ML20093N795

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Rev 0 to Procedure 1.3.37, Post-Trip Reviews
ML20093N795
Person / Time
Site: Pilgrim
Issue date: 02/28/1984
From:
BOSTON EDISON CO.
To:
Shared Package
ML20093N760 List:
References
TASK-1.C.6, TASK-TM 1.3.37, NUDOCS 8408020149
Download: ML20093N795 (18)
v • d • e


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I NUCLEAR OPERATIONS DEPARTMENT

' PILGRIM NUCLEAR POWER STATION Procedure 1.3.37 POST TRIP REVIEWS List of Effective ~Pages l 1.3.37-1 1.3.37-2 1.3.37-3 1.3.37-4

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1.3.37-5 1.3.37-6 1.3.37-7 l

l List of Attachments I

1.3.37A-1 1.3.378-1 1.3.37B-2 k 1.3.378-3 Approved _ /J&wo QA Punager 1.3.37B-4 1.3.378-5 .

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I. PURPOSE 5 To provide instructions to personnel to perform Post TRIP Reviews (PTR) following unplanned reactor trips. Adherence to this procedure will ensure that consistent data will be collected, and that a uniform an-alysis and decision process will be applied after a reactor trip 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 INPO have resulted in recommendations aimed at standardizing post trip reviews. BEco commitments 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. i.

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

  1. to return to power atter an unscheduled reactor trip. Station per-sonnel must reasonably determine the cause of the trip, verify proper functioning of safety related and other equipment during the trip, and ensure that the trip 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 j

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

8. US NRC Generic Letter 83-28 " Generic Implications of Salem ATWS Events" C. BECo letter " Response to Generic Letter 83-28" D. Nuclear Operations Pracedure NOP 3301 " Conduct of Operations" ,

i -

E. PNPS Operations Manual procedures:

l  : 1. 1.3.3 ' Authority to Shutdown and Startup Station"

$ 2. 1.3.9 ' Reports" i

l

! 3. 1.3.12 " Notification and Recall of Personnel" 4 2.2.17 " Communications" 1

i l-1.3.37-2 Rev. 0 4

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IV. APPLICABILITY i

The requirements of this procedure will apply to all unplanned trips from the "RUN" mode. The Chief Operations Engineer or the Nuclear Operations Manager may require that elements of this procedure be fol-lowed for other unit problems f rom other operating conditions, such as unexplained power reductions.

V. PREREQUISITES A. The post trip review (PTR) will be initiated after 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.

8. Sufficient information shall be collected from personnel involved i.

in the unit trip prior to permitting relief by the oncoming shif t.  :

6

. >- 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.

8. 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-ments from operating personnel and others involved in the trip.

C. Shift Technical Advisor is responsible with the NWE and NOS for trip investigation. This includes interpretation of the Process Computer Data Recall tog and Sequence of Event Log. Additionally the STA will record the data required on Attachment 8.

D. Shif t Administrative Assistant is responsible for collecting the .

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required recorder charts and obtaining statements of invcived per sonnel. Additionally the SAA is responsible to ensure that the PTR package is retained for records and distributed to the Nuclear

- Operation: Manager and the OSS&P Group Leader.

g 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.

F. The Nucicar Operations Manager or his designate is responsible to authorize restart of the reactor.

1.3.37-3 Rev. 0

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G. The ORC shall review all post trip reviews. In the case of trips classified as Type 1 or 2, the ORC shall review the trip at its next scheduled meeting. If the trip is classified as Type 3, the NOM will convene ORC to provide independent assessment of the event.

VII. POST TRIP REVIEW PROCEDURE NOTE: Event notification to appropriate agencies 'or persons shall be made consistent with procedure 2.2.17 " Communications".

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

1. Data Collection
2. Trip Investigation }-
3. Event Classification / Safety Assessment
4. Restart Authorization
5. Information Feedback B. Data Collection
1. The object of the data collection is to assemble enough infor-mation to reconstruct the trip, assess the response of sys-tems, and identify the root cause of the event.
2. The SAA shall collect the following hard copy data:
a. Process Computer Data Recall Log (Attachment F identifies points on this Log)
b. Process Computer Sequence of Events Log and Alarm Typer Output *
c. Recorder charts f rom:

1

1. One APRM recorder

! ii. Feedwater flow iii. Wide (or) narrow range reactor pressure iv. Reactor level

v. Other recorders as identified by the STA or NWE l
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.

1.3.37-4 Rev. 0 O

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4. The SAA in conjunction with or under the direction of the NOS shall record statements from each individual involved in the event. These statements shall be obtained oniv after the plant is in a stable condition. Statements should include facts concerning the event relative to pretrip conditions or activities, initial indications of a problem, initial actions taken, equipment automatic and manual operation, observed malfunctions or procedural deficiencies. -
5. It may be appropriate for the WE to interview involved per-sonnel or to collect information from assembled individuals as a group.
6. The STA shall complete part 1 of the PTR DATA Sunmary (Attachment B). If additional information is needed it shall be obtained at this time. _

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7. Collected data and personnel statements shall be assembled
  • into a package and delivered to the WE to begin the trip
  • investigation.

C. Trip Investigation (Attachment C)

1. The WE, NOS, and STA shall reconstruct the event by preparing a chronology of the event. ( Attachment C, part 1)
2. The W E 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 i

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 Analysis. Peak reactor pressure, lowest water level, drywell pressure, steamline radiation are examples of what

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shall be recorded for this analysis. Departures from the bounds of the FSAR or reload analysis shall be noted on part 3;.

of Attachment C and t,rought to the immediate attention of th'e WE. A potential unreviewed safety question may exist.

i 4. The WE will complete the Scram Report (Attachment A) and sum-martre the event. The scram report shall identify 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.

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

1.3.37-5 Rev. O i

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D. Event Classification / Safety Assessment

1. The preliminary Safety Assessment (Attachment D) form shall be completed by the STA and Watch Engineer.
2. The Event shall be classified as Type 1, 2, or 3 by the fol-lowing criteria:
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 er 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-ir-cess of correction or a Tech. Spec. Constraint does not 5 prohibit startup.

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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, e the malfunction cannot be readily corrected, or startup is precluded due to Tech. Spec. Constraints, or the transient did r.ot 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.

t E. Restart Authorization

. 1. The NOM will be informed of the results of the PTR and classi-

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fication. The NOM may then authorize restart if the event is classified as Type 1 or 2.  ;.

If the NOM is not satisfied with the results of the

2. PTR he

- will take actions necessary to satisfy his concerns.

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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.

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4. In the case of Type 1 or 2 events the DRC will review the PTR l at the next scheduled meeting.

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. 0 L

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b. Copy to ORC Secretary for ORC review

, c. Copy to OSS&P Group Leader.- for operating experience review per NOP 8401

2. Af ter ORC review is complete the OSS&P Group Leader will review the PTR package, summarize it and' route the information within the Nuclear Organization. Additionally any information of general interest to the industry will be entered into the operating experience category of NUCLEAR NOTEPAD.

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

r-B. Event Classification will key appropriate actions prior to restart *

[ of the unit. -

IX. ATTACHMENTS A. Scram Report B. PTR Data Sununary C. Investigation :nd Evaluation D. Preliminary Safety Assessment /Startup Authorization L

E. Sample PTR Package Organization F. Data Recall Log Point ID Sunnary e .

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PILGRIM STATION ,

RTYPE C7 SCRAM REPORT Date Time Number

-CAUSE- -MODE SWITCH- -REACTOR STATUS- -FLOW CONTROL STATUS-0 Operator Error POSITION O Critical O Loop Manual O Teoti.,o stror O Run O Equipment Failure O Start-Up MWT O Master Manuai O other O Refuel O Subcritical O Master Automatic

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

O wie - Ps Reactor Pressure (640 27)

O Shutting Down O nom available r-Steam Flow (640-27)

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' O Changing Power Average Sec.

D Vessel Level (640-26) Inch O Steady State Fastest Rod Time Sec.

O S eveillance Testing Off Gas Actl.ity Level Mrli Slowest Rod Stack Gas Activity Level Ci O Other Time Sec. Building Vent Activity Level C; Generator Output MWe (gross) O(

Prompt Notification Via Telephone Required TILimit T/ Notified Person Contacted On CallIndividual Required 1 hr. __

NRC Required 1 hr.

L Brief Description of Scram Corrective ActionTaken Watch Engineer SECo. FORM X5006 ORIGINAL STATION MANAGER TO D C.C.

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PNPS PTR DATA

SUMMARY

Date of Occurrence Time of Occurrence By STA: l

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Date Time Part 1 INITIAL CONDITIONS The status of safety systems and a selected set of important plant i.

parameters, pump running combinations control switch ositions, *

"- chemistry results, and radiation readings that exist. prior to the I 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 prior to the trip Examotes (a) Reactor Power (tawth)*

k (b) Unit Generator Load * ,

k (c) Mode Switch Position *

(d) Reactor Vessel Pressure *

.I (e) Reactor Feed Pumps Operating (Circle) A B C (f) Vessel Level

  • on Instrument in (g) Main Circulating Water Pumps j

Running (Circle), A 8 Attachment B l

  • Also contained on Scram Report Page 1 of 5 l

1.3. ,78-1 Rev. 0

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'i (h) Status of Control Stations (Circle)

1. Recirculation pump control Master Local mode (Circle) Manual Manual
2. Vessel level control (Circle) Auto Manual
3. Turbine Control Pressure Setpoint Load Limiter Setpoint ,

E-(i) Torus Temperature (C7) 0F $

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

From Dr.er. 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 .

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Test Number Status / Step I'

Attachment B Page 2 of 5 1.3.378-2 Rev. 0 l.

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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 o printouts from devices such as the process control computer, alarm printer, and event recorders o safety systems activations and performance information o manual, radiological, and control system actions I-Examples a s-

  • (a) .0btain a copy of the applicable parameter plots given below for every Event:

Panel 905 1. 1 Channel APRM (recorders 750-10A, B, C or D)

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, 905 2. Recorder 640 Reactor Vessel Level (black pen), 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 ,

l Event: p Panel Parameter l

? j 905 1. Core Flow FR 263-110 C-1 2. Control valve position ZR 9027 903 3. Recirc. pump suction TR 151 A and B 904 4. Drywell pressure TRU 9045 905 5. Torus Level LR 5038 l

. Attachment B Page 3 of 5 '

1.3.37-B-3 Rev. 0 l

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6. Relief / Safety Valve Temperature TR 260-20 1
7. Reactor Water conductivity CAS-129-25
8. Main Steam Line Radiation PR 1705-11
9. ECCS Systeer Performance (if system actuated)

Parameter HPCI Flow RCIC Flow LPCI Flow CS Flow

10. Condenser Vacuum PR 3392

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11. Vessel Metal Temperature
  • (c) Obtain a printout from:
1. Process Computer Sequence of Events Log
2. Process Computer Data Recall Log
3. Alarm Printer Output (with c.1 above)
4. Control Rod Position (program 0D-7 output)

(d) Safety System Actuation and Performance

1. Reactor Protection System RPS Trips giving scram Actuation Time _:_
2. Containmaat ?colations -

Cause Time h Group I i Group 11 Group III Group IV Group V Group VI Attachment B Page 4 of 5 1.3.378-4 Rev. O l

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Actuation r s 3. ECCS Signal Time HPCI RCIC CS ,

LPCI ADS _

(e) Control Systems (Circle)

1. Turbine Runback Yes No
2. Turbine Trip Time Signal Yes No }.
3. Recirculation Pump Runback Yes No a
  • 4. Recirculation Pump Trip Yes No Cause (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 (h) Chemistry Response 1. Reactor Coolant Chemistry

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Other STA Consnents:

Attachment B

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.. coe c.. o . e INVESTIGATION AND EVALUATION Date Time .

Part 1 Chronology of Event Attach written copy Part 2 PROBABLE CAUSE OF TRIP Comments:

Attach statements of involved personnel.

By Watch Engineer NOS STA Part 3 UNEXPECTED ASPECT OF TRANSIENT BEHAVIOR (if event compared with previous similar I-transient note, the transierit with which

, compared)

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 4

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WE Signature Date Time

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

Attachment C Page 1 of 1 1.3.37C-1 Rev. 0

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l PRELIMINARY SAFETY ASSESSMENT STARTUP AUTHORIZATION Part 1 TRANSIENT DATA FOR PERTINENT PLANT PARAMETERS Maximum Minimum (a) RCS pressure Loop A 8 Loop A B (b) Reactor vessel water level (c) Reactor coolant flow Loop A B Loop A B (d) Reactor core thermal powar Part 2 PRELIMINARY SAFETY ASSESSMENT (Circle)

(a) RCS pressure remained above 880 Yes No a- (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 temp (h) Torus water level temp Part 3 EVENT CD:!DITION Classify trip as 1, 2 or 3 according to guidelines in procedure section.

The event on at  : is a condition Date Time I, II, III Signature indicates agreement with  ;

P condition.

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Watch Engineer Date Time g

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STA Date Time NOM COE Notification NOM notif.1d of event classification /

Date Time Attachment D Page 1 of 2 1.3.370-1 Rev. 0

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If Type 3 COE notified to take over investigation .

Time By Part 4 STARTUP AUTHORIZATION Class 1, 2 EVENTS Plant manager notified and permission granted to startup the reactor.

/

Watch Engineer Date Timt E-t

( Comments:

Class 3 EVENT ORC review of event on , meeting numbe.'

Minutes of the meeting (s) are attached.

ORC Chairman Date Permission is granted to startup the reactor

/

NOM Date Time

{ Connents:

I.

1-h Attachment 0 Page 2 of 2 1.3.370-2 Rev. 0

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SAMPLE PTR PACKAGE ORGANIZATION

1. Scram Report (Attachment A)
2. PTR Data Summary (Attachment B)
3. Recorder Chart Copies
4. Investigation Evaluation (Attachment C)
5. Chronology of Event (separate sheet)
6. Statement of Involved Personnel (separate sheet)
7. Preliminary Safety Assessment /Startup Authorization (Attachment D) J.

a.

i-l Attachment E l

1 1.3.37E-1 Rev.0 t

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DATA RECALL LOG POINT ID

SUMMARY

LINE 1 B000 APRM "A" (% PWR) 8001 APRM 'C' (% PWR)

B013 Rx. Pressure (PSIG) -

8014 Core Plate D/P (PSI)

B015 Rx. Core Flow (M/H)

I B017 CRD Flow ( M /H) 8018 Rx. FW Inlet Flow "A" (M/H)

B019 Rx. FW Inlet Flow "B" ( M/H) 8024 Rx. Water Level 8025 Rx. Outlet Steam Flow (M/H) 8028 FW Inlet Ternp. "Al" (OF ) _

, B030 FW Inlet Temp. "B1' (DF) E-

, 8038 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 4 T (OF)

M071 Torus Level LINE 2 L 6012 Stator Cooler Inlet Temp (DF) 6013 Stator Cooler Outlet temp 6019 Alternator Air to Cooler F) ({0F) 6020 Alternator Air from Cooler (OF)

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

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

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F013 East Condenser Pressure (In Hg)

F028 RBCCW Loop 'A' Flow -

F029 RbCCW Loop 'B' Flow i;.

F030 RBCCW To RHR Hx. Loop "A" Flow F031 RBCCW To RHR Hx. Loop "B" Flow F077 RBCCW Hx. Outlet Temp. 'A' i F078 RBCCW Hx. "B" Outlet Temp M034 Torus Pressure M035 Drywell Pressure l M042 SSW Flow Loop 'A' l M043 SSW Flow Loop 'B" l

l Attachment F Page 1 of 1

! 1.3.37F-1 Rev. 0 l

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