ML20127A250
| ML20127A250 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/25/1985 |
| From: | Chase J CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20127A233 | List: |
| References | |
| OI-22, NUDOCS 8506210074 | |
| Download: ML20127A250 (40) | |
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1 s CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PIANT UNIT 0 PLANT INCIDENT AND POSITRIP INVESTIGATION OPERATING INSTRUCTION: OI-22 VOLUME VII l Rev. 009 Approved By: C.m U.General beManager / Date: /-a -87 ; j Manager - Operations 8506210074 850419 PDR ADOCK 05000324 O PDR L..
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LIST OF EFFECTIVE PAGES OI-22 Pane (s) Revision 1 7 2-4 8 5' 9 6-7 8 8 6 9 7 10 8 11-13 6 14 8
. 15 6 16 7 17 6 18 8 19-25 6 26-33 8 34-36 6 37 8 1
4 1 BSEP/Vol. VII/0I-22 i Rev. 9
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1.0 Purpose i To establish a formal plant level instruction whereby plant incidents are properly investigated to ensure:
-l- 1.1 The root cause of the incident is well understood and documented. , 1.2 The incident has not resulted in an unsafe or unacceptable condition for the operation of the power plant.
1.3 Appropriate corrective action is taken to prevent a recurrence of the incident. . 1.4 A continuing unanalyzed condition does not exist. 2.0 Definition 7 Plant incident - a plant incident is a(an): 2.1 Unplanned release to the environs of radioactive materials, or radiation exposures, in excess of allowed limits. 2.2 Anytime that E0P-01 is used. l 2.3 Release of radioactive materials within the power plant due to incorrect operation of a system. 2.4 Unplanned actuation of a safety system. 2.5 Failure of a safety system to actuate or remain actuated. 2.6 Degradation of a system / component such that power plant operation requires use of the Abnormal Operating Procedures (AOPs). l 2.7 Fire within the protected area of the power plant which lasts ten minutes or longer. 2.8 Discovery of a condition which would result in the failure of a safety system to perform its intended function. This shall include deliberate and inadvertent acts (reference IE Information Notice 83-27). 2.9 Event that Plant Management determines warranting a formal investigation. [ 3.0 Responsibilities 3.1 Shift Operating Supervisor Provides central direction of incident investigation until relieved i by the Engineer - Operations.
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3.2 Engineer - Operations 3.2.1 Assumes direct responsibility for investigating plant incidents, excepting radiological incidents. 3.2.2 -Provides written report of incidents as provided herein. 3.3.3 Forwards report to Manager - Operations. 3.3 Shift Technical Advisor Assists Shift Operating Supervisor in incident investigation. Aids Engineer - Operations as duti'es permit. 3.4 Manager - Operations 3.4.1 Determines when incident is terminated. 3.4.2 Recommends restart (if applicable) of the unit (s). 3.4.3 Forwards, with approval, incident investigation to Director - On-site Nuclear Safety. i 3.5 Fire Protection Specialist 3.5.1 Provides fire reports, as required, to Engineer - Operations. 3.5.2 Investigates fires; makes recommendations. 3.6 Manager - Environmental & Radiation Control (E&RC) 3.6.1 Assumes responsibilities for radiological incidents. 3.6.2 Generates other written reports of incident as may be assigned by the plant General Manager. 3.7 Manager - Technical Support Provides engineering support, as requested, to aid in determining root causes of occurrence, and makes engineering recommendations. 3.8 Director - On-site Nuclear Safety 3.8.1 Provides independent assessment of the incident, if required, by Attachment 1. 3.8.2 Conducts postrestart review of incidents, comparing it to similar previous incidents and/or FSAR transients. Rev. 8 ,' BSEP/Vol. VII/01-22 2 A
m e e i_ 3.9 Director - Quality Assurance Reviews incident and provides assessment of corrective actions if nonconformances are identified. 3.10 Plant General Manager Reviews incident and signs indications report is accepted. 4.0 References 4.1 RCI-06.5 4.2 E0P-01 4.3 RCI-06.1 5.0 Discussions All plant incidents, as defined in this standard instruction, are of such a nature so as to warrant a formal, documented investigation to discover and remove the root cause of the incident. Additionally, this instruction may be put into effect for any incident as determined by the plant General Manager or Manager - Operations. Each incident has two distinct phases in an investigative mode. 5.1 First, there is an Lamediate investigation to: 5.1.1 Obtain written or verbal first-hand reports of the event. 5.1.2 Retrieve and adequately label all pertinent data regarding' the event (i.e., charts, computer printouts, log entries). Take photographs, as applicable. 5.1.3 Determine the safety of the power plant and that all safety systems functioned as required. 5.1.4 Make an immediate recommendation regarding continued operation or restart of the power. plant. 5.1.5 If a scram resulted, prepare a scram report (Attachment 3). 5.1.6 Assess immediate reportability (RCI-06.5) and implementa-tion of the Emergency Plan (PEP-02.1). 5.2 Second, there is a follow-up, in-depth investigation which: BSEP/Vol. VII/01-22 3 Rev. 8
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,* i 5.2.1 Ensures initial determinations regarding causes of the incident and safety of the unit are adequate.
5.2.2 Ensures that an ATWS event did not occur. This review shall include verification that an incomplete automatic scram signal is not hidden by quick operator response such as a manual scram or mode switch positioning. 5.2.3 Determines the root cause of the event and makes appropriate corrective recommendations. 5.2.4 Reviews reportability. For simple incidents, the immediate investigation may also serve as the follow-up investigation. This determination will be made by the Manager - Operations. Figure 1 indicates the incident investigation flow path. 6.0 Instruction 6.1 When a plant incident occurs or upon achieving a stable condition following a plant incident, the Shift Operating Supervisor shall contact the appropriate Engineer - Operations (or in the absence of the Engineer - Operations, the Principal Engineer - Operations) or other appropriate managers, and inform that person of the event. 6.2 The Engineer - Operations or other designated individual will normally report to the site to direct the immediate investigation. If conditions do not warrant this action, the Manager - Operations may waiver this. This investigation shall address all the questions, as a minimum and as applicable, on Attachment 1. NOTE: In the event of a fire, the Fire Protection Specialist shall normally perform this step. However, if the fire is other than minor in nature, the Engineer - Operations shall also report to the plant and assess nuclear safety. This is defined as a fire of sufficient significance as to warrant a " notification of significant event" per 10CFR50.72. (See RCI-06.5) 6.3 Prior to the Engineer - Operations arriving, the Shift Operating Supervisor shall collect written statements from those involved regarding the event. (See Attachment 9.) Also, any other pertinent information (specified in Attachment 3) shall be gathered. 6.4 At all times, the function of the Shift Operating Supervisor and his staff shall be, foremost, the stable and safe configuration of the power plant (see E0P-01 for more detail). Attachment 3 shall be completed. i BSEP/Vol. VII/0I-22 4 Rev. 8 ,! j i
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NOTE: SRV failure / challenge /actuations shall be documented on Attachment 5. Each actuation, or attempted actua-tion, of each SRV shall be documented, as well as duration of each actuation. Maximum torus temperature shall be recorded. 6.5 Upon reporting to the power plant, the Engineer - Operations shall review the collected information and, with the aid of the Shift Operating Supervisor and Shift Technical Advisor, develop initial conclusions and recommendations and complete Attach-ment 10. Interviews, if deemed appropriate, shall be conducted. Engineering support will be requested if required. Within three working days, a formal written report in the format of Attach-ment 1 will be presented for PNSC review and plant General Manager approval. Other time deadlines may be established by the plant General Manager. A brief corporate report (see Attachment 4) shall be prepared by the Engineer - Operations and submitted to the Manager - Operations within three working days of the day of the event. After review by the Manager - Operations and the plant General Manager, Attachment 4 shall be transmitted to the Vice President - Brunswick Nuclear Project within six working days of the day of the event for further transmittal. An approved copy of the SRV failure / challenge /actuations report (Attachment 5) shall be forwarded to the Performance Engineering group on the day following the day of the event. 6.6 If completion of Attachment 1 indicates that the cause of the event is known and understood, immediate corrective action has reduced the probability of recurrence, there are no safety concerns, all safety systems functioned as required, and that an ATWS event was not hidden by prompt operator action, then continued unit operation or immediate restart may be conducted with the verbal permission of the Manager - Operations and General Manager. 6.7 If the cause of the event is not determined, if there are safety concerns, or the event has been repetitive, the Manager - Operations shall cause the Engineer - Operations to initiate a follow-up investigation. Follow-up investigations shall be similar in format to Attachment 2. 6.8 Follow-up investigations shall be directed by the Engineer - Operations. Technical support, as required, shall be supplied to the Engineer - Operations. 6.9 Follow-up investigations shall continue until such time as the PNSC and plant General Manager accept the follow-up report as being final. Until the final follow-up report is complete, an updated interim report in the format of Attachment 2 shall be supplied to the following personnel: 6.9.1 General Manager ,- BSEP/Vol. VII/0I-22 5 Rev. 9
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. o i 5 6.9.2 Vice President - Brunswick Nuclear Project 6.9.3 Manager - Operations l
The interim reports are due each month by the monthly date of the event. 6.10 In the event that cause of the unscheduled reactor shutdown cannot be determined, the General Manager may concur or approve a restart based on the following conditions. l 6.10.1 Any further reasonable actions to determine cause is considered unproductive. 6.10.2 No physical damage was done by the event and a determination made that the plant had not operated beyond the boundaries established by approved plant safety and transient analysis. 6.10.3 Any reasonable action to gain additional information has been considered. 6.11 All reports shall contain sufficient detail (diagrams, simple flow diagrams or schematics, pictures, descriptions) to be readily understood. 6.12 The final investigative reports following final review by the PNSC and approval by the plant General Manager shall be distri-buted per Attachment 11. l l 6.13 Report numbers are maintained by the Operations clerk. 6.14 Reports shall be considered a QA record and, when complete, shall be sent to Document Control. 6.15 Document Control shall distribute copies of the completed investigation to the personnel specified in 6.12. Distribu-tion will be documented by completing Attachment 11. BSEP/Vol. VII/0I-22 6 Rev. 8 f
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7.0 Tra'ining and Qualifications 7.1 Responsible personnel, as delineated in Section 3.0 of this instruction, shall be trained in accordance with BSEP Training Instructions 200, 201, and 202. 7.2. Responsible personnel, as delineated in Section 3.0 of this instruction shall be qualified in accordance with the BSEP Administrative Procedures (Volume I, Book 1, Operating Manual). l l l l t l l 1 BSEP/Vol. VII/0I-22 7 Rev. 8 l I
F-w , . ATTACHMENT 1 _ . Immediate Incident Investination Report No. Unit No. - 3 Time Date C
- 1. Status of the power plant immediately prior to the event (to include plant power levels, primary temperature, primary pressure, MWe [ gross),
significant equipment / systems 00S).
- 2. Evolutions in progress when event occurred.
- 3. Effect of the event on the power plant. ~
- 4. Attach completed copy of Attachment 3, if applicable.
- 5. Attach a completed copy of Attachment 10.
- 6. If incident involves failure of a component, describe the component (use illustrations, if useful). Include manufacturer, type, ratings, system identification, etc. Contact Maintenance and instruct them to retain and control the failed component for further assessment, if required .
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) ATTAC101ENT 1 (Cont'd)
Immediate Incident Investigation
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i ! i I 7. Is the cause understood? ( ) Yes ( ) No i
,' If the answer is yes, describe.
l If the answer is no, a follow-up investigation is required (complete
, Attachment 2).
- 8. Is the cause repetitive to require more extensive corrective action?
( ) Yes ( ) No NOTE: This is a judgement by the Engineer - Operations.
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l If yes, a follow-up investigation is required (complete Attachment 2).
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- 9. a. Does a potentia 1 saftty concern regarding operation or restart exist? (Complete Attachment 6.) ( ) Yes ( ) No Complete a follow-up ' investigation (Attachment 2), if required by
- Attachment 6.
- b. Were there any safety systems that did not function as required?
l , (Complete Attachment 8) ( ) Yes ( ) No
- 10. If any of the above questions (9.a or b) is answered yes, then ONS must be notified to conduct an independent assessment of the event for PNSC review and General Manager approval prior to conducting restart or continuing operations (see Attachment 7, Section II).
- 11. If incident is a result of an automatic scram, was an ATn*S event hidden by quick operator response such as a manual scram or mode switch
, positioning (review charts and computer printouts, reports, individual statements, etc.)? ( ) Yes ( ) No If yes, a safety evaluation must be written for PNSC review and General Manager approval prior to unit being returned to service (see
- Attachment 7, Section II).
- 12. If all the above questions (9-11) are answered No, operation of the power plant may continue with approval of the Manager - Operations and the
! General Manager. Complete Section I of Attachment 7.
l , 13. Is the event reportable? () Yes ( ) No (Refer to RCI-06.1) l l If yes, what type of report? i i I ! l 1 BSEP/Vol. VII/0I-22 9 Rev. 7 ! l l l 4
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e . ATTACHMEhT 1 (Cont'd) Immediate Incident Investigation
- 14. Does the event warrant a verification that a similar condition does not exist on the sister unit? ( ) Yes ( ) No If yes, a follow-up investigation is required (Attachment 2).
- 15. Follow-up report required: ( ) Yes ( ) No
- 16. The Shift Operating Supervisor on shift at the time of occurrence of the event concurs with the Engineer - Operations' responses to Steps 1 through 15. ( ) Yes ( ) No .
By: l Shift Operating Supervisor on shift If the Shift Operating Supervisor does not concur, a follow-up investigation must be conducted (Attachment 2). Submitted: Date: Engineer - Operations Concurrence: Date: Principal Engineer - Operations Recommended: Date: Manager - Operations 4 i Reviewed: Date: ONS Reviewed: Date: 1 l Director - QA/QC Reviewed: Date: PNSC Accepted: Date: General Manager BSEP/Vol. VII/01-22 10 Rev. 8
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. ATTACHMENT 1 (Cont'd)
.d Immediate Incident Investination NOTES: 1. After any scram, the following charts will always be attached to Attachment 3.
- Recirculation Pump Flow A & B
- Core Flow and Core Delta-P
- APRM Flux - All Recorders
- Feed Flow -
- Reactor Water Level
- Reactor Pressure
- Computer Printouts -
- 1. Sequence of events log from alarm typer
- 2. NSSS Post-trip log 3.-BOP Post-trip log *
- 4. CRD Position *
- SRM Flux -
- Main Condenser Vacuum
- SJAE Radiation Monitor
- MSL Radiation Monitor
- Reactor Building Ventilation Radiation Monitor
- Drywell Pressure
- 2. Other charts or computer printouts may be attached at the discretion of the Operations Engineer.
BSEP/Vol. VII/01-22 11 Rev. 6
s . ATTACEMENT 2 - Follow-Up Incident Investination Report No. Unic No. J Time Date ( ) Interia Report ( ) Final Report 1 Cause of Follow-Up Incident Investigation: ( ) Repetitive Event ( ) Safety Concern ( ) Clarity . ( ) Verification of sister unit
- 1. Briefly state initiating event (attach copy of Immediate Incident .
Investigation).
- 2. State result of e.he investigation to resolve item (s) above.
- 3. Proposed corrective action to resolve No. 2.
- 4. State conclusions regarding safe operation of the power plant.
- 5. Reportability evaluation (refer to RCI-06.1). l
- a. Event is 10CTR21 reportable. ( ) Yes ( ) No If Yes, notify Regulatory Compliance of evaluation.
( ) Notification made BSEP/Vol. VII/01-22 12 Rev. 6 l
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. ATTACHMENT 2 (Cont'd) .
Follow-Up Incident Investination
- b. Event is reportable other than 10CFR21. ( ) Yes ( ) No ,
If Yes, a copy of all regulatory reports are to be attached to this report. List reports t
- 6. Did inadequate procedures significantly contribute to the event? If so, specifically describe the inadequacies and procedures. ,
- 7. Is an operating experience report to be made? ( ) Yes ( ) No
- 8. Is an entry into NOTEPAD desirable? Entries should be made if the l incident provides industry useful information. ( ) Yes ( ) No i
If Yes, notify the Manager - Technical Support who will coordinate NOTEPAD entries. - Date Notified:
- 9. If sister unit was checked to see if similar condition exists, state:
1 a Who conducted check.
- b. When check was performed.
- c. What results were of check.
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! BSEP/Vol. VII/0I-22 13 Rev. 6 I
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m_ i g% .4 ATTACHMEhT 2 (Cont'd) Follow-Up Incident Investination Submitted: Date: Engineer - Operations Recommended: Date: Manager - Operations Reviewed: _ Date: ONS Reviewed: Date: ' Director - QA/QC Reviewed: Date: PNSC PNSC Action Item List ( ) Yes ( ) Not required Accepted: Date: General Manager BSEP/Vol. VII/0I-22 14 Rev. 8
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. A'ITACHMENT 3 Unit No. -.
BRUNSWICK STEAM ELECTRIC PLANF SCRAM No. . Date of SCRAM Time of SCRAM SCRAM Report
- 1. Conditions Prior to SCRAM:
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- a. Mode Switch Position d. Rx Pogpr %
- b. Total Core Flow lbs/hr e. Rx Press. psig
- c. Unit Gen. Load (net) MWe f. Rx Vessel Level in. ?
- 2. Parameters which directly caused SCRAM:
Channel A Time / min-see cycle Channel B Time / . min-sec cycle
- 3. Describe plant operation in progress at the time of SCRAM, including events causing SCRAM:
- 4. Corrective Action / Remarks:
- 5. BOP post-trip los attached. (Initia')
- 6. NSSS post-trip los attached (post-trip log from computer room).
(Initial)
- 7. SCRAM printout attached (alarm typer). (Initisi)
- 8. CRD position after SCRAM; printout attached (computer).
(Initial)
- 9. Following charts attached. (Initial)
A. Recirculation Pump Flow A & B B. Core Flow and Core Delta-P C. APRM Flux D. Main Steam Flow E. Feed Flow F. Reactor Water Level G. Reactor Pressure H. SRM Flux I. Main Condenser Vacuum . BSEP/Vol. VII/0I-22 15 Rev. 6 l l l l
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e i ATTACHMENT 3 (Cont'd)
! J. SJAE Radiation Monitor K. MSL Radiation Monitor l L. Reactor Building Ventilation Radiation Monitor M. Drywell Pressure I
- 10. Cycle the SDV vent and drain valves after the SCRAM reset and leave open.
(Initial)
- 11. Verify that the North and South SDV Not Drained alarms reset following draining of SDV. (Initial)
- 12. SCRAM / Incident Critique sheets attached. (Initial)
- 13. SRV Failure / Challenge Report (if necessary). (Initial)
- 14. Recorders marked to indicate time of SCRAM, chart speed, and time scale.
. (Initial)
- 15. On-call personnel notified (Who)
- 16. The following personnel notified:
A. Manager - Operations i B. General Manager l C. NRC Resident Inspector l
- 17. Plant Personnel Statements (Attachment 9) completed and attached.
- 18. A copy of all portions of E0P-01 that were used. The flowchart should be marked to indicate routes taken. All end path manual procedures used l should be checked off.
l SHIFT FOREMAN: / OPERATIONS ENGR.: / Initials Date Initials Date
. SHIFT OPERATING SUPV.: /
Initials Date i I i i I
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. A'ITACHMENT 4 .
Plant Incident Summary BACKGROUND DATA (Identify) Incation: c Time Frame: System Primarily Involved: Report Number: TRENDS (Fill in Blank)
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Plant Incidents Year-to-Date: Total Lost MWH: Plant Incidents Prior Year: Total Lost MWH Prior Year: EVENT (Describe) , IMPACT ON SAFETY (Describe) IMPACT ON OPERATIONS (Describe) REASONS EVENT OCCURRED (List) ROOT CAUSE ASSESSMENT (Check Appropriate Categories) Personnel Error: Improper Training Failure to Follow Instructions or Procedures Other (Explain in Comments) y Design External Cause (Explain in Comments) Program or Procedure Deficiency: Unclear Procedure Unavailable Procedure Not Covered by Procedure Not Covered by Program Component Failure CORRECTIVE ACTIONS (List) CORRECTIVE ACTION RESPONSIBILITY (Identify) COMMENTS (List) REPORTABILITY (Yes or No) PREPARED BY APPROVED BY APPROVED BY APPROVED BY BSEP/Vol. VII/0I-22 17 Rev 6
. . Figure 1 e
. as a End NO pit. incide. occurred
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YES Immed. inves.
- Ops Engr. = .
Immed. Inves.
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. Scram h - Ops Engr.-
- 30S *
- STA -
Cause Cogcur. , Obvious gr.-Ops. Cause' YES Obvious
. v NO safety
. NO IW oncern or epetitive Immed. place Safety plant into Y. config. to .Y Safety NO Concern- remove safety Concern L concern 9 Return to
- I service I,
T' NO Investigation granted by Mgr -Ops. and by Gen. Plnt. Mgr. Ops. Engr. Ops. Engr. heads Continues ._ k investigative Follow-up team investigation a o
'y Operations Engineer i
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p Resolved Continue Resolved , h No. investigation until NO l I
; Ir3 resolved Concerns e
Return to YES service i
, ranted by Return Unit gr.-Ops, and N,0 to normal Gen. Mgr.
configuration; per o Mgr.-Ops, Review report o b PNSC,0hs,qa I, J , ' ' r pproved by GeneraLPlant Manacer j Distribute copies per /- > Att hment BSEPVol. VII/01-22 18 Rev. 8
( l ATTACHMElff 5 , . SRV FAILURE / CHALLENGES REPORT TO: Performance Engineering Group FROM: Operations g DISCUSSION: Failure of SRV's to close is to be reported immediately to NRC (within one hour), while a summary of the challengers (when SRV's open) is to be reported annually. Effective as of April 1, 1980. Date: Time of occurrence: Time NRC person notified if necessary: Details of SRV's failure: - Maximum torus temperature: Copy to NRC coordinator: Copy to Performance Engineering:
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Shift Foreman Date Managet - Operations cc: Scram Report BSEP/Vol. VII/0I-22 19 Rev. 6
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. ATTAC}0fENT 6 .
.d Screening Guidelines for Transients And Safety System Challenges EVENT TITLE: EVENT DATE:
.c A. Criteria for determining event significance YES NO
- 1. TWo or more failures occur in redundant systems during the same event.
- 2. Two or more failures due to a common cause occur during the same event.
- 3. Three or more failures occur that easily would have escaped detection by testing or examination. ,
- 4. Component failures occur that easily would have escaped detection by testing or examination.
- 5. An event proceeds in a way significantly different than would be expected.
- 6. An event or operating condition occurs that is not enveloped by the plant design bases.
- 7. An event occurs that could have been a greater threat to plant safety w*th different plant conditions, the advent of another credible occurrence, or a different progression of occurrences.
- 8. Administrative, procedural, or operational errors are committed that result from a fundamental misunder-standing of plant performance or safety requirements.
- 9. A single failure occurs in a nonredundant system.
- 10. Two apparently unrelated failures occur during the same event.
- 11. A problem results in an off-site radiation release or personnel exposure.
- 12. A design or manufacturing deficiency is identified as the cause of a failure or potential failure.
- 13. A problem results in a long outage or major equipment damage.
BSEP/Vol. VII/0I-22 20 Rev. 6
ATTACHMENT 6 (Cont'd) . YES NO
- 14. An ESF actuation and injection occurs.
- 15. A particular occurrence is recognized as having c a significant recurrence rate.
- 16. A particular occurrence is recognized as posing a safety concern not specifically delineated above. '
- 17. If any of the above are answered yes, then a safety concern exists, and a follow-up investigation is required. (Attachment 2)
C f i t t BSEP/Vol. VII/0I-22 21 Rev. 6 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ I
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, ATTAC10fENT 7 -
1 ' ' ~ Permission To Start Up I M - SECTION I i General Manager notified and permission granted to start up the reactor. f l 1 Shift Supervisor Date Time W
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Duty Engineer Date Time I COMMENTS: < i, SECTION II PNSC Restart / Continued Operation Criteria
- 1. PNSC review of event on , meeting number l .i
- 2. PNSC findings and reccamendations to General Manager:
- a. Cause of trip or incident. *
- b. Maintenance or testing necessary prior to restart / continued operation.
BSEP/Vol. VII/OI-22 22 Rev. 6 m .. _ _ _ _ _ _ _ . - _ _ _ _
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$ 4 ATTAC10fE.VT 7 (Cont'd) -
- c. Additional monitoring / trending required after restart / continued operations. <
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- d. Necessary briefings / training for Operations / Maintenance concerning specific equipment indicctions or possible malfunctions.
- e. Conditions necessary for a reactor restart / continued operations.
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- f. Any unsafe or unacceptable conditions as a result of the reactor trip / incident.
- g. Ar.r axisting unanalyzed conditions.
l l l h. Additional comments. l I l 4-BSEP/Vol. VII/0I-22 23 Rev. 6
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- c ATTAC10fEST 7 (Cont'd) ,
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- 3. Permission is granted to start up or continue operation of the reactor.
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General Manager Date Time ; C0tefEhTS: l l 1 i BSEP/Vol. VII/0I-22 24 Rev. 6 1
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_~ t ATTACHMFXT 8 . Transient Response Verification
- 1. Review the plant's transient response from process computer data and chart recorders and determine the following:
YES NO A. Did a vessel level transient occur? (If yes, complete Attachment 8a.)
- 3. Did a reactor pressure transient occur?
(If yes, complete Attachment 8b.) C. Did a reactor power / flux transient occur? (If yes, complete Attachment 8c.) , D. Did a drywell pressure transient occur? (If yes, complete Attachment 8d.) E. Did a condenser vacuum transient occur? (If yes, complete Attachment 8e.) F. Did an off-gas activity transient occur? - (If yes, complete Attachment 8f.) G. Did a loss of electrical power occur? (If yes, complete Attachment 8g.) H. Was a scram setpoint reached? (If yes, complete Attachment 8h.) BSEP/Vol. VII/0I-22 25 Rev. 6
ATTACHMENT 8a Vessel Level Transient
- 1. From computer and RTGB recorders, obtain the lowest and highest levels l reached during the transient. High = Low =
- 2. If the below listed technical specification setpoints were reached or exceeded, verify the appropriate automatic actions listed below occurred.
Verification Automatic Setpoint Action Tech. Spec. Reached Occurred Setpoint Setpoint Automatic Action (Yes/No) (Yes/No) r Yes a. +208 inches HPCI, RCIC*, RFPT, main turbine trips No b. +192 inches High level alarm j No c. +182 inches Low level alarm Recirculation : runback permissive - Yes d. +162i inches Scram *, . GRP 2, 6, and 8 isolation
- ADS permissive
- Yes e. +112 inches GRP 1 and 3 isolation
- Secondary containment isolation
- 4 HPCI* (1) and RCIC*
Yes f. +2) inches CS* and RHR* initiation ADS
- actuation permissive Yes g. -53 inches Containment spray permissive *
- Technical Specification required NOTES: (1) Injection valve opening requires the level signal to remain for 6 seconds.
(2) Permissives without associated actions (trips initiations, etc.) need not be verified. , l (3) Lowest available recorded level is 150 inches unless both reactor i recirculation pumps are tripped. I BSEP/Vol. VII/0I-22 26 Rev. 8 ) i i
ATTACHMENT 8b Vessel Pressure Transient
- 1. From computer and RTGB recorders, obtain the highest and lowest pressures reached during the transient. High = Low =
- 2. If any of the below listed setpoints were reached or exceeded, verify the appropriate automatic actions listed below occurred:
NOTE: Recorder / instrument accuracy may be significant on pressure transients. . Verification Automatic Setpoint Action Tech. Spec. Reached Occurred l Setpoint Setpoint Automatic Action (Yes/No) (Yes/No) Yes a. 1325 psig None (Safety Limit)8 l Yes b. 1125 psig 2 3 SRVs _ Yes c. 1120 psig ATVS recire pump trip Yes d. 1115 psig 2 4 SRVs i Yes e. 1105 psig 2 4 SRVs i Yes f. 1045 psig Scram Yes g. 1020 psig None (Technical j specification limit normal operation) Yes h. 825 psig Group I in "run" Yes i. 395-425 RHR/CS injection psig valve permissive l . Yes j. 295-325 Recire discharge psig valve isolation if l initiation signal present Yes k. 140 psig SDS isolation 2 Consider recorder / instrument accuracy and setpoint tolerance per Technical l Specification 3/4.4.2. 8 If the RTGB recorder is upscale high, then check Reactor Building 50' elevation pressure instruments for peak pressure (B21-PI-R004A and B). BSEP/Vol. VII/0I-22 27 Rev. 8
3 . ATTACHMENT 8c Power / Flux Transient
- 1. From computer and RTGB recorders, obtain the highest power / flux level reached during the transient. High = I
- 2. If any of the below listed set points were reached or exceeded, verify ,
the appropriate automatic actions listed below occurred: 1 Verification Automatic Setpoint Action Tech. Spec. Automatic Reached Occurred j Setpoint Setpoint Action (Yes/No) (Yes/No) I l Yes a. 120% (fixed) APRM trip Scram l Yes b. Flow biased (.66W+54)T Scram Yes c. 90% (Unit 2 if APRM Scram setdown occurred) Yes d. 15% (not in "run" mode) Scram 3 BSEP/Vol. VII/0I-22 28 Rev. 8
D ATTACHMENT 8d Drywell Pressure - Transient
- 1. From computer and RTGB recorders, obtain the lowest and highest drywell pressures reached during the transient. High = Low =
- 2. If any of the below listed technical specification setpoints were reached or exceeded, verify the automatic actions listed below occurred:
Verification
. Automatic Setpoint Action Tech. Spec. Reached Occurred Setpoint Setpoint Automatic Action (Yes/No) (Yes/No) n Yes 2 psig HPCI initiation GRP 2, 6, 8 isolation Yes -0.5 psig Suppression pool to Reactor Bldg vacuum breaker isolation valves open No 40 psig CAC/ CAD drywell inlet valves isolate BSEP/Vol. VII/0I-22 29 Rev. 8
1 . ATTACHMEhT 8e Condenser Vacuum Transient
- 1. From computer and RTGB recorders, obtain the lowest and highest condenser !
vacuums reached during the transient (in inches of Hg). High = Low =
- 2. If the below listed technical specification setpoints were reached or l i
exceeded, verify the associated automatic actions listed below occurred: Verification Automatic Setpoint Action Tech. Spec. Reached Occurred Setpoint Setpoint Automatic Action (Yes/No) (Yes/No) No a. 22 inches Hg Turbine trip Yes . b. 7 inches Hg GRP 1 isolation Turbine bypass valves close BSEP/Vol. VII/0I-22 30 Rev. 8
m - - -
-1 \ m A'lTACHMENT Bf Off-Gas Activity Transient
- 1. From computer printouts and RTGB recorders obtain the highest values of:
- a. MSL Rad Monitor (A-D) A: B: C: D:
- b. SJAE Rad Monitor A: B:
- c. RB Roof Vent Rad Monitor A: B:
- 2. If the below listed technical specification setpoints were reached, verify ,
the associated automatic actions listed below occurred: Verification Automatic , Setpoint Action Tech. Spec. Automatic Reached Occurred Setpoint Setpoint Action (Yes/No) (Yes/No) Yes a. ( ** ) MSL Rad Monitor GRP 1 isolation l Mechanical vacuum pump isolation , Yes b. ( ** ) SJAE Rad Monitor Timer starts: 15 minutes later A0G HCV-102 isolates. l Yes c. 11 (mr/hr) RB Roof Vent Rad Monitor GRP 6 isolation SBGTS initiation Reactor Building ventilation isola-tion NOTE: ** Calculated setpoint. Check current instrument setpoint. BSEP/Vol. VII/0I-22 31 Rev. 8 i
ATTACHMEhT 8g Loss of Electrical Power Transient
- 1. Determine which electrical buses (4 kV) lost power and indicate below.
- 2. If the below listed technical specification setpoints are reached or exceeded, verify the associated automatic actions listed below occurred:
Verification f Automatic ; Bus (es) Action i Tech. Spec. Automatic That Lost Occurred l (Yes/No) Action _ Power (Yes/No) ; Yes a. Loss of power to BOP bus ! E bus feeders l Yes b. Loss of power to E bus trip and DG starts and j , energizes E bus. v BSEP/Vol. VII/0I-22 32 Rev. 8
ATTACHMENT 8h Scram Setpoint Reached
- 1. From computer printouts and RTGB recorders, determine if any of the following scram setpoints were reached or exceeded:
Verification Automatic Setpoint Action ,
- Automatic Reached Occurred I Setpoint Action (Yes/No) (Yes/No)
- a. IRM --
120% of scale Scram
- b. APRM -- 120% (fixed) Scram
(.66 W+54)T Scram (flow biased) 15% (not in "run") Scram
- c. Pressure high -- 1045 psig Scram
- d. Vessel water level low -- Scram
+162.5
- e. MSIV closure -- 10% closed Scram
- f. MSL high rad -- 3X normal
- Scram .
- g. Drywell pressure high -- Scram 2 psig
- h. TSV close -- 10% closed Scram
- 1. TCV fast closure (850 psig Scram EHC pressure)
- j. Mode switch in shutdown Scram
- k. SDV level high -- 109 gals Scram
- 2. If any of the above setpoints were reached or exceeded in both RPS divisions, verify that an automatic scram resulted.
- 3. If a manual scram occurred during a transient, verify no scram setpoints were exceeded prior to the manual scram. .
- Calculated setpoint. Check current instrument setpoint.
BSEP/Vol. VII/0I-22 33 Rev. 8
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[. 9 , . . -t
. NITAC} DENT 9 .
Plant Personnel Statements ; Attach a written statement from each person associated with the trip < concerning the events that preceded and followed the trip. Each individual should write the statement the way he remembers the event rather than including any discussion items with other people after the event. Example:
- Name Position BSEP/Vol. VII/0I-22 34 Rev. 6
6 . i " = c ATTACHMENT 9 (Cont'd) . Prepare a haadwritten statement describing the trip event sequence and plant response as you remember it. Include the plant conditions prior to the trip, your indications that a problem existed, noted equipment malfunctions or inadequacies, and any identified procedural deficiencies. Also include any ) information you consider important to review this unscheduled reactor trip. (Use additional sheets, if necessary.) C
/
Signature Date Time BSEP/Vol. VII/0I-22 35 Rev. 6
g . ATTACHMENT 10 ,
?
Operations Engineer Written Narrative of Scram / Incident The unit Operations Engineer shall, utilizing all available data, written statements, and results of interviews, reconstruct the event and draft a written narrative including sequence of events leading up to the ; scram / incident, automatic actions, operator-initiated actions, causes of scram / incident, conclusions, and recommendations, m l 1 l BSEP/Vol. VII/0I-22 36 Rev. 6 '
.2M
o e b t .6 , ATTACHMENT 11 Completed Investigation Report Distribution The attached' completed incident investigation has been distributed to the following personnel:
- 1. Chairman of the Board - (Attachment 10 only)
- 2. Vice President - Brunswick Nuclear Project
-3. Vice President - Corporate Nuclear Safety
- 4. General Manager
- 5. Manager - Operations -
- 6. Manager - Maintenance
- 7. Manager - Technical Support
- 8. Director - QA/QC
- 9. Director - On-Site Nuclear Safety i
- 10. Director - Regulatory Compliance '
- 11. Principal Engineer - Operations f
i Document Control Specialist Date , BSEP/Vol. VII/0I-22 37 Rev. 8}}