ML20155H665
| ML20155H665 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, 05000000 |
| Issue date: | 05/15/1985 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20155H523 | List: |
| References | |
| RTR-NUREG-1195 AP.28, NUDOCS 8605160326 | |
| Download: ML20155H665 (27) | |
Text
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05-15-85 e
Rev. 8 WP0075P 0-0061P AP.28 POST TRIP TRANSIENT REPORT 1.0 PURPOSE
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.I' Pre-Startuo Actions To provide a systematic met' hod for diagnosing the cause(s) of a reactor trip, ascertaining the proper functioning of safety-related and other important equipment during the trip, determining any detrimental effect on plant equipment caused by the trip, and making the determination that the plant can be restarted safely.
.2 Subsecuent Actions (Usually Post-Startuo)
To provide for a detailed account of the trip, and to develop and adopt long-term corrective actions to be taken.
2.0 REFERENCES
.1 NO-001 Coordinated Commitment Log t
.2 Adapted from INPO Good Practice OP-211
.3 Engineering and Quality Control Guidelines, MEG.203 8*
.4 Technical Specifications 6.2.2.g. 6.5.1.6, 6.9.4.1. 6.10.2.f
.5 Administrative Procedures 1, 3, 17, 22, 39 i
.6 Plant Operations Manual. Procedure A.75 -- Plant Computer i
.7 Code of Federal Regulations,10 CFR 50.72 l
[
8*
.8 R. 'J. Rodriguer to J. F. Stoir (NRC):
letter of 1/21/85 (RJR 85-40) r 3.0 PROCEDURE l
8*
.1 Sequence / Summary H,0H.:
0 These major steps are performed in sequence.
Actions within major steps may be performed concurrently.
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0605160326 860306 PDR ADOCK 05000312 g
n
.(Continued)
PROCEDURE 3.1
.1 Plant trip or transient occurs.
.2 The Shif t Supervisor irittiates this procedure.
.3 Personnel statements gathered and machine generated data is A
collected.
c v m
.4 The STA reviews the assembled information for t.u..ipletiness.
1
.5 Trip witnesses'-shift may end now, at the earliest.
.6 The Shift Supervisor and STA perform a nuclear safety assessment.
8 ++
.7 Trip is " classified".
j
.8 The Plant Superintendent 'may allow restart here (at the earliest).
.9 Engineering and Quality Control Engineer writes a Trip Report.
i
.10 The Plant Review Cormittee discusses and amends the Trip Repo'rt and its recommended corrective actions, i
.11 The Plant Superintendent reviews and approves the amended report and its recommended corrective actions.
.2 Plant Trip or Transient 4
.1 For the purposes of this procedure, a plant trip is when *a generator, turbine, or reactor trip occurs that is not a part of a planned plant progression.
.2 At the discretion of the Shift Supervisor, this procedure may be initiated for non-trip transients.
I
.3 At the discretion of the Engineering and Quality Control Superintendent, this procedure may be entered for a non-trip transient.
If their shift had ended, the witnessing Operations crew has no requirement to complete their sections of this procedure.
.3 Initiation (Pre-Startup)
Notify the NRC in accordance with 10 CFR 50.72.
l.
84 If any of the Primary System Safety Valves, PSV-21506 PSV-21507, or PSV-21511. 11fted or failed to lift when required, an Occurrence Description Report must be initiated in accordance with AP.22.
(Reference Technical Specification 6.9.4.1.j) g 9
Rev. 8
}
AP.28-2 i
}......
l i
PROCEDURE (Continued) 3. 3' (Continued) once the plant is in a condition where the Shift Supervisor judges that data gathering will not impede the Operator's control of the plant, this procedure is initiated.
If plant conditions change, any licensed Operator may interrupt this procedures' performance temporarily.
If it'is the Engineering and Quality Control Superti.Undent's decisten to enter this procedure, see Step 3.2.3.
8+
4 Data Gathering (Pre-Startup)
.1 Attempt to record CR0 breaker positions and STAS initiation status T
before lhy reset.
(Enclosure 4.1)
)
.2 As simultaneously as possible for all Control Room recorders, time-mark the charts adjacent to the pen /printhead. A precise, narrow mark soon after the transient ends can aid later sequence-of-events resolution.
.3 For chart gathering, there are several acceptable methods.
Charts may be snipped, photocopied, then spliced to the original.
Or, charts may be snipped and the originals attached to the Enclosures.
If a photocopier is closeby, charts may be wholly removed, a portion copied, and the roll wholly replaced.
The control Room oper'ator can be consulted for preferences.
Those charts showing a single downward trend may be left o'n the recorder.
Label each collected chart with its source.
CAUTION:
i Bailey computer " Sequence of Events' and ' Memory Trip Review' functions, if cancelled, are lost.
Because the utility typer must be shared for real-time monitoring and trip data gathering, call up only one Bailey function at one time.
i j
.4 Retrieve computer printouts.
I
.5 Complete Enclosures 4.1 and 4.2.
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.6 Plant Personnel Statements After the plant is in a safe, stable condition, the Shift Supervisor shall ensure each individual involved in the trip (e.g.,
- reactor operators, maintenance technicians, etc.) provides a statement concerning his/her involvement in the reactor trip.
These statements may be obtained'in one of the following ways:
1) self-written statements;
- 2) interviews with personnel involved in the reactor trip
- 3) critique with all involved personnel.
Rev. 8 f
AP.28 3 i
-1
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, - L...
l s..
PROCEDURE '(Continued) 3.~4 6.
(Continued) t 1
Use Enclosure 4.3.
Read and initial statements of. nearby personnel so that, with your statement's assistance, an accurate and complete picture results.
Restrict stat.ertents to facts concerning the event; the facts should b.e state'd chronologically, if possible.
f, The written statements shall be included in the reactor trip data package to assist in the event reconstruction..
.5 Data Completeness Doublecheck (Pre-Startup)
.1 The STA will review the collected data.
i
.2 Any followup questions are best asked at this time, while the' events are fresh in everyones' mind.
Use ' Remarks
- areas for the i
documentation space.
i
.6 Time to Release Individuals From Shift I
The goal is to finish the completeness doublecheck before releasing individuals f, rom their shift.
Exceptions are:
.1 The Shift Supervisor determines that the individual will otherwise work beyond limits of Technical Specification 6.2.2.g.
l
.2 By exemption by the Operations Superintendent or higher management.
.7 Nuclear Safety Assessment (Pre-Startup)
.1 Analysis of Transient The Shif t Supervisor, or SA0 designate, and the duty STA will reconstruct the transient by completing Enclosure 4.4 using the collected data. A chronological description of the event will be
[
developed, using all available data.
Pertinent alarms, trips, i
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}
actuations, and isolations will be listed or marked on the sequence-of-events or alarm type printout.
Pertinent plant parameters should be t'ncorporated into the chronological list of.
i 1
events during the reconstruction.
l
.2 The trip and transient shall be. compared to the expected trip l
l response, based on the training and experience of the duty licensed operators.
CAUTION:
FSAR Iransients are ' worst case' or limiting conditions.
Do not assume that because a transient i
did not result in peak parameters exceeding the FSAR values that the plant response was' acceptable.
i i
Rev. 8 AP.28-4 C.
_ = - _
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j... &
7, PRCCEDURE (Continued) 3.7. 3 If found desirable by the Shift Supervisor, a comparison with previous trip reports or FSAR transients should be made.
.4 ' Analysis of Equipment 8ehavior
$]1,:
look t,vyond the obvious indications to dias use the cause 1
of the trip and evaluate the plant response.
Review the available information thoroughly.
Look for (1) abnormal indications or degraded trends in equipment performance, (2) events occurring out o'f the normal or anticipated sequence, (3) failed or degraded response of equipment to control signals. (4) unusual chemistry results or radiation readings, and (5) unanticipated alarms.
.1 Determine:
.1 If all major safety-related and other important equipment involved in the. trip operated as anticipated or expected, and
.2 If the trip / transient caused any detrimental effects on plant equipment, and 8+
.3 The most probable cause of the trip.
.8 Trip Classification
.1 8ased on the results of the analysis and evaluation of the plant.
trip and subsequent response, the Shift Supervisor or SR0 designate, and the Duty STA shall classify the event as one of the following conditions (ori Enclosure 4.5):
.1 Type !
i At the time of trip classification, by consensus of the Shift Management, (Shift Supervisor, Senior Control Room Operators, Shift Technical Advisor), the plant is in a condition to be safely restarted and operated with all the screening criteria in Enclosure 4.5 circled 'Yes.'
1
.2 Type !!
Knowledge or suspicion exists that some important equipment behaved abnormally or failed.
(Technical-specification.
required or as judged by the Shif t Supervisor as essential for i
safe and reliable operation.)
1 Rev. 8 AP.28-5 y
m s.
. PROCEDURE (Continued) a 3.8 i2 3, Management Notifications ll Once the reactor trip event is classified, the Shift Supervisor L
i shall inform the Plant Superintendent.
If the event is classified i'
8-
- as a Type II, the Shift Supervisor shall also inform the Operations h
Superintendent and the Duty STA shal.1 inform the Superintendent of s
i:
Engineering and Quality Control.
i.
.9 Restart Decision (Pre.Startup) a 8
.1 Type ! Event Sased upon this classification, the Shift Supervisor can recommend restart.of the reactor.
Use Enclosure 4.5 ' remarks'.to document y
?
pre-startup corrections..
.2 Type !! Ivent N
l Once classified as a Type'!!, no change in classification is i
allowed.
The Operations Superintendent and Engineering and Quality Control.
Superintendent will direct the further investigation of the trip to determine necessary corrective action before restart.
N ggt [,r Sources of expertise that should be considered include 8-nuclear steam supply vendors, vendor engineers, onsite I
engineering staff, corporate engineering staff, and
)
other experienced operations and maintenance personnel.
i The Operations Superintendent and Engineering and Quality Control a
l Superintendent will analyze the event reconstruction, emphasizing L
8*
the most probable cause of the trip and the resolution of abnormal p.
or degraded indications.
Use available expertise to resolve 8+
questions concerning the cause and plant response.
Su,pply and document on Enclosure 4.5 the following information to the Plant Superintendent:
9
.1 The actual or most probable cause of the trip; p
4 i
3 b
Rev. 8 AP.28 6 0
s,
, m.
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'~
PROCEDURE (Continued) 3 '. 9
.2
.2 The maintenance and testing necessary Lefore reacter restart including additional measures to verify the most probable 6
cause; g
.3 Additional monitoring or trending required during and/or after reactor restart; s
.j
.4 Necessary briefings to operations and/or maintenance personnel concerning specific equipment indications or possible '
malfunctions; and
.5 The conditions necessary for a reactor restart.
8 *+
.3 Plant Superintendent Evaluation and Startup Decision The Plant Superintendent shall evaluate the recommendation made by 8+*
the personnel performing the trip investigation.
For Type II Events, the Plant Superintendent should consider convening the PRC to review the trip investigation prior to reactor restart.
The Plant Superintendent shall ensure the following was done before allowing reactor restart:
.1 The most probable cause of the trip is known and corrected.'
.2 Major safety-related and other important equipment functioned properly during' the transient, or corrective maintenance and satisfactorf testing has been performed or will be completed when plant l conditions permit.
1 The plant, response during the event has been analyzed and the
.3 plant responded as anticipated, or abnormalities are i
understood and corrected as required by Technical Specifications except as described below.
e If the cause of the trip has not been positively identified, the l
Plant Superintendent shall determine if the cause and the j
circumstances surrounding the cause have been analyzed adequately.
4 Adequate measure must be implemented to prevent repetitive j
challenges to safety systems during future power operations.
.10 Trip Repor,t Writing (Usually Post-Startup) l The Engineering and Quality Control Superintendent appoints the Trip Report Writer.
Once the Operations Superintendent allows, the Writer 4
acqqires custody of all collected data:
either originals or xerographic
?
copies are a'cceptable.
The Trip Report Writer shall ensure that snipped sttip charts are later stored in the same boxes as the original rolls.
q
- 1 Rev. 8 AP.29-7
- i. -. -.. - -,. - -. -, _.. -... - -
U"
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PROCEDURE (Continued)-
3.10 A trip report number will tie obtained f' rom the Surveillance Scheduler or the Engineering and Quality. Control Secretary and. entered on the Trip Data Form.
The number issuer will-complete a Comitment Transferral Form and forward it to the Nuclear Operations Department Ccmmitment Coordinator for a three (3) wee,k. followup on the Coordinated Comitment Log.
The post-trip review data packEge will be reviewed to determine its significance to plant safety and reliability.
01 :ng the generation of the' Trip Report, the event will be evaluated to produce recommendations for corrective actions (e.g., procedure changes, design
- l modifications, operator and plant staff training).
'A comparison should be made with previous similar trips in order to identify abnormal or degraded conditions.
Engineering and Quality Control Guideline MEG.203 contains ir} formation on the Trip Report itself.
.11 PRC Review
'8*
.1 Type I Event The Trip Report shall be reviewed by the PRC.
This review is,not required prior to reactor restart.
l 8*
.2 Type II Event f
If directed by the Plant Superintendent, the PRC will review a Type
~
~II Event before a reactor restart is comenced.
In any case, a i
condition II Trip Report will receive a PRC review.
l l
.3 PRC amendments may be made on Enclosure 4.7 without retyping the report.
Justify amendments to recommended corrective actions.
.12 Plant Superintendent Review and Approval (Usually Post-Startup)
Submit the amended report to the Plant Superintendent for approval of I
the recomended corrective actions.
NOTE:
Quality Assurance tracks the implementation of the l
recomended corrective actions.
.13 Dissemination to the Industry (Time Frame Independent of Startup)
The Supervisor of Regulatory Compliance will determine what, if any, L
[
information on the trip will be useful to the industry, and is responsible for its dissemination via Nuclear Network, subject to Plant Superintendent approval.
t 6
0 i
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Rev. 8 AP.28-8 l
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PROCEDURE (Continued) f 3.14'Reten' tion The' trip report and data package shall be retained for the life of the plant.
(
Reference:
Technical Specification 6.10.2.f) 4.0 ENC 1.05URES
.1 Trip Data
.2 Secondary Safety and Dump Valves Temperature Stickers Data Sheet
.3 Plant Personnel Statements
.4 Analysis and Evaluations 8*
.5 Trip Classification a'nd Restart Appro, val
.6 Trip Report Amendment and Approval l.
[
h j
e t
"Rev. 8 AP.28-9 f.
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ENCLOSURE 4.1 s
f.
TRIP OATA p
I Trip No,'
8*
(E&QC use)
,w NOTE:
- e. u i
Of pr'ime importance are those' values er observations l
that are only obtainable from witnesses.
Operator's P
' statements.must be completed before.the end of'their shift except as waived by the Operations Superintendent or higher management or Technical Specification 6.2.2.g I
would be violated.
1 j
1.
Plant Conditions immediately Prior to Trio:
A/Ipwee 75?o : 7/o male. : Tam 582 :
fu~a ICS Auro, A?cs pre ru<e#=.7/Som%:
2 AhV's/&De 6Evro r
i a
i I
2.
Testino/ Maintenance or Contributino/Comolicatino Factors Prior to Transient:
8*
NONE65FOLLOWl(Circle)
SA/ ff b/7/PU7foC har Of
<$fdu/df ~ NouBLf)NOc7hM4 j
k Psoneess I
f O
1 Rev. 8 AP.28-10
/
[
/.
l4 ENCt.05URE 4.1 (Continued) f),
TRIP DATA
- 3. ~ Standby Eouioment Resconse/ Status:
CAUTION:
j
',1 Circle only one status per period.
I During
/ Ou' ring Pre-Trip At Post-Trip Transient Trip Transient How, Why, and Remarks
).
g HPI Pump A Onh On@
@0ff
(
hshsved dles 4 fors hoff
@/Off Onh ofs'cN%
i MU Pump u
i L
HPI Pump B Onh Onh hoff 8++
Open for minutes Ope h Ope h h Shut STV-23811 HPI Nozzle
\\
Open Q Ope Q Q Shut STV-23809 HPI Nozzle
'i STV-23812 f
'HPI Nozzle Open h Open h h Shut
+
t n
(
1
.STV-23810 Ope h Open h Q Shut HPI Nozzle l
Aux Feed Pump P-31'8 On@
On@
@0ff Aux Feed fj Pump P-319 Onh Onh
@0ff
~
I Rev. 8 AP.28-11 v
Y
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t_
e'
. ENCLOSURE 4.1 (Continued)'
TRIP DATA 3.
Stdndby Ecuiement Response / Status:
(Continued)
Ouring
..juring Pre-Trip At Post, Trip j
I Transient Trip Transient Mcw Why, and Remarks Secondary Safety Open Q Q Shut Open @
l Valves 5,
. r-Atmospheric
%us Mu.:
! XM
[
Dump Open @ h Shut Q Shut d //o6.hd h_
Valves 4re
//
9 1
3 Ud2+wn/A > Jss4Yhd Turb1ne Bypass n h @ Shut C0pefu/
M
+w /h i
8-Valves Manual Auto / Manual Auto nua
/ p r W &,4//s e g b/10 /ddbsbx k d Ed3 F6aK-e re
. MJ.
A-0TSG
% M W 4tdwanc.
In h h / Normal In morNd/A Bru //wh BTU Limits i
i
. k'dacfth */ /e/AixW e
B-0TSG BTU Limits InM h' Normal Inh erMAW A/BTE //-edr Big Boiler E-360 Run @
Run@
h 0ff i
Small Boiler l:
E-365 Run @
Run h
@ 0ff Throttle 3
Stops
@ Cl CNotCPCl Not C1/Cl i
a On H1SS Indication l
l
'8*
Governor CliapId' Slow /.
Valves No/ Movement Not Cl/Cl Not Cl/Cl j
\\
On HISS Indication j
1 Rev. 8 l
.1 AP.28-12
.j
-m
. -. -.. - - =
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ENCLOSURE 4.1 (Continued)
TRIP DATA 3..
Standby EcuiDment Res00nse/St5tus:
(Continued)
During.
s During Pre-Trip At Post-Trip Trans1ent Trip Transient Hnw. Why, and Remarks-l RH Stops Not Cl/Cl Not C1/Cl Not Cl/Cl Dn H1SS Indication i
RH Intercepts Not C1/Cl
'Not Cl/Cl Not C1/Cl l
On H1SS Indication 4.
Other Ecuipment Response:
Taken'to I
Manual Performance Anytime Acceptable Remarks l
Jeda WIC5 MML f%2.aMd
~atr&A 16 & n/dreafz. '
Rx - SG Master Yesh h/No Ide7a,Au # m A d2a.du a/I 1/ahmt 4 M4mha/ne m In4AN/Aff.
I i
h/No j
~
ATc controller N/A Rx Master Yesh
~
~
No Yesh hNo Diamond CR0 i
Loop "A FW Yesh h/No Demand
.l.
Y Rev. 8 AP.28-13 j
1 y.
L
/ '.
J
L 1
s
.ENCI.05URE4.1 (Continued)
TRIP DATA 1
I.
4.-
Other Ecuipment Response:
(Continued)
Taken to A
- Manual, Performance Anytime Acceptable Renarks i
See DAf2m JktoL
/
Loop
'B' FW h/No Demand Yes I.
Yesh h/No
'A' Feed Pump
?.
i.
i
'B' Feed Pump Yesh
/No 4
('
"A" Main Feed Yesh hNo Valve "B" Main Feed Valve Yesh hNo I
,l.
V P&urf m do fo 16td
- A* Startup 67sn '.s n + / m' &f Reus Yesh hNo
,_ M h /ne i
Valve V
h; 5" ' "" 8 35 !Mr^*
'B' Startup Valve Yes@
hNo f
f f
"A" Turbine j
Bypasses Yesh ho i
s h
Rev. 8 AP.28-14 s
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0 p,...
ENCLOSURE 4.1 (Continued)
TRIP OATA i
9 4.,
Other Eouiement Response:
(Continued)
Taken to Manual Performance Anytime Acceptable Renu rks See m z.o c.
/3 M J v dt.J'.s
'B' Turbine Bypasses Yes@p)
@ No
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8*
~
- /V mm m.inutes open i
'A' Atmospheric Yes h h No Dump Valves i
I 8*
^'/v m m minutes open I
"B" Atmospheric Dump Valves Yes h h No
?:,
1/
I J.'
8+
4 V2.
minutes open Secondary Code Safeties
'N/A h No Condenser h.
Vacuum N/A No
)
5.
Transient Reactor Coolant System Boundino Values:
l NOTE:
If initial value, signify with an asterisk (*).
j RCS Parameter Maximum Minimum 8*
RCS Temp (Max Th, Min.Tc) 003%
f RCS Loop A Pressure (PSIG) tose ~d.
j RCS Loop 8 Pressure (PSIG) loso5w rel Pressurizer Level (INCHES) 230 eMrea.l e l ou.-
Observed Subcooling Margin (*F)
/60 4 /-M RCS Average Cooldown Rate 'F/hr 8*
(Tc in 1/2 hour after trip) f dote k'. O k 3 MR b s55ccES pg.,, g $ se.le
\\ c a - y g, y,c,,,,[,,,
l' AP.28-15 s k o, vee ( u. n _ pesg u ae of
\\05075 c
i
f....
ENCLOSURE 4.1 (Continued)
I TRIP OATA
. 84 6.
SFh Initiation Information 4
If RCS pressure decrease below 1650 psig or RB pressure maximum is above i
2 ps)g, complete this section.
SFAS '
Analog HPI Trip Lamp LPI Trip Lamp
[
?
Channels f
Channel A (Tr~Iq'hD/ dim
[TrichWdim Griabdim Ch'armel 8 Cbrichf dim Channel C M dim 6 rich & dim i
9 r
SFAS 1
Digital Analog A
. Analog B Analog C Channel Trip Lamps.
Channels Trio.Lamo Trio Lamo Trio Lamo 18-HPI M dim (bEIq' Iib / dim (brichl/ dim (Drichrydim 28-l.P I CbrighT_Ydim (brighTydim f_brichTydim (brichTfd im
,69
~ ~ '
8+
[TrighBdim
('ETIcIit7 dim
$ richTydim fbrichT7 dim 2A-LPI M dim (brichD/ dim o t7 dim rio dim
+
'7.
Control Rod Breaker Assessment:
i If the Bailey Computer " Sequence of Events' or the alarm typer printouts
)
1 is irretrievable, explain why:
j f
Not ADolicable/As Follows (circle) 3
~~B A t tei ComPOTEG Owr OF S e cuc.r
%c.E V700 f
1
\\1-25-85 1
RPS Channel Trip Time (From Bailey Computer Group 4) EAiLEY OO S
~
Initial Trip Channel: Channel Parameter' Time 0/A j
j.
Second Trip Channel:
Channel Parameter Time M/A 0/A Fr.om Alarm ~Typer:
CR0 Trip Confrim (2004):
Time Rev. 8 AP.28-16 c.
G...
}
ENCLOSURE 4.1 (Continued)
TRIP OATA l
7..
Control Rod Breaker Assessment:
(Continued) i NOTE:
i Acceptance criterion is:
not more than 2.
seconds difference.
The Bailki Computer's design can allow an indicated two second difference for simultaneous events.
Control Rod Breaker Opening Time:
Subtract Time second Second Trip Channel from Trip Confirm 3AM.E3 banPOT6e?.Oc)$
8+
CR0 Breaker tiocal)
A-AC Breaker Indica.tes Trip Yes/No l
B-AC Breaker Indicates Trip Yes/No I
C-0C Breakers (CB1 and CB2) Both Indicate Trip Yes/No 0-DC Breakers (CB3 and CB4) Both Indicate Trip Yes/No Logic Cabinet Ammeters All Read Zero (Contactors E.F)
Yes/No
+
8.
Transient Secondary Plant Bounding Values:
84*
Lowest OTSG Startup Range Levels (Observed - not from charts or computer)
I i
OTSG A:
OTSG B:
l 8*
Lowest OTSG Pressures (Observed - not from charts or computer) l 0TSG A:
3 7 5 p si, -
Worst Condenser Vacuums Observed U"
W LP:
HP:
REMARKS:
4 4
i 9
i l
l Rev. 8
}
AP.28-17
-w (Continued)
ENCLOSURE 4.1 TRIP OATA 8.
LisE Ecuiement Damaced Duriric the Transient:
NAV60P vMP i
[
t.
9.
RE' MARKS: Work Request No.
completed to replace
[
temperature recording stickers on Secondary Safety Valves and t
Atmospheric Dump Valves.
t 4
t 10.
Recuired Computer Printouts:
Attached 1.
Bailey:
Alarm Summary
. -- Function 13, Group 1 Memory Trip Review
' -- Function 13, Group 3 Sequence of Events
-- Function 13, Group 4 Contact Status Sunrnary
-- Function 13, Group 5 g.MEY Bad Input Summary
-- Function 13, Group 7 g g 3sucG.
Deleted Point Sunrnary
-. Function 13, Group 8 j Deleted Monitor Summary
-- Function 13, Group 9' (Long Term Data Collection -- OP A.75 Attachment 4.3 2.
Com Rm PET:
Secondary Code and ADV Printout i
1.
i j
11.
Strip Charts Retrie:ved (C.ircle)
NOTE:
Retrieving all before startup is not required.
~
8*
Trip Report writer to retrieve the balance.
XR-00403.
Percent FP
' PR-21092 RCS Press WR LR-21503 Przr Ly1 XR-00205 Log N TR-21023 Te PR-20543 Ndr Press i
FR-21027 RC Flow FR-30119 Stm F1ms l
TR-21031 Th FR-20535 FW Flow A FR-20536 FW Flow 8 TR-21025 Tave
)
PR-21038 RCS Press B LR-20504 SG B Lv1 i
i XR-30504 Turbine 1
f, AP.28-18 Rev. 8 y
- w..
f,.
lh'
- s
ENCluSURE 4.2
.. SECONDARY SAFETY AND bump YALVES TEMPERATURE STICKERS DATA SHEdT I
("X out" Heat-tilackened_ dots)
ATMOSPHERIC DUMP VALVES PV-20571 PV-20562
'0000/
/0000/_ -
A
/0000/
/0000/
/
A 3
/0000/
/0000/
B
/0000/
/0000/
C
/0000/ '/0000/
C
/0000/
/0000/
SECONuARY SAFETIES C
B A
C B
A
/P 16u-250F 310-340F Setpoint PY-205/1
. PV-206o2 i
J-P1155m5 f
PSV-20545
-?-j Q000Ap
/CecG/
1050 psi
~~
g PSV-20547
/00u0/
/00u0/
1050 psi i
PSV-20549 --$d
/Cocc/
/6000/
1070 psi PSV-20551
/00u0/
/Ouuv/
1070 psi d
l
~ PSV-20553
-f
/000J,
/0u00/
1090 psi
-PSV-20533
-p-
/0000/
/u0001 1102 psi PSV-20555
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ENCLOSURE 4.3
)
PLANTPERSONNE[ STATEMENTS l
l Attach statements from personnel invol'ved with the trip concerning the events that preceded and followed the trip.
Indtvidual or group statements on how j
they remember the trip events are acceptable. _
?
Position Name I
Include the plant conditions prior to the trip, your indications that a f
problem existed, actions taken as a result of those indications, noted equipment malfunctions or inadequacies, and any identified procedure-deficiencies.
Also, incfude any inferination you consider important to review this unscheduled reactor trip and actions to prevent recurrence.
t l-(Read and initial those statements of the individuals that were near your location.
If there are conflicts, provide more detail of what you saw.. All
- observations of equipment malfunctions will be investigated.
Use additional sheets if necessary.)
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Rev. 8 AP.28-20
u ENCLOSURE 4.4 ANALYSIS AND EVALUATIONS
.73/d PROSABLE CAUSE OF TRIP M
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i ledan/er sAuks /% task doAAne i/ d m
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A, haf 'om i both %&a% m>,ws 4 s4w o%c.
AabeAk ec' 2dhafu mutkd m 2P.S 3n$ numa /nA v p Coments:
UNEXPECTED ASPECT OF TRANSIENT BEHAVIOR (if event compared with previous similar transient, note the transient with which compared)
Comoared With Previous trip on
/
Date Time FSAR Transient page number IDENTIFICATION OF SYSTEMS WITH INADEQUATE PERFORMANCE System /comoonent Description of Problem Rev. 8 AP.23-21
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i-ENCLOSURE 4.4 (Continued)
' ANALYSIS AND EVALUATIONS t
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a SRO Name
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Date Time 1
Signature
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q Duty STA Name Date Time Signature 1
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ENCLOSURE 4.5 8+
TRIP CLASSIFICATION AND.STARTUP APPROVAL j
TYPE I SCREENING CRITERIA.(Circle)
J i
l l
(a)
RCS Pressure Remained Above Setpoint Y'e t For Automatic SFAS Trip l
f (b)
RCS Pressure Remained.Below Setpoint' h
'No For PZR Code Safety Valve Actuation (c)
RCS Te.mp. Decreased Less than 100*F/hr Yes (Tech Spyc)
N (d)
Reactor Coolant Was Contained Within h
No The Primary RCS and PRT 8*+
(e)
Indicated PZR Level Remained On Scale
- Yes (f)
Indicated SG Level Remained Between Yes 18" indicated (S/U) and 93%
(Operate Range)
(g)
RPS Channels Did Not Fail t'o trip Properly @
No
~
(h)
On RPS Trip All CRD Breakers Tripped Yes No and the Reactor tripped wi. thin two gy Anime c.g. ~B4 w
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seconds of demand (Enclosure 4.1) gg gg
( 1)' Adequate Subcooling Margin Existed No' I
8+
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(j)
Post-trip, the Electrical Distribution System Functioned properly No If Pressurizer Code Safety (tes) passed subcooled water, take plant to depressurized cold shutdown (Reference 8).
Comments:
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Rev. S AP.28-23 3
84 ENCLOSURE -4'. 5 (Continued)
TRIP CLASSIFICATION AND STARTUP APPROVAL Classify ' trip as Type I or II according to definition in procedure.
The event on at is a type
.I.
II Date Time
/
Shift Supervisor
- Date, Time
/
STA Date Time Notification Plant Superintendent notified of event classification
/
Date Time 8*
Remarks:
l PERMISSION TC START UP i
Plant Superintendent notified and permission granted to start up the reactor.
/
I Shift Supervisor.
Date Time
/
i Outy STA Date Time e
i Rev. 8 f
AP.28-24
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0 ENCLOSURE 4.6 4
TRIP REPORT AMEN 0MENTS AND APPROVAL t
I.
Prepared By Date
'l Engineering / Quality control PRC Amendments and Remarks:
i 1.
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L 1
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PRC review of event on
, meeting number Minutes of the meeting (s) are attached
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PRC Chairman Date j
Plant Superintendent Remarks:
t I
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Approved Plant Superintendent Date
.END Rev. 8 AP.28-25 l
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION q
3 3
i 8
OFFICE OF PUBLIC AFFAIRS, REGION V e,
%,,,,,+
1450 Maria Lane, Suite 210, Walnut Creek, CA 94596 NRC:V-6585 FOR IMMEDIATE RELEASE
Contact:
Greg Cook (Mailed - Tuesday, December 31.19E Bus: 415/943-3809 Home: 707/644-2428
-l NRC UPGRADES INVESTIGATION OF INCIDENT AT RANCHO SECO NUCLEAR PLANT 1
The Nuclear R'egulatory Comission has upgraded an investigation of an incident at the Rancho Seco Nuclear Generating Plant, 25 miles southeast of Sacramento, California, which oc, curred on December 26. The facility is operated by the Sacramento Municipal Utility District.
Based on preliminary infomation obtained from its initial investigation of the incident, the NRC has upgraded the investigation and fomed an Incident -
l Investigation Team (IIT).
On December 27, an NRC Augmented Inspection Team of headquarters and j
regional personnel, reporting to the Regional Administrator in Walnut Creek, a
California, was sent to the site.
Initial results of their investigation indicate that this incident is complex and has potentially significant implications for other nuclear power plants.
j^
The IIT, comprised of HRC Headq' arters personnel, will report to NRC's u
Executive Director for Operations. Frederick Hebdon, Chief. Program Technology Branch Office for Analysis and Evaluation of Operational Data, will head the y
team.
It includes specialists in reactor syste.ns, reactor operations, human i
factors, and instrumentation and control systems.
The IIT is to find the facts of the incident, investigate the probable cause, and make appropriate findings and conclusions whic.h would form the basis for any necessary follow-on actions. A specific focus of the team will be on the design and the response of the Integrated Control System (ICS), an instrumentation and control system which assists in the plant operation at j
power. The team will also give special attention to operator performance and j
training as they related to the loss of the ICS during the event.
i The plant was operating at about 70 percent power when electrical power to the ICS was lost. Preliminary infonnation indicates that when the automatic system failed, operators had difficulty controlling the water supply to the plant's steam generators, leading to an excessively rapid cooldown of the reactor pressure vessel. Also, while stabilizing the plant, a pump was damaged, resulting in a spill of about 450 gallons of radioactive water within the plant's auxiliary building. Some of this water escaped the building as steam.
The release to the atmosphere was within regulatory limits and did not pose a health risk to plant workers or the public.
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NRC:V-6585 i On December 26, the Regional Administrator of NRC's Region V office at Walnut Creek, California, sent two confirmatory action letters to SMUD confinning the NRC's understanding that the District will hold in abeyance any work in progress and any work planned on equipment that was involved in the incident until the District and NRC have had an opportunity to evaluate i
the event. The letters also confinn that SMUD.will maintain the reactor in j
a shutdown condition until concurrence is received from the NRC to return to power.
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p aY Date:
12/30/85 I
PRELIMTNARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE--PNO-V-85-92 Thio preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public The information presented is as initially received withoat verificatio interest significance.
cr cvaluation and in basically all that is known by Region V staff on this date.
FACILITY: Sacranento Municipal Utility District Licensee Emergency Classification X Notification of Unusual Event Rancho Seco Nuclear Power Plant
-~
Alert Sacramento County, Califr,rnia Site Area Emergency l,
Docket No. 50-312 General Emergency Not Applicable
SUBJECT:
STATUS REPORT TROM AUGMENTED INVESTIGATION TEAM Tha Augmented Investigation Team (AIT) arrived onsite on December 28, 1985 to begin its investigation of the events surrounding the December 26, 1985 loss of electrical power to the Integrated Control System (ICS) - Reactor Trip and cooldown. The sequence of ev:nts was initiated by a loss of ICS power. This resulted in plant heatup and a The plant then experienced a rapid cooldown of cubsequent trip on RCS over-pressure.
i almost 200*F degrces in 26 minutes followed by plant stabilization at approximately 440*P and 714 psig. During the cooldown the plant reached a minimum pressure of 1047 psig ct 456*F and subsequently repressurized due to auto-actuation of HPI to 1413 psig at 386*F. Other significant highlights include possible emptying of the pressurizer, cntrance into the Interim Brittle Fractica region by 800 psig, potential overfill of j,
cua steam generator, failure of a makeup pump due to lack of water supply, and auxiliary fasdwater valve control problems. The full impact of these items on the transient is y:t to be determined.
An of 8:00 PM PST, December 29, 1985, the team had complaced all initial interviews.
I Th;se interviews included all of the operation staf f including management significantly
+
I involved in the event.
i This information is current as of 9:00 AM PST.
h CONTACT:
A. Chaffee (209) 748-2791 DISTRIBUTION H St.
MNBB Phillips E/W Willste Air Rights Mail:
Chairman Pallidino EDO NRR IE NMSS ADM:DMB Comn. Zach PA OIA RES DOT:Trans Only Comm. Bernthal MPA AEOD Comm. Roberts ELD Corcm. Asselstine Regions:
SECY INPO NSAC ACR$
Lice'nsee:
CA (Reactor Licensees)
REGION V: FORM 211 PDR Resident Inspector _
(Revised 3/14/83) 2DN i
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