ML20155J984
| ML20155J984 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/02/1986 |
| From: | Collins M FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20155J977 | List: |
| References | |
| AI-704, NUDOCS 8605270268 | |
| Download: ML20155J984 (28) | |
Text
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Rev. 0 04/24/86
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ADMINISTRATIVE INSTRUCTION AI-704 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 REACTOR TRIP REVIEW AND ANALYSIS 6?)
THIS PROCEDURE ADDRESSES NON-SAFETY RELATED COMPONENTS i
REVIEWED BY:
Plant Review Committee
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Date 04/24/86 Meeting No.
86-16 APPROVED BY: Nuclear Plant :'anager
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)Ye 8605270268 860521 PDR ADOCK 05000302 P
PDR Date 8. /#6
( ;4 77 INTERPRETATION CONTACT: Nuclear Safety Supervisor
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n TABLE OF CONTENTS
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Section Page 1.0 PURPOSE,
1
2.0 REFERENCES
1 3.0 RESPONSIBILITY 1
4.0 IMPLEMENTA1 ION 2
4.1 INTRODUCTION
2 4.2 POST-TRIP REVIEW 3
4.3 RESTART DECISION.
5 4.4 INDEPENDENT REVIEW.
6 4.5 SUBSEQUENT EVALUATION.
8 ENCLOSURES, Post-Trip Review and Restart Justification.
12
..........., UOER Distribution 25
............, Unplanned Operating Event / Transient Assessment Report Flow Diagram.
26
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a AI-704 Rev.f Page i
.s 1.0 PURPOSE g
hY5 The purpose of this procedure is to establish the administra-tive requirements following transient events for implementation of a Post-Trip Review program.
2.0 nrrrarutvc a.
- Post-Trip Review' and FPC Responses b.
" Documenting, Reporting, and Reviewing Non-Conforming Operations Reports" c.
CP-125, " Corrective Action Procedure" d.
Babcock & Wilcox ' Transient Assessment Program (TAP) Guide-r :,.
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- 3.0 nFcDONSIBTLTTY
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i The Nuclear Safety Supervisor is responsible for the content of 1
this procedure and shall act as the interpretation contact for any questions regarding its content.
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b rTh x N?bO AI-704 Rev. L 0 0 :
e 4.0 IMPLDIENTATION Na W
4.1 INTRODUCTION
4.1.1 This procedure establishes guidelines for a systematic method of conducting the technical review and analysis of plant performance associated with reactor trips in order to:
Determine the immediate, intermediate, and root causes of the trip.
Identify unexpected or abnormal response to the trip by plant systems, equipment, or personnel.
Assess the impact of identified abnormalities on' nuclear
- safety, equipment reliability, system performance, and plant availability.
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Develop corrective action recommendations to prevent the recurrence of the trip and to mitigate abnormal responses.
Satisfy reporting requirements.
Document observed plant response for future study and comparison.
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AI-704 Rev. L 0 0 Page 2
4.1.2 The. reactor trip review program consists of four distinct O
%,aC phases:
Post-Trip Review Restart Decision Independent Review Subsequent Evaluation All unplanned reactor trips will be subject to full review and evaluation.
Planned reactor trips with no identified abnormal-ities need not proceed to the subsequent evaluation phase unless it is deemed necessary by the Nuclear Safety Supervisor.
4.1.3 The Nuclear Plant Manager (NPM) or Man On Call (MOC) is the team leader in assessing and justifying reactor restart in accordance with Section 4.3.1 of this procedure.
,-g 4.2 POST-TRIP REVIEW 4.2.1 Immediately following plant stabiliration after a reactor trip, the Shift Operations Technical Advisor (SOTA) will complete Enclosure 1,
' Post-Trip Review and Restart Justification".
This data collection effort should be completed as rapidly as possible following the event but should. not interfere w.tth performance of the SOTA's duties if subsequent events occur or further complications develop.
1 h
AI-704 Rev.,
Page 3
4.2.1.1 If the Recall System is unavailable or malfunctioning, obtain 4U; photographs or photocopies of pertinent plant parameters.
Ensure that any such photographs or photocopies are clearly labeled with date, time, and tag number of the device, and are included in the restart package.
4.2.1.2 Following a reactor trip the Computer Alarm System printout switches to the line printer.
4.2.1.3 Any member of the Nuclear Safety Section may assist in the data collection effort as designated by the Nuclear Safety Super-visor.
Additional personnel may assist in the data collection or analysis effort as designated by the NpM/MOC or Shift Super-visor on Duty (FSOD).
4.2.1.4 All times noted on Enclosure 1 should be clearly labeled with the source of that time (Annunciator, Recall, Computer Alarm, Nuclear Operator or Shift Supervisor log, etc.).
This will assist in assembling a sequence of events for the transient.
E 4
O AI-704 Rev.pm00 Page 4 t
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4.2.1.5 When assembling the Sequence of Events, it is necessary to O\\
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correct the various data collection system times to a common time reference.
This time reference should be chosen on the basis of convenience (i.e.,
if most of the data is taken from the Annunciator printout, use the Annunciator system time as the reference; if most of the data is taken from the Computer Alarm system, use this time as the reference, etc.).
Once a reference system is selected and noted, a data point common to all data collection systems (such as CRD Trip Confirm / Reactor Trip) will be found.
Time corrections based on the ref erence time are determined and noted. The Sequence of Events can then be assembled into a single coherent chronology.
4.3 RESTART DECISION m.
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4.3.1 When Sections I through V of Enclosure 1 are complete, the SOTA and NPM/MOC will review and discuss the document.
This review should ensure that the root cause of the event has been deter-mined (if possible) and should ensure that adequate corrective actions have been proposed to prevent recurrence of the event.
' Plant transient response should be compared to expected responses to verify proper system performance.
In addition all identified performance anomalies should be assessed for impact on nuclear safety, equipment reliability, system performance, or plant availability.
AI-704 Rev. { Q Q Page 5
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4.3.2 The above review will be documented by completing Item A of m
~ '2 Section VI.
This section formally documents completion of the data review.
4.4 INDEPENDENT REVIEW e
4.4.1 Upon completion of Item A of Section VI, the Plant Review Committee (PRC) shall be convened to review the restart evalua-tion.
The PRC quorum shall include representatives from Licensing, Maintenance, Nuclear Safety and Reliability, Opera-tions, and Engineering.
The PRC Chairman will determine what additional representation from other departments is necessary.
4.4.2 The PRC shall review the restart evaluation with particular attention to Section IV, " Automatic System Challenges and Plant Response".
The committee shall also review the corrective actions proposed for any significant equipment malfunctions.
Primary consideration must be given to determining what troubleshooting and/or repairs need to be completed prior to restart.
4.4.3 The PRC Chairman will indicate the committee's recommendation 1
'to restart the reactor by signing Item B of Section VI.
Any 3
recommendations of the PRC should be attached to that enclosure I
for consideration by the NPM/MOC.
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7 AI-704 Rev.( Q O Page 6
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t 4.4.4 Upon completion of the Independent Review by the PRC, the m
kh NPM/MOC shall review all sections of Enclosure 1 and any PRC recommendations attached.
When the NPM/MOC is satisfied that all corrective actions required prior to restart are complete and that no outstanding safety concerns remain, he shall sign Ites C of Section VI.
This signature authorizes restart of the reactor. Enclosure 1 will then be given to the SSOD for review and signature prior to restart of the reactor.
The restart package, consisting of the completed Enclosure 1 of this proce-dure and all supporting documentation, should be forwarded to the Nuclear Safety Supervisor for further evaluation.
NOTE:
In the interim between trip and approval for recovery, the Nuclear Shift Supervisor may authorize the with-
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drawal of Safety Group 1 provided a 11 delta-k/k shut-down margin is maintained and rod withdrawal is not prohibited by any RPS ' Action Statements' of Standard Technical Specifications.
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AI-704 Rev. L00 page 7
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4.5 dUBSEQUENT EVALUATION m
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m' 4.5.1 Overview 4.5.1.1
~ 7.stther evaluation of reactor trip events will be conducted to satisfy reporting requireisents, to inform other utilities of our experience and lessons learned from the event, and to
. document observed plant response for future study and compari-son.
4.5.1.2
' For reactor trips, a site visit by -Babcoes and Wilcox (B&W)
^
personnel should be requested under the Transient Assessment Program. This visit, arranged by the-Nuclear Safety Supervisor through the B&W Resident Engineer, will assist Nuclear Saf ety Group personnel in investigating the event and in preparing an r.,
initial written assessment of the event.
The site visit, when
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requested, should hecirwithin twenty-four hours of the event or on the first normal working dai following the event if it occurs on a weekend or holiday.
I 4.5.1.3 The following reports are prepared in'the course of subsequent evaluation of a reactor trip:
[-
Nuclear Network Entry Licensee Event Report Unplanned 05.ezating Event Report Transient Assessment Progran Report Preparation of each report is described below.
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AI-704 Rev. L 0 0 Page 8
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4.5.2 ruclear Network Entry
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A Nuclear Network Entry should be made covering the reactor trip event. The entry should be made within one working day of the event and may be prepared by any member of the Nuclear Safety Group.
The entry should be a brief summary of the
- event, including contributing unusual pre-trip
- lineups, cause(s) of the event if known, and significant post-trip abnormalities.
The entry should be placed on Nuclear Network in the Operating Plant Experiences (OE) topic if the event is of interest to the entire nuclear community or in the B&W Owners Group (BW) topic if the event is of interest only to-other B&W designed plants. Any entry concerning the event must have Nuclear plant Manager or Man On Call approval prior to placing it on the Nuclear Network.
4.5.3 Licennee Event Report (LER) t As required by 10CFR50.73, any unplanned actuation of the Reactor Protection System (i.e., reactor trip) must be repor ted in writing to the Nuclear Regulatory Commission within thirty l
days of the event.
This report is prepared in accordance with CP-111, ' Documenting, Reporting, and Reviewing Non-Conforming l
Operations Reports", and NUREG-1022,
- Licensee Event Report System", and its supplements.
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Rev.l 0 0 i AI-704 Page 9
o 4.5.4 Unnlanned Operatino Event Report (DOER) ru 4.5.4.1 Upon receipt of the restart package for a reactor trip event, the Nuclear Safety Supervisor will assign a UOER number by year and numerical sequence (e.g., DOER 85-3 for the third event of 1985).
He will then assign responsibility for preparation of the UOER to a member of the Nuclear Safety Group and deliver the package to that person.
4.5.4.2 The Nuclear Safety Group member assigned responsibility for preparation of the UOER will fully investigate the event.
If deemed necessary, Corrective Action Assignments (CAAs) may be made in accordance with CP-125, ' Corrective Action Procedure".
In order to preclude duplication, all CAAs assigned should carry the designation of the Non-Conforming Operations Report
..rq (NCOR) associated with the event rather than the LER or UOER designation.
When invertigation is complete, the responsible member will prepare the UOER using guidance from the Babcock &
Wilcox ' Transient Assessment Program (TAP) Guidelines".
4.5.4.3 When the UOER is complete, it will be reviewed and approved by the Nuclear Safety Supervisor, Nuclear Safety and Reliability Superintendent, Nuclear Plant Technical Support Manager, and Nuclear Plant Manager.
4.5.4.4 Upon approval by the Nuclear Plant Manager, the UOER will be distributed per Enclosure 2, 'UOER Distribution".
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AI-704 Rev.E O O Page 10
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4.5.4.5 The DOER preparation process is summarized in Enclosure 3,
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" Unplanned Operating Event / Transient Assessment Report Flow Diagram *.
a 4.5.5 Tran=ient Assessment Procram Report (TAP)
Following approval of a UOER by the Nuclear Plant Manager, it may be forwarded to the Babcock & Wilcox Resident Engineer for inclusion in the Babcock and Wilcox Owners Group Transient Assessment Program.
I ENCLOSURES Post-Trip Review and Restart Justification
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UOER Distribution Unplanned Operating Event / Transient Assessment Report Flow Diagram i
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AI-704 Rev. i 0 0 Page 11
ENCLOSURE 1 O
(Page 1 of 13)
POST-TRIP REVIEW AND RESTART JUSTIFICATION Shutdown Report Number:
(From AI-500 Enclosure 11) 3 Trip Date:
Trip Time:
SSOD During Trip:
NO During Trip:
SOTA During Trip:
ANO During Trip:
MOC During Trip:
I.
DATA COttFCTION A.
Gather the following information as appropriate or available:
Annunciator Events Logs:
Shift Relief Checklist:
I Computer Alaras NSS NSS Recall Tapes NO NO Post-Trip Review Other Other Shutdown Report OOS/ Links RC Inventory Tracking:
i STI (if applicable)
Clearance Level Operator Interviews Voids a
j B.
Determine which actuation (s) occurred:
3 1.
EEE First, each actuation / trip must be identified and CRD trip response determined.
Each RPS cabinet should be checked for 1) all RPS actuations, and 2) CRD Breaker Open light which indicates proper CRD breaker and elec-tronic trip response.
Second, the actuation times must be determined.
The Main Control Board will indicate
" first out".
All data sources should be examined for times in the following priority:
A(nnunicator),
R(ecall), C(omputer).
Indicate the data source used for time in the Time column.
4 Data Sources:
[
Annunciator Printout-(A) Computer Alarms Printout (C)
Recall Tapes (R)
RPS Caninet (RC)
Trip Channel A Channel B Channel C Channel D Parameter YN Time YN Time YN Time YN Time 1
RCPPM
(/6f/ Flow Hi Press RB Bighp Hi Press RCS Lo Press RCS Var Lo Press j
High T ot h
ART MFWP ART Turbine Manual CRD Bkr Open a
AI-704 Rev. L 0 0 J Page 12
s 9 y
ENCLOSURE 1
'N (Page 2 of 13)
D 2.
II As above for RPS, all ES actuations/ trips and times must be determined using the available data sources.
Data sources:
Annunciator Printout (A) Computer Alarms Printout (C)
Recall Tapes (R)
ES Panel / Cabinets (E)
Train A Train B Channel YN Time YN Time HPI RC1 RC2 RC3
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LPI RC4 RCS RC6 RBIC RB1 RB2 RB3 BS RB4 RB5
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RB6 3.
Other ESFAS As above, examine the available data sources to determine if and when any of the following systems actuated:
ESFAS YN Time Comments Rad Mon EFIC EW EFIC MSLI EFIC MRI i
Other As above, examine available data sources to determine if and when any of the following systems actuated:
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Main Turb Trio YN Time Comments Low Vacuum a.
Low Cont Oil
[-
Thrust Brg
- P Solenoid Manual k
ICS Runback k
RCS Flow MWP's Trip
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MWBP's Trip i
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Assyn Rods r;
Revl. 0 0 f AI-704 Page 33
ENCLOSURE 1
]
(Page 3 of 13) f,*)
II.
PRE-TRIP / EVENT REVIEW A.
Major plant parameter / component status:
Mode Main Turbine-Generator Rx Power MWth Mode MAN /AUT0/ICS AUTO RCP's on ABCD Output Breakers OPEN/ CLOSED MFWP's on AB Generated MW MWe MUP on ABC B.
ICS and other control station status:
H = Hand A = Automatic
.g.
.B' H A MFW H A H A Pressurizer H A ULD MFWP Level S/G Rx Master _ _ MBV Spray Rx Demand LLBV Heater Rx Diamond LLCV Delta T-Cold SUBV PORV Open Clos Auto TBV SUCV PORV Block Open Clos ADV Spray Control Open Clos Auto Spray Block Open Clos m
J, C.
Maintenance or testing in progress:
D.
Equipment Availability:
A = Available D = Degraded N = Not Available 1.
Safety-Related equipment / component degraded or out of service:
[
Elec Pwr Other 2.
Important non-safety related equipment / component degraded /
00S:
A D N Comments
'a ICS fe Turbine t
other AI-704 Rev.
00T Page 14
ENCLOSURE 1 O
(page 4 of 13)
$5 III. CAUSE DETERMINATION A.
SEOUENCE OF EVENTS 1.
Time correction Using an event known to be common to the Annunciator, Recall, and Computer Alarm systems, determine the relationship between Annunciator time, Recall time, and Computer Alarm time. Record these time relationships below (select the most convenient of the three systems and ref erence the other two to that system).
+/- min:see Annunciator time ~:
Recall time:
Computer Alarm time:
2.
Secuence of Events In the spaces below, assemble a sequence of events for the major pre-trip and post-trip occurrences.
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Data Corrected Time Source Time Event T
AI-704 Rev.L00 Page 15
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ENCLOSURE 1 (Page 5 of 13)
.s 2.
seauence of Events (continued):
ti Data Corrected e!
t Time Source Time M
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ENCLOSURE 1 6]
(Page 6 of 13)
B.
POTENTIAL CAUSES In the spaces below, list all potential causes and/or contrib-uting factors leading to the reactor trip.
When all potential causes are listed, determine what testing needs to be performed to prove or disprove each cause or contributing factor. From the results of this testing, determine the degree of involvement of each cause (i.e.,
whether each cause/ contributing factor is a root cause, intermediate cause, immediate cause, or not involved in the trip). Finally, list any corrective actions deemed neces-sary to preclude similar events in the future.
Root Cause = (R)
Intermediate Cause = (IN)
Immediate cause = (IM)
Not Involved = (N)
Cause or Test of Degree Necessary Contributine Factor Possible cause Involved Corrective Action 1.
2.
3.
4.
5.
6.
Root Cause = underlying condition or factor which when corrected minimizes
^
the probability of recurrence of the event or similar events.
Intermediate cause = condition or factor which led from a root cause to an immediate cause.
j Immediate Cause = condition or factor which directly led to the event.
IC NOTE: Refer to NUREG-1022 Supplement 2, Pages 8 through 11. for examples of the above definitions.
AI-704 Rev.L.00 Page 17
t ENCLOSURE 1
t (Page 7 of 13)
L'E IV.
AUTOMATIC SYSTEM CHALLFNGES AND ptANT RESPONSE A.
SAFETY SYSTEM CHALLENCES AND RESPONSE Using the data sources available, determine the maximum values, minimum values, and minimum margins that occurred throughout tne event.
Compare these values to the actuation setpoints, deter-mine if the safety systems were challenged, and using the prior recorded actuations determine if the expected response was achieved.
Value Recall Pts Setpoints RPS Chall-Expected Parameter Max Min Marc (Note 1)
ES EFIC enged Response Y N Y N RCPPM 129-132 2 Off p/dp/ Flow (Note 2) 0-3,31,32 STS Figure 2.2.1 _ _
58-61 Hi Press RB 82,83 4 psig High Flux 0-3 104.9*5 or 79.92*5 __
High Press 4-6 2300 psig Low Press 4-6 1800 1500/500 Var Lo Pres (Note 3) 6,14,15 (11.59 x Thot)
- 5037.8 psig High That 14,15 618 F 7
ART MFWP 2 Off at >20*5
,i ART Turbine 170 Trip at >205 OTSG Press 104, 105 600 psig OTEG Level 90,91 6 inches NOTE 1: The following Recall points can be put into User-Defined Groups O for ease in data collection:
Points 0,1,2,3,4,5,6, 14, 15, and 170.
L Points 82, 83, 90, 91, 104, 105, 129, 130, 131, and 132.
O Points 31, 32, 58, 59, 60, and 61 (if 0F 0/ Flow plot is needed).
NOTE 2:
If no loss of RCS flow occurs, this parameter may be checked by selecting the highest power and the largest (in absolute value) f imbalance and comparing this point to STS Figure 2.2.1.
If this point of maximum power and maximum imbalance lies within the region for acceptable operation, the margin may be noted as I..
' Satisfactory *.
If the point of maximum power and maximum o
imbalance lies outside the region for acceptable operation, then a plot of power versus imbalance is required.
If a loss of RCS flow occurs without a O/.* 0/ Flow trip, then a plot of power, RCS ilow, and imbalance is required.
NOTE 3: This parameter may be checked by visual inspection of the SPDS 7.
Post-Trip Display.
If the trace did not cross the variable low pressure trip line, the margin may be noted as " Satisfactory *.
If the trace crossed the variable low pressure trip line, or if the 5
SPDS Post-Trip Display was not available for visual inspection, D,
then a plot of variable low pressure is required.
AI-704 Rev. L00f' Page 18
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,a ENCLOSURE 1
N (Page 8 of 13)
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B.
PLANT RESPONSE 1.
Reactivity Control s
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k RPS - Determined above in IV.A.:
SAT UNSAT L
b.
Boration:
i Required for Used Flow Length Reactivity Control j
Source I__H Rate Of Time Amount Y
N BWST CBAST c.
15 Shutdown Margin:
Was it shutdown margin achieved and maintained? YES NO If "NO*,
explain below.
e 2.
Thermal Control a.
Core Heat Removal Mode:
Method Y N Comments OTSG with RCPs __ __
OTSG w/o RCPs HPI Flow Rate Amount Temp V
LPI Flow Rate Anount Temp is
[
Core Flood Amount
's kh b
% rl
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AI-704 Rev.,8"" ~0 ()
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Page 19 c
ENCLOSURE 1 y
(Page 9 of 13) b.
Major parameters:
Limit or Parameter Ein Bag Setooint comments RCS Pressure 2500 1500 PORV Opened YN 500 psig Safties Opened Y N RCS Temp RCS Couldown 100 F/ Hour PTS Concerns if if Teold Tcold ( 500F
< 500 F Pzr Level 2nd MUP or Used:
HPI MUV-23 24 25 26 MUP-1A 1B IC RCS P/T Subcooled 50/20 F Post-Trip Per SPDS Display OTSG Pressure "A"
1010 600 Open ADV Y N
'T? %,
Rescat ADV Y N 1050 psig open MSV 33 34 Rescat MSV 33 34 1070 psig Open MSV 37 38 Rescat MSV 37 38 1090 psig open MSV 42 43 Rescat MSV 42 43 1100 psig Open MSV 40 46 Rescat MSV 40 46
- B' 1010 600 Open ADV Y N Reseat ADV Y N 1050 psig Open MSV 35 36 Rescat MSV 35 36 1070 psig open MSV 39 41 Rescat MSV 39 41 1090 psig open MSV 44 45 Rescat MSV 44 45 1100 psig Open MSV 47 48 Rescat MSV 47 48 OTSG Level
'A' 35' 98%
EFIC EFW Y N Flow: A1 gpm A2 gpa
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- B' 35' 98%
B1 gpa B2 gpa AI-704 Rev. - () 0 j page 20
2:
ENCLOSURE 1 (Page 10 of 13)
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3, Radioactive Inventory Control Alarm Systen W
Comments Main RM-A12 f-Steam RM-G's 1
Reactor RM-A1 Bidg RM-A6 RM-G's Sump RCS RM-L1 RCDT i
Aux RM-A2 Bldg RM-A3 RM-A4 RM-A11 5
RM-L2 MWST r
Control RM-A5 Complex RM-A14 RM-G1 l-
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AI-704 Rev.! 0 0 page 21 l-
ENCLOSURE 1 m
y'.h; (Page 11 of 13) 3.
Eauiement Availability List any equipment that did not perform as expected.
Include any Work Request or NCOR numbers in the Corrective Action column.
Place a check in the appropriate column to indicate whether completion of the Corrective Action is required before restart or whether the Corrective Action may be completed af ter restart.
Any Corrective Actions marked as required before restart must be completed before restart is a
authorized.
Action Taken Proposed Before/After Number Malfunction Corrective Action Restart
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P AI-704 gey L 0 0 f vage 22
ENCLOSURE 1 C7)
(Page 12 of 13)
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V.
Suf91ARY In the space below provide a brief narrative summary of the event.
Include as a minisua all significant abnormal pre-trip lineups, I
discussion of the cause or causes of the trip, and significant post-l trip abnormalities.
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AI-704 Rev. g 0 0 i Page 23 f
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ENCLOSURE 1
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?I (Page 13 of 13)
VI.
RMTART REVIEW AND APPROVAL A.
Post-Trip Review completed:
SOTA Date B.
Restart recommended:
4 PRC Chairman Dste Meeting C.
Restart authorized:
NPM/MOC Date
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D.
Reviewed prior to restart:
SSOD Date t..
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AI-704 Rev.> 0 0 J Page 24
ENCLOSURE 2
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k' UOER DISTRIBUTION o
Vice President, Nuclear Operations Director, Nuclear Operations Engineering & Licensing o
o Nuclear Plant Manager o
Nuclear Operations Manager o
Nuclear Operations Superintendent o
Nuclear Maintenance Superintendent Nuclear Chemistry / Radiation protection Superintendent.
o o
Nuclear Plant Technical Support Manager o
Nuclear Engineering Superintendent o
Nuclear Safety & Reliability Superintendent o
Chairman, Nuclear General Review Committee
't o
Nuclear Operations Training Manager o
Plant Quality Files o
Manager Site Nuclear Licensir.g o
Manager Nuclear Licensing 7
AI-704 Rev. L 0 0 '
Page 25
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8 ENCLOSURE 3 m
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UNPLANNED OPERATING EVENT / TRANSIENT ASSESSMENT REPORT FLOW DIAGRAM Shift Operations Gathers Data i
Technical Using 4
Advisor
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Nuclear Safety Logs Report, Reviews Distributes Supervisor Assigns and UOER Responsibility, Approves Using Requests Assistance UOER
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a Nuclear Safety Investigates Group Member and Prepares UOER l
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Nuclear Safety Reviews and and Reliability Approves Superintendent UOER N/
Nuclear Plant Reviews and Technical Support Approves i5 Manager UOER I:
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g Nuclear Plant Reviews and D
Manager Approves 3
UOER r,
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-1 AI-704 Rev.[ Q Q Page 26 (LAST PAGE)
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