [ .
- g. LT o Leng Tcra STATUS DATE 02-14-86 f PE = Power Escalation DECEMBER 26, 1985 TRANSIENT ! TIME 0800 f NA = Not Applicable Page 12 of 19 ST - Short Ters - ACTION LIST - ;
b 0$fNh!$0{N/h r*Nufhebes Per GAC 85-1001 Apply DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ l COMMENTS h ETC. I hg , 7,C* b)* 1. Flooding / Filling of J. Field Complete I SU NOM 86-17 Helium leak test of WGS y . Waste Gas Header / completed. Action Plan d Surge Tank approved 01-13-86. Work complete. (i NOM 86-85 Closure Report 2-5-86.'
? NOM 86-86 Transmittal to NRC-V,
{,' ,
< 2-5-86.
k ** 2.. Waste Gas Compressor M. Price SU Complete WR 107518 Investigation shows no ki. Investigation ' damage. P
- NOM 86-119 Transmitted Closure Rpt. to
% NOM 86-120 NRC 2-11-86.
(' c
, Distributed CR 2-11-86.
i
@ c) Usefulness of RJR-013 V. Lewis ST Complete NOM 86-50 Ldng. term or. Living c; Schedule. Interim trend k points on H3TMI. 8 weeks
@ 1ead time on PROMS. @ NOM 86-65 T 0(ansmittedtoNRC-IIT $ , 1-31-86. 4 , i ! J, 8. Emergency Plan - EP Activation Analysis R. Myers SU Closure Rpt Troubleshooting Plan f l a) due 2/.18 approved 01-14-86, Incorporating Redeker /c.. , comments of 2/12. 1 3 TSC Fire Sprinkler Closure co b) V. Lew'is SU NCR S-5060 Test of effected $ g Actuation due 2/14 ECN R-0197 annunciators done 2/5. Y g.. - i (
. i k
' LEGENO: SU = Startup R2 quired . .
L LT = Long Tcra STATUS DATE 02-14-86 PE = Power Escalation DECEMBER 26 1985 TRANSIENT TIME 0800 , MA ='Not Applicable Page 13 of 19 ST = Short Tem - ACTION LIST - , i kNNN!ho!k/hN*Nuk!eWnes Pe GAC 85-1001 Apply I DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ I COMENTS , ETC. -
- 8 ', c) State Auto-notification R. % ers SU Closure Rpt. System worked properly.
y upon SFAS due 2/18 , Incorporating comments
'y in Reports.
t 9 Training
'a) Sequence of Events J. Mau SU. Complete BOOGItem5. RJR 86-19, Awareness Item 9. All operators received training by 1/24. a
' NOM 86-102 Distributed Closure Report 238-86. .
b) New Procedures and Mods J. Mau SU Interim Identificationand Complete - scheduling of training. RJR 86-19 Item 9. 8WOG , Ites 1. Review must be -o
~
documented. NOM 86-102 Interim Closure Report - distributed 2-8-86. Final dde 3-7-86 following completion of training. NOM 86-103 Transmitted to NRC-2-8-86. c) 'Short Tem Training J. Mau ST Interim Edent Training 1 week / shift' Complete required. . i d) Simulator Training J. Mau SU Interin RJR 86-19. Item 10. Complete Simulator training Required. Interim o) Command and Control J. Mau SU Tkaining simultaneous with' Training Complete that of Item 9b.
'l
. ! H
- ---___--______ ___ __ ___-_-_-__-_______: . _ d
- . -. -- _ .-- = - - . - - - .---_ --_-.-- - .- _. ..
LEGEND: SU = Startup R::quirsd LT = Long Tsrm STATUS DATE 02-14-86 PE = Power Escalation DECEMER 26, 1985 TRANSIENT TIE 0800 l NA = Not Applicable Page 14 of 19 l, , ST = Short Tern - ACTION LIST - ; N $Ye!h$ $ing/Np N *Nuk nes Per GAC 85-1001 Apply DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ COMENTS ETC. g 10.' Operational Review f
. a) Procedure Adequacy B. Ford SU Due 2/14 SJ Redeker preparing -
7 Summary, due 2/14. Includes y[ ATOG Review per item 3.e.3. q5l b) Communications B. Ford . SU Due 2/14 y.: . (( c) Operator Performance B. Ford SU Due'2/14 { i d)' Local / Manual Valve B. Ford SU Due 2/14 l;g Operations -
** e) Effect of Loss of ICS V. Lewis ' SU Complete L during SFAS R$R86-19. Item 4.
NOM 86-124 Closure Rpt. issued 2-12-86.
- c 11** System / Component Response .
a)* Aux Steam PSV-36012A J. Field , SU To be Revised NON 86-11 R1 Closure Report revised. . Due 2/25 ' 1417-86. Revision to. address loss of ICS causing A4x Steam overflow. i 1 l l ,
L - li LEGEND: SU = Startup R; quired LT = Leng Tcra h STATUS DATE 02-14-86 p PE = Power Escalation DECEMBER 26, 1985 TRANSIENT TIME' 0800 y ' NA = Not Appifcable Page 15 of 19
% ST = Short Tem - ACTION LIST -
d . he!h ! ng/h$ *NuNebes Per GAC 85-1001 Apply j DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ i C0f# TENTS
; ETC.
I 11**'b)* 1) Failure investigation J. Field SU Closure Rpt. WR107855/856 Ts Action Plans issued per
! of Aux FW, FV-20527, Due 2/21 NOM 86-06/01-10-86. Repair i FV-20528 -
NOM 86-74 Action Plan for FV-20527/8 [;; , issued 1-21-86 per NOM q , 86-41. , fl ' 4 NOM 86-76 Distribution of Engr. ?; .- Report 2/1. k~ K NOM 86-89 Removed from Q-List per NRC, 2-6-86: f;^'. NOM 86-101 Distribution of Repair Plan, 2-8-86. [,"J - J. Field L' 2) FWS-063, FWS-064 , SU Closure Rpt. NOM 86-75 FWS-063 had rusted b Due 2/21 bearings. Others OK. & . NOM 86-77 Distribution of Engr. F Report 2/1. NOM 86-89 Removed from Q-List per t NRC, 2-6-86. [, NOM 86-100 Distribution of Repair Plan, 2-8-86. t c)* Main Steam, PSV-20544 J. Field SU Complete NON 86-11 R1 Closure Report revised 01-17-86 to NRC on 1-21-86. NOM 86-70 Velve returned to Operations per NRC 1-30 86. d) RCP Seal Injection Flow J. Field SU Complete NON 86-07 No maintenance required. } NOM 86-66 Transmitted to NRC-IIT 9 01-31-86. M i d , _ __ __' _g
LEGEND: SU = Startup R:quircd j' LT = Long Tcra f STATUS DATE 02-14-86 PE = Power Escalation DECEMBER 26, 19E5 TRANSIENT i TIME OC NA = Not Applicable Page 16 of 19 ST = Short Term - ACTION LIST - I kNNYe!h!$$ing/h!b*Nuf0ebes Per GAC 85-1001 Apply ' DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ i CO M NTS f
' '- ~
ETC. 11** o) DHP Lo 011 in Auto Feeder M. Price SU Complete NOM 86-09 No problems'found. Closure v Report 01-10-86. . NOM 86-67 Transmitted Closure Report , to NRC 2-8-86. n- f)* Feedwater Heater Shell J. Field SU Closure Rpt. NOM 86-40 Reworked control valve S ' Reliefs opened being revised seats. Due 2/17 ' t S g)* SFV-23811 Position J. Field , SU Complete NOM 86-60 Z9020 indicated " Closed" Indication
~
although F9003 continued to iridicate flow.
, MI issued, due 2/3.
NOM 86-104 Closure Report distributed, 2-8-86.
' NOM 86-105 Closure Report transmitted to NRC, 2-8-86.
h) ' PM program for M. Price SU Mge't Review complete
- manual isolation valves 2-10-86 1.) Deterinine List of B. Spencer SU Complete Issued to Maintenance Values to PM/ Stroke
' 2-13-86. .
for SU.
.y 2.) PM Lists of 102 Values Price / Spencer Scheduled t
W SU Due 3/5 Maintenance / Ops Joint Project. 2 3.)~ Enter into PM Program M. Price ST Due 4/1/86 Fjllow-up Scheduling.
- 1) HPI Flow Bhlance Test, J. Field SU Complete NOM 86-71 Report sent to NRC-IIT per STP-085. -
H. Bailey requested and ~' sent 01-31-86. s I i, .
~
l %. _ i .. _ .- _ -__________m
LEGENO: SU = Startup Rsquirsd 4 + LT = Lcng T:rs STATUS DATE 02-14-86 , PE = Power Escalation DECEMBER 26, 1935 TRANSIENT . ' TIME 0800 , NA = Not Applicable
- Page 17 of 19. ,
ST = Short Ters - ACTION LIST - .
. I T bye!ho!!$N/Nh*b0ebes'Pe GAC 85-1001 Apply i DESC'RIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ , CO M NTS ETC.
g}2, Quarantined Equipment List
- I
'a) PSV-20544, Main Steam J. Field SU Off List NOM C6-11 R1 . Closure Report, Rev. I-h.
es Code Relief to NRC 1-21-86..
'j , ,
NOM 86-70 Returned to Operations per
., NRC 1-30-86. . .. b) FV-20527, A-AFW Control J. Field f SU off List Hold on repairs pending NRC Valve will tear down valve 4 internals and replace hand i NOM 86-44 jack.
l; ' Engineering Report to NRC. , IJ21-86.
, NON 86-74 Revised. Engr. Report.
',* NON 86-41 Hold on repairs pending NRC concurrence. NOM 86-89 Removed from Q-List 2-6-86. . c) FV-20528, B-AFW Control J. Field SU Off List Will rebuild hand jack.
' Valve. NOM 86-44 Engineering Report to NRC' 1-21-86.
NOM 86-74 Revised Engr Report.- NOM 86-41 Hold on repairs pending NRC i
' . concurrence.
HOM 86-89 Removed from Q-List 2-6-86, i ' I J. Field d) FWS-063, A-AFW Manual SU off List NOM 86-75 Action Plan and Isolation Valve mis completed 1-21-86. NOM 86-89 Removed from Q.-List 2-6-86. I
- j. -
l
! v
! ] 'w. t . _
. 1 4
LEGEND: SU = Startup R:; quired i n . LT = Lcng Tcra STATUS DATE 02-14-86 PE = Power Escalation DECEMBER 26, 1985 TRANSIENT l TIME 0800 C 3 NA = Not Applicable Page 18 of 19 ST = Short Term - ACTION LIST - ' s h0$NIh!$$iN/k!kh*NufheknesPe GAC 85-1001 Apply 1 . "' . DESCRIPTION RESPONSIBILITY SCHEDULE STATUS WR No./NCR/ COMENTS
! ETC. ) ,
[ ;12. e) FWS-064, B-AFW Manual J. Field SU 'Off List NON 86-75 Adtion Plan and 5- Isolation Yalve mis completed 1-21-86. j , NOM 86-89 Removed from Q-List 2-6-86. U i. h , f) ICS asse tated Power C. Linkhart SU Off List Closure Report completed. p "c - Supplw Sde items 3.b.1/2. . g . NOM 86-84 Returned buses C. E, F, and w - 2
.~
J ito Operations 2-4-86. g)- ICS Power Distribution N. Brock SU OFF List Troubleshooting complete. Engr Report to.NRC 02-07-86.
+
l NON 86-94 i
~
NOM 86-132 Removed from Q-List 2-14-86. ,
. Switches S1 and S2 :
separately Quarantined.
? y I p .
3 13. Procedure Changes Tracking J. Field SU Draft List Steve Luke maintain list of l ' Published al.1 related procedure < l 1/27 changes.. Update due j [ 2/5 ; ; I
- q !
- 14. Carryover Items from October 2, '
n 1985 Event Action List: i , I
- 1. Item II: Root Cause S. Crunk ST Pending NA CiosureRequired Final R'ep' ort Due 2/24 ! l
? ! l . s a , l ) n t g . _. l . . n n
LEGEND: SU y Startup R:quircd p LT - Long T;rm STATUS DATE 02-14-86 s . PE - Power Escalation DECEMBER 26, 1985 TRANSIENT ' TIME 0800 4 NA - Not Applicable . Pdge 19. of.19
! ST.= Short Term - ACTION LIST - l t
' l Obfe!hho!fng/hN'NulaeMesPei'GAC85-1001 Apply
. DESCRIPTION RESPONSIBILITY SCHEDULE STATUS NR No./NCR/ COMENTS
)f ((.
- ETC.
pe- .:14, 2. . Item VIb: Check common N. Brock LT Complete Original Commitment
! -' . failure possibility on 12-15-85.
Pjd* controller ifnkages. NOM 86-121 Distributed Closure gj r Report 2-11-86. I Item VIIII: Correct / M. Price LT O'verdue CCL 85-0856 Original Commitment f upgrade documentation on e l 3, turbin~e trip block. 01-01-86.
! U'~
f Due2/%-
, Y 9 J 4.' Item. Villa: Design V. Lewis LT Complete To Living Schedule by l .-
improvements to reduce ' 01-30-86.
+
MSR Relief Valve leakage ; p 5. . Item Xb: Perform ICS N. Brock . ' PE- . Needs Closure NA Original ' Commitment h 7 tuning at power Due 2/9 12-01-85.
; m , .<
- 6. Item XIIlj:- S. Crunk LT Pending NA Need Final Report
. P-319 Bearing Failure Due 2/24 -
l
. Root Cause Report H
f, 7. Item XIVb: S Crunk LT Pending CCL 85-0862 Original Commitment f, Loss of Aux. Steam Event
. Due 2/24 12-15 85 Root Cause Report i j 8 Item XVIb: D. Comstock
LT Due 3-1-86 NA Procedure Adequacy 4 Lessons Learned i . ! 9. Item V.f: V. Lewis LT In Progress NEP 50205.14 is in final i Establishcriteriafor review 2-5-86. Will close 4 process setpoint by 3-15-86. .) ' j determination f I ! I i . s . z _ _n .. L_ - __ . _ _ - _ _ _ _
, (,
4 i. 1 l ATTACHMENT 4 REGION V 00NFIRMATORY ACTION LETTER I \ l l t G 9 6 s l e a 1
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/ .a ase ,Io UNITED STATES <f . g NUCLEAR -REGULATORY COMMISSION i 3 j REGION V
! %
- 14e0 MARIA LANE, SUITE 210
/ wALauf CREEK.CALtPORNIA 94006 l DEc 2 61985 .
Docket No. 50-312 License No. OPR-54 CONFIRMATORY ACTION LETTER
~ ~
Sacramento Municipal Utility Distrht p.0. Box 15830 Sacramento, California 95813 N Attention: Mr. R. J. Podriguez ~
Subject:
Return of Rancho Seco to Power Operations Gentlemen: 1 In a Confirmatory Action letter that'I provided to you on December 26. 1985, I stated it was my understanding that prfor to return of Rancho Seco. to power operations, you will conduct a toot e.ause analysis of the reactor trip which occurred on Secober 26, 1985. In addition, and prior to return to power, you will provide Lhv NRC d britefls:V tsf your assessment of the root cause and your justification as to why the Rancho Seco facility is ready to resume power operations. Based upon additional conversations we held on De' camber 26,1985, and upon the~ fact that the NRC is sending an Augmented Inspection Team to Rancho Seco to review the December 26, 1985 reactor trip and resulting.cooldown. I further understand that you will hold in abeyance any repair work planned on equipment that malfunctioned during the incident, sucsi as dismantling existing evidence, until the SMUD and NRC insps,ction teams have hao an opportunity to I evaluate this event. This restriction does not cover any repair needed to place systems and equipment in sersice that are required to monitor and maintain safe conditions of the facility, i.e., the containment radiation I monitor that failed during the cour:e of the event. l If sty understanding concerning your proposed actions, as sumarized above, is not correct, you should promptly notify this office in writing. i 1
. Sincerely.
t
] m // O '
y J. B. Martin Regional Administrato j i cc: L. G. Schwf agar, .W00
~
G. Coward SMUD Suu of m. - ff p ?![2 ?;6 k _ , _ f -
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Y 9 6 D D , ATTACHMENT S POST TRIP PARAMETRIC DATA - C e E O e O O
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o H t I a l GRAPils FOR:
!: RANCHO SECO REACTOR TRIP
, DECEMBEW 26, 1985 h
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D p w , E s. r: (i M 1 c .,1 g F. L R January 3, 1986 4
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r TABLE OF CONTENTS
. a
- 1. IDADS Computer Point Identifications
- 2. RCS Pressure / Temperature (Tave)
.3. 'RCS Pressure B-Loop
- . :4. RCS Average Temperature U
- 5. RCS Pressure /PZR' Level
, 6. RCS Hot Leg Temp ~erature vs. TSAT
~
- 7. RCS Sub-Cooling Margin .
'8. HPI & Make Up F10w........................Rev. 1 .1/6/86
- 9. Total HPI &'Make Up Injection
- 10. OTSG Pressures-
- 11. High Resolution of 0TSG Pressures' .
- 12. Startup'0TSG Levels
- 13. . Full Range OTSG Levels
- 14. Extended Trend.of A OTSG Levels
- 15. Extended Trend of B OTSG Levels
- 16. Main Feedwater Flowrates -
- 17. . Aux Feedwater:Flowrates- -
- 18. . Main Feed Pump Speed (RPM) >
- 19. Main Feed Pump 1 Discharge ~ Pressure .
20 Rx Trip Comparison Dec. 26, 1985 vs Mar. 20, 1985L 21.--Pressure ' Temperature to Tech. Spec. Cooldown Curve..........Rev.;1 1/6/86
- 22. Make Up Tank Level
, 23. ' Aux 81dg Vent Stack Rad Monitor W
- 24. High Resolution of M.U.T. Level
- 4
,q
_____m _ _ . _ _ . _ _ _ _ _ . _ _ _ _ . _ _ _ _ _.__.___.______.__.________.__.___.__a______. . . _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ m . '_.,1 ..-J._ ._1M
I v e IDADS COMPUTER POINT IDENTIFICATIONS { l . . PT ID PT SOURCE DESCRIPTION UNITS INSTRUENT . RANGE . 00f9ENTS.
'E0603 'LoveJoy. VOLTS DC BFP A SPEED TO LOVEJ0Y (MS) VOLTS DC 0.000 10.000
~E0614 .LoveJoy . VOLTS DC BFP B SPEED TO LOVEJ0Y (MS) VOLTS DC 0.000 10.000 .
F1660 FT-31802 .AFW FLOW TO OTSG A GPM 0.00000 :1300.0-
'F1663 FT-31903 AFW FLOW TO OTSG B GPM 0.00000 1300.0 F9000 FT-23603- MAKEUP FLOW GPM 0.00000 .250.00. ,
F9001 FT-23805 HPI LOOP A FLOW GPM -0.00000 600.00 ' FT-23306 HPI LOOP B FLOW 0.00000 "+ F9002. . GPM 600.00 ~ ~ F9003 . FT-23807: -HPI LOOP A FLOW GPM 0.00'K)0 600.00' ' F9004 FT-23808- HPI LOOP B FLOW GPM 0.0' 00 600.00. '
?
F9600 'FI-20563 MFW FLOW B FFi/HR 0. 00t. 0 6.5000 F9601 FI-20535- W W FLOW A WBL/HR 0.00000 ~ 6.5000 L1805 LT-20503C' .OTSG A SU RNG IN H20 0.00000 250.00- . UNCOMPENSATED-L1807 LT.-20504C OTSG B SU.RNG IN H20 0.00000 250.00 UNCOMPENSATED L9000 LT-23502A MAKEUP ~ TANK LEVEL IN H2O 0.00000 100.00- UNCOMPENSATED. ;
.L9005 LT-21503A PZR LEVEL IN H2O. 0.00000 320.00- UNCOMPENSATED ~
L9801 LT-20501 OTSG A LEVEL' FULL.RNG . IN H2O 0.00000 600.00' UNCOMPENSATED , L9802 LT-20502 0TSG B LEVEL FULL RNG IN H2O 0.00000 600.00 UNdOMPENSATED
. P9000 PT-21050 RCS. PRES LOOP B PSIG 0.00000 3000.0- -
P9300' PI-205208 OTSG PRESSURE B PSIG .0.00000 -1200.0 t P9301 PI-205198 OTSG PRESSURE A PSIG 0.0000 1200.0 T9010 TY-21031C RCS LOOP B HOT LEG TEMP DEG F 120.00 .920.00
.T9016 ..TI-21024A RCS LOOP B COLD LEG TEMP DEG F 50.000 650.00 ,'
9 4
~
'12-30-85 Fig. #1-s s pp g ,_
._ - ---~~__ _____i,_? '
REACTO R ~~ RIP DEC. 26, 985 l RCS PRESSURE / TEMPERATURE (Tave) 600 - l 2.3 590 - ~ 580 - O J 570 - 2.1 - p 560 - 2-
-m 550 - .hj 540 - 1.9 -
g 530 - 1.8 - Rx Trip til 5 2 0 - b 510 - T 1.7 - p 500 D 490 -Di e o} 1.6 - , L o 1.5 - - h 480 ' til 470 - 8 ~ 3,4 _
% 460 .
13-h 450 - , 440 - 1.2 - 430 - ' 420 - 11- ' 410 - 1-400 - 0.9 : - 390 - . 380 - O.8 - , ,, , , , 04:00:00 10:00 .20:00 30:00 40:00 50:00 05:00:00 TIME .
-- O - RCS PRESS B-LOOP -O- RCS-Tave B-LOOP
. IDADS P9000 AVERAGE.OF IDADS T9010 AND T9016 ' Fig. #2
V
. REACTOR TRIP DEC. 26, 1985 -
RCS PRESSURE B-LOOP 2.3 2.2 - " 2.1 - 2-1.9 - 8 1.8 - ur SE 1.7 - E 1.6 - U Rx TRIP O
$o 1.5 -
[ 1.4 - 1.3 - 1.2 - . 1.1 - - 3_ . O.9 , , , , , === 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME RCS-PRESS IDADS P9000 Fig. #3 4 - . . _ .. . . . . - . . - . . . . . - ~ . . . .
L R EACTOR TR P DEC. 26, ' 985-ROS AVERAGE TEMPERATURE 600 590 - l 580 - 570 - 560 - 550 - C 540 - 530 - E 520 - 510 - Rx TRIP Q O 500 . g 490 - E 480 - . h E 470 - g 460 -
- n. 450 -
440 - 430 - 420 - 410 - 400 - 390 - 380 , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME RCS Tove B-LOOP AVERAGE OF IDADS T9010 AND T9016 Fig. #4
^ d
REAC~~O R TRl 3 DEC. 26, 1985 RCS PRESSURE /PZR LEVEL - 2.4 , 2.2 - g 2- . 1.8 - 8W n 1.6 -
^
E n. " 1.4 - O l v v"c . o Wo o em 1.2 - o s
]me@g g 1-Q- O.8 - .
O.6 - Rx TRIP Q O.4 - s O.2 -- n O , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME
. - B -- RCS PRESSURE -O- PZR LEVEL (UNCOMPENSATED) ;
, IDADS P9000 IDADS L9005 Fig. #5 .;
n - - - a
~
R EACTO R TRIP DEC. 26, 1985 RCS HOT LEG TEMPERATURE VS. TSAT
. 660 -
r, , 640 - . 620 - n ; 600 - , 580 - 0 m- ' 560 - tr . O g 540 - ,
- W 520 - . ,
3 500 - Rx TRIP O W 480 - n. 460 - / 440 - 420 - 400 -- 380 , ,- , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 . TIME i -G-- SATURATION TEMP. -O- RCS-B LOOP HOT LEG - IDADS T9010 FROM ASME SlEAM TABLES Fig. #6 .?
. - _ . - EXF810K1RLMEUNIED - _
REACTOR TRI:3 DEC. 26, 1985 RCS SUB-COOLING MARGIN 210 200 - 190 - 180 - 170 -
' 160 -
h 150 -
@ 140 -
$ 130 -
hl 120 - Rx WP Q a: , g 110 - e 100 - E 90 - i 80 -
- 70 -
60 - 50' - 40 30 , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME Tsat MINUS Thot CALCULATED FROM T VS T GRAPil 10T MT Fig. f7 ci
R EAC- O R TR P DEC. 26, 1985
~
HPI & MAKE UP FLOW 300 280 - Rx TRIP (i A 3. ggpg Spy A 260 - CloSto (fUtt) 240 - "A" IPI SrV Ri1 OPEN- "B" IIPI SFV -
"B" IIPI PUMP STOP TilR0 TILED 220 - "B" IIPI PUMP START-- "B" IIPI SFV CL n -
"A" IPI PUMP STOP "B" llPI PUMP START 200 - M g 180.-
C "C" IPI SeV CLOSED MAKE UP PUMP TRIPPED-O. e 160 - D 'N (iULL) AT 0500
\ - D" IIPI SFV g 140 - .A" llPI PUMP START--m -
t
, CLOSED (f0LL) "
B" llPI e PUMP E 120- A-B-C-D llPI SFV's - STOP/ START
. n o m ED POSITION A g. A 100 - j K
H I il r
-
- Indicates the
- c C c beginning of 60 -
possible invalid H a 7 , n g , computer data, 40 - actual flow was }" c" _ _ p D l umry T . D n, D -. D , probably zero. 20 - _.D
- H M M __
M A O .
^ ,' ' A .
04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00
~ TIME A B C -
D M/U ilPI IIPI HPI IIPI INJECTION pig, gg r9003 09001 09004 09007 , F9000 a
O REACTOR TR P DEC. 26, 1985 , TOTAL HPl & MAKE UP INJECTION 900 800 -
~
700 - Rx TRIP Q 'k 600 - 3 1 500 - .
~
400 - N. 300 - " 200 - - /A f J 100.-N { 1 -- x,. _ _.-_ , O , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME A , B -- C D M/U TOTAL _ rM rll r%'i . r%'7 __ "'MN " EMHIPN -
l , 1 REACTOR TRIP, DECEV BER 26, 1985 OTSG PRESSURES ' 1.1 1-O.9 0.8 - .. E un O.7 - j gj Rx Trip Q - c O.6 - M o 0.5 -
@L P
O.4 - { 0.3 - O.2 - 0.1 - , 0- , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME A-OTSG B-OTSG ,, IDADS P9301 IDADS P9300 - Fig. #10 .j _ -_____e__-___-______-___-_ - - _ _ _ . &
- REACTOR TRIP DECEV BER 26, 1985 '
HIGH RESOLUTION OF OTSG PRESSURES 1.08 1.06 - 1.04 - 1.02 - g , 1.00 - . O.98 - g O.96 - m O.94 - ae v"c O.92 - em 0.90 0 g
- u) 0 0.88 - '
O.86 - I E L O.84 - O.82 - O.80 - 0.78 - 0.76 - 0.74 - O.72 , , , 04:13:00 04:14:00 04:15:00 04:16:00 04:17:00
^
BME (SECONDS) O A OTSG PRESS -- O - B OTSG PRESS - IDADS P9301~ IDADS P93N) 1 Fig. #11 -
^
s _%.. b -- _ . _ _ _ . _ _ _ _ _ _ ______._m --
! REACTOR TR! 3 , D ECEV BER 26, 1985
- STARTUP OTSG LEVELS 300 280 -
260 - Off SCALE IIIG1 240 - h 220 - a 200 - 180 - 160 - 22_31c- f O u-o en 140 - . u.i
$ 120 - ,
- 100 - .
80 - SO - 40 _ 20 - O , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00. 05:00:00 TIME 1 -G- A-OTSG -O- B-OTSG ' Ji I 13 ADS L1805 IDADS L1807 (UNCCTENSATED) (UNCOMl'ENSATED) Fig. #12 L ________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ n ____-__CM
i REACTO I - R 3 J EC. 26, 1985 FULL RANGE OTSG LEVELS 800 l 700 - Rx TRIP . Q 600 - WATER BEGINS TO SPILL OVER TOP Of SilR000 INTO STEAM ' ANNIX.US SPILT. AGE STOPS gh f o 500 - y c -- b 400 - s 300 - _ n_. ., - 200 - 100 - O- , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME - M
-El- A-OTSG LEVEL - &-- B-OTSG LEVEL -
IDADS L9801 IDADS L9802 Fig. #13 _ _ _ _ _ _ _ _ _ - _ . _ _ _ _ _ .A
- o l .
- REACTOR TRIP DEC. 26, 1985 .
EXTENDED TREND OF A OTSG LEVELS ' 600 3
. Rx TRIP Q 500 -
o l 1 400 -
. O 1
N I
- -b 300 -
- 200 -
C 1 100 - ';
^ .
.'l
- i i O . . .
04:00:00 04:30:00 05:00:00 05:30:00 06:00:00 N TIME O START UP LEVEL -O- FULL RANGE LEVEL IDADs L1805 IDADS L9801 fi 9 Il4 l q a __ _. 4
i g-0 0: 5 1 0 0: g i . 6 F N 0 5 ' 8 ~ 9 L E 1 0 V 0: E S 0 L
,L ,3
- E _
6EV 5 0 G N 2 2EL A0 R8 9 _ G L
. S L S T L CO D _
- UA D F EB I 0 0: DFO 0 E-O M
,0 I D : T PNE I
r 5 0 . RTR TDE ,
- D
_ RNE - 0 0: _ T OX L _ 0 E _ TE ,3 4 V E C 0 L P 7 ._ A ~ U0 8 . F O TL RS 1 R q 1 AD P I O C TA D R T - 0 SI . R x 0: m 0 O
- - - - - 0: . _
0 0 0 0 0 0 O4 . 0 0 0 0 0 0 0 - 6 5 4 3 2 1 oNI b I_.I. k . o l ii i
w REACTOR TRIP, DECEM 3ER 26, 1985 MAIN FEEDWATER FLOWRATES 4.5
&;NCh;=% '
4- , 3.5 - 3-E I Rx Trip Q
\ 2.5 -
3 - 3 g 2- ' o d 1.5 - 1- , 0.5 - O , , , , . 04:00:00 10:00 20:00 30:00 40:00 50:00. 05:00:00 TIME ' A MN FD WTR B MN FD WTR 1 IDADS F9601 IDADS F9600 Fig. #16 f _____ _ _ _ , , . , m-
REACTOR TR P, DECEV 3ER 26, 1985
- AUX FEEDWATER FLOWRATES 1.4 r 0ff SCatE ilIGi m 1.3 -
f H 1.2 - U Il RxTRIP( ) 3,3 _ 1- O 6-O.9 - w gu O.8 -
*8 o
a O.7 - k e wg 1 m
*~ I Es .
o O.s - O.4 - l ~ Il O.3 - . j O.2 - POSSIBLY INVALID DAM
. L-0.1 - q PROBABLY ZERO FLOW
-z__m_______,..
= . ,
04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME '
- El-- A AUX FEED FLOW -O- B AUX FEED FLOW Fig. 917 i M660 IDADS F1663 .
_ m
REAC- O R T R 3 DEC. 26, 1 9~8 5 MAIN FEED PUMP SPEED (RPM) , 4.5 4-4 3.S - 3-O. Rx TRIP vg g3 2.5 - g Os . go 2-Eb m I 1.5 - 1-0.5 - O O 04:00:00 10:00 20:00 30:00 40:00 MM NW TIME
-El-- A-FEED PUMP -&- BgEEgP IDADS E0603 (x750) Fig. #18 n a
l ,
- +
. REACTOR TRIP DEC. 26, 1985 MAIN FEED PUMP DISCHARGE PRESSURE , 1.2 1.1 - ~- -g =
~ - -C--- = = -
1- , , W P DISCHARGE PRESSURE CALCULATED FROM W P RPM , ^ O.9 - Q AND DATA FROM STP-186.
- 1 j Rx TRIP o8 pi e 0.8 -
n_ g , 0.7 -
- I I O.s - .
r, l 0.5 - . t _ _ 4 O.4 , , , , , 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 TIME O A-FEED PUMP -O- B-FEED PUMP CALCULATED CALCULATED p gg _ _ _ -_ J
d Rx TR P COM PARISON DEC. 26,1985 vs Mar. 20,1978 - 600 l
~
580 - , 1
~
~
560 - _ , , 540 - , :; .
~
, '~ ~
520 - r ' t h. e 500 - 4 I 480 - E 460 -
- O w 440 - , g _
" 420 -
400 - E 380 - f 360 - 340 - ? 320 - 300 - 280 - - 260 .....................................................................,......... .........
~
00:00 10 20 30:00 40 50 60:00 70 80 90:00 TIME - Minutes from Rx Trip :' O 12/26/85 TRIP-Tc B --O- 3/20/78 TRIP-Tc B IDADS T 9016 8 & W REPORT BAW-1514 ~ Fig. f20 _ _ ------- _ ---- - - _ . . - - _ _ _ - - - _--- - - _ _ - . - _ _ - - - . - - - - - . - - - - . - o m
4 Rx TR 3 CO V PAR SO NI PRESS-TEMP to TECH SPEC COOLDOWN CURVE 2.4 IIAM LIMII Rx IRIP 2.2 -
- CROSS 88M LIMIT AT "04??:00 INIERIM llRITILE
** REIURN 10 IIORMAL MANGE AI "0505:00 iRACilmE ClRVE s
J k 1.8 -
$ 1.6 -
to a 1.4 - EV DS . yg 1.2 - td o *
$5 3 1-u, ..
o- mit LlHI g 170 M!NilIES O.8 - - Ar1[R Rx 1RII' E 10011. SPEC. EtRVE
~ O.6 - \
O.4 - , _ , SATURATION N CLEVE A O.2 - , O r ~ O 200 - 400 600 TEMPERATURE Tc DEGREES F . Fig. #21
- O-- RCS --m- NDT 1.lMIT -A- 13&W l.lMIT -$- SKFURKilOII:
iOA05 isois ircii. Si tc. ric. 3.i.z-2 PoM i:.03 ric. i . rRoMAsMcSiiAMrAiiii)
?
REACTO R TRIP DEC. 26, 1985 MAKE UP TANK LEVEL 120 - 110 - 0FF SCALE HIG1 ; 100 - 90 - 80 - o y ~70 - b 60 - J 50 - 40 - RX TRIP () 30 - . 20 - 10 - 3 0 . . . . . j 04:00:00 10:00 20:00 30:00 40:00 50:00 05:00:00 i
' TIME f MAKE UP TANK LEVEL 3 ms tm Fig. #22 l
I -- __ __ :_M 1-_ _ _
~
')
r ~ i i REACTOR. ~~ RIP D ECE \A B ER 26, 1985 r i . AUX. BLDG VENT STACK RAD. MONITOR li, 17.0
~
I 16.0 - j 15.0 - .
~
lr 14.0 - - ( o 13.0 - r o ~ t o 12.0. - 3 r o r L
\ 1 * .O -
l 7 10.0 -
- G 9.0 -
2 ;
; o 8.0 -
a v o e 7.0 - r
; 6.0 -
i 5.0 - U l 4 4.0 - a I f 3.0 - t. F 2.0 - i' 1.0 - h [ 0.0 ...................,,.......... ,,,,... 05:00:00 05:05:00 05:10:00 s t TIME l STACK L(. R. MONITOR 5 SEC. INTERVALS) l Fig. #23
+ -
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. . ., y ATTACHMENT 6 i
e HUMAN FACTORS ITEMS 4 e e
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SUMMARY
OF HUMAN FACTORS OBSERVATIONS l 0F TE 26 DECEMBER 1985 TRIP i i Item lAddressed by. IControl Room l I j umb:rl HE0 Number
- l Related l HFE Observations l Disposition 1 1 I l' 1 IGN-0-0324 I Yes IExcessive noise in Control Room IReduce noise level l l l l from essential HVAC I I I I 2 l l No ITime Delays in accessing valves I Analyze accessibility I of valves I 3 IGN-S-0464 l No IPoorly Located emergency / lAdd more different sizes l l l protective equipment I of equipment IGN-S-0466 l l l Clearly mark equipment g i I l l sizes 4 IGN-S-0468 l l l Provide maps indicating i i i l location of equipment IGN-S-0469 l l l Locate lead sealed 1 l l l l lockers with equipment
! I l IDevelop procedure to I l 1 I analyze usefulness l l l l of equipment after each l I 1 -
I trip l 1 1 I 1 l 4 I I Yes . l Time delays thru Control Room IImprove Control Room i l 1 door l Security interface i I I I 5 l i Yes l Difficulty monitoring plant lInstall long cord on NRC l 1 I status idiile on red ph'one I red phone 5 1 -l Yes ISpurious ringing on NRC phone IRepair system to l I I I I alleviate I I I 7 lGN-S-0293 I Yes IPoor consnunication bethen i l Provide new wireless l l Control Room & plant I communication system i I I I 3 lAN-0-0225 l Yes lNo MSLF alarm' l Provide alarm l 1 1 I I
) l I No IPoor local valve status IProvide accurate valve l l l' indication 'l status indication, l
l l l l including highly l l l visible labels I i l 1 - 3 l Yes l .lNo Control Room control of l Investigate providing l ICS components upon ICS failure I redundant controllers 1 l
! ! l l on ICS components
! I I I i IPS-0-0348 l Yes -
l Difficulty tracking status of IReplace indicator with IPS-W-0474 1 1 Pressurizer Relief Tank l trend recorder I .I I I E0's referenced are from the submitted CRDR Summary Report. .
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m3 Item l Addressed by ICcntrol Room I i Numberl'HE0 Number
- l Related l HFE Observations l Disposition-1 I I .
I 12 l l Yes lInconsistent indication between l Modify systems to l l l l SPDS & OTSG 1evel trend recordersl improve consistency I I I l 13 l l Yes lInsufficient status information l Provide important l l 1 on diesel generators l parameter status in the l l l l Control Room ! 14 IGN-0-0278 l Yes IUnre11able/ unreadable recorders (Replace trend & impact , IGN-0-0305 l l I recorders ! IXX-S-0214 l l l lPS-S-0212 l l l PS-S-0209 XX-S-0213 155-S-0215 l l l lGN-S-0208 1 l l < . I l l l 15 IGN-S-0068 l Yes iDue to loss of power, ILabel all indicators '{ lGN-S-0074 l l inconsistency between feedwater I with power source-lGN-S-0075 l l flow meters and trend recorders I and train i I I I i 16 I l No l Poor labeling of breakers S1/S2 l Relabel i I I i 17 i l No - l Difficult to use elementary IImprove nomenclature
. l I I drawings 1.A provide operator .
l l l l training on elementaries I I I i 18 l l Yes IConfusing alarms on IDADS- IImprove operator /IDADS . I I I I interface i I I 1 19 ISF-V-0396 l Yes IAFW components' scattered- Iconsolidate with EFIC. IGN-S-0258 l , I throughout. Control Room l
. I I I I 20 lGN-W-0513 l Yes IDelays in gaining control IRelabel lenses lGN-0-0283 l 'l of SFAS valves l IGN-S-0253 l l "' I I I l' -
l 21 l l No IE0P's were vague IModify E0P's i I l l 22 l l No IDamage to pump IInvestigate feasibility I l l l of adding low suction i I l l trips to important i I I I pops - 23 l l No l Inadequate' personnel support l Provide 24 hr support- ! I I I I i 24 lRI-V-0392 l Yes IDifficulties recognizing loss l Yellow band. failure
-l I l of power l point on indicators l I i 1 25 l l- No lNo information on feedwater iInvestigate, providing i l l heater reliefs '
I relief infprmation i . I I . l l 26 IMR-S-0493 l Yes IRadiation information scattered Ice'ntrally located ' ! I thru l l throughout Control Room I within the RM-ll IMR-S-0511 I I I -
- E0'S referenced are from the submitted.CRDR Samary Report.
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ATTACHNENT 1 CONTROL ROOM EQUIPMENT AFFECTED BY LOSS OF ICS DC POWER I i i l l I i 1 9
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, , y The following list complies all of,the Control Room equipment' affected by the loss of ICS power.
- 1. Hand / Auto Stations Unit Demand Steam Generator / Reactor Reactor Demand Steam Generator ATc Turbine Header Pressure Setpoint "A" Feedwater Demand '8" Feedwater Demand "A* Feedwater Pump Speed '8" Feedwater Pump Speed "A" Main Feedwater Valve '8" Main Feedwater Valve "A" Startup Feedwater "B" Startup Feedwater -
l Valve Valve "A" Turbine 8ypass '8" Turbine Bypass Valves Valves
'A' Atmospheric Exhaust "B" Atsiospheric Exhaust Valves Valves
'A' Auxiliary Feedwater '8' Auxiliary Feedwater
,- Valve . Valve
. . Auxiliary' Steam Reducing Station
- 2. Indications
'A' Loop Feedwater. Flow '8" LoopL Feedwater Flow Recorder .
- Recorder .
Generator Frequency (H1SS) , 1 ICS Indicator Lights on Load Control Panel H-1RI ! (above Diamond Panel) !
- 3. Annunciation (H2PSB) l M4 ICS or Fan Failure l
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