ML20134D444

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Rev 1 to Procedure 1.3.37, Post-Trip Reviews
ML20134D444
Person / Time
Site: Pilgrim
Issue date: 08/24/1984
From:
BOSTON EDISON CO.
To:
Shared Package
ML20134D433 List:
References
1.3.37, NUDOCS 8508190222
Download: ML20134D444 (17)
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NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure 1.3.37 POST TRIP REVIEWS List of Effective Paaes 1.3.37-1 1.3.37-2 1.3.37-3 1.3.37-4 1.3.37-5 1.3.37-6 1.3.37-7 4

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List of Attachments" 1.3.37A-1 1.3.378-1 Approved w~

1.3. 7B-3 QA Manager 1.3.37B-4 1.3.378-5 1 3 37C-1 1 2 Approved -~ OR

// frman 1.3.37E-1 1.3.37F-1 ,1f/fW Date

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8508190222 DR 850813

( ADOCK 05000293 PDR 1.3.37-1 Rev. 1

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. I. PURPOSE

[h yg To provide instructi::ns to perstnnel to perform Prst TR1Y'R following unplanned reactor trips or unexplained power reductiG t

Adherence to this procedure will ensure that consistent data wU collected, and that a uniform analysis and decision process will be

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ap$ lied after a reactor trip or tmexplained power reductidt and prior to granting permission to restart. ,

II' . DISCUSSION -

The analysis of the Salem Nuclear Station event of 1983 revealed that an unrecognized failure to scram (ATWS) event took place. Evaluations of the event by both the NRC and INPD have resulted in recommendations aimed at standardizing post trip reviews. BECo consnitments to both INPO and NRC are directed to proceduralizing and standardizing post trip reviews. This procedure will formalize the existing post trip review method at PNPS.

The purpose of a post trip review is to determine the plant's readiness

  • to return to power af ter an unscheduled reactor trip or unexplained power reduction. Station personnel must reasonably determine the cause of the trip, verify proper functioning of safety related and other I equipment during the trip, and ensure that the trip / reduction did not have a detrimental effect on the plant.

Post trip reviews can also serve to provide lessons learned to the plant staff and other utilities.

III. REFERENCES A. INPO Good Practice OP-211 " Post Trip Reviews" Draft.

B. US NRC Generic Letter 83-28 " Generic Implications of Salem ATWS Events" C. BEco letter " Response to Generic Letter 83-28" O D. Nuclear Operations Procedure.NOP 8301 " Conduct of Operations" E. PNPS Operations Manual procedures:

1. 1.3.3 " Authority to Shutdown and Startup Station"
2. 1.3.9 " Reports" 3, 1.3.12 " Notification and Recall of Personnel"
4. 2.2.17 " Communications"

- 1.3.37-2 Rev. 1

IV. APPLICAB1GTY, l7 g f O MWh I dilUH Th2 r:;quirements cf this pr:ctdure will apply to all and unexplained power reduction from the "RUN" mode. The Chi V

'- Operations Engineer or the Nuclear Operations Manager may req lat elements of this procedure be followed for other unit problems from f' other operating conditions. g V. PREREQUISITES

A. The post trip review (PTR). Will be initiated af ter plant conditions

.are stabilized. The PTR shall not distract the Watch Engineer, the Operating Supervisor, the STA or operating personnel from their

- primary responsibility of monitoring plant parameters and main-taining the plant in a safe condition.

B. Sufficient information shall be collected from personnel involved in the unit trip prior to permitting relief by the oncoming shift.

C. A Scram Report shall be prepared by the Watch Engineer on duty at  :

the time of the scram. (Scram Report is Att. A of this proc.). A scram report shall be prepared whenever the control rods are scramed in from a situation where fuel is in the reactor vessel, 1

and more than one control rod is withdrawn. A Scram Report is required for all conditions, planned or unplanned. Distribution of Scram Reports shall be per section VII.F of this procedure.

VI. RESPONSIBILITIES A. Nuclear Watch Engineer is responsible for ensuring that the post trip review is initiated. He is also responsible, along with the NOS and STA, for the investigation phase of the PTR and review of the results. '

B. Nuclear Operating Supervisor is responsible as part of the investi-gation phase to record the actions taken and preliminary informa-tion relating to the initiating event. He is also responsible for directing the Shift Administrative Assistant in obtaining state-4 ments from operating personnel and others involved in the trip.

C. ShiftTechnicalAdvisorIsresponsiblewiththeNWEandNOSfor trip investigation. This includes interpretation of the Process Computer Data Recall Log and Sequence of Event Log. Additionally

' the STA will record the data required on Attachment 8.

D. Shift Administrative Assistant is responsible for collecting the required recorder charts and obtaining statements of involved per-sonnel.. Additionally the SAA is responsible to ensure that the PTR package is retained for records and distributed to the Nuclear Operations Manager and the OSS&P Group Leader.

E. Operations Personnel and others (i.e., I&C Technicians) involved in the unplanned trip are responsible for providing the SAA with ob-jective statements that describe their observations of and/or par-ticipation in the trip.

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1.3.37-3 Rev. 1

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e F. The Nuclear Operatiens Manag2r cr his designate is TinM auth rize r; start cf the r:act:r. llUll

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G. The ORC shall review all post trip reviews. In the case o classified as Type 1 or 2, the ORC shall review the trip a

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, next scheduled meeting. If the trip is classified as Type 3, the

( NOM will convene ORC to provide independent assessment 0cf the 1 event, prior to authorizing restart. _,

VII. POST TRIP REVIEW PROCEDURE -[

..-NOTE: Event notification to appropriate agencies or persons shall be made consistent with procedure 2.2.17 " Communications".

A. General Post trip review is a 5 step process as follows:

1. Data Collection
2. Trip Investigation
3. Event Classification / Safety Assessment l
4. Restart Authorization 1
5. Information Feedback B. Data Collection
1. The object of the data collection is to assemble enough infor-( mation to reconstruct the trip / reduction, assess the response of systems, and identify the root cause of the event.

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2. The SAA shall collect the following hard copy data:

- a. Process Computer Data Recall Log (Attachment F identifies points on this Log) (not applicable to unexplained g

C reduction) ,

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b. Process Computer Sequence of Events Log and Alarm Typer Output
c. Recorder charts from:
1. One APRM recorder
11. Feedwater flow iii. Wide (or) narrow range reactor pressure iv. Reactor level
v. Other recorders as identified by the STA or NWE
3. When recorder charts are collected, photo copies of the ori-ginal strip may be made. Chart speed, reference time, date, scale values and pen channels will be recorded on the chart or copy.

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Rev. 1 1.3.37-4 v ~

4. The SAA in ccnjuncticn with cr under the dir

, shall rcccrd statements f rom each individual cy:nt. These statements shall be obtained o -

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- . olant is in a stable condition. Statements sh:uld i

  • facts concerning the event relative to pretrip condi r activities, initial indications of a problem, initial actions

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  • taken, equipment automatic and manual operation,a9bserved malfunctions or procedural deficiencies. .
5. It may be appropriate for the NWE to interview invfved per-
sonnel or to collect information from assembled indhviduals as a group.
6. The STA shall complete part 1 of the PTR DATA Summary (Attachment B). If additional information 1s needed it shall be obtained at this time.
7. Collected data and personnel statements shall be assembled into a package and delivered to the NWE to begin the trip investigation. , ;

C. Trip Investigation (Attachment C)

1. The NWE, NOS, and STA shall reconstruct the event by preparing a chronology of the event. (Attachment C, part 1)
2. The NWE and NOS shall review the data package for proper sys-tem performance and note that appropriate automatic functions and equipment operation took place. The reviewers should look beyond the obvious indications to diagnose the cause of the trip and determine acceptability to restart the unit.
3. The STA shall analyze the data to determine if critical para-meters remained within the bounds of the FSAR or the Cycle Reload Analyses. Peak reactor pressure, icwest water level, drywell pressure, steamline radiation are examples of what shall be recorded for this analysis. Departures from the bounds of the FSAR or reload analysis shall be noted on part 3 9 of Attachment C and brought to the imediate attention of the NWE. A potential unreviewed safety question may exist.
4. The NWE will complete the Scram Report (Attachment A) and sum-marize the event. The scram report shall iden ify the prob-able cause of the event, and identification of systems with inadequate performance (if applicable). This Scram Report shall become the cover sheet for the PTR package. The Scram Report is not used for reviews of unexplained power reductions.

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5. Any equipment or processes identified with inadequate per-formance or abnormal response shall be reported on separate Fa'1ure and Malfunction Reports according to procedure 1.3.24.

1.3.37-5 Rev. 1

D. Ev;nt Classification /S fety Assessment

1. Th2 prelininary Safety Assessment (Attachment D) fcrm labe l

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'. completed by the STA and Watch Engineer. l The Event shall be classified as Type 1, 2, or 3 by the fol- I

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2.

lowing criteria:

  • I NOTE: Classifications are also applicable to unexplained i power reductions. 1

..- a. Type 1 - The cause of the trip is positively known and has been or is in the process of being, corrected; all safety related and other important equipment functioned properly during the trip.

b. Type 2 - The cause of the trip is positively known and has been or is in the process of being, corrected except some safety related or other important equipment did not func-tion properly. The malfunction has been or is in the pro-  :

cess of correction or a Tech. Spec. Constraint does not prohibit startup.

c. Type 3 - The cause of the trip is not positively known and/or some safety related or important equipment func-tioned abnormally during the trip, or the malfunction cannot be readily corrected, or startup is precluded due to Tech. Spec. Constraints, or the transient did not remain within the bounds of the FSAR or Reload Analysis.
3. The NOM and the COE will be notified of the classification of

(' the event and their concurrence shall be obtained. If the event is classified as Type 3, the COE or his designate will take charge of the investigation until the cause of the event and corrective action has been determined.

E. Restart Authorization 0 1. The NOM will be informed of the results of the PTR and classi-fication. The NOM miy then authorize restart if the event is classified as Type 1 or 2.

2. If the NOM is not satisfied with the results of the PTR he will take actions necessary to satisfy his concerns.
3. In the case of a Type'3 event the NOM will convene the ORC for further evaluation and independent review of the event prior to authorizing restart.
4. In the case of Type 1 or 2 events the ORC will review the PTR at the next scheduled meeting.

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F. Information Feedback

1. After the PTR is completed, the SAA will ensure that' the as-sembled package is forwarded as follows:

1.3.37-6 Rev. 1

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Original to the NOM a.

b. Copy to ORC Secretary for ORC review
c. Copy to OSS&P Group Leader - for operating experience

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  • review per NOP 8401 ,
2. After ORC review is complete the OSS&P Group Leaderwill ,

review the PTR package, summarize it and route the fpformation 1

within the Nuclear Organization. Additionally any 1Information of general interest to the industry will be entered into the operating experience category of NUCLEAR NOTEPAD. A final

- copy shall be sent to Document Control Center.

3. A Scram Report prepared for a planned trip not requiring a PTR shall be routed as follows:

I The original of the Scram Report with control rod position computer printout (00-7) or the scram attached, should be l r1uted through the Station Manager, a copy filed in the i Cmtrol Room and a copy routed to the Reactor Engineer.

VIII. ACCEPTANCE CRITERIA A. The PTR will be performed as described in the procedure.

B. Event Classification will key appropriate actions prior to restart of the unit.

IX. ATTACHMENTS A. Scram Report B. PTR Data Sumary C. Investigation and Evaluation D. Preliminary Safety Assessment /Startup Authorization 9

E. Sample PTR Package OrganiIation F. Data Recall Log Point ID Summary

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1.3.37-7 Rev. 1

BOSTON EDISON COMPANY PILGRIM STATION FORNORMATION mPE .C7.49 DE

.. SCRAM REPORT  ;

( s Date Time Number .

-CAUSE- -MODE SWITCH- -REACTOR STATUS- -FLOW CONTROL STATUS-POSITION I O operstarError O Loop Manual O Critical O Testing.Er7r O Run O Equipment Failure O Start up MWT O MasterManual O other O Refuel O Suberitical O Master Automatic

-STATION STATUS- -SCRAM TIMES- -STATION CONDITIONS-(f r m nit red CRD'S) Core Flow (263-10) Lb/Hr.

O Sm-@ Reactor Pressur,e (640 27)

Palg.

O Shutting Down O none available Steam Flow (640 27) LblHr.

O Changing Power Average Sec.

Vessel Level (640-26) Inches O SteadyState Fastest Rod Time Sec. Mr/Hr.

O SurveillanceTesting Off Gas Activity Lever Slowest Rod Stack Gas Activity Level Cps.

O other Time Sec. Building Vent Activity Level Cps.

Generator Ou'put MWe (gross) O Off Prompt Notification Via Telephone Required T/ Limit T/Notifled Person Contacted on CallIndividual Required 1 hr.

NRC Required 1 hr.

w Brief Description of Scram Corrective Action Taken

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Watch Engineer aEco. FORM X5066 ORIGINAL . STATION MANAGER TO D.CC

FORINFORMATION PNPS PTR DATA

SUMMARY

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-4 Date of Occurrence Time of Occurrence

~l I By STA:

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Date Time Part 1 INITIAL CONDITIONS The status of safety systems and a selected set of important plant parameters, pump running combinations control switch positions, chemistry results, and radiation readings that existed prior to the unscheduled reactor trip must be recorded. The data to be selected ,

should be based on the following considerations,:

o the data is not directly available on control room strip charts or computer printouts o the data is necessary to ascertain the cause of the trip or abnormal response and proper functioning of safety-related equipment o the data is necessary to effectively reconstruct plant status I prior to the trip or unexplained power reduction Examples (a) Reactor Power ( wth)*

(b) Unit Generator Load

  • 4 (c) Mode Switch Position *

(d) Reactor Vessel Pressure

  • l (e) Reactor Feed Pumps Operating (Circle) A B C on Instrument in l (f) Vessel Level *

! (g) Main Circulating Water Pumps Running (Circle) A B Attachment B

  • Also contained on Scram Report Page 1 of 5 1.3.378-i Rev. 1

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FORNORMATION (h) Status of Control Stations (Circle)

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1. Recirculation pump control Masterf? Local mode (Circle) Manual ,

Manual

2. Vessel level control (Circle) Auto -j Manual
3. Turbine Control Pressure Setpoint Load Limiter Setpoint f (1) Torus Temperature (C7) __

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(j) Off normal status of any trains / l portions of a safety system prior to event :

From Oper. 28 -

Details

1. RPS
2. ECCS
3. SBGTS
4. Emergency Buses / Diesels
5. DC Buses
1. Testing /Surveillances in Progress from Oper. 28 or other
  • Test Number Status / Step Attachment B Page 2 of 5 1.3.378-2} Rev.1

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FORNOREll0N

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Part 2 PLANT RESPONSE

. Data selected to be documented for determining plant response

(' should include the following:

o list of strip charts to be retained -

.i I o printouts from devices such as the process control computer,

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alarm printer, and event recorders o safety systems activations and performance information o manual, radiological, and control system actions Examples (a) Obtain a copy of the applicable parameter plots given below for every Event:

Panel 905 1. 1 Channel APRM (recorders 750-10A, B, C or 0) 905 2. Reccrder 640 Reactor Vessel Level (black per) . Feedwater Flow (red pen) 905 3. Reactor Vessel Pressure

a. Narrow range - 640-28 (red pen)
b. Wide range - 640-27 (black pen)

Steam flow is also on this recorder.

(b) Optionally the following may be requested depending on the Event:

Panel Parametef 905 1. Core Flow FR 263-110 C-1 2. Contrdi valve position IR 9027 903 3. Recirc. pump suction TR 151 A and B 904 4. Drywell pressure TRU 9045 905 5. Torus Level LR 5038 Attachment B Page 3 of. 5 1.3.37-B-3 Rev. 1

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Relief /S fety Valve Temperat 6.

7. Reacttr Water ccnductivity CAS-129-2

, 8. Main Steam Line Radiation PR 1705-11 L -

9. ECCS System Performance (if systte actuated)

Parameter -

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HPCI Flow

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RCIC Flow LPCI Flow CS Flow f

10. Condenser vacuum PR 3392
11. Vessel Metal Temperature l (c) Obtain a printout from: .
1. Process Computer Sequence of Events Log (not applicable to unexplained reductions) l 2. Process Computer Data Recall Log (not applicable to unexplained reductions)
3. Alarm Printer Output (with c.1 above)
4. Control Rod Position (program 00-7 output)

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5. Scram time output (d) Safety System Actuation and Performance
1. Reactor Protection System RPS Trips giving scram Actuation Time _: _
2. Containment Isolations Cause Time Group I Group II Group III Group IV Group V Group VI _

Attachment B

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Page 4 of 5 1.3.378-4 Rev. 1

3. ECCS l HPCI

- RCIC n

CS 7 LPCI ~l

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.- ADS (e) Control Systems (Circle)

1. Turbine Runback Yes No I
2. Turbine Trip Time Signal Yes No
3. Recirculation Pump Runback Yes No l
4. Recirculation Pump Trip Yes No  !
1 Cause l (f) Manual Actions Were any control stations taken from Yes No auto to manual? (Specify station time at time / sequence)

(g) Radiological Response (include abnormal area radiation monitoring, process radiation monitoring and environmental radiation monitoring indications O (h) Chemistry Response 1. Reactor Coolant Chemistry (1) Scram Times . Data copied for this report original output to the Reactor Engineer check g

that available computer printout scram times meet Tech. Spec. 3.3.C.3.

Other STA Comments ~:

Attachment B Page 5 of 5 1.3.378-5 Rev. 1

INVESTIGATION AND EVALUATION Date Time Part 1 Chronology of Event

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Part 2

. Attach written copy PROBABLE CAUSE OF TRIP

~l Conments: .

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.- Attach statements of involved personnel.

NOS STA

' By Watch Engineer NOTE: Unexplained power Part 3 UNEXPECTED ASPECT OF TRANSIENT BEHAVIOR (if event compared with previous similar reductions will not have transient note, the transient with which correlation with analyzed compared) transients or prior scrams. Examine para.

. meters appropriate to 'the g event to ensure safety limits not jeopardized.

Reload Analysis or FSAR Transient Page Number

-or- Previous Trip on. /

Date Time By: STA F&M issued if required by:

Part 4 IDENTIFICATION OF SYSTEMS WITH INADEQUATE PERFORMANCE System / Component Description of Problem s

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Signature Date Time WE

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Signature Date Time NOS Note: A separate F&M is required for each malfunction above.

Attachment C Page 1 of 1

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Rev.1 1.3.??r 1

PRELIMINARY SAFETY ASSESSMENT QQ u DN STARTUP AUTHORIZATION Un lM1f Part 1 TRANSIENT DATA FOR PERTINENT PLANT PARAMETERS Minimum

( , Maximum (a) RCS pressure Loop A B Loop A 8

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(b) Reactor vessel water level

..- (c) Reactor coolant flow Loop A B Loop A B (d) Reactor core thermal power Part 2 PRELIMINARY SAFETY ASSESSMENT (Circle)

(a) RCS pressure remained above 880 Yes No (b) Reactor isolation occurred Yes No (c) RCS pressure increased to safety /  :

relief valve operating pressure . Yes No (d) RCS temperature decrease less than 1000F/hr Yes No (e) HPCI/RCIC initiated Yes No (f) ADS timer initiated Yes No (g) Primary containment press tenip (h) Torus water level temp Part 3 EVENT CONDITION Classify trip as 1, 2 or 3 according to guidelines in procedure section VII.D.

I The event on at  : is a type Date Time 1, 2 or 3 Signature indicates agreement with

+ type classification

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Watch Engineer Date Time

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STA Date Time NOM COE Notification NOM notified of event classification /

Date Time Attachment D Page 1 of.2 1.3.370-1. Rev.1

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.,' NOM concurrenc37 YES NO

-If Type 3 COE n:,tified to take over investigatirn Time By Part 4 STARTUP AUTHORIZATION

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Type 2 EVENTS Plant manager notified and permission granted to startup the reactor.

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Watch Engineer Date Time Coments:

g) Type 3 EVENT

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ORC review of event on , meeting number Minutes of the meeting (s) are attached.

ORC Chairman Date Permission is granted to startup the reactor

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NOM Date Time Coments:

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f Attachment D Page 2 of 2 1.3.370-2 Rev. 1

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  • ^"nuam:^" FORNORMATION
1. Scram Report (Attachment A)

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PTR Data Sunsnary (Attachment B) w

3. Recorder Chart Copies ".
4. Investigation Evaluation (Attachment C)
5. Chronology of Event (separate sheet)
6. Statement of Involved Personnel (separate sheet)

~1. Preliminary Safety Assessment /Startup Authorization (Attachment D)

Attachment E 1.3.37E-l' Rev.0

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DATA RECALL LOG POINT ID

SUMMARY

LINE 1 B000 APRM "A" (% PWR)

( ~ B001 APRM "C" (% PWR) It i

B013 Rx. Pressure (PSIG) "

B014 Core Plate D/P (PSI) .

8015 Rx. Core Flow (M/H) -l B017 1 CRD Flow (M#/H) .

8018 - Rx. FW Inlet Flow "A" (M/H)

B019 ' Rx. FW Inlet Flow "B" (Mf/H)

B024' Rx. Water Level 8025 Rx. Outlet Steam Flow (Mf/H)

B028 FW Inlet Temp. "Al" (OF )

B030 FW Inlet Temp. "Bl" (DF )

B038 Recire. Flow Loop "A2" (M#/H) 8039 Recirc. Flow Loop "B1" (M#/H)

C002 Rx. Saturation Temp. (OF) .

CO23 Seawater Flow (GPM)

CO27 Hotwell Outlet Temp. (OF)

C050 Condenser d T (OF)

M071 Torus Level i LINE 2 G012 Stator Cooler Inlet Temp (OF) 6013 Stator Cooler Outlet temp (OF )

6019 Alternator Air to Cooler (OF) 6020 Alternator Air from Cooler (OF)

( F002 F007 Condensate Demineralizer D/P RFP Suction Pressure (PSIG)

F011 Condensate Pump Header Pressure F012 West Condenser Pressure (In Hg)

F013 East Condenser Pressure (In Hg)

F028 RBCCW Loop "A" Flow l F029 RBCCW Loop "B" Flow l + F030 RBCCW To RHR Hx. Loop "A" Flow .

F031 / RBCCW To RHR Hx. Loop "B" Flow -

F077 RBCCW Hx. Outlet Temp. "A" F078 RBCCW Hx. "B" Outlet Temp M034 Torus Pressure M035 'Drywell Pressure M042 SSW Flow Loop "A" M043 - SSW Flow Loop "B" Attachment F Page 1 of 1 1.3.37F Rev.1

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