ML17346B067
| ML17346B067 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/03/1985 |
| From: | Williams J FLORIDA POWER & LIGHT CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| GL-83-28, TAC-52810, TAC-52811, NUDOCS 8506040327 | |
| Download: ML17346B067 (25) | |
Text
REGULATOR Y I -
RMATION OI'S TR IBu T ION S VS T (R IOS)
ACCESSIOtf NBR: 8506O4O327 DOC. DATE; 85/05/03 NOTARIZED:
NO DOCKET FACXL'50-250 Turkey Point Planti Unit 3i Florida Power and Light C
05000250 50-251 Turkey Point Planti Unit 4i Florida Power and Light C
05000251 AUTH~NAME AUTHOR AFFILIATION WILLIAMS'
~ H, F I or ida Power L Light Co, RECIP,NAME RECIPIENT AFFILIATION VARGAiS ~ AD Op~rating Reactors Branch 1
SUBJECT:
For wards response to NRC 850416'equest for addi info re methods 8 criteria for comparing post trip review info,w/
known or expected plant behavior<per response to Generic Ltr 83 28
'ISTRIBUTION CODE:
A055D COPIES RECEIVED:LTR ~
ENCL SIZE; r
TITLE; OR/Licensing Submittal:
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FLORIDAPOWER & LIGHTCOMPANY NY 3 s I985 L-85-214 Office of Nuclear Reactor Regulation Attention:
Mr. Steven A. Varga, Chief Operating Reactors Branch 7j I Division of Operating Reac'tors U. S. Nuclear Regulatory Commission
Dear Mr. Varga:
Re:
Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-25 I Generic Letter 83-28, Item I. I (Post Trip Review)
TAC Numbers 528 I 0 and 528 I I Attached is our response'to your April l6, l985 request for additional information regarding the methods and criteria for comparing post trip review information with known or 'expected plant behavior.
It addresses the concerns and recommendations in Sections III.C and III.Eof your Safety Evaluation.
If you have any questions regarding this response, please call us.
Very truly yours, J. W. Williams, Jr.
Group Vice President Nuclear Energy JWW/TCG/eab 8506040327 850503 PDR ADOCK 05000250 P
PDR PEOPLE... SERVING PEOPLE
Turkey Point Units 3 and 4
Docket Nos.
50-250, 50-251 ENCLOSURE
SUBJECT:
REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS (Generic Letter 83-28, Item 1.1 - Post Trip Review)
By letter dated November 8,
- 1983, which was forwarded in response to the NRC Generic Letter 83-28, Florida Power and Light provided information regarding its Post-Trip Review Program and Procedures for the Turkey Point Plant Units 3
and 4.
In its Draft Safety Evaluation (SE) for Item 1.1 of Generic Letter 83-28, the NRC identified a concern with the Florida Power and Light response to Item 1.1 of Generic Letter 83-28 and requested additional information.
The fol 1 owing informat i on addresses the NRC '
concerns and recommendat i on s identified in Sections III.C and III.E of the Draft Safety Evaluation.
NRC SE CONCERNS AND RECOMMENDATIONS I II.C.
III.E.
The licensee has not addressed the methods and criteria for comparing the event information with known or expected plant behavior.
We recommend that the pertinent data obtained during the post-trip review be compared to the applicable data provided in the FSAR to verify proper operation of the systems or equipment.
Where
- possible, comparisons with previous similar events should be made.
The licensee has provided for our review a systematic safety assess-ment program to assess unscheduled reactor trips.
We recommend that this program be revised to include methods for comparing the event information with known or expected plant behavior as described in Section III.C above.
RESPONSE
Since submittal of FPL's November 8, 1983 response to Generic Letter 83-28, a
number of revisions to both Administrative Procedure (AP) 0103.16, "Duties and Responsibilities of The Shift Technical Advisor (STA)",
and Off-Normal Operating Procedure (ONOP) 0208. 1, "Shutdown Resulting from Reactor Trip or Turbine Trip", have been made to address the methodology and criteria used to perform post-trip reviews.
The improvements made to the way the post-trip reviews are handled at Turkey Point used INPO's Good Practice Procedure OP-
- 211, "Post-Trip Reviews",
as a
model in developing these procedural revisions.
All requirements governing the post-trip review process have been removed from the Off-Normal Operating Procedure 0208.1 and have been incorpor-ated into Attachment 2, Appendix B, of Administrative Procedure 0103. 16, dated April 24, 1985.
A copy of this latest revision of Attachment 2, Appendix B, to Administrative Procedure 0103. 16 has been enclosed for the NRC 's informa-tion.
Turkey Point Units 3 and 4
Docket Nos. 50-250, 50-251
Enclosure:
GL83-28 (Item 1. 1) Response In the April 24, 1985 revision to Procedure AP 0103.16, the STA has been assigned the responsibility to perform a post trip review for any automatic or required manual (due to a malfunction) reactor trip or turbine trip, including subcritical reactor trips.
These post trip reviews will be initiated in accordance with the methodology and criteria of Attachment 2,
Appendix B, of Administrative Procedure 0103.16 and will be documented and retained as perma-nent plant records on completed copies of Appendix B.
Utilizing the new post-trip review guidelines delineated in the latest revision of Section 6.0 of Appendix B, the STA compares the trip to similar transients, either a previous trip and/or FSAR delineated transients in order to verify the proper operation of systems and equipment.
Provisions have been made in Appendix B to document the date of any previous trips and/or page number(s) of applicable FSAR trans-ients (Appendix B, Subsection 6.d).
Appendix B, Subsection 6.d, also requires the STA to describe any unexpected aspect of the transient behavior of the plant and to document the resolution of any variances which were identified during the post trip review process.
A file of previous post-trip reviews is kept in the control room for the purposes of performing this comparison.
A copy of the FSAR is also kept in the control room for this purpose.
The criteria used to perform a preliminary safety assessment as part of the post-trip review process are identified in Section 7.0 of Appendix B to AP 0103.16.
These criteria are as follows:
a.
RCS pressure remained above setpoint for automatic SIS Actuation b.
RCS pressure remained below setpoint for pressurizer code safety valve actuation c.
RCS temperature decreased less than 100'/hr.
d.
Reactor Coolant was contained within the primary RCS and the pressur-izer relief tank e.
Indicated pressurizer level remained on scale f.
Indicated S/G level remained on scale The STA's have been instructed on the above methodology and criteria for performing a post-trip review and are familiar with the requirements of the latest revision of AP 0103.16.
ADMINISTRATIVE PROCEDURE 0103. 16, PAGE ll DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 ATTACHMENT 2 STA REPORT OF PLANT ABNORMAL OCCURRENCES Prepared by Reviewed by Reviewed by Reviewed by I.
Title:
S ngsneer STA Lead ng>neer peration upport ng>neer uperv>sor Techn>ca Department Supervisor Date Date Date Date II Report Type:
(Check One)
STA Report of Plant Significant Events.
(Complete and attach Appendix A)
Copies to:
(Marked Prel iminary)
Plant Manager - Nuclear Operations Supervisor - Nuclear Operations Superintendent
- Nuclear Maintenance Superintendent
- Nuclear guality Assurance Superintendent Technical Department Supervisor Regulation and Compliance Supervisor NRC Resident Inspector Post Trip Review:
(Complete and attach Appendix B)
Copies to:
(Marked Preliminary)
Plant Manager - Nuclear Operations Supervisor - Nuclear Operations Superintendent
-'Nuclear Maintenance Superintendent
- Nuclear guality Assurance Superintendent Technical Department Supervisor Regulation and Compliance Supervisor NRC Resident Inspector STA Report of Plant Occurrences (Complete'nd attach Appendix C)
Copies to:
(Marked Preliminary)
Regulation and Compliance Supervisor Technical Department Supervisor
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 20 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADYISOR 4/24/85 APPENDIX B
POST TRIP REVIE'A Unit:
Date of Trip:
Time of Trip:
1.
Initial Plant Conditions:
a, Plant Status Prior to Reactor Trip (circle one):
(1)
(2)
(3)
(4)
(5)
Unit Startup Operations Steady State Operations Load Changes during Routine Power Operations Unit Shutdown Operations Other (describe) b.
Reactor Power c ~
e.
Reactor Coolant System Pressure PORY Block Valves MOV-*-535 MOV-*-536 Status of Control Stations:
(ci rcl e)
(circle)
Open Open Slg Closed Closed (1)
(2)
(3)
(4)
(5)
Loop A Feedwater Control Loop B Feedwater Control Loop C Feedwater Control Steam Dump to Condenser Pressurizer Pressure Control (ci rcl e)
(ci rcl e)
(circle)
(ci rcl e)
(ci rcl e)
Manual Manual Manual Manual Manual Auto Auto Auto Auto Auto f.
Off Normal Status of any Trains/Portions of Safety Systems:
(1)
Reactor Protective System (2)
Safety Injection System
ADtlINISTRATIVE PROCEDURE 0103.16, PAGE 21 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX 8 (Cont'd) beet 2 of 12)
POST TRIP REVIEW (3)
Emergency Containment Cooling and Filtering (4)
Containment Isolation (6)
Diesel Generators g.
Test Surveillances in Progress:
Test Number Status/Step
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 22 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX 8 (Cont'd)
POST TRIP REVIEW 2.
Plant
Response
a.
Trip Sequence Time Intervals (obtained from DDPS-SOE)
- NOTE: If DDPS is Out of Service (OOS) indicate status of 'components marked with asterisk.
1.
- Logic Matrix Actuated:
2.
RT Relays Dropout Time 0 3.
- Reactor Trip Breakers (OPEN)
A B
NOTE:
The Reactor Trip Breakers opening time is checked on a monthly basis using Operating Procedure 1004.2.
4.
- RPI Rod Bottom 5.
- Turbine Stop Right Valves -
(CLOSED)
Left 6.
- Generator Breakers A
(OPEN)
Note the first-out annunciator on each panel.
C E
A<<(i
)
0 available).
b.
Safety System Actuation (N/A if not applicable)
Actuation Time (1) Reactor Protection System Initiating Signal (2) Safety Injection System (3) Emergency Containment Cooling and filtering System (4) Containment Spray System (5) Containment Isolation c ~
Safety Valve Actuation:
(1) Pressurizer Code Safety Valves (circle if lifted)
(2) Main Steam Code Safety Valves (circle)
A 8
C Yes No
ADHINISTRATI VE PROCEDURE 0103.16, PAGE 23 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
POST TRIP REVIEM d.
Manual Actions:
(1) Mere any control stations taken from AUTO to HANUAL mode Yes No If yes, speci fy stati on and time.
(2) Other manual actions (specify and include time):
e.
Radiological
Response
(include abnormal ARM and PRIS indications):
f.
Other Comments and Responses:
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 24 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4i24/85 APPENDIX B (Cont'd)
Sheet 5 of 12)
POST TRIP REVIEW
- g. ackup Documentation:
attach a copy.of all applicable logbook entries, recorder charts, log sheets, etc...
Information sources to.be collected.
Item to be collected Check if copied and collected 2 ~
Sequence of Events (from Units high speed printer)
Post Trip Review files (from DDPS)*
/
Appropriate log books, list as applicable (refer to Aministrative Procedure 0103.2, Responsibilities of Operators and Shift Technicians on Shift and.
Maintenance of Operating Logs and Records, Section 8.3 and 8.4 for listing of Control Room log books.)
log book og book og book log book og book 3.
Appropriate recorder charts, list as applicable (refer to Administrative Procedure 0103.2, Responsibilities of Operators and Shift Technicians on Shift and Maintenance of Operating Logs. and Records, Section 8.3 and 8. 4 for partial listing of Control Room recorder charts)
Recorder No.
Recorder No.
Recorder No.
Recorder No.
Recorder No.
4.
Other information (list)
Measures Measures Measures easures Measures The following is a list of channels that should be retrieved following a Unit trip:
PRIORITY 1: (Should include after every trip if possible) 009, 042, 046, 062, 081, 083, 085, 088,
- 141, 142, and 145 PRIORITY 2: (Include if time and conditions permit) 024, 033, 052, 053, 064, 076, 096,
- 100, 105,
- 143, and 144.
NOTE:
If a particular channel is out of service, or provides erroneous information, substitute it with another channel reading similar information (i.e.,
Ch.
042 is out of service, substitute with Ch.
41 or 43).
~
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 25 DUTIES AND RESPONSI8ILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 3.
PLANT PERSONNEL STATEMENTS ATTACHMENT 8 (Cont'd)
Sheet 6 of 12)
POST TRIP REVIEW a.
Operations Shift on Duty During Event (number):
The following personnel were directly involved with the event and may be able to supplement the information in this report if clarification is needed:
b.
Attach statements from personnel involved with the trip concerning the events that perceded and followed the trip.
Each individual should submi.t a
statement concerning the way they remember the event.
Personnel statements attached (Circle one)
YES NO Example Name:
Position:
If handwritten statements are prepared, include the plant conditions prior to the trip, your indications that a problem existed, your action as a
result of those indications, noted equipment malfunctions or inadequacies, and any identified procedure deficiencies.
Also, include any information you consider important to review this unscheduled reactor trip and actions to prevent recurrence.
(Use additional sheet if necessary) signature ate 1me
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 26 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd) 2)
POST TRIP REVIEW 4.
Sequence of Events:
Prepared by STA:
Reviewed by:
P ant Superv>sor
- Nuc ear Date Date Briefly describe the sequence of events preceding and following the reactor trip.
TIME EVENT DESCRIPTION
I, l'
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 27 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
POST TRIP REVIEM 5.
Cause of Reactor Trip:
a.
Logic that directly resulted in reactor trip:
b.
Probable cause of the reactor trip:
6.
Post Trip Evaluation:
a.
Limiting Conditions for Operation (list any deviations):
b.
Identification of Systems and Equipment with Inadequate Performance:
SYSTEM COi4lPONENT DESCRIPTION OF PROBLPilS NOTE:
If appropriate, have system or equipment walkdown inspections or damage assessment conducted and documented.
Attach results
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 28 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
Sheet 9 of 12)
POST TRIP REVIEW C ~
Verify unit is stable or controlled to desired conditions, consider:
CHECK X
Nuclear Instrumentation RCS Fluid Inventory RCS Pressure RCS Temperature RCS Flow Containment Sump PRMS ARMS S/G Level S/G Pressure Turbine Gen. (Vib., hydrogen status, etc.)
Secondary Systems REMARKS:
I Sa ety Inject>on was actuated verify Emergency Operating Procedure 20000 is implemented.
d.
Unexpected Aspect of Transient Behavior (i,f event compared with previous similar transient; note the transient with which compared) ompared
>th Previous Reactor Trip on Resolution of Difference(s)
Date 1me e.
Transient Data for Pertinent Plant Parameters MAXIMUM MIN IMUM RCS Pressure RCS Temperature SG Pressure SG Level Loop A
B Loop A
A Loop A
B Loop A
B Loop A
B Loop A
Loop A
B Loop A
B
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 29 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
POST TRIP REVIEW 7.
Preliminary Safety Assessment (circle) a.
RCS pressure remained above setpoint for automatic SIS Actuation b.
RCS pressure remained below setpoint for pressurizer code safety valve actuation c.
RCS temperature decreased less than 100'/hr.
d.
Reactor Coolant was contained within the primary RCS and the pressurizer relief tank e.
Indicated pressurizer level remained on scale f.
Indicated S/G.level remained on scale COMMENTS:
Yes Yes Yes Yes Yes Yes No No No No
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 30 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
POST TRIP REVIEW 7.
Hake appropriate notifications:
Operating Procedure 0103.12 Notification of Significant Events to NRC Emergency Procedure 20103 Classification of Emergencies 8.
a.
If Maintenance Department support is required to investigate the cause of a
reactor/turbine trip, the appropriate maintenance department(s) shall provide the results of their investigation:
NOTE:
If maintenance support is not required, mark this step N/A.
)
PWO(s) attached.
PWO Nos.
Fill in if additional clarifying information is available or if a PWO is not attached.
Work performed and findings:
Signed Mech. Maint.
Elect. Maint.
I 8
C Haint.
Appropriate Maint. Dept. Supervisor(s) or designees).
(Hark N/A for departments not involved in investigation) b.
Failure of the Maintenance Department investigation to support the most probable cause listed in item 4b above shall be discussed in item 9 below.
ADMINISTRATIVE PROCEDURE 0103.16, PAGE 31 DUTIES AND RESPONSIBILITIES OF THE SHIFT TECHNICAL ADVISOR 4/24/85 APPENDIX B (Cont'd)
POST TRIP REVIEW 9.
Discuss trip cause, any safety considerations, and immediate corrective actions.
A)
If cause of reactor trip is known (see Item 4), acceptably corrected and any safety considerations are resolved, and the PS-N, STA, and Operations Supervisor,and Plant Manager - Nuclear or his designee concur; authorize unit startup in accordance with OP 0202.2, Unit Startup, Hot Shutdown to Power Operation.
8)
If cause of reactor trip is not known (see Item 4) or any safety considerations are not resolved, The Plant Manager - Nuclear or his designee shall be consulted and he shall determine the need for additional investigation, technical support and PNSC review prior to authorizing unit startup.
The STA's Signature on the Report of Plant Abnormal Occurrences shall be complete prior to restart.
REMARKS: - Reactor startup authorized pending resolution of the following items iL>st as applicable):
NOTE:
Attach completed copy of all applicable PS-N checklists.
Completed Date:
Signed:
Plant Supervisor - Nuc ear REMARKS'ime:
Signed:
Shi t Technica Advisor Transmit package with attachments to the Operations Supv. for further review and routing.
SIGNED:
perations Supervisor - Nuc ear DATE REVIEWED:
Transmit package with attachments to the STA Lead Engineer (Tech. Dept.) for further review and routing.
Turkey Point Units 3 and 4
Docket Nos.
50-250, 50-251 ENCLOSURE
SUBJECT:
REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EYENTS (Generic Letter 83-28, Item 1.1 - Post Trip Review)
By letter dated November 8,
- 1983, which was forwarded in response to the NRC Generic Letter 83-28, Florida Power and Light provided information regarding its Post-Trip Review Program and Procedures for the Turkey Point Plant Units 3
and 4.
In its Draft Safety Evaluation (SE) for Item 1.1 of Generic Letter 83-28, the NRC identified a concern with the Florida Power and Light response to Item 1.1 of Generic Letter 83-28 and requested additional information.
The fo1 1 owing informat i on addresses the NRC '
concerns and recommendat i on s identified in Sections III.C and III.E of the Draft Safety Evaluation.
NRC SE CONCERNS AND RECOMMENDATIONS III.C.
III.E.
The licensee has not addressed the methods and criteria for comparing the event information with known or expected plant behavior.
We recommend that the pertinent data obtained during the post-trip review be compared to the applicable data provided in the FSAR to verify proper operation of the systems or equipment.
Where
- possible, comparisons with previous similar events should be made.
The licensee has provided for our review a systematic safety assess-ment program to assess unscheduled reactor trips.
We recommend that this program be revised to include methods.for comparing the event information with known or expected plant behavior as described in Section III.C above.
FPL RESPONSE Since submittal of FPL's November 8,
1983 response to Generic Letter 83-28, a
number of revisions to both Administrative Procedure (AP) 0103.16, "Duties and Responsibilities of The Shift Technical Advisor (STA)",
and Off-Normal Operating Procedure (ONOP) 0208.1, "Shutdown Resulting from Reactor Trip or Turbine Trip", have been made to address the methodology and criteria used to perform post-trip reviews.
The improvements made to the way the post-trip reviews are handled at Turkey Point used INPO's Good Practice Procedure OP-
- 211, "Post-Trip Reviews",
as a
model in developing these procedural revisions.
All requirements governing the post-trip review process have been removed from the Off-Normal Operating Procedure 0208.1 and have been incorpor-ated into Attachment 2, Appendix B, of Administrative Procedure 0103. 16, dated April 24, 1985.
A copy of this latest revision of Attachment 2, Appendix 8, to Administrative Procedure 0103. 16 has been enclosed for the NRC 's informa-tion.
Turkey Point Units 3 and 4
Docket Nos. 50-250, 50-251
Enclosure:
GL83-28 (Item l. 1) Response In the April 24, lg85 revision to Procedure AP 0103.16, the STA has been assigned the responsibility to perform a post trip review for any automatic or required manual (due to a malfunction) reactor trip or turbine trip, including subcritical reactor trips.
These post trip reviews will be initiated in accordance with the methodology and criteria of Attachment 2,
Appendix B, of Administrative Procedure 0103.16 and will be documented and retained as perma-nent plant records on completed copies of Appendix B.
Utilizing the new post-trip review guidelines delineated in the latest revision of Section 6.0 of Appendix B, the STA compares the trip to similar transients, either a previous trip and/or FSAR delineated transients in order to verify the proper operation of systems and equipment.
Provisions have been made in Appendix B to document the date of any previous trips and/or page number(s) of applicable FSAR trans-ientss (Appendix B, Subsection 6.d ).
Appendix B, Subsection 6.d, also requires the STA to describe any unexpected aspect of the transient behavior of the plant and to document the resolution of any variances which were identified during the post trip review process.
A file of previous post-trip reviews is kept in the control room for the purposes of performing this comparison.
A copy of the FSAR is also kept in the control room for this purpose.
The criteria used to perform a preliminary safety assessment as part of the post-trip review process are identified in Section 7.0 of Appendix B to AP 0103.16.
These criteria are as follows:
a.
RCS pressure remained above setpoint for automatic SIS Actuation b.
RCS pressure remained below setpoint for p ressuri zer code safety valve actuation c.
RCS temperature decreased less than 100'/hr.
d.
Reactor Coolant was contained within the primary RCS and the pressur-izer relief tank e.
Indicated pressurizer level remained on scale f.
Indicated S/G level remained on scale The STA's have been instructed on the above methodology and criteria for performing a post-trip review and are fami liar with the requirements of the latest revision of AP 0103. 16.
e
~e