ML030870025
| ML030870025 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/06/2003 |
| From: | Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 | |
| Download: ML030870025 (105) | |
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Nuclear Mangement CompanyP Page 1 o of 3 STATE CHANGE HISTORY
- _L Conduct Work Review & Quality Check Initiate Assign Approved Assign Work Work Complete Approval 7/212002 2/5/2002 6/5/2002 7/2/2002 2.32 59 PM 4.05 16 PM 5 02.45 PM 1:51.58 PM Owner PBNP Owner RICK Owner RICK by RICK by RICHARD by RICK Owner RICK by RICK CAP Admin WOOD WOOD WOOD WOOD WOOD FLESSNER WOOD Complete and Close Done 7/10/2002 12.07.37 PM Owner (None) by MARYBETH ARNOLD SECTION 1 Activity Request Id: CA003704 Activity Type: Corrective Action Submit Date: 2/5/2002 4:05:16 PM Site/Unit: Point Beach Common Activity Requested: Evaluate if an Engineering Supplemental Guideline is the appropriate procedural method for controlling PRA updates, or if a higher tier document such as a Nuclear Procedure (NP) should be used considering the interfaces involving other departments. Initiate any procedure changes resulting from that evaluation.
0 CATPR: N Initiator: FLESSNER, RICHARD EPP Engineering Initiator Department: EX Engineering Responsible Group Code:
Programs PRA PB Processes PB RICK WOOD l Responsible Department: Engineering Activity Supervisor:
Activity Performer: RICK WOOD SECTION 2 Priority: 3 Due Date: 7/3/2002 Management Exception From P5?: N
" Mode Change Restraint: (None)
"0Licensing Review?: N "0QA/Nuclear Oversight?: N NRC Commitment?: N "0NRC Commitment Date:
SECTION 3 Activity Completed: 1/18/2002 12:52PM - LARRY PETERSON:
Due date extended as requested and approved by F. Cayia in prior update. Retruned to R.
flessner for completion.
1/18/2002 12:54PM - LARRY PETERSON.
Reassigned to R. Flessner for completion following extension.
6/5/2002 5:02:45 PM - RICK WOOD:
The NP is in draft. Additional comments from technical reviewer need to be incorporated.
6/19/2002 10:00:52 AM - RICK WOOD Additional comments from the reviewers need to be incorporated. The expected issue date of https://nmc.ttrackonline.comltmtrackltmtrack.dll?IssuePage&Tableld=1000&Recordld= 1 (... 9/20/2002
Nuclear Management Company Page 2 of 3 the procedure is 7/26/2002. The Eng Director has approved the extension 7/2/2002 1:51:58 PM - RICK WOOD:
NP 7.7.20 Probabilistic Risk Assessment was issued 6/26/2002. This procedure includes the interface requirements.
7/10/2002 12.07:37 PM - MARYBETH ARNOLD:
NP 7.7.20, Revision 0 was issued on 06/26/02. The Purpose notes this CA and the Bases contains this CA as to why the procedure was created. CLOSED.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 "f* Project: CAP Activities &
Actions =ý
" State: Done @Active/Inactive: Inactive (None) AR Type: Daughter
" Owner:
Assigned Date: 6/5/2002
"* Submitter: RICHARD FLESSNER 6
"* Last Modified Date: 7/10/2002 12.07:37 0 Last Modifier: MARYBETH ARNOLD PM
"* Last State Change Date: 7/10/2002 12:07:37 D Last State Changer: MARYBETH ARNOLD PM "OClose Date: 7/10/2002 12:07:37 PM
" One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 1
References:
CR 01-2278 RCE 01 -069 GOOD CATCH NP 7.7.20, Revision 0 Update:
Import Memo Field:
PBNP CAP Admin Site: Point Beach CAP Admin:
OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS
&9* ACE00314: Probabilstic Risk Assessment PRA For AuxiliaryFeedwaterSys~tefAFW
!.CAP001415" Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https://nmc.ttrackonline.com/tmtrack/tmtrack.dil?IssuePage&TableId= 1000&Recordld=l 1H... 9/20/2002
Nuclear Management Company Pagle 3 of 3
&. CE000316" PRA information would improve training (tracking) https://nmc.ttrackonline.comltmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld= 11 ... 9/20/2002
ESG 5.1 PRA MAINTENANCE AND UPDATE GUIDELINE DOCUMENT TYPE: Administrative REVISION: 2 EFFECTIVE DATE: September 20, 2002 APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Nuclear Safety Analysis OWNER GROUP: Engineering
POINT BEACH NUCLEAR PLANT ESG 5.1 ENGINEERING SUPPLEMENTAL GUIDELINES NNSR Revision 2 PRA MAINTENANCE AND UPDATE GUIDELINE September 20. 2002 1.0 PURPOSE 1.1 This document provides overall guidance for updating the Probabilistic Risk Assessment (PRA) model on an on-going and routine basis.
1.2 This guideline applies to any updates to the controlled PRA documents and models. This guideline does NOT apply to Safety Monitor database changes that conform to the documented PRA model and notebooks. Documents and models within the scope of this procedure include:
1.2.1 PRA Notebooks 1.2.2 WinNupra Fault Trees and Data Files identified in the PRA Notebooks 1.3 The PRA model may be used to support evaluation of proposed procedure changes, technical specification, surveillance interval changes, system configuration changes, and evaluation of nuclear safety issues. However, such analyses are NOT considered as updates unless changes to the models or databases are actually implemented.
NOTE: The intent of this guideline is to provide a framework for accomplishing changes to the PRA model starting with the 1999 PRA Update effort. Closeout steps in Sections 4.3.3 and 4.3.4 will be preformed following completion of the initial model changes for this update effort (PRA Update Phase I) and for all model changes thereafter.
2.0 DISCUSSION 2.1 Regulatory requirements have been established for each licensee to perform a PRA for their respective plants. The purpose of this is for the licensee to develop an appreciation of severe accident behavior, to understand the most likely severe accident sequences that could occur at its plants, to gain a quantitative and qualitative understanding of the overall probability of core damage and radioactive material release, and to reduce the overall probabilities of core damage and radioactive material release by modifying, where appropriate, hardware and procedures that would help prevent or mitigate severe accidents.
2.2 An Individual Plant Examination (IPE) has been performed for Point Beach to meet the regulatory requirements found in Generic Letter 88-20. The initial PRA models for each of the units was created in support of the IPE. PRA models continue to be used. The model is used as the basis for the risk monitoring program (Safety Monitor). In addition, the model is used in support of many activities including the following list. With the many uses of the PRA model, it is important to maintain an updated PRA model for the plant.
"* On-Line Maintenance Risk Evaluations
"* Modification Prioritization Page 2 of 6 INFORMATION USE
POINT BEACH NUCLEAR PLANT ESG 5.1 ENGINEERING SUPPLEMENTAL GUIDELINES NNSR Revision 2 PRA MAINTENANCE AND UPDATE GUIDELINE September 20, 2002
- Significant Event Assessment
- Issue Management
"* Maintenance Rule Assessments
"* Severe Accident Management
"* Shutdown Risk Evaluations 3.0 RESPONSIBILITIES 3.1 Nuclear Safety Analysis (NSA) Supervisor - The NSA Supervisor is the person who has line responsibility for the NSA group. The NSA Supervisor (or designee) has the authority to sign as the approver on the document update form.
Responsible to ensure analyst and reviewer are qualified to perform the task 3.2 PRA Analyst - The PRA Analyst is responsible for performing the following:
3.2.1 Evaluation of changes identified on the PRA Facility Change Impact Form to
.determine the potential impact of the change on the PRA model and determine the acceptable time frame for documentation and model update that may result.
3.2.2 Analysis of plant changes using the latest update of the PRA model and making modifications to the PRA model and documentation using the PRA software.
3.2.3 Review of changes made to the PRA model and documentation.
3.2.4 Tracking pending changes to the PRA model and documentation.
4.0 GUIDELINE 4.1 PRA Model Review and Change Form 4.1.1 Any plant personnel can initiate a PRA model Review and Change Form.
PBF- 1626 can also be initiated as directed in Design Engineering Procedure D6-P03.
4.1.2 Section 1 of the form should be filled out by the initiator fully describing the plant change and providing references to any documentation that would be useful in evaluating the change. (Modification numbers, Procedure numbers, etc.) Form can also be initiated to suggest potential enhancements not related to an actual facility change.
Page 3 of 6 INFORMATION USE
POINT BEACH NUCLEAR PLANT ESG 5.1 ENGINEERING SUPPLEMENTAL GUIDELINES NNSR Revision 2 PRA MAINTENANCE AND UPDATE GUIDELINE September 20. 2002 4.1.3 The initiator should send the form to a PRA Analyst or NSA supervisor.
4.1.4 The PRA analyst should provide an initial evaluation of the change and determine appropriate disposition using industry guidance. (i.e., EPRI TR-105396, "PSA Applications Guide," August 1995)
- a. No Impact - Should be marked for those issues that do not require a PRA model or document change.
- b. Immediate Change - Should be marked when the change could have a significant impact on the use of the model for PRA applications. Change should be implemented within the next 90 days of just prior to the completion of the actual plant change - whichever is later.
- c. Minor Impact (Change within the next 3 years) - Should be marked when the change has only minor impact on use of the model for PRA applications.
4.1.5 After completion of the disposition (no impact or model change implemented), the form should be routed to NIM for filing as a plant record.
4.2 PRA Data Analysis Periodic Update 4.2.1 Periodic data analysis will be performed on the PRA model approximately every 3 years.
4.2.2 The periodic update will include:
- a. Updating Basic Event data resulting form current plant equipment availability and reliability data.
- b. Updating Initiating Event frequencies considering plant history for these initiating events.
- c. Reviewing plant procedures that may impact Human Error Probability (HEPs) used to support the PRA analysis.
- d. Reviewing Operating Experience associated with the PRA systems and documenting any changes performed as a result of this review in the appropriate system or data analysis notebook.
- e. Reviewing changes to Technical Specifications and Design Basis Calculations that may affect assumptions used in the PRA model. Any changes identified should be documented in the appropriate system or data analysis notebook.
Page 4 of 6 INFORMATION USE
POINT BEACH NUCLEAR PLANT ESG 5.1 ENGINEERING SUPPLEMENTAL GUIDELINES NNSR Revision 2 PRA MAINTENANCE AND UPDATE GUIDELINE September 20, 2002 4.2.3 Periodic Update Process
- a. Data Collection Phase - During this phase, data sources will be identified and pertinent data extracted.
"* Calculations that are the basis for the PRA assumptions will be reviewed for changes.
"* Operating procedures used as input into the HRA analysis.
"* Equipment performance data will be extracted.
"* Surveillance Test Procedures will be reviewed for changes in test frequency.
"* Key personnel, such as Maintenance Rule Owner, Operations personnel, and System Engineers may be contacted as necessary.
"* Operating Experience associated with PRA systems should be collected.
- b. Data Screening and Analysis Phase - The PRA Analyst will screen the data to determine if model changes are warranted and data analysis should be.
performed.
- c. Any changes identified by the Periodic Update will be performed per the guidance contained in Section 4.3.
4.3 PRA Model and Documentation Update 4.3.1 The PRA model and documentation will be updated as necessary due to changes identified by the PRA Facility Change Impact Forms, changes identified by the Periodic Data Analysis, or any other changes identified by a PRA Analyst.
4.3.2 PBF-0026a should be used to document the review and approval of changes made to the PRA notebooks. Since this is a generic form for document review and approval, there are some sections that do not apply to the PRA notebook updates. Questions in Section Ifi associated with validation and safety evaluations should be marked as follows:
"* Validations required: marked NO
"* Changes pre-screened: marked YES Page 5 of 6 INFORMATION USE
POINT BEACH NUCLEAR PLANT ESG 5.1 ENGINEERING SUPPLEMENTAL GUIDELINES NNSR Revision 2 PRA MAINTENANCE AND UPDATE GUIDELINE September 20, 2002
"* Screening Complete: marked NA
"* Training or Briefing Required: marked as appropriate
- Training assistance desired: marked as appropriate
- QR/MSS Review NOT Required should be marked 4.3.3 Prior to the Release for Distribution the following should be performed:
- a. Revise any other PRA Notebooks affected by the change
- b. Update the Safety Monitor Database with any related changes
- c. Revise the CDF baseline, if necessary, for use in trending.
- d. Inform Safety Monitor users of any model changes that will significantly affect results or will impact how Safety Monitor can be used (e.g., addition of a new surveillance test effect). Initiate a Training Request, PBF-6101, if formal training is appropriate.
4.3.4 Following the Release for Distribution, these steps should be performed:
- a. Review the impact of the change on the overall PRA model and determine if new vulnerabilities should be addressed. GL 88-20 and NEI 91-04 can be used as a guide. New vulnerabilities which need to be addressed should be documented in the Corrective Action Program.
- b. Perform the additional actions specified in NP 7.7.20, Probabilistic Risk Assessment, to inform the plant staff of new PRA results and to determine any impact on programs that utilize those results.
5.0 REFERENCES
5.1 Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities," November 23, 1988 5.2 EPRI TR-105396, "PSA Applications Guide," August 1995 5.3 NEI 91-04, "Severe Accident Issue Closure Guidelines," Revision 1, December 1994 5.4 PRA Notebooks 5.5 NP 7.7.20, Probabilistic Risk Assessment 6.0 BASES None Page 6 of 6 INFORMATION USE
NP 7.7.20 PROBABILISTIC RISK ASSESSMENT DOCUMENT TYPE: Administrative REVISION: 0 EFFECTIVE DATE: June 26,2002 APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Group Head OWNER GROUP: Engineering
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE .............................................................................................................. 3 2.0 RESPONSIBILITIES ..................................................................................................... 3 3.0 DISCUSSION ........................................................................................................... 3 4.0 PROCEDURE .......................................................................................................... 3 5.0 REFEREN CES ........................................................................................................... 5 6.0 BASES ........................................................................................................................... 5 Page 2 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT 1.0 PURPOSE The procedure establishes interface requirements between Programs Engineering - PRA and Training, Licensing and Operations.(B-1) 2.0 RESPONSIBILITIES 2.1 PRA staff: Ensure that the appropriate memos are developed following update of the PRA model. Identify risk significant Human Interactions and forward to Operations and Training as they are identified.
2.2 Supervisor PRA: Review the memo and information sent to applicable groups.
2.3 Operations Training Supervisor: Incorporate information from PRA into Licensed Operator Training.
2.4 Operations Procedures Supervisor: Review information from PRA and identify procedure changes 3.0 DISCUSSION 3.1 The update of the PRA model is controlled via ESG 5.1 PRA Maintenance and Update Guideline.
3.2 Human Interactions are classified as three types: Type A are interactions occurring before the initiating event; Type B are interactions associated with the initiating event; Type C are interactions associated with response to the initiating event. The focus of this procedure is Type C Human Interactions.
3.3 The EOP Verification and Validation Matrix was developed with a cutoff of an Initiating Event frequency greater than 1 E-3 /year and a Core Damage Frequency of 1 E-6/year.
4.0 PROCEDURE 4.1 Following periodic update of the PRA model, notify the Training Group of significant changes to:
4.1.1 System Importance 4.1.2 Initiating Event frequency 4.1.3 Human Error Probabilities and Importance 4.1.4 EOP Verification and Validation Matrix (OM 4.3.2, Reference 5.3).
Page 3 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT 4.2 Send a memo to the Training Manager documenting these changes. Simulator training should focus on the important Human Error Probabilities. Scenarios should be developed to ensure that these specific items are taught and practiced. Training should compare new results with those contained in TRPR 33.0 Appendix D and G (Reference 5.4).
4.3 Copy the Operations Manager and supervisor in charge of Operations procedures on the memo.
4.4 Copy the Maintenance Rule Coordinator on the memo and notify of any changes that may affect the list of risk significant systems of list of components that should be considered risk significant.
4.5 If Human Reliability Analysis suggests that an important procedure can be improved to significantly reduce the human error probability, then submit a PBF-0026p. Document Feedback, to process the recommended changes. (B-2) 4.6 If training on particular human interaction would significantly improve the performance of the action, then submit a PBF-6101, Training Request. (B-2) 4.7 Review PRA model changes for impact on other PRA applications and risk informed programs.
4.8 Submit PBF-6101, Training Request, to provide training on changes to the PRA model which had significant changes in overall results or risk significant rankings as a minimum, to the following personnel:
- Senior Plant Management
"* Operations Control Room Staff
"* Shift Technical Advisors
"* Maintenance Rule Coordinator
"* Work Week Supervisors
"* System Engineering Page 4 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT
5.0 REFERENCES
5.1 ESG 5.1, PRA Maintenance and Update Guideline 5.2 RCE 01-069, Increased CDF in AFW PRA Model due to Procedural Inadequacies related to Loss of Instrument Air, May 14, 2002 5.3 OM 4.3.2, EOP/AOP Verification/Validation Process 5.4 TRPR 33.0, Licensed Operator Requalification Training Program.
6.0 BASES B-1 CA003704, Probabilistic Risk Assessment PRA for Auxiliary Feedwater AFW (Procedures)
B-2 CA003705, Probabilistic Risk Assessment PRA for Auxiliary Feedwater AFW (Forms)
Page 5 of 5
Page I of 3 Nuclear Management Company STATE CHANGE HISTORY SWork Quality Check Assign Conduct Complete Review & Approved Initiate Assign Work Work Approval 7/2/2002 2/5/2002 615/2002 7/2/2002 2 33.21 PM 4 09 18 PM 504 19 PM 1.54 50 PM Owner PBNP Owner RICK Owner RICK by RICK by RICHARD by RICK Owner RICK by RICK CAP Admin WOOD WOOD WOOD FLESSNER WOOD WOOD ** WOOD T
Complete and Close Done 7/10/2002 12.14 10 PM Owner (None) by MARYBETH %IV ARNOLD SECTION 1 Activity Request Id: CA003705 Corrective Action Submit Date: 2/5/2002 4:09:18 PM Activity Type:
Site/Unit: Point Beach Common Activity Requested: Revise the procedure governing PRA updates to include identification of the formal methods to be used for providing information to other groups. Use of existing processes, such as training work requests and procedure feedback forms, should be used whenever possible.
N Initiator: FLESSNER, RICHARD 0 CATPR:
EPP Engineering Initiator Department: EX Engineering Responsible Group Code:
Programs PRA PB Processes PB Activity Supervisor: RICK WOOD fý Responsible Department: Engineering =
Activity Performer: RICK WOOD 9 SECTION 2 Priority: 3 Due Date: 7/3/2002 N
(None) Management Exception From Pi?:
0 Mode Change Restraint:
"0Licensing Review?: N 0 QA/Nuclear Oversight?: N NRC Commitment?: N "0NRC Commitment Date:
SECTION 3 Activity Completed: 1/18/2002 12:52PM - LARRY PETERSON:
Due date extended as requested and approved by F. Cayia in prior update. Retruned to R.
flessner for completion.
1/18/2002 12:54PM - LARRY PETERSON:
Reassigned to R. Flessner for completion following extension.
6/5/2002 5:04:19 PM - RICK WOOD.
New NP is in draft. Comments from technical reviewer will be incorporated 6/19/2002 10:03.13 AM - RICK WOOD.
The procedure is in typing following incorporation of reviewrs' comments. The procedure is expected to be issued on 6/26/2002. The Engineering Director has apprved the 2 week https://nmc.ttrackonline.com/tmtrack/tmtrack.dll?IssuePage&TableId= 1000&Recordld= I I (... 9/20/2002
Nuclear Management Company Page 2 of 3 extension 7/2/2002 1:54:50 PM - RICK WOOD:
NP 7.7.20 Probabilistic Risk Assessment was issued on 6/26/2002. This procedure includes description of the existing processes for making changes to procedures and training 7/10/2002 12:14:10 PM - MARYBETH ARNOLD.
NP 7.7.20, Revision 0 was issued on 06/26/02 and Steps 4.5 and 4.6 specifically tie to this CA and the Bases include this CA as to why this procedure was written. CLOSED.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 SProject: CAP Activities &
Actions
- State: Done SActive/Inactive: Inactive (None) AR Type: Daughter 0 Owner:
RICHARD Assigned Date: 6/5/2002
" Submitter:
FLESSNER
" Last Modified Date: 7/10/2002 12:14:10 0 Last Modifier: MARYBETH ARNOLD PM 0 Last State Changer: MARYBETH ARNOLD
" Last State Change Date: 7/10/2002 12:14:10 PM
" Close Date: 7/10/2002 12:14:10 PM 0 One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 1
References:
CR 01-2278 RCE 01-069 GOOD CATCH NP 7.7.20, Revision 0 Update:
Import Memo Field:
PBNP CAP Admin Site: Point Beach CAP Admin:
OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS
£. ACE000314_ Probabilistic _Risk Assessment PRA For Auxdiiar _FeedweaterS m_AFW dý' CAP001415 Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https:Hlnmc.ttrackonline.comltmtrackltmtrack.dil?IssuePage&Tableld= 1000&Recordld= If ... 9/20/2002
Nuclear Management Company Page 3 of 3 https:H/nmc.ttrackonline.comntmtrackltmtrack.dll?IssuePage&TableId= 1000&RecordId= 11 (... 9/20/2002
5o NP 7.7.20 PROBABILISTIC RISK ASSESSMENT DOCUMENT TYPE: Administrative REVISION: 0 EFFECTIVE DATE: June 26, 2002 APPROVAL AUTHORITY: Department Manager PROCEDURE OWNER (title): Group Head OWNER GROUP: Engineering
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT TABLE OF CONTENTS SECTION TITLE PAGE 1.0 PURPOSE ...................................................................................................................... 3 2.0 RESPON SIBILITIES ..................................................................................................... 3 3.0 D ISCU SSION ................................................................................................................ 3 4.0 PROCEDURE ................................................................................................................ 3 5.0 REFEREN CES ......................................................................................................... 5 6.0 BASES ........................................................................................................................... 5 Page 2 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT 1.0 PURPOSE The procedure establishes interface requirements between Programs Engineering - PRA and Training, Licensing and Operations.(B-1) 2.0 RESPONSIBILITIES 2.1 PRA staff: Ensure that the appropriate memos are developed following update of the PRA model. Identify risk significant Human Interactions and forward to Operations and Training as they are identified.
2.2 Supervisor PRA: Review the memo and information sent to applicable groups.
2.3 Operations Training Supervisor: Incorporate information from PRA into Licensed Operator Training.
2.4 Operations Procedures Supervisor: Review information from PRA and identify procedure changes 3.0 DISCUSSION 3.1 The update of the PRA model is controlled via ESG 5.1 PRA Maintenance and Update Guideline.
3.2 Human Interactions are classified as three types: Type A are interactions occurring before the initiating event: Type B are interactions associated with the initiating event; Type C are interactions associated with response to the initiating event. The focus of this procedure is Type C Human Interactions.
3.3 The EOP Verification and Validation Matrix was developed with a cutoff of an Initiating Event frequency greater than I E-3 /year and a Core Damage Frequency of 1 E-6/year.
4.0 PROCEDURE 4.1 Following periodic update of the PRA model, notify the Training Group of significant changes to:
4.1.1 System Importance 4.1.2 Initiating Event frequency 4.1.3 Human Error Probabilities and Importance 4.1.4 EOP Verification and Validation Matrix (OM 4.3.2, Reference 5.3).
Page 3 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT 4.2 Send a memo to the Training Manager documenting these changes. Simulator training should focus on the important Human Error Probabilities. Scenarios should be developed to ensure that these specific items are taught and practiced. Training should compare new results with those contained in TRPR 33.0 Appendix D and G (Reference 5.4).
4.3 Copy the Operations Manager and supervisor in charge of Operations procedures on the memo.
4.4 Copy the Maintenance Rule Coordinator on the memo and notify of any changes that may affect the list of risk significant systems of list of components that should be considered risk significant.
4.5 If Human Reliability Analysis suggests that an important procedure can be improved to significantly reduce the human error probability, then submit a PBF-0026p. Document Feedback, to process the recommended changes. (B-2) 4.6 If training on particular human interaction would significantly improve the performance of the action, then submit a PBF-6101, Training Request. (B-2) 4.7 Review PRA model changes for impact on other PRA applications and risk informed programs.
4.8 Submit PBF-6101, Training Request, to provide training on changes to the PRA model which had significant changes in overall results or risk significant rankings as a minimum, to the following personnel:
"* Senior Plant Management
"* Operations Control Room Staff
"* Shift Technical Advisors 0 Maintenance Rule Coordinator
"* Work Week Supervisors
- System Engineering Page 4 of 5
POINT BEACH NUCLEAR PLANT NP 7.7.20 PROCEDURES MANUAL Revision 0 June 26, 2002 PROBABILISTIC RISK ASSESSMENT
5.0 REFERENCES
5.1 ESG 5.1, PRA Maintenance and Update Guideline 5.2 RCE 01-069, Increased CDF in AFW PRA Model due to Procedural Inadequacies related to Loss of Instrument Air, May 14, 2002 5.3 OM 4.3.2, EOP/AOP Verification/Validation Process 5.4 TRPR 33.0, Licensed Operator Requalification Training Program.
6.0 BASES B-i CA003704, Probabilistic Risk Assessment PRA for Auxiliary Feedwater AFW (Procedures)
B-2 CA003705, Probabilistic Risk Assessment PRA for Auxiliary Feedwater AFW (Forms)
Page 5 of 5
Nuclear Management Company Page 1 of 3 STATE CHANGE HISTORY L. Work Review & Reject Conduct Work Initiate Assign Work Assign Conduct Work Complete Approval 3/12/2002 3/14/2002 4/8/2002 4/9/2002 9.32 54 AM 12.07:58 PM 3 28.02 PM 7.43.27 AM Owner Owner DON Owner by DENNIS by DENNIS Owner DON by JULIE DENNIS PETERSON by DON DENNIS HETTICK HETTICK PETERSON KREIL HETTICK T't* PETERSON HE'TICK V:
9 L Work Review & ZL Complete an d 5 Approval Approved Quality Check Close Done Complete 5/1/2002 4/11/2002 4/12/2002 11:54 50 AM 11.29 44 AM 11:31:52 AM Owner Owner PBNP Owner by DENNIS (None) by DON DENNIS CAP Admin by MARYBET H HETTICK
"[77 ARNOLD P PETERSON HETTICK SFr SECTION 1 Activity Request Id: CA003982 Activity Type: Corrective Action Submit Date: 3/12/2002 9:32:54 AM Site/Unit: Point Beach - Common Activity Requested: Per CARB Meeting of 3/05/2002 (NPM 2002-0112), Review SEN 174 response (from RCE 01 069, which is ACE000314 in tTrack). This SEN is discussed on page 26 of RCE 01-069. Re Open the OE items if questions about the procedures for ensuring adequate pump flow is maintained, are no fully addressed, including pumps other than AFPs.
0 CATPR: N Initiator: MASTERLARK, JAMES t EPN Engineering Responsible Group Code: AP Performance Initiator Department:
Programs Nuclear Assessment PB ZZ Safety Analysis PB Activity Supervisor: DENNIS HETTICK Responsible Department: Assessment Activity Performer: DON PETERSON 6ý SECTION 2 Priority: 3 Due Date: 4/2512002
"<Mode Change Restraint: (None) Management Exception From PI?: N
" QA/Nuclear Oversight?: N 0 Licensing Review?: N NRC Commitment?: N 0 NRC Commitment Date:
"0Significance Level: A SECTION 3 Activity Completed: 3/17/2002 1:59PM - DON PETERSON:
SEN 174 has six completed actions directed at the need to develop procedures for off- mormal events, to restore power and recover equipment for non-vital 4160 & 480 V busses and associated MCCs. Action six was closed out to CR 98-0050 action item #43. Action #43 was closed to the issuance of AOP-18. Pump flow concerns were not directly identified in the action items for SEN 174.
https://nmc.ttrackonline.com/tmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld=24(... 9/20/2002
Page 2 of 3 Nuclear Management Company 418/2002 3:28PM - DON PETERSON:
items, CR 97-1992, CR 98-0050, The following documents were reviewed SEN 174 actions Pump flow concerns were not directly identified in any of AOP-18, AOP-18A and RCE 01-069.
with Mr. Mark Rinzel, Corrective the above documentation. This concern was discussed in t-Track should be issued he was in atgreement, that an action Action Liaison for Operations, on adequate pump flows 174 with special focus to Operations to revisit the issues of SEN 4/8/2002 3:31PM - DON PETERSON:
on "How does PBNP maintain Issue an action to Operations; Review SEN 174, focusing SEN 174. This action was discussed adequate pump flow, under the conditions descnbed in him.
with Mr. Mark Rinzel, he has requested that it be sent to 4/11/2002 11:54AM - DON PETERSON:
Mr. Duane Schoon.
CA004279 was created and sent to the Operations Group!
5/1/2002 11:31AM - MARYBETH ARNOLD:
action created (CA004279) for The response to SEN 174 was reviewed with one follow up Operations to look at a specific item. CLOSED.
SECTION 4 (None) Licensing Supervisor: (None)
QA Supervisor:
SECTION 5
"* Project: CAP Activities & Actions Done O Active/Inactive: Inactive
" State: Parent (None) AR Type:
"* Owner: 3/14/2002 JULIE KREIL Assigned Date:
"0Submitter: MARYBETH ARNOLD 5/1/2002 11:31:52 AM 0 Last Modifier:
0 Last Modified Date:
MARYBETH ARNOLD 0 Last State Change Date: 5/1/2002 11:31:52 AM 0 Last State Changer:
0 Close Date: 511/2002 11:31:52 AM System AFW 0 One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater CR 01-3595 NUTRK ID:
Child Number: 0 CR 01-2278
References:
RCE 01-069 GOOD CATCH SEN 174 CR 98-005 AOP-18 CR 97-1992 CR 98-0050 AOP-18A NPM 2002-0112 Part I, Revision 0, of CR 01 Update: b\(20011204 PB2171 JMK1) Operability Determination (OD) 11/30/01. Operable But Degraded - or Operable But Nonconforming 3595 was approved on measures ARE required.
meets the minimum required level of performances, compensatorywas approved on 12/01/01.
\\Operability Determination (OD) Part 1, Revision 1 of CR 01-3595
- meets the minimum required level Operable But Degraded - or Operable But Nonconforming of performances, compensatory measures ARE required.
1000&Recordld=24(... 9/20/2002 https://nmc.ttrackonline.com/tmtrack/tmtrack.dli ?issuePage&Tableld=
Nuclear Management Company Page 3 of 3 Accepted into group and assigned priority 3. This questions the adequacy of an SEN applicability determination and evaluation. Per NP 5.4.1, SEN are to be priority 3.
Import Memo Field:
CAP Admin: PBNP CAP Admin Site: Point Beach OLDACTIONNUM:
Cartridge and Frame:
NOTES/COMMENTS
- *'l*it,*¸ * * ......
Note created during 'Reject' transition by DENNIS HETTICK (4/9/2002 7:43:27 AM)
Specify the action number that was created to perform the review discussed in the action section.
ATTACHMENTS AND PARENTICHILD LINKS 4- Linked From CAP001415 E CA004279: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https:Hlnmc.ttrackonline.com/tmtrack/tmtrack.dll?lssuePage&Tableld= 1000&Recordld=24(... 9/20/2002
INPO SOERs, SERs, SENs, OEs SEN 174 F CLOSED UNIT: 0 SYSTEM: XX INITIATED: 11/10/97 CLOSED: 10/24/00 MSS #:
IR: HENRY JOYCE ADMINISTRATOR: JAMES PULVERMACHER ISSUE MANAGER: BRIAN OGRADY NI.t JF OPEN ACTIONS : 0 NUMBER OF CLOSED ACTIONS : 6 TOTAL NUMBER OF ACTIONS : 6 LOSS OF NONVITAL BUS CAUSES DUAL UNIT SCRAM AND DEGRADED AUXILIARY FEEDWATER SYSTEM DESCRIPTION:
- SEE E-MAIL CONF "NP-INPO-NETWORK-IS" FOR FULL TEXT *
Subject:
SEN 174, Loss of Nonvitat Bus Causes Dual Unit Scram and Degraded Auxiliary Feedwater System Novembter 10, 1997 Description On Septemrber 6, 1997, both McGuire units automatically scrammed from 100 percent power when the alternate supply breaker to nonsafety-related 120- volt AC instrument and control power bus KXA opened, stripping control power to several important plant components in each unit. The loss of nonvital power caused Unit 1 main feedwater pumps to trip, resulting in a turbine trip and automatic reactor scram.The loss of nonvitaL power on Unit 2 caused the main steam isolation valves to close, resulting in an automatic reactor scram on high pressurizer pressure. Aboutan hour later, power was restored to bus KXA, and affected secondary systems were subsequently returned to service.
At the time of the event, bus KXA was energized from its alternate transformer power supply while the inverter battery was undergoing an equalizing charge.This alternate alignment is normally used only during this annual equalizing charge.The breaker supplying the bus that was feeding both units opened because a Loose cable connection on the Load sideof the breaker generated enough heat to actuate the thermal trip unit. No preventive maintenance had ever been performed on the breaker. Station personnel believe that the Loose connection had existed since construction.
Power was Lost to various equipment in each unit.The most significant effects were in Unit 1 and included the following:
oPower was Lost to the solenoid-operated recirculation valves for all three auxiliary feedwater (AFW) pumps and to the main control board indication for these vaLves.To provide adequate pump cooling, the motor- driven AFW pumnps require a minimum of 100 gallons per minute (gpm) flow, and the turbine driven AFW pump requires 200 gpm. As water level is recovered in the steam generators and the operator manually throttles back AFW flow, the recirculation valves are designed to open automatically to solenoid provide the minimum flow through each pump. However, these valves fail closed by design.With power lost to both the both
'yes and their associated indicators on the main control board, the AFW pumps were operated for 20 to 60 minutes with recirculation valves and main flow control valves closed. However, leakage through the flow control valves resulted in period of operation was AFW pump
.. roximately 12 gpm flow through each pump.Only because of this Leakage and the limited damage caused by overheating precluded.
oPressurizer power-operated relief valve automatic control was Lost, but manual operability was not affected.
oNormal and excess Letdown flow capability was lost, potentially affecting the ability to prevent overfilling the'pressurizer.
oCapability to perform normal containment air releases was lost, resulting in slight containment pressurization.
Significant aspects of this event include the following:
automatic oAn installation deficiency on the alternate power supply breaker resulted in both units experiencing simultaneous scrams and lost availability.
oThe design of the AFW power supply represents a common cause failure mechanism where a Loss of power to the nonsafety-related on the main bus resulted in both a loss of power to the AFW pump recirculation valve solenoids and the associated indication control board.
with oThere was no procedure specific to the Loss of nonvitat buses for the operators to use during the event. Consequently, valve position indication, operators were not aware of the potential for damaging the AFW pu*ps.A the toss of recirculation list of loads supplied by the nonsafety-related bus was not readily available for operators.
oOperators had received classroom training on the effects of a toss of nonvital buses in initial licensed operator continuing training.
training.However, they had notreceived subsequent simulator training on a loss of nonvital buses during final procedure had The need for a procedure to help mitigate this transient had been identified, and a draft written, but a not been issued.
or associated breakers oNo preventive maintenance activities had been established for the auxiliary control power system bus was considered but not because both units would have to be shut down to deenergize the bus.On-line preventive maintenance performed because of personnel safety issues.
bus (KXB), resulting oA similar event occurred at Unit 2, on September 6,1987, involving a Lossof power to the other nonvitat fault in an instrument air compressor. However, the investigation into that event concentrated on from an overcurrent in the alternative venting similar compressor motor faults. An opportunity was missed to identify the risk of being the AFW pumps jnment and the need for preventive maintenance, operating procedures, and training.The potential for damage to of battery
. the common-causefaiLure was identified as a concern during station blackout (potential foreventuat Loss operating procedures.
power); however, operator actions for other nonvital bus failure scenarios were not addressed in abnormal to Losing the alternate power oDuring work preparation and planning activities, station personnel focused on minimizing risks supply during the maintenance activity. Placing the bus on its alternate power supply wasnot considered a significant risk
prejob brief evolution and the prejob brief primarily emphasized protecting bus KXA from being bumped by station personnel.The did not adequately inform the operators of the necessary contingency actions needed if bus KXA was lost.
referenced The event described in this significant event notification (SEN) was screened significant by INPO.The documents
"-'ow are sufficiently detailed such that INPO does not intend to publish a separate significant event report (SER); therefore, ities should review this event notification and implement corrective actions where necessary to avoid similar events.
and Operation References 1. NRC Licensee Event Report (LER) 369/97-09, "Reactor Trip on Both Units Due to an Equipment Failure Prohibited by Technical Specifications Due to Failure to Comply with Required Action Statement," October 6, 1997 Motor - Caused Loss
- 2. NRC LER 370/87-16, revision 1, "Reactor Trip Due to Overcurrent Faults in an Instrument Air Compressor of Power to a Main Turbine Control System Relay," December 16, 1987 Reactor Type Plant Information Unit:McGuire Nuclear Station Unit 1 (Duke Power Company) Year Commercial: December 1, 1981 Power Company (Size): PWR (1,180 HMe) Reactor Manufacturer:Westinghouse Turbine Manufacturer:Westinghouse Plant Designer:Duke Event Date:September 6, 1997 Pump Equipment Information Name and Size: Two Motor-Driven Centrifugal Pumps (450 gpm capacity) One Steam-Driven Centrifugal (900 gpm capacity Event Criteria Unusual Plant Transient Deficiency Installation Deficiency Maintenance Deficiency Procedure Deficiency Training Deficiency Design not scheduled/performed) Written Cause Categories Construction (improper installation) Work Organization/PLanning (maintenance Procedure (lack of procedure)
Training/Qualification (lack of training) Design Configuration (inappropriate layout of systems or subsystems) as AMl) due to failing the plant Level Malfunctioning Systems The 240/120-volt AC auxiliary control power system was classified performance criterion for reactor trips.
Attachments Station Unit 1). Utilities and This document is based on technical information provided by Duke Power Company (McGuire Nuclear are requested to provide feedback on similar occurrences and solutions at theirplants or on their equipment to the participants
"* ormation contact listed below.
.nited Distribution use. Reproduction of this report Copyright 1997 by the Institute of Nuclear Power Operations. Not for sale nor for commercial is a violation of applicable Law.
without the prior written consent of INPO is expressly prohibited. Unauthorized reproduction and participant may reproduce this document for its business use. This document should not be Each INPO member without the prior agreement otherwise transferred or delivered to any third party, and its contents should not be made public, of INPO. All other rights reserved.
Notice (INPO). Neither This information was prepared in connection with work sponsored by the Institute of Nuclear Power Operations participants, nor any person acting on behalf of them: (a) makes any warranty or representation, INPO, INPO members, INPO contained in this document, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information method, or processdisclosed in this document may not infringe on privately owned or that the use of any information, apparatus, from the use of any information, rights; or (b) assumes any liabilities with respect to the use of, or for damages resulting apparatus, method, or process disclosed in this document.
Distribution System, Bus Keywords Auxiliary Feedwater System, Auxiliary Feedwater Pump, Loss of AC Power, Electrical Tetecopy No.: (770) 644-8594 Information
Contact:
Brett Kruse, (770) 644-8729, krusebaainponn.org
- SEE E-MAIL CONF. "NP-INPO-NETWORK-IS" FOR FULL TEXT
- STATUS UPDATE:
4; and wi[l be tracked under the (20001024 WE1384 JRPI) Changes to plant Abnormal Operating Procedures have been submitted referenced Condition Reports.
SCREENED BY : DATE: COMMITMENT ................ (Y/N): N TS VIOLATION .............. (Y/N): 10 CFR 21 ................. (Y/N):
REGULATORY REPORTABLE ..... (Y/N): (A N P R W):
(Y/N): OPERABILITY IMPACT PER TS.(Y/N): ACTION .............
TS LCO ENTRY .............
SIGNIFICANCE ........ (A B C 0): OPERABILITY DETERMINATION.(Y/N):
" REVIEW REQUIRED ...... (Y/N):
b.r. ING DETERMINATIONS:
CR 97-1992 #2 CR 98-0050 #43 CR 98-0050
REFERENCES:
TWR 97-337 ACTION NUMBER 1 D". DUE DATE: 01/10/98 PRIORITY: -100 EXTENSIONS MADE: 0 CREATED : 11/10/97 OER TOM SHELEY RECEIVED: 11/14/97 EIS TOM JESSESSKY WORK DONE: 12/10/97 KELLY HOLT APPROVED: 01/06/98 TOM JESSESSKY VERIFIED : 01/07/98 HENRY JOYCE CLOSED : 01/07/98 TODD COOPER Evaluate for applicability to PBNP in accordance with NP 5.3.2. Identify and initiate any necessary corrective actions. coordinate response with Operations.
(11/12/97 TPS) Issued to Group: EIS Per Conversatioin with Kelley Holt. Kelley will evaluate the McGuire station event aginst PBNP's configuratioin. Note action item has a 60 day due date.
(11/14/97 TJJ) Received Action into Group: EIS Responsible Person: KJH:KELLY HOLT Due Date: 01/10/98 (12/10/97 KJH) Passed to TOM JESSESSKY for acceptance of work.
(01/06/98 TJJ) Passed to HENRY JOYCE for Verification.
Point beach has four safety-related instrument bus trains for each Unit, which are normally supplied from static inverters.
These instrument busses are designed to be automatically transferred to a non safety-related backup supply upon an inverter failure. An 8-hour LCO is in effect whenever a safety related instrument bus is being supplied from the backup source. The backup source is designed to be used only to prevent the Loss of power to an instrument bus in the event of an inverter failure. The backup source supplies power only until the affected busses are manually aligned to an inverter supply. Safety related alternate inverters are available to take the place of the normal inverters during routine maintenance or repair of the normal inverters. A swing safety-related battery is available to take the place of any of the normal safety-reLated batteries to allow for discharge testing or equalizing charges.
Point Beach has two non safty-related instrument busses for each unit. These busses are supplied from offsite power through transformers. One of these busses for each unit is supplied from a bus with a diesel backup supply.
Tabulations of the loads supplied from the safety-related and non safety-related busses are available to control room personnel. Operators receive training on the effects of the loss of power to these busses.
- of the instrument bus breakers have been replaced within the last four years. Connections were torqued upon breaker lacement. A program is i n place to perform breaker testing every five years. An analysis to determin the effects of the
.s of power to many instrument bus loads was completed as part of the breaker replacement effort.
A detailed analysis of the effects of the loss of power to each instrument bus at Point Beach is not available to Operations if personnel. A new action item could be initiated to complete this analysis and provide additional Operator training, necessary. The PLA should review the operations evaluation to determine if Operations needs the detailed analysis.
This evalutation item is complete. No further actions required (with possible exception noted in paragraph above).
(01/07/98 HAJ) PLA Closure of Item.
see update field. Need for further action will be determined based on Operations response.
initial (10/22/98 TPS) After commxnicating with the new system engineer (J. Matloy), and Tom Garotit it was recognized that the was response within action item #1 from the system engineer was adequate in identifying that no immediate corrective action required outside of a non essential "recommendation" that a Loss of instrument buss would be "beneficial" to the operators.
Both the new system engineer and the electrical crew DSS (Tom Garot) agree that a recommendation for a procedure to be developed to support the loss of an instrument buss is important but is not necessary to support closure of this SEN evaluation.
the To generate the development of such a procedure a procedure feed back form has been generated by this evaluator to track recommendation:
"Develop a procedure for operator response to a toss of instrument buss. Reference SEN 174 event."
ACTION NUMBER 2 DONE DUE DATE: 01/10/98 PRIORITY: 3 EXTENSIONS MADE: 0 CREATED : 11/10/97 OER JIM SCHWEITZER RECEIVED: 12/05/97 OPS JOHN ANDERSON
' NE: 08/17/98 THOMAS GAROT APPROVED: 08/18/98 RICHARD MENDE 10/02/98 TODD COOPER CLOSED : 10/02/98 MICHAEL ROACH corrective actions. Coordinate Evaluate for applicability to PBNP in accordance with NP 5.3.2. Identify and initiate any necessary response with Engineering.
(11/24/97 JGS) Issued to Group: EIS
I Tom please assign this for evaluation. Information on the event is included in the parent document.
(12/05/97 KMY) Received Action into Group: EIS Responsible Person: WAH:BILL HENNIG Due Date: 01/10/98
'10/97 WAH) Changed Responsible Person: From (WAH) to (HAJ)
.nged Responsible Group: From (EIS) To (OER).
Changed Responsible section: From (SEN) To (OAS).. Action item 2 is supposed to be assigned to Operations with the intention to coordinate with Engineering (Kelly Holt has been assigned action item 1). This is per direction of TJ Jessessky.
(12/11/97 FPH) Changed Responsible Person: From (HAJ) to CRGM)
Changed Responsible Group: From (OER) To (OPS).
Changed Responsible section: From (OAS) To (PRD)..
(12/15/97 TPS) Changed Responsible Person: From (RGM) to (TWG). Task assigned per "E" mail communication from engineering (K.
Holt).
of NP 5.4.1,Attachment (05/15/98 SJN) Set Work Priority to 3. Significance level 3 assigned by the BST based on the guidance B. The reason for the assigned significance is that this item is an evaluation of a SEN.
restore power and (08115/98 TWG) Have not seen evaluation on this issue. However, I recommend procedures be developed to for the vital busses and recover equipment for the Non-vitaL 4160 + 480 volt busses and associated MCCs. We have AOPs in place all the Vital and also for the DC system. The procedure upgrade project is developing System Operating Procedures for instrument busses 1+2 Y05 are non-vital busses but these are designed for planned outages not off-normal events. The non-vital powered from B-41s which are non-vital MCCs.
(08/17/98 TWG) Passed to RICHARD MENDE for acceptance of work.
(08/18/98 RGM) Passed to MICHAEL ROACH for Verification.
Completed review and provided recomendations.
(10/02/98 TAC) PLA Closure of Item.
evaluation. No further actions Action #4 has been created and sent to OPS for the creation of the procedures discussed in this were identified as being required. This action item may be closed.
ACTION NUMBER 3 DUE DATE: 01/10/98 PRIORITY: -100 EXTENSIONS MADE: 0 DONE OER HENRY JOYCE RECEIVED: 11/13/97 TRPSA LARRY EPSTEIN CREATED : 11/10/97 WORK DONE: 01/06/98 MARK RINZEL APPROVED: 01106/98 MARK RINZEL HENRY JOYCE CLOSED : 01/08/98 HENRY JOYCE VERIFIED : 01/08/98 evaluate "Prevent Events" section of this SEN for training appLicability.
(11/13/97 LDE) Received Action into Group: TRPSA Responsible Person: LDE:LARRY EPSTEIN Due Date: 01/10/98 (12/03/97 MDR) Changed Responsible Person: From (LDE) to (MDR).
in the (12/03/97 MDR) TWR 97-337 has been issued and actions 1 + 2 under the TWR will evaluate Training needs/enhancements Operations and ESP areas. Recommendations for action will be made based on these evaluations.
(01/06/98 MDR) Passed to LARRY EPSTEIN for acceptance of work.
(01/06/98 MDR) Passed to HENRY JOYCE for Verification.
TWR 97-337. The results of this This item was evaluated for applicable and potential inclusion into Training programs under for both SEN and evaluation show that there is some applicability to PBNP and it warrants inclusion into the group meetings it is recommended NES. This will be accomplished at the February group meetings as part of the OE discussions. Therefore, of SEN 174 are Events that an action item be issued to R. Bauer, with a due date of 3/31/98, to ensure that the Prevent for this item and it may be included in the upcoming SEN and NES group meetings/discussions. No further action is needed closed.
(01/08/98 HAJ) PLA CLosure of Item.
see update field and NUTRK TUR 97-337
REFERENCES:
TWR 97-337
ACTION NUMBER 4 DUE DATE: 12/31/98 PRIORITY: 4 EXTENSIONS MADE: 0 10/02/98 DER TODD COOPER RECEIVED: 10/22/98 OPS BRIAN OGRADY
- 6. NE: 06/24/99 STEPHEN GUCWA APPROVED: 06/24/99 JOHN ANDERSON VERIFIED : 06/24/99 JOHN ANDERSON CLOSED : 08/04/99 TODD COOPER Based on the evaluation conducted in child records #1 + #2, develop procedures, for off-normal events, to restore power and recove equipment for non-vital 4160 + 480 V busses and associated MCCs. Document actions taken in response to this item.
(10/22/98 TPS) Received Action into Group: OPS Responsible Person: TPS:TOM SHELEY Due Date: 12/31/98 (10/22/98 TPS) Set Work Priority to 4. INPO SEN for evaluation.
(10/22/98 TPS) After communicating with the new system engineer (J. Malloy), and Tom Garot it was recognized that the initial response within action item #1 from the system engineer was adequate in identifying that no immediate corrective action was required outside of a non essential "recommendation" that a toss of instrument buss and recovery procedure for non vital AC busses would be "beneficial" to the operators.
that PBNP safety related instrument busses are supplied by safeguards The differences between PBNP and the McGuire station is power and battery backup and McGuire's were not.
Although there is an alternative non safety related (non battery supported power supply to support instrument bus auto of a LCO transfers, this system is not employed unless a safeguards inverter fails. If this transfer occurs a IS declaration would be required.
(10/22/98 TPS) Passed to JOHN ANDERSON for acceptance of work.
(10/27/98 RGM) Returned to TON SHELEY for additional work.
this (10/27/98 RGM) I believe that ANSI requires procedures for these type of anticipated operational occurences and as such, item should not be closed.
(10/28/98 TPS) Changed Responsible Person: From (TPS) to (SGG). Attempts to close this action to procedure feed back submitta (not required for SEN closure) was not accepted. This is a AOP issue and needs to be resolved or corrected by the EOP / AOP procedure group.
'24/99 CAWI) Passed to JOHN ANDERSON for acceptance of work.
(06/24/99 JRA1) Passed to JOHN ANDERSON for Verification.
See CR update for additional information that justifies that all SEN corrective actions have been completed.
and recover equipment SEN 174 action item #4 identified a need to develop a procedure for off- normal events to restore power This action came out of SEN 174 action item# 2. After discussing for non vital 4160 and480 V busses and associated MCC's.
operators action item #2 closure with the responsible person (Tom Garot) he agreed that his recomnmendation was more of an of non safeguards / nor opinion rather then an action that must be completed to address the SEN 174 event (MCGuire station Loss battery supply inverters and critical control perimeters were Lost). Tom agreed that the submittal of a procedure feed back form would be adequate to support his recommendation.
A procedure feed back has been submitted to:
Develop an operations procedure for off-normal events to restore power and recover equipment for non vital 4160 and 480 V busses and associated MCC's.
This action can be closed.
ADDITIONAL INFORMATION:
Operations has evaluated this item and has determined that it is a long-term project that will take three years to complete.
to Operations for tracking It is recommended that this action item be closed and two more created. One action item will go purposes. The other action item should go to Engineering for support of this project.
(06/24/99 JRA1) Passed to TODD COOPER for Final Close Out.
Verified.
(08/04/99 TAC) PLA Closure of Item.
Additional child records opened as required by Issue Manager.
"CES: TWR 97-337
ACTION NUMBER 5 DUE DATE: 12/31/99 PRIORITY: 4 EXTENSIONS MADE: 0 r
DER TODD COOPER RECEIVED: 08/06/99 OPS BRIAN OGRADY 08/04/99 TOM SHELEY APPROVED: 08/12/99 BRIAN OGRADY Wt, jNE: 08/06/99 BRIAN OGRADY CLOSED : 08/13/99 TODD COOPER VERIFIED : 08/12/99 SYSTEM LOSS OF NONVITAL BUS CAUSES DUAL UNIT SCRAM AND DEGRADED AUXILIARY FEEDWATER off-normal events, to restore power and Based on the decision of the Issue Manager in child record #4, develop procedures, for V busses and associated MCCs. Document actions taken in response to this item.
recover equipment for non-vitaL 4160 + 480 (08/06/99 TPS) Received Action into Group: OPS Responsible Person: TPS:TOM SHELEY Due Date: 12/31/1999 (08/06/99 TPS) Set Work Priority to 4. Action supports station and department goals.
(08/06/99 TPS) Passed to BRIAN OGRADY for acceptance of work.
(08/12/99 BJ01) Passed to BRIAN OGRADY for Verification.
appears to be limited value in tracking an action Discussions with the new Operations Manager has identified that this there developing procedures for recovery of non safety related busses if it has already beenidentified in action item 91#2 item for is required to support the SEN, and the action is only tracking a procedure feed back recommendation.
that no corrective action related buses is already being tracked in The recommendation for the development of recovery procedures for NON safety operations to support other concerns / investigations. Both CR 97-1992 #2 (no AOP for Seismic Events) priority #4, and CR need for such procedure development.
98-0050 #43 (loss of offsite power) priority #2. are targeting the An update has been placed in both CR's identifying a This action can be closed to both CR 97-1992 #2 and CR 98-0050 #43.
reference to SEN 174 #5 as a reference.
This item can be closed.
(08/12/99 BJ01) Passed to TODD COOPER for Final Close Out.
This item can be closed.
(08/13/99 TAC) PLA Closure of Item. No additional actions are required.
-, needed procedure development wilt be tracked under CR 97-1992 #2 and CR98-0050 #43.
CR 97-1992 #2 CR 98-0050 #43
REFERENCES:
TWR 97-337 ACTION NUMBER 6 DUE DATE: 09/15/00 PRIORITY: 4 EXTENSIONS MADE: 2 DONE TODD COOPER RECEIVED: 08/05/99 SDE MICHAEL ROSSEAU CREATED : 08/04/99 OER MICHAEL ROSSEAU APPROVED: 08/04/00 MICHAEL ROSSEAU WORK DONE: JAMES PULVERMACHER BRIAN OGRADY CLOSED : 10/24/00 VERIFIED : 08/18/00 FEEDWATER SYSTEM LOSS OF NONVITAL BUS CAUSES DUAL UNIT SCRAM AND DEGRADED AUXILIARY in the development of procedures, for off-normat Based on the decision of the Issue Manager in child record #4, assist Operation Document actions taken in to restore power and recover equipment for non-vital 4160 + 480 V busses and associated MCCs.
events, response to this item.
(08/05/99 LJA1) Received Action into Group: SDE No Priority Assigned Responsible Person: KJN1:KEN NETZEL Due Date: 12/31/1999 (08/30/99 KJN1) Set Work Priority to 4.
(12/20/99 KJN1) Changed the Due Date from: 12/31/1999 to 04/01/2000 are working on modifications or higher priority NUTRK This item will not be worked in the near term. All electrical personnel items.
(03/31/00 KJN1) Changed the Due Date from: 04/01/2000 to 09/15/2000 (08/04/00 MJR1) Passed to BRIAN OGRADY for Verification.
for non-vital bus recovery. The OPS action item for SEN This action item states to assist OPS in the developement of new AOPs This action item may be closed with no further actions required as a NUTRK item is not 174 was closed to CR 98-0050 #43.
-essary for one group to support another.
,/18/00 BJO1) Passed to DAVID GARCIA for Final Close Out.
Close item (10/24/00 JRPI) PLA Closure of Item.
Close to actions of the referenced Condition Report CR 98-0050.
REFERENCES:
TWR 97-337 CR 98-0050 CR 98-0050 #43 Page 1 of 4 Nuclear Management Company STATE CHANGE HISTORY 0 1Ll Conduct Work Review & Approved Initiate Assign Complete Approval Quality Check E*5 Assign Work 4/111/2002 11:52.53 AM Work 4/12/2002 6.40 15 AM 413012002 4:37.57 PM 5/13/2002 2.22:07 AM Owner PBNP Owner DUANE by TOM Owner DUANE by DUANE by DON Owner MARK by MARK CAP Admin SCHOON SHELEY RINZEL SCHOON SCHOON PETERSON RINZEL p,
NZ IF, Complete and Close Done 5/13/2002 2 04:47 PM Owner (None) by JULIE KREIL SECTION 1 Activity Request Id: CA004279 Activity Type: Corrective Action Submit Date: 4/111/2002 11:52:53 AM Site/Unit: Point Beach - Common Activity Requested: Re-Open the evaluation of SEN 174, ensuring that questions about the procedures for ensunng adequate pump flow is maintained, are fully addressed, including pumps other than AFPs Action is out of CA 3982 where CARB (3/5/02) while reviewing RCE 01-69 /ACE 314 requested a reopening of SEN 174 to specifically adress a question if procedures for ensuring adequate pump flow is maintained (possibly this point was not adequatly documented in the SEN) and discuss other pumps other then AFPs. TPS N Initiator: MASTERLARK, SCATPR: JAMES PO PB Operations PB Initiator Department: EPN Engineering Responsible Group Code:
Programs Nuclear Safety Analysis PB Activity Supervisor: DUANESCHOON Responsible Department: Assessment Activity Performer: MARK RINZEL 1*
SECTION 2 3 Due Date: 5/10/2002 Priority:
Management Exception From P1?: N "OMode Change Restraint: (None)
N "0QA/Nuclear Oversight?: N " Licensing Review?:
NRC Commitment?: N " NRC Commitment Date:
0 Significance Level: A SECTION 3 Activity Completed: 3/17/2002 1:59PM - DON PETERSON:
SEN 174 has six completed actions directed at the need to develop procedures for olf- mormal events, to restore power and recover equipment for non-vital 4160 &480 V busses and associated MCCs. Action six was closed out to CR 98-0050 action item #43. Action #43 was https://nmc.ttrackonline.con/tmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld=94(". 9/20/2002
Nuclear Management Company Pace 2 of 4 closed to the issuance of AOP-18. Pump flow concerns were not directly identified in the action items for SEN 174.
4/812002 3:28PM - DON PETERSON:
The following documents were reviewed: SEN 174 actions items, CR 97-1992, CR 98-0050, AOP-18, AOP-18A and RCE 01-069. Pump flow concerns were not directly identified in any of the above documentation. This concern was discussed with Mr. Mark Rinzel, Corrective Action Uaison for Operations, he was in atgreement, that an action in t-Track should be issued to Operations to revisit the issues of SEN 174 with special focus on adequate pump flows.
4/8/2002 3:31PM - DON PETERSON.
Issue an action to Operations; Review SEN 174, focusing on "How does PBNP maintain adequate pump flow, under the conditions described in SEN 174 This action was discussed with Mr. Mark Rinzel, he has requested that it be sent to him.
4/30/2002 4:36PM - MARK RINZEL Corrective Action (CA) 4279 re-opened an evaluation of INPO SEN 174, "Loss of Non-Vital Bus Causes Dual unit SCRAM and degraded Auxiliary Feedwater System". The evaluation was re-opened based on a CARB request from 3/5/02 review of RCE 01-069, "Increased CDF in AFW PRA Model Due to Procedural Inadequacies Related to Loss of Instrument Air". The CARB requested this evaluation be re-opened to examine additional pumps, other than the AFW pumps, to ensure that adequate flow or recirculation flow would be maintained via procedures through these pumps to prevent damage.
To re-examine this issue, reviews of AOP-5B, "Loss of Instrument Air", EOP 0.1, "Reactor Trip Response" and EOP 1.3, "Transfer to Containment Sump Recirculation" were performed. In addition, conversation with three Licensed SROs were performed to identify where in the procedures adequate pump flows were addressed.
The re-examination focused on safety related pumps necessary for unit shutdown or to dissipate decay heat and maintain core cooling. It was discovered that the AFW pump recirculation valves are unique in the fact that their recirculation valves fail closed upon loss of instrument air. (This was an original plant design function to ensure all flow going to the steam generators, and has since been rectified with the addition of a backup nitrogen supply to ensure the valves ability to be opened and stay open. This was done via the modification process).
Safety Injection system recirculation valves are locked to the open position. This is stated in AOP-5B, Attachment D, Part 2, "System Response", which states:
"Test line valves SI-897A and SI-897B are fail open with IA isolated. This maintains a recirc flow path for the SI pumps."
Feed and Condensate pumps and valves are covered in AOP-5B Attachment T.
"*CS-2180, CS-2188, Main Feed Pump mini-recirc valves fail open, if doesn't go open, instructed to use the manual gag override to open the valve" "CS-2252, Condensate Pump mini-recirc valve fails open, instructed to use the manual gag override to open the valve if it doesn't go open."
RCP Thermal Barriers are covered in AOP-5B Attachment H, Component Cooling.
"RCP thermal barrier isolation valves fail open to maintain thermal barrier cooling" AOP-5B, Attachment E covers the RHR system discharge and recirculation valves. These also fail open upon loss of instrument air. This will ensure adequate cooling to the pumps, however, creates a different issue. Due to the RH-624 and RH-625 (RHR Heat Exchanger Outlet valves) failing open, the potential exists for the RHR pumps to go into a runout condition when Containment Sump recirculation is put into operation. This is because of the supplies to and discharges from the Spray and SI pumps, as well as the RHR pumps, being maximized.
This has been a known issue for some time and has been addressed within both the AOP-5B and EOP-1.3 procedures. To ensure that the RHR pumps do not go into a runout condition, the RH-624 and RH-625 outlet isolation valves, RH-716A and RH-716B are throttled to ensure a miximum RHR flow of 2200 gpm. In addition, in EOP 1.3, the SI to RHR supply valve, SI 857 (either A or B depending on the RHR train being used/lined up for sump recirc) is throttled https://nmc.ttrackonline.con/tmtrackltmtrack.dll?IssuePage&TableId= 1000&Recordld=94(". 9/20/2002
Nuclear Management Company Page 3)of 4 to maintain RHR pump discharge pressure less than 130 psig Therefore, the AOPs and EOPs address the issues of RHR and SI pump having inadequate flow, as well as preventing pump runout conditions, to ensure no damage to the pumps.
Based on what was discovered through these reviews and conversations, it appears that the AFW pumps were in a unique situation, which has since been resolved. All other safety related/high profile pumps are protected from low or no flow damage, or pump runout, through steps built into the current EOPs and AOPs.
Based on this information, the SEN and CARB concerns are believed adequately addressed No further actions are recommended at this time, and this action item may be closed.
4/30/2002 4:37PM - MARK RINZEL:
Evaluation completed, see above update.
5/13/2002 2.22.07 AM - DUANE SCHOON:
Action complete. Closed.
5/13/2002 2.04:47 PM - JULIE KREIL:
SEN 174 evaluation was re-evaluated Based on what was discovered through these reviews and conversations, it appears that the AFW pumps were in a unique situation, which has since been resolved. All other safety related/high profile pumps are protected from low or no flow damage, or pump runout, through steps built into the current EOPs and AOPs. The SEN and CARB concerns are believed adequately addressed. No further actions are recommended.
CLOSED CA004279 to completion of Requested Activity.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 0 Project: CAP Activities & Actions 0 State: Done 0 Active/Inactive: Inactive AR Type: Parent 0 Owner: (None)
DON PETERSON f Assigned Date: 4112/2002 0 Submitter:
- Last Modified Date: 5/13/2002 2:04:47 PM 0 Last Modifier: JULIE KREIL 5/13/2002 2:04.47 PM 0 Last State Changer: JULIE KREIL Q Last State Change Date:
0 Close Date: 5/13/2002 2:04.47 PM 0 One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
CR 01-2278 RCE 01-069 GOOD CATCH SEN 174 CR 97-1992 CR 98-0050.
AOP 18 AOP 18A EOP 0.1 EOP 1.3 AOP 5B Update: E\(20011204 PB2171 JMK1) Operability Determination (OD) Part I, Revision 0, of CR 01 3595 was approved on 11/30/01. Operable But Degraded - or Operable But Nonconforming -
https://nmc.ttrackonline.comltmtrackltmtrack.dll?IssuePage&Tableld= 1000&RecordId=94(... 9/2012002
Nuclear Management Company Page 4 of 4 meets the minimum required level of performances, compensatory measures ARE required
\\Operability Determination (OD) Part I, Revision 1 of CR 01-3595 was approved on 12/01/01.
Operable But Degraded - or Operable But Nonconforming - meets the minimum required level of performances, compensatory measures ARE required.
Accepted into group and assigned pnonty 3. This questions the adequacy of an SEN applicability determination and evaluation. Per NP 5.4.1, SEN are to be prionty 3.
Pnonty = This is a reflash question towards the adquacy of a SEN closure from engineering TPS.
Import Memo Field:
CAP Admin: PBNP CAP Admin Site: Point Beach OLDACTIONNUM:
Cartridge and Frame: .
ATTACHMENTS AND PARENT/CHILD LINKS E CA003982: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW S CAP001415" Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https://nmc.ttrackonline.comltmtrackltmtrack.dil?IssuePage&Tableld= 1000&Recordld=94(.. 9/20/2002
Nuclear Management Company Page I of 3 STATE CHANGE HISTORY p.
Assign Conduct Work Review & Approved Initiate Assign Work Complete Quality Check Work Approval 5/8/2002 4126/2002 4/30/2002 5/8/2002 11 "06 25 AM 11 14 26 AM 4:22 30 PM 11.14 03 AM Owner PBNP by RICHARD Owner RICK by RICK Owner RICK Owner RICK by RICK WOOD by RICK WOOD CAP Admin FLESSNER WOOD WOOD WOOD V WOOD p,
Complete and Close Done 5/28/2002 2 18 57 PM Owner (None) by JULIE KREIL SECTION 1 Activity Request Id: CA004388 Activity Type: Corrective Action Submit Date: 4/26/2002 11:06:25 AM Site/Unit: Point Beach - Common Activity Requested: Review operator action assumptions in PRA Model for validity for the top risk-significant systems prior to NRC regulatory conference on 4/29/2002.
0 CATPR: N Initiator: MASTERLARK, JAMES (* I EPN Engineering Responsible Group Code: EPP Engineering Initiator Department:
Programs Nuclear Programs PRA PB Safety Analysis PB Responsible Department: Engineering Activity Supervisor: RICK WOOD Activity Performer: RICK WOOD I)
SECTION 2 Priority: 4 Due Date: 5/10/2002 Mode Change Restraint: (None) Management Exception From PI?: N N "0Licensing Review?: N 0 QA/Nuclear Oversight?:
NRC Commitment?: N "* NRC Commitment Date:
SECTION 3 Activity Completed: 4/30/2002 4:14PM - RICK WOOD:
Operator actions assumed in the PRA model for the Component cooling water system, service water, aux feedwater, ECCS and the instrument air system were identified and forwarded to Operations. The risk rank of the actions and the probability that the action would be performed incorrectly was also included.
5/8/2002 11:02:22 AM - RICK WOOD.
Operations (T. Vandenbosch) identified the following problems with the HEP forwarded to them:
Listed below are the comments associated with the HEPs:
CCI-AOP9B-73 ..... We do not take credit for crosstie of U1 and U2 CCW pumps.
CCW-AOP9B-73....We do not take credit for crosstie of U1 and U2 CCW pumps.
https://nmc.ttrackonline.comltmtrack/tmtrack.dll?IssuePage&Tableld=l 000&Recordld=97... 9/18/2002
Nuclear Management Company Page 2 of 3 HEP-SW-RE-C-0011....Should be P32B.
HEP-SW-RE-C-0012....Should be P32C.
HEP-SW-RE-C-0013....Should be P32D.
HEP-SW-RE-C-0014....Should be P32E.
HEP-SW-RE-C-0015....Should be P32F.
HEP-SW480AOP10C5....AOP 0.0 Step 6.1 does not align anything to B08/B09.
AF-HEP-START1TD ...Procedure guidance is given to start the TD AFW pump I'm not sure how this fits into the actions not accomplished by the Operators AF-HEP-START2TD ...Procedure guidance is given to start the TD AFW pump I'm not sure how this fits into the actions not accomplished by the Operators.
Guidance is given for the following and I am not sure how this fits into the actions not accomplished by the operators:
RHR-ISO-RHRA RHR-ISO-RHRB RHR-OP-7A-01 SI-ACCUM-IS 5/8/2002 11:14:03 AM - RICK WOOD:
The review is complete. Correction of the HEPs is tracked via OTH00451 0.
5/28/2002 2.18:57 PM - JULIE KREIL:
Action completed as documented above. OTH004510 will track correction of the HEPs.
CLOSED CA004388.
9/18/2002 6:06:49 PM - RICHARD FLESSNER:
Additional details on the HEP review are provided in attached document CA4388.doc.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 0 Project: CAP Activities & Actions 0 State: Done "0Active/Inactive: Inactive (None) AR Type: Parent 4D Owner:
Assigned Date: 4/30/2002 "0Submitter: RICHARD FLESSNER
" Last Modified Date: 9/18/2002 6:06:49 PM " Last Modifier: RICHARD FLESSNER "0Last State Change Date: 5/28/2002 2.18:57 PM 0 Last State Changer: JUL IE KREIL
" Close Date: 5/28/2002 2.18:57 PM
" One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
https://nmc.ttrackonline.com/tmtrack/tmtrack.dll?IssuePage&Tableld= 1O0O&Recordld=97:... 9/18/2002
Nuclear Management Company Page 3 of 3 Update:
Import Memo Field:
CAP Admin: PBNP CAP Admin Site: Point Beach OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS ACE0_00314. Probabilistic Risk Assessment PRA For Auxiliar_Fe edwae ystemAFW S/_z Linked From CAP001 415 Human Error Probabilities in PRA model (48640 bytes)
Linked from OTH004510: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW CA4388 doc (77824 bytes) https://nmc.ttrackonline.comltmtrackltmtrack.dll?IssuePage&TableId= 1000&Recordld=97... 9/18/2002
R. Flessner asked me to provide more detail regarding the review of HEPs performed by Operations and PRA. I provided the following list to Operations (T. Vandenbosch) in April 2002 to determine if the HEP was correctly described and if there are procedures directing the performance of the action.
Human Error Probabilities for top risk significant systems Instrument Air Event Name HEP Value F-V Description HEP-IA-FO-04748 1.OOE-03 1.59% Operator fails to reopen 3047 or 3048 to re establish IA supply to containment following SI
/ containment isolation signal HEP-IA-FO-START 6.90E-04 0.40% Operator fails to restart IA or SA compressor following a loss of offsite power HEP-IA--AOP5B-74 2.OOE-02 0.15% Operator fails to isolate IA header rupture (for the fraction of pipe breaks that can be isolated)
HEP-OCC-EOP01-04 1.50E-02 0.00% Operator fails to control charging/letdown following a loss of IA AF--HEP-MDP-FLOW 4.40E-02 3.54% Failure to manually control MDAFW flow after a loss of IA AF-HEP-RECIRC-1 4.30E-02 1.16% Failure to manually control recirc flow on same unit TDP P-29 after a loss of IA AF-HEP-RECIRC-2 4.30E-02 0.03% Failure to manually control recirc flow on opposite unit TDP P-29 after a loss of IA AF--HEP-RECIRC2F 2.84E-02 2.33% Dependent failure to manually control 2 AFW pumps recirc flow after a loss of IA AF--HEP-RECIRC3F 2.56E-02 42.20% Dependent failure to manually control 3 AFW pumps recirc flow after a loss of IA AF--HEP-RECIRC-A 4.30E-02 0.19% Failure to manually control recirc flow on MDP P-38A after a loss of IA AF--HEP-RECIRC-B 4.30E-02 0.21% Failure to manually control recirc flow on MDP P-38B after a loss of IA Component Cooling Water Event Name HEP Value F-V Description CCI-AOP9B-73 6.6E-2 Failure To Crosstie U1 & U2 CCW After Failure (renamed HEP-CCI Of The U1 Pumps AOP9B-73)
CCI-AOP9B-74 5.0E-2 Failure To Isolate A Rupture In The CCW And Restore CCW To An Operable State CCI-01-71-42 1.5E-2 0.18% Failure To Align Standby CCW Hx After Failure Of The Normal CCW Heat Removal System CCW-AOP9B-73 6.9E-2 Failure To Crosstie U1 & U2 CCW During Another Accident CCW-AOP9B-74 5.4E-2 0.31% Failure To Isolate A Rupture In The CCW During Another Accident
Event Name HEP Value F-V Description CCW-EOP13-03 1.2E-4 0.04% Failure To Start CCW Pumps After A Concurrent SI Signal LOSP Or An SI Signal Followed By A LOSP (Prior To Resetting SI)
CCW-O1-71-42 3.OE-2 Failure To Align Standby CCW Hx After Failure Of The Normal CCW Heat Removal System During Another Accident Service Water Event Name HEP Value F-V Description HEP-SW--AOP9A-63 5.2E-2 2.21% Operator fails to isolate SW header rupture HEP-SW--EOP-0-9A 1.9E-02 Operator fails to isolate non-essential SW loads HEP-SW-RE-C-0010 Failure to manually close SW P-32A isolation valve SW-1 0 HEP-SW-RE-C-00i 1 Failure to manually close SW P-32A isolation valve SW-i 1 HEP-SW-RE-C-0012 Failure to manually close SW P-32A isolation valve SW-1 2 HEP-SW-RE-C-0013 Failure to manually close SW P-32A isolation valve SW-13 HEP-SW-RE-C-0014 Failure to manually close SW P-32A isolation valve SW-14 HEP-SW-RE-C-0015 Failure to manually close SW P-32A isolation valve SW-15 HEP Operator failure to align to B08/09 per AOP 0.0 SW480AOP10C5 Step 6.1 HEP-SWI-AOP9A-61 3.6E-04 Operator fails to start standby SW pumps AFW Event Name HEP Value F-V Description AF--HEP-CST-LOW 3.90E-04 9.27% This event estimates the probability that the operator will fail to respond to a low level CST alarm (Pc only), therefore failing the option for long-term auxiliary feedwater use.
AF--HEP-TDAFISOL 5.75E-03 0.20% Failure of operator to isolate the Turbine-Driven Auxiliary Feed Water (TDAFW) pump from a faulted steam generator.
AF--HEP-MDP-FLOW 4.40E-02 3.54% Failure to manually control Motor-Driven Auxiliary Feed Water (MDAFW) pump after a loss of IA.
AF--HEP-START-MD 1.1 E-03 This event estimates the probability that the operator will fail to manually start the correct motor driven pump MDP P-38A or MDP P-38B after the pump's auto start logic fails.
AF--HEP-CST-FW-- 1.10E-02 2.86% This event estimates the probability that the operator will fail align firewater as an alternate feed source to the acorooriate steam
Event Name HEP Value F-V Description generators (Pe only).
AF--HEP-RECIRC-1 4.30E-02 1.16% This event estimates the probability that the operator fails to initiate recirculation for 1 P29 upon a loss of instrument air.
AF--HEP-RECIRC-2 4.30E-02 0.03% This event estimates the probability that the operator fails to initiate recirculation for 2P29 upon a loss of instrument air.
AF--HEP-RECIRC-A 4.30E-02 0.19% This event estimates the probability that the operator fails to initiate recirculation for P38A upon a loss of instrument air.
AF--HEP-RECIRC-B 4.30E-02 0.20% This event estimates the probability that the operator fails to initiate recirculation for P38B upon a loss of instrument air.
AF--HEP-RECIRC2F 2.84E-02 2.33% Dependent failure to manually control 2 AFW pumps recirc flow after a loss of IA AF-HEP-RECIRC3F 2.56E-02 12.20% Dependent failure to manually control 3 AFW pumps recirc flow after a loss of IA AF--HEP-START1TD 1.1 E-03 This event estimates the probability that the operator will fail to manually start TDP 1P-29 after the pump's auto start logic fails.
AF--HEP-START2TD 1.1 E-03 This event estimates the probability that the operator will fail to manually start TDP 2P-29 after the pump's auto start logic fails.
AF-HEP-CST-SWTD 9.20E-03 0.07% This event estimates the probability that the operator will fail to align service water to the turbine-driven pump as an alternate feed source to the appropriate steam generators (Pe only).
AF--HEP-CST-SWMD 1.50E-02 1.79% This event estimates the probability that the operator will fail align service water to the motor-driven pump as an alternate feed source to the appropriate steam generators. (Pe only)
ECCS Event Name HEP Value F-V Description HEP-RHR-EOP13-23 2.45 E-02 11.60% Failure to align SI for low containment sump recirculation RHR-ISO-RHRA 6.OE-1 Failure to isolate a rupture in the A train of RHR (rupture caused by failure of RH-720 and subsequent overpressurization)
RHR-ISO-RHRB 5.4E-1 Failure to isolate a flow diversion from the B train of RHR to the RWST through a failed open MOV (RH-742)
RHR-OP-7A-01 8.8E-02 Failure to place the Residual Heat Removal system into operation per OP-7A SI-ACCUM-IS 1.7E-01 Failure to isolate a ruptured accumulator by (renamed HEP-SI closing the isolation MOV ACC-AISOL)
Event Name HEP Value F-V Description HHR-EOP-RECIRC 5.4E-03 Operator fails recirc switchover to high head This list was compiled from an earlier listing of HEPs and a number of these events have been deleted from the current model. The deleted HEPs are:
Instrument Air Event Name HEP Value F-V Description HEP-OCC-EOPOI-04 1.50E-02 0.00% Operator fails to control charging/letdown following a loss of IA AF--HEP-RECIRC2F 2.84E-02 2.33% Dependent failure to manually control 2 AFW pumps recirc flow after a loss of IA AF--HEP-RECIRC3F 2.56E-02 '12.20% Dependent failure to manually control 3 AFW pumps recirc flow after a loss of IA Component Cooling Water Event Name HEP Value F-V Description CCI-O1-71-42 1.5E-2 0.18% Failure To Align Standby CCW Hx After Failure Of The Normal CCW Heat Removal System CCW-AOP9B-73 6.9E-2 Failure To Crosstie U1 & U2 CCW During Another Accident CCW-AOP9B-74 5.4E-2 0.31% Failure To Isolate A Rupture In The CCW During Another Accident Service Water Event Name HEP Value F-V Description HEP-SW-RE-C-001 0 Failure to manually close SW P-32A isolation valve SW-10 HEP-SW-RE-C-001 1 Failure to manually close SW P-32A isolation valve SW-11 HEP-SW-RE-C-0012 Failure to manually close SW P-32A isolation valve SW-12 HEP-SW-RE-C-0013 Failure to manually close SW P-32A isolation valve SW-1 3 HEP-SW-RE-C-0014 Failure to manually close SW P-32A isolation valve SW-1 4 HEP-SW-RE-C-001 5 Failure to manually close SW P-32A isolation valve SW-15 HEP Operator failure to align to B08/09 per AOP 0.0 SW480AOP10C5 Step 6.1
AFW Event Name HEP Value F-V Description AF--HEP-RECIRC2F 2.84E-02 2.33% Dependent failure to manually control 2 AFW pumps recirc flow after a loss of IA AF--HEP-RECIRC3F 2.56E-02 12.20% Dependent failure to manually control 3 AFW pumps recirc flow after a loss of IA ECCS Event Name HEP Value F-V Description RHR-ISO-RHRA 6.OE-1 Failure to isolate a rupture in the A train of RHR (rupture caused by failure of RH-720 and subsequent overpressurization)
RHR-ISO-RHRB 5AE-1 Failure to isolate a flow diversion from the B train of RHR to the RWST through a failed open MOV (RH-742)
RHR-OP-7A-01 8.8E-02 Failure to place the Residual Heat Removal system into operation per OP-7A T. Vandenbosch had comments on the following items that were not deleted from the model:
HEP-CCI-AOP9B-73 This item is not connected in the model and therefore has no effect on the model.
AF-HEP-STARTITD and AF-HEP-START2TD have been produced using current Human Reliability Analysis techniques. These HEPs have been reviewed by a PRA engineer and by a reviewer with considerable Point Beach Operations experience.
HEP-SI-ACC-AISOL: The ruptured accumulator initiating event is not included in the model.
Items not included in the original Operations review and are in the current model are as follows:
Instrument Air Event Name HEP Value F-V Description HEP-IA--RE-01207 5.OE-03 Operator fails to restore IA-01207 after T/M HEP-IA--RE-01210 5.OE-03 Operator fails to restore IA-01210 after T/M Service Water Event Name HEP Value F-V Description HEP-RP--AOP9B-63 1.1 E-4 2.06E-3 Operator fails to manually trip reactor (Tccrrsw)
AFW Event Name HEP Value F-V Description AF--HEP-MINI-GAG 3.4E-3 2.25E-05 Failure to gag mini recirc valve >lhr into event AF--HEP-RECIRC4F 5.06E-3 1.85E-05 MEX event fail to manually control 4 AFW pumps ECCS Event Name HEP Value F-V Description HEP-HHR-EOP13-23 1.25E-2 1.71 E-01 Failure to align for high head recirculation HEP-SI-SD--DRN 1.OOE-0 Shutdown model HEP HEP-ESF-EOP-0-04 3.25E-3 3.46E-04 Failure to manually initiate SI HEP-RCS-CSPH1-12 2.36E-2 1.26E-01 Failure to establish Feed & Bleed (No SI)
HEP-RCS-CSPH1-13 2.05E-2 3.19E-02 Failure to establish Feed &Bleed (With SI)
The deleted HEPs have no impact on the current PRA model.
J. Masterlark and I have reviewed the list of added HEPs. The two instrument air HEPs are type A and are preinitiators. They are not associated with operator action or Operations procedures. The new SW HEP is associated with AOP 9A Service Water System Malfunction or AOP 9B Component Cooling Water System Malfunction. Both of these procedures direct the operator to trip on loss of SW or CCW. The new Auxiliary Feedwater HEPs have been recently created based on the latest Operations procedures.
The importance of these two HEPs is low. The ECCS HEPs generally have high importance. All of these HEPs have procedures that specifically direct the performance of each of these activities. The Shutdown model HEP is not used since the shutdown model is not complete or being used.
The Human Error Probabilities (HEPs) identified in this corrective action item are no longer included in the PRA model except for AF-HEP-START1TD and AF-HEP START2TD. These two items have been updated in the latest model and will be issued with the AFW system PRA notebook in the near future. These Human Error Probabilities estimate the probability that the operator fails to manually start the turbine driven auxiliary feedwater pumps after the pumps auto start fails to start the pump. The question identified in the original Action Request was that the Operations reviewer was not sure if PRA had credited the fact that there is procedure guidance to start the pump following an auto start failure.
The Human Reliability Analysis for these Auxiliary Feedwater system HEPs does factor in the procedure step directing start of the pump. It also recognizes the ability of the STA to diagnose lowering steam generator levels and prompt the operator to start the pump.
The original question posed by the Operations reviewer has been answered for the HEPs remaining in the current PRA model. As the project to update the PRA model progresses, existing HEPs will be revised and new operator actions will be identified and HEPs for these actions calculated.
Appendix B CALCULATION OF TYPE C HUMAN ERROR PROBABILITIES
AF--HEP-STARTXXX REVISION: 4/17/2001 EVALUATOR: James Masterlark REVIEWER: Paul Knoespel OPERATIONAL REVIEW: John Sell SCENARIO: Failure to manually start Auxiliary Feed Water (AFW) pump after auto-start fails.
INTERVIEWS: Interviews were conducted with Jim Fouse, Sr. Training Specialist.
BASIC EVENT(S): AF--HEP-START1TD, AF--HEP-START2TD, AF--HEP-START-MD AF--HEP-START12T AFW pump DESCRIPTION: This HEP calculates the probability to fail to manually start an the Unit after its associated auto-start fails. An average dependency is assumed to exist between pump since 1 Turbine Driven Auxiliary Feed Water (TDAFW) pump and the Unit 2 TDAFW these two events separate control room operators would be controlling these pumps. Therefore, in the same cut-set are replaced with the following formula:
AF--HEP-START1TD
- AF-HEP-START2TD= (I+6*START1TD)/7*START2TD
= 2.57E-4 (AF--HEP-START12T) the two Motor (Same formula for Unit 2). A complete dependency is assumed to exist between basic AF--HEP Driven Auxiliary Feed Water (MDAFW) pumps. Therefore, both use the same START-MD).
RESULTS:
Pc: 5.40E-4 Pe: 1.10E-3 ""
TOTAL: 1.64E-3 (AF-HEP-STARTlTD, AF--HEP-START2TD, AF-HEP-START-MD) 30 minutes before TIMING ANALYSIS: It is assumed that the initiation must take place within to perform the actions is the intact steam generator dries out. The start of the time window action verification upon assumed to be 10 minutes to allow time for diagnosis (part of automatic in Attachment A to a trip or SI). The action required (manually start pump) are also located than 5 minutes) to EOP-0. Therefore, this action will only take a few minutes (assumed less Status Trees at 15 perform. The STA is assumed to start monitoring the Critical Safety Function 15 minutes (30 minutes into the event. Therefore, recovery time by the STA is limited to these actions, this minutes - 15 minutes). Due to the short time needed to diagnosis and complete scenario is not considered time limited when determining recovery probabilities.
SUCCESS CRITERIA: Success is upon manually starting an AFW pump.
INITIATING EVENT EFFECTS: These HEPs are used for most initiating events. Since AFW initiation is expected near the start of all of these events, the same assumptions would apply and same recoveries would apply.
PROCEDURES:
Pc: EOP-0, Step A3, Rev 34 dated 10/30/2000 Pc Revovery: CSP-H.5, Step 4, Rev 8 dated 6/9/1999 EOP-0, Step 6a Response Not Obtained (RNO), Rev 34 dated 10/30/2000 Pe: EOP-0, Step A3 RNO, Rev 34 dated 10/30/2000 Pe Recovery:. CSP-H-5, Step 5 RNO, Rev 8 dated 6/9/1999 EOP-0, Step 6a RNO, Rev 34 dated 10/30/2000 ASSUMPTIONS:
Pc:
- 1. ERE Not credited for recovery due to the short length of time available (< 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Pc Recovery:
- 1. Self Review - Self Review is not credited.
- 2. Extra Crew - Extra crew members will also be following EOP-0 and credited for recovery of Pce by verifying that the AFW pumps have started in Step 6a of EOP 0.
- 3. STA Review - The STA's will start CSFST review within 15 minutes of the event. By failing to initiate an AFW pump manually after automatic start fails, the steam generators will eventually reach a low level. At this point, the STA will identify this through the CSP's and enter CSP-H.5 for low level. This procedure contains a step (Step 4) to verify AFW flow of at least 50 gpm. Therefore, credit is taken for the STA to identify that the AFW pumps have not been started.
Normally credit is not taken for the STA chance to recover from a missed procedure step (Pce) because they do not follow the same procedures. However, in this case credit could be taken since similar steps are included in the CSP's that would be involved if the EOP steps would be missed.
- 4. Pce - As described above, recovery is credited for Extra Crew and STA Review.
is Since multiple recoveries are included, an override value of 5E-1
- IE-1 = 5E-2 included. Due to potential dependencies within these recoveries, the screening values from Table 4-1 are used instead of the independent recovery HEP's.
Pe Recovery:
- 1. The STA's will start CSFST review within 15 minutes of the event. By failing to initiate an AFW pump manually after automatic start fails, the steam generators will eventually reach a low level. At this point, the STA will identify this through the CSP's and enter CSP-H.5 for low level. This procedure requires the initiation of AFW if flow is less than 50 gpm (step 4). Therefore, credit is taken for the STA to recover from a failed step in EOP-0.
- 2. Two separate operators will be performing EOP-0. One will be verifying automatic actions with Attachment A, and an additional operator will be performing the main body of the procedure. The main body contains additional steps to verify SG level and to manually start AFW if it has not automatically started. Since time is available for recovery, credit is taken for this additional operator to manually start AFW if the first operator failed to do so correctly. Due to the time frame available, a high dependency is assumed.
- 3. Since the STA has 15 minutes or less for recovery there is an assumed high dependency between the STA's reviews in the CSFST's operations use of the EOP's. Due to the limited time frame, the recovery is assumed to be 5E-1.
Nuclear Management Company Page I of 3 STATE CHANGE HISTORY Initiate l g Work Assign L. Assign ;6 S~ 5/8/202 11:09 56 AM Conduct Work Owner. RICK WOOD 7/24/2002 4.03.55 PM Owner: RICK WOOD by RICK WOOD . Fr by RICK WOOD SECTION 1 Activity Request Id: OTH004510 Activity Type: Other Submit Date: 5/8/2002 11:09.56 AM Site/Unit: Point Beach - Common Activity Requested: Correct the problems identified with the Point Beach HEPs. If the HEPs no longer exist in the model, then the description in our data base should be eliminated.
George Baldwin - Kewaunee PRA will perform the work.
0 CATPR: N Initiator: MASTERLARK, JAMES Initiator Department: EPN Engineering Responsible Group Code: EPP Engineering Programs Nuclear Programs PRA PB Safety Analysis PB Z Responsible Department: Engineering Activity Supervisor. RICK WOOD f*
Activity Performer: RICK WOOD SECTION 2 Priority: 4 Due Date: 12/20/2002 Mode Change Restraint: (None) Management Exception From Pi?: N 0 QA/Nuclear Oversight?: N 0 Licensing Review?: N NRC Commitment?: N 0 NRC Commitment Date:
SECTION 3 Activity Completed: 4/30/2002 4:14PM - RICK WOOD:
Operator actions assumed in the PRA model for the Component cooling water system, service water, aux feedwater, ECCS and the instrument air system were identified and forwarded to Operations. The risk rank of the actions and the probability that the action would be performed incorrectly was also included.
518/2002 11:02.22 AM - RICK WOOD:
Operations (T. Vandenbosch) identified the following problems with the HEP forwarded to them:
Usted below are the comments associated with the HEPs" CCI-AOP9B-73 ..... We do not take credit for crosstie of U1 and U2 CCW pumps.
CCW-AOP9B-73....We do not take credit for crosstie of U1 and U2 CCW pumps HEP-SW-RE-C-0011....Should be P32B.
HEP-SW-RE-C-0012....Should be P32C.
HEP-SW-RE-C-0013....Should be P32D.
https:llnmc.ttrackonline.com/tmtrackltmtrack.dll?ssuePage&Tableld= I 000&Recoi did= 12 ... 8P28/2002
Nuclear Management Company Page 2 of 3 HEP-SW-RE-C-0014....Should be P32E.
HEP-SW-RE-C-0015....Should be P32F.
HEP-SW480AOP10C5....AOP 0.0 Step 6.1 does not align anything to B081B09.
AF-HEP-START1TD....Procedure guidance is given to start the TD AFW pump. I'm not sure how this fits into the actions not accomplished by the Operators.
AF-HEP-START2TD....Procedure guidance is given to start the TD AFW pump. Im not sure how this fits into the actions not accomplished by the Operators.
Guidance is given for the following and I am not sure how this fits into the actions not accomplished by the operators:
RHR-ISO-RHRA RHR-ISO-RHRB RHR-OP-7A-01 SI-ACCUM-IS SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 o Project: CAP Activities & Actions 0 State: Conduct Work "0Active/Inactive: Active 0 Owner: RICK WOOD 1 AR Type: Parent "O Submitter: RICK WOOD 6ý Assigned Date: 7/24/2002 "0Last Modified Date: 7/24/2002 4:03:55 PM "* Last Modifier: RICK WOOD "0Last State Changer: RICK WOOD
" Last State Change Date: 7/24/2002 4:03:55 PM "0Close Date:
"0One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
Update:
Import Memo Field:
CAP Admin: SCOTT PFAFF 6) Site: Point Beach OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS SLinked From CAP001415 CHANGE HISTORY https://nmc.ttrackonliine.com/tmtrack/tmtrack.dIl')I"-suePage&Tablcld= I000&Rccordld=l 2 ... 8/28/2002
Nuclear Management Company Page 3 of 3 5/812002 11:10:09 AM by admin Last Modified Date Changed From 5/8/2002 11:09:56 AM To 5/8/2002 11:10 09 AM Last Modifier Changed From RICK WOOD To admin Attachment Added. Linked From CAP001415 7/24/2002 4:03:55 PM by RICK WOOD Due Date Changed From 11/20/2002 To 12/20/2002 Stale Changed From Assign Work To Conduct Work Via Transition Assign Assigned Date Changed From 4/30/2002 To 7/24/2002 Last Modified Date Changed From 5/812002 11:10.09 AM To 7/24/2002 4:03:55 PM Last Modifier Changed From admin To RICK WOOD Last State Change Date Changed From 5/8/2002 11:09 56 AM To 7/24/2002 4-03 55 PM https://nmc.ttrackon i ne.com/tmtrack/tmtrack.dli ?ISsuePage&Tabl eld= 1OOO&Record [d= 12 ... 8/28/2002
Nuclear Management Company Page I of 2 STATE CHANGE HISTORY
.. - C L i .L L Conduct Work Review & Quality Initiate Assign Work Assign Work Complete Approval Approved Check 9/3/2002 9/3/2002 9/3/2002 9/3/2002 6:36:35 PM 6 37:42 PM 6 45"08 PM 6"45 49 PM Owner Owner Owner Owner by RICHARD RICHARD by RICHARD by RICHARD RICHARD by RICHARD RICHARD FLESSNER PBNP CAP FLESSNER FLESSNER FLESSNER FLESSNER FLESSNER Admmn FLESSNER V
SECTION 1 Activity Request Id: CA026223 Activity Type: Corrective Action Submit Date: 9/3/2002 6.36:35 PM Site/Unit: Point Beach Common Activity Requested: CA#17: Update the PBNP simulator to model AFW pump failure due to less than required minimum recirculation flow.
Q CATPR: N Initiator: MASTERLARK, JAMES Initiator Department: EPN Engineering Responsible Group Code: EXC Engineering Programs Nuclear Processes Continuous Safety Analysis PB Improvement PB Z2 Responsible Department: Engineering Activity Supervisor: RICHARD FLESSNER Activity Performer: RICHARD FLESSNER SECTION 2 Priority: 3 Due Date: 9/3/2002
" Mode Change Restraint: (None) Management Exception From Pt?: N "0*OAJNuclear Oversight?: N 0 Licensing Review?: N NRC Commitment?: Y 0 NRC Commitment Date:
SECTION 3 Activity Completed: 9/3/2002 6:45:08 PM - RICHARD FLESSNER:
Simulator Discrepancy Report SDR 02-0046 was initiated on 4/22102 to "Install auto pump trips per plant direction". T. Kendall provided technical direction (documented in the SDR). The software change was made and tested on 4/22/02. The SDR was closed by R. Parlato on 4/23/02.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 https://nmc.ttrackonline.com/tmtrackltmtrack.dll?IssuePage&TableId= 1000&RecordId=26'... 9/20/2002
Nuclear Management Company Page 2 of 2 O Project: CAP Activities &
Actions 0State: Quality Check 0 Active/Inactive: Active O Owner: PBNP CAP Admin AR Type: Parent
" Submitter: RICHARD FLESSNER Assigned Date: 9/3/2002
" Last Modified Date: 9112/2002 10:12:43 0 Last Modifier: RICHARD FLESSNER AM Q Last State Change Date: 9/3/2002 6"45:49 PM 0 Last State Changer: RICHARD FLESSNER "0Close Date:
" One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
CR 01-2278 RCE 01-069 GOOD CATCH LER 266/2001-005-00 Update: This CA is being issued to document a completed action.
Import Memo Field: LER 266/2001-005-00 made the committment that" Simulator modifications to enhance modeling the potential failure of the AFWS pumps following loss of instrument air scenarios are being pursued.*
CAP Admin: PBNP CAP Admin Site: Point Beach OLD_ACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENTICHILD LINKS Subtask from CAP001415: Probabilistic Risk Assessment PRA For Auxiliajy Feedwater Syqtem AFW E -1Linked to ACE000314: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https://nmc.ttrackonline.com/tmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld=26:... 9/20/2002
SimulatorDiscrepancyReport SDR Num. Title Orig Date 02-0046 Install auto pump trips per plant direction rdp 412212002 Status System Closed By Close Date COMP PMP rdp 4/23/2002 Description MODEUNG OF MULTI-STAGE PUMP FAILURES ON THE SIMULATOR It was recently determined that the PBNP simulator needed to model pump failure due to low flows. Particularly vulnerable to these kinds of failure are horizontal split case multi-stage centrifugal pumps. These are the SI and AFW pumps.
Failure occurs when water being pumped is reduced to the point that pump energy is not removed at a sufficient rate and shows up as thermal energy rather than pressure x volume work Ifthe thermal energy raises the liquid to the saturation point localized or generalized boiling occurs that upsets the hydro-dynamic supporting of the shaft and causes severe vibration of the rotating element The vibration of the shaft causes contact between the rotating element and the casing, severe localized heating, and shaft seizure due to the rapid expansion of the rotating element.
Shaft seizure may cause the prime mover (turbine or motor) to stall, or it may cause the shaft/coupling to break allowing the prime mover to continue running. Failure mode is indeterminate and could be modeled as either shaft seizure or shaft breakage without loss of fidelity.
The following Approach uses several simplifications to arrive at a reasonable figure for modeling low-flow Induced pump failures. Industrial experience demonstrates that failure under these conditions is very rapid. However, there is little verifiable empirical data to establish a close correlation for predicting low flow failures. The heat balance approach used below is judged to be as good as any method for the practical purposes of modeling expected conditions in a training simulator.
Cavitation (boiling) will occur when the liquid enthalpy reaches the saturation point at the eye of the pump impellor. Each stage boosts pressure sequentially, so In theory only the first stage is a concern. Ifsufficient flow Is maintained to remove the energy Imparted by the first stage without reaching saturation, the pump should continue to function However, other factors (such as Internal recirculation) come into play that can also lead to severe vibrations and rapid failure.
Therefore, to simplify matters (and to be consistent with anecdotal observations of pump failure), it is assumed that all of the pump work is deposited In the liquid at atmospheric pressure. This is a reasonable assumption that offsets the likelihood of a higher suction head due to an elevated head tank level (RWST or CST) and low piping friction head losses with the high pressure drop at the suction of a pump.
Based on the above assumption, when sufficient energy Is imparted by the pump to raise the flow through it to 212 deg F, cavitation (and therefore failure) are assumed to occur.
Although higher temperatures are permitted by various plant designs, Technical Specifications, and procedures, It will be assumed that the pump inlet temperature is 70 deg F. This is judged to be a representative ambient temperature for normal plant operations. Ifsuction temperatures are lower the predicted minimum flow that will prevent failure will decrease, and vice versa The difference In liquid enthalpy between 212 deg F and 70 deg F is -141 Btu/lbm The heat imparted by the pump is a function of both the brake horsepower and the pump efficiency at a given flow rate, and both are taken from the manufacturer's pump curves. The specific heat input is then q = W'(1-h)/m' Where:
- q is the specific heat input in Btu/Ibm (equivalent to the change in enthalpy)
W' Is the break horsepower from the pump curve (Hp) converted to Btulmin h Is the pump efficiency from the pump curve m' is the mass flow rate through the pump (gpm) converted to Ibrrmin Substituting 141 Btu/lbm for q, rearranging to solve for the critical mass flow rate, and using appropriate conversion factors, this equation becomes q'cntical = 0.361 Hp(1-h)
Where q'cntical is in gpm. Because both pump efficiency and Horsepower are (strictly speaking) not linear, this equation must be solved iteratively. However, for all practical purposes the functions are reasonably linear within the small range of interest and the solution converges very quickly.
Motor Driven AFW Pumps:
From the pump curves for P-38A&B. the pump efficiency at -6 gpm is 3 6%, and the brake horsepower Is -150 Hp Substituting into the above equation gives a figure of 5 2 gpm. This is as close a solution as can be obtained by reading the curves. Use 6 gpm as the failure flow for these pumps.
Turbine Driven AFW Pumps:
From the pump curves for 1P-29 and 2P-29, the pump efficiency at -5 gpm Is -2%, and the brake horsepower is -200 Hp.
Substituting into the above equation gives a figure of 7 gpm Trying 10 gpm (4% efficiency and 200 Hp) gives a figure of 6.9 gpm. These two figures bracket the actual value and are virtually identical Use 7 gpm as the failure flow for these pumps SI Pumps From the pump curves for 112P-1 5A&B, the pump efficiency at -10 gpm is 2.5%, and the brake horsepower is -300 Hp Substituting into the above equation gives a figure of 10.6 gpm. This is as close a solution as can be obtained by reading the curves. Use 11 gpm as the failure flow for these pumps.
Suggestions for Modeling Failures It is not reasonable to instantaneously fail the pumps immediately upon dropping to less than the flows listed above. A realistic failure would be for the pumps to suffer unrecoverable failure 30 seconds after dropping below the listed flows. Timely restoration of flow above the threshold could be modeled to avert ultimate failure, but if this is not done within -15 seconds, severe degradation should be modeled as a penalty (50% degraded pump curve).
T. C. Kendall (the previous letter was created per plant management direction to provide modeling information to the simulator..)
References:
Outage: Mod Number: Priority:
SHWR: HW Change: HW Spec: Date:
HardwareScope SW Change: y S W Eng. rdp Date: 4/22/2002 Software Scope Created subroutine autotrip with calls from Intlkp and Intlkpu2 for P29,P15 and P38 (both units except for P38). Setpoints as directed by the included letter from Ton Kendal. Database modification for new subroutine.
OPCSpec: A. Morris Date 4/2=002 OperationsScope
LIMITED SCOPE SIMULATION TEST Test being conducted in support of SDR No.: 02-0046 Date of Test: 4/22/2002 BriefDescription of Test Tested unit AFW and SI pumps In accordance with the pump trip cnteria outlined in engineenng transmittal attachment "MODEUNG OF MULTI-STAGE PUMP FAILURES ON THE SIMULATOR.
Test Objectives Verify pump tnps on a 30-second timer when low flow cntenon is met Verify that the pump does NOT trip if flow Is recovered within the 30-second time alloted. Verify that the newly Installed trip does not adversly Impact other failure modes for each pump potentially disrupting training Verify that after pump trip that the pump is NOT recoverable.
InitialConditionsfor Test, IC No. orplant conditions established(ReferenceProcedure).
I/C 24 (adequate conditions to opeate the Turbine-Dnven AFW pump at full capacity)
LOAs andEquipment Overrides requiredfor Test.
LOA for-. Pump recirc vlaves, pump discharge valves, pump suction valves.
Malfunctions/ComponentFailuresentered during conduct of test, including severity, time of activation,and ramp time.
Malfunctions for. Broken Shaft, Head Capacity, Beanng failure, and Shaft Seizure. Component failures for-. applicable FTs and PTs, TD-AFW pump trip-throttle valve.
Test Results Each pump (the HHSI. MD-AFW, TD-AFW) met the test objectives outlined for that pump, satisfactorily.
F Test Unsatisfactory 0 SPFSubmitted: E]
I Test Satisfactory Test ConductedBy: A. Morms N -
Nuclear Management Company Page 1 of 2 STATE CHANGE HISTORY L_
Assign Work Conduct Work Review & Approved Quality Check Initiate Assign Work Complete Approval 9/312002 6 51:49 PM 9/3/2002 9/3/2002 E:> 9/3/2002 6 52.39 PM 6 58.41 PM 6 59.04 PM Owner by RICHARD by RICHARD Owner Owner by RICHARD Owner RICHARD RICHARD by RICHARD RICHARD FLESSNER FLESSNER FLESSNER FLESSNER FLESSNER PBNP CAP FLESSNER FLESSNER Admin SECTION 1 Activity Request Id: CA026224 Activity Type: Corrective Action Submit Date: 9/3/2002 6:51:49 PM Site/Unit: Point Beach Common Activity Requested: CA#18: Revise the EOP validation process to include PRA involvement.
0 CATPR: N Initiator: MASTERLARK, JAMES Initiator Department: EPN Engineering Responsible Group Code: EXC Engineering Programs Nuclear Processes Continuous Safety Analysis PB Improvement PB Z Responsible Department: Engineering Activity Supervisor: RICHARD FLESSNER Activity Performer: RICHARD FLESSNER SECTION 2 Priority: 3 Due Date: 9/3/2002 "0Mode Change Restraint: (None) Management Exception From Pl?: N "0QAINuclear Oversight?: N 0 Licensing Review?: N NRC Commitment?: N 0 NRC Commitment Date:
SECTION 3 Activity Completed: 9/3/2002 6.51:49 PM - RICHARD FLESSNER:
This CA is being issued to document a completed corrective action.
9/3/2002 6:58"41 PM - RICHARD FLESSNER:
The EOP/AOP Verifcation and Validation processes were combined into one process with issuance of Rev 2 of OM 4.3.2 on 5/13/2002 and cancellation of OM 4.3.3. OM 4.3.2 step 4.2.4 requires involvement of the PRA Group to review technical changes. Attachmnet D contains a PRA Core Damage Risk Matrix listing Procedures and applicable Events to be used for validation.
SECTION 4 CA Supervisor: (None) Licensing Supervisor: (None) https://nmc.ttrackonline.com/tmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld=26:... 9/20/2002
Nuclear Management Company Page 2 of 2 SECTION 5
" Project: CAP Activities &
Actions "
"0State: Quality Check 0 Active/Inactive: Active D Owner: PBNP CAP Admin AR Type: Parent 0 Submitter: RICHARD FLESSNER Assigned Date: 9/3/2002 0 Last Modified Date: 9/3/2002 7:16"46 PM 0 Last Modifier: RICHARD FLESSNER RICHARD FLESSNER
" Last State Change Date: 9/3/2002 6:59.04 PM 0 Last State Changer:
"0Close Date:
" One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
CR 01-2278 RCE 01-069 GOOD CATCH Update:
Import Memo Field:
CAP Admin: PBNP CAP Admin Site: Point Beach OLDACTION_NUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS Subtask from CAP001415: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW E-2 Linked to ACE00431-4 Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https://nmc.ttrackonline.com/tmtrackltmtrack.dllIssuePage&TableId= 1000&RecordId=26.... 9/20/2002
Vrc ..
Nuclear Power Business Unit DOCUMENT REVIEW AND APPROVAL N\'ote: Refer io NP 1.1.3 for requirements.
I - INITIATION Unit PBO Usage Lescl Ih:formation Proposed Rev No 2 Doc Number OM 4.3.2 Classification NA Title EOP/AOP VerilicationrValidation Process
[D Revision El Cancellation El New Document 0l Other (eg. periodic review. admin hold)
List Tempora.y Changes/Feedbacks Incorporated:
Description of AlterationlReason (If nece~sary. continue descnption of changes on PBF-0026c and attach)
OM 4.3.3, EOP Validation (Rev. 0)
Total rewrite, remised format per Procedure Writers' Guide, incorporated information and PRA Core Damnace Risk data. See PBF-0016c for details.
L:ist other documents required to be effective concurrently with *"herevision (cr, other procedures. forms, drawings, etc.):
(Rev. 0). 2103a (Rev. 0). 2103b (Rev. 0). 2103c (Rev. 0)
PBF-2102a (Rev. 0). 2102b James G Green i - Date S [ t Document Preparer (print/sign) doct,mente-d and resolved lrd;cates Jraft prepared according to NPI .I .3,any commitmrentsase;icees have been I1 - TECHNICAL REVIEW (Cannot be the 1rreparcr or Approval Authority.
. )z A J" A;2 V #! -4j /
Technical Reviewer (print/sign) requirements of NP 1.1.3 completed.
S.... draft technically correct. consistent ith remfcrences/bae*scappcr tier requirements.
Indicais l III - DOCUMENT OWNER REVIEW Required Reviewers/Organizations:
Validation Requirc$7 IM NO El YES [] AWED (Group Head Approval andA Reason Required) POTC: ?
l<..0-a. I Reason Validation Waived: LOflW1UC on S. I lalidation Waiver Apý-'val:
C~ontinue on1 rLjr*.,Ex; Is nCCCSur.
VIVO,2 "GroupHead Sagnature Ch~inge pre-scrcened according to NP 5.1.87 O NO [a YES (Provide documentation according to NPS 1.S)
Scrtering zompleted according to NP 5.1.8? O NA 0l YES (Attach copy) Safety evaluation required? [] NO 0l YES If YES, training or briefing required before issue? O NO 0] YES TrairinEor briefing required.& IO 91YES Training assistance .eied*;fN g YES ifYES, Training Coordinator contacted/date: 9 ? & s.
.- I 5$-7 /06 0 QR Review Required 0l MSS Review Required (reference NP 1.6.5)
] QRIMSS Review NOT Required (Admin or NNSR onl)
Date ]>-x DocumentOvriw -print/sign) 0(11 .~ .~.
reviews bay;-been indicates document 6i technically corrcet, can be performed as %vtiekdoes not adversely affect personnel or uear safety. appropriate affected co~mments have been resolved an IU.rated as appropriate, petfortned (i.e.. technical. cross-disciplinaty. %alidation and 50-59M7.48).
Control notified if emenrt issuance required (e g may . be documerants/ traininyjbricfing have been identified and word processing completed. Document less than 2 das for procedure issuance)
IV- APPROVAL (The Preparer, Qualified Reviieiir ( ), and Approval Authority shall be difftrent indihldual5)
QR[MSS (printlsign) ) 76 I Date determination mad a,; to %shetheradditional cross Indicat-s S.-39fl2.48 applicability assessed, any necessvty scret'runpcvaluations perfomed.
dAipiro y review required, and it required. performcd.
MSS Mfeeting No. _______
Date Aproravl Authority (print/sign)
V - RELEASE FOR DISTRIBUTION word processing, etc.).
] NA 0l YES Pre-implementation requircmcnts compktc (c.g, training/briefings, affccted documents, 0 Specific effective date not required. Issue per Document Control schedule.
0l Required effective date: Coordinate date %%thDocument
(_________
Doue!Owner/Desi-nee (print/sign) IC ~ \ " .L Ji...L.. Date
_t_ 1 3
_y_ _ _
Effectie Date (tobc entered byDocumcntControl).
nc n MfA!XY 1 20OZ MICROFILMED MV..JIJ a~ir , - - -- R~cfcr.ccs" .NP1.12-3. NP 1.1.5. NP 1 16 PBF-00261 NP12 3 NP 1 2 1. NT 1.2 6 r~csi,:in23 i1116'2 MAY 3 12002 Imp I-I-M11 ocmiili wasmtuiqut~suim
MAY I 23'.'7 Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Pa -e T___5 OM 4.3.2 Revision I Unit PB0 Doc Numbcr rile EOPIAOP VerifikationfValidation Process Temporary Change Number Description of Changes:
Step
- Change/Reason Co. er Sheet Added cover sheet per Procedures Writers' Guide. I This is pre-screened to Criteria #1 -Editora!.
Simplified PURPOSE statement and incorporated information from OM 4.3.3. EOP Validation. I This is 1.0 pre-screened to Criteria #2-Administrative Procedure.
(DISCUSSIONI per Procedures Writers' Guide. /This is pre-screened to Criteria #1 2.0 Editoral.new section Added 2.1 Incorporated information from the old PURPOSE sections of OM 4.3.2 and OM 4.3.3. / This is pre through screened to Criteria #1-Ed;toral.
2.2 to Criteria Added steep 2.3 in reference to the PRA Risk matrix in Attachment D. /This is p,e-screened 2.3 #2-Administrative Procedure.
2.4 Added steps. I This information is clarifying in nature. This is pre-scrcened to Criteria #2-Administrative through Procedure.
2.7 3.1 Incorporated information from the old RESPONSIBILITIES sections of OM 4.3.2 and OM 4.3.3. / This through is pre-screened to Criteria #I -Editoral.
3.3 Added (Nuclear Engineering I responsibility. I This is pre-screened to Criteria 42-Administrative Procedure.
3.4 Added (Reactor Engineering) responsibility. IThis is pre-screened to Criteria #2.Administrative Procedure.
be located Added (General) step to Section 4.0. I This allows a place for non-specific information to 4.1 together. / This is pre-screcried to Criteria #2-Administrative Procedure.
4.1.1 Incorporated information from the old PROCEDURE sections of OM 4.3.2 and OM 4.3.3./This is pre through screened to Criteria #1-Editoral.
4.1.2 Added step to cross reference other procedures in the z%,L'.t standard steps are revised. /This is pre 4.1.3 screened to Criteria #2-Administrative Procedur,..
to Criteria #2 Added step to reference the EOPSTPT for applicable changes. / This is pre-screened
- 4. 1.4 Administrative Procedure.
section of OM 4.3.2. / This is pre-screened to 4.1.5 Incorporated information from me old PROCEDURE 4.1.5 _ Criteria #I -Editoral.
Added step referencing the Deviation Document. /1 his is pre-scrccned to Criteria #2-Administrative 4.1.6 Procedure.
Added NOTE./This information is clarifying in nature. This is prc-scrcencd to Criteria #2 4.2.1 NOTE Administrative Procedure.
Other Cninments formrition or %%hennot bcneficiat to re'.cv-'arrM I Nowc Rcc,.ding of Step Numbers, is not required tor multiple occurrences of tdent:.:al -
PI3F-0026e Rcftiret,c¢c" NP I.I 1. *P 1 21 RciisionI 6041h/01 z L7
Point Beach Nuclear Plant MAf 1 3 2,?7, DOCUMENT REVIEW AND APPROVAL CONTINUATION Page.* Jof 5__
Doc Number OM 4.3.2 Revision I Unit PBO Title EOPIAOP VerifictionlValidation Process Temporary Change Number Description of Changes:
Step
- Change/Rea-on Added new step to diffirentiate between Technical changes and Editorial changes to EOPsIAOPs. /This 4.2.1 is pre-screened to Criteria #2-Administrative Procedure.
Added new step to reference new Attachment A for Tech. Evaluation Guidelines. I This is pre-screened to Criteria #2-Administrative Procedure.
4.2.3 Incorporated information from the old PROCEDURE section of OM 4.3.2. /This is pre-screened to Criteria # l-Editoral.
Incorporated information from the old PROCEDURE section of OM 4.3.2. Revised (Verification Teamn 4.2.4 requirements. Added PRA Group reference. / Expanded Team member requiremer:s and PRA Group involvement allows for more accurate evaluation. This is pre-screened to Crittera #2- Administrative Procedure.
4.2.5 Incorporated information from the old PROCEDURE section of OM 4.3.2./ This is pr*-screened to thrg Criteria #1-Editoral.
4.2.8 Added NOTE. I This information is clarifying ia nature. This is pre-screened to Criteria #2 Administrative Procedure.
4.2.9 Added step referencing a safety review. /This is pre-screened to Criteria #2-Administrative Procedure.
4.2.101Note Added this NOTE and steps to ensure specific groups reviewlevaluate procedure changes that effect their through areas of responsibility. /This is pre-screened to Criteria #2-Administrative Procedure.
4.2.13 Incorporated information from the old PROCEDURE section of OM 4.3.2, and revised and reformatted Section 4.3 the steps within the section to be consistent with the Validation steps. / This is pre-screened to Criteria
- 1 -Editoral.
Section 4.4 Incorporated information from the old PROCEDURE section of OM 43.3. /This is pre-screened to Criteria #t-Editoral.
4A.1rug Incorporated information from the old PROCEDURE section of OM 4.3.3. /This is pre-screened to through Criteria #1-Editoral.
4.4.L.b Added new step referencing Table-top validation method. I This is pre-screened to Criteria #2 Administrative Procedure.
Added new step defining the Validation Team Leader qualifications. / This is pre-screened to Criteria #2 Administrative Procedure.
4.4.3.a Incorporated informatior. from the old PROCEDURE section of OM 4.3.3. / This is pre-screcne4 to 4 Criteria Wl-Editoral.
Other Comments
" Note: Recording of Step Numbens) is not required for multiple occurrences of identical information or w,,hen not benefictal to rec'ic, ers PBF-00Z6c Pe,.ision6 0411/1 ecfercnces. NP 1 1.3. NP 1.2 3
IPAg oI _5 _
Print Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUAT!ON Page _Aiof *5 Doc Number OM 4.3.2 Revision I Unit PB0 Title EOP/AOP VerificationfValidation Process Temporary Change Number Description of Changes:
Step
- Change/Reason 4.4.3.b Added new sub-Ntcps dcaning the requirements of the Validation Team. I This is pre-screcned to Criteria through #2-Administrative Procedure.
4.4.3.d 4.4.3.e Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to Criteria #1-Editoral.
Added new step dirccting the Validation Team to review the Verification Teams work. I This is pre 4.4.5 _ screened to Criteria #2-Administrative Procedure.
Incorporated information from the old PROCEDURE section of OM 4.3.3./This is pre-screened to thrug Criteria #1-Editoral.
4.4.6 4-5.1.a Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to through Criteria #1-Editoral.
4.5.1 .d Added new step for the evaluation of the Simulator response./ This is pre-screened to Criteria #2 4.5.1 .e Administrative Procedure.
4.5.1.f Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is prc-screened to Criteria # 1-Editoral.
Added new step to define the course of the simulator scenario pcrformance. / This is pre-screened to 4.5.2.a Criteria #2-Administrative Procedure.
4.5.2.b Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to through Criteria #1-Editoral.
4.5.2.c Added new step to direct use of alternative methods of validation for parts of the procedure that are not evaluated by the simulator. /This is pre-screened to Criteria #2-Administrative Procedure.
Section 4.6 Added new section to dcfrine the steps to be followed during a Walkthrough Validation. / This is pre screened to Criteria #2-Administrative Procedure.
Added new section to define the steps to be followed during a Table-top Validation. / This is pre screened to Criteria #2-Administrative Procedure.
"Section4.8 Added NOTE to direct the re-performance of portions of the verification or validation processes. I This is NOTE pre-screened to Critcria #2-Administrative Procedure.
4.8.1 Incorporated information from the old PROCEDURE section of OM 4.3.3. /This is pre-screened !o
.4 through Criteria #Il-Editoral.
4.8.2 4.8.3 Added new step for the evaluation of the Simulator response. /This is pre-screened to Criteria #2 Administrative Procedure.
Other Comments
- Note: Recording of Step Ncmbcrts) is not required for rnultiplc occurrences of identiczi mnformz+/-ioa or %%hcnnot bcneficiaI to reviecers PBF-0026c Revision6 04t8fI0l Rcferenccs NP i.1.3. NP I 2.3 P I R 8,M1
Point Bcach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page .Iof_5 Doc Number OM 4.3.2 Revision I Unit PBO Title EOP/AOP Verification/Validation Process Temporary Chaisge Number Description of Changes:
Step
- Change/Reason 4.8A Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to through Criteria #I -Editoral.
4.8.5 4.8.6 Added new step describing the post-validation responsibilities of the Team Leader. I This is pre-screened to Criteria #2-Administrative Procedure.
Added NOTE. / Tiis information is clarifying in nature. This is pre-screened to Criteria #2 Administrative Procedure.
4.9.1 Added new step describing the final approval process. / This is pre-scrcened to Criteria #2 Administrative Procedure.
4.9.2 Incorporated information from the old PROCEDURE section of OM 4.3.3.1 This is prc-screened to Criteria # l-Editoral.
4.9.3 Added new step describing the Operations Manager responsibilities. I This is pre-screened to Criteria #2 Administrative Procedure.
5.1 Incorporated information from the old PROCEDURE section of OM 4.3.2. Deleted references to INOM through EOP), (NP 1.2.2) and {PBNPEOP}./ (NOM EOP) and {NP 1.2.2) have been canceled. {PBNPEOP) 5.9 is a redundant reference. This is pre-screened to Criteria #1-Editoral.
5.10 Added references to new forms developed from the forms in the old OM 4.3.3. / This is pre-screened to through Criteria #2-Administrative Procedure.
5.12 6.0 Added new scztion (BASIS) per Procedures Writers' Guide. /This is pre-screened to Criteria #1 Editoral.
Attachment A Added new Attachment A to provide guidance for Technical Evaluations. / This is pre-screened to Criteria #2-Administrative Procedure.
Added new Attachment B to provide guidance for Status Tree Evaluations. I This is pre-screened to Criteria #2-Administrative Procedure.
Attachment C Incorporated information from the old TABLE I of OM 4.3.3. /This is pre-scrcened to Criteria #1 Editoral.
Attachment C Added new step for guidance in the validation of actions taken outside the Control Room. / This is pre step 2.3 screened to Criteria #2-Administrative Procedure.
Attachment D Added new Attachment D to provide guidance for PRA Core Damage Risk Assessment. /This is pre screened to Criteria #2-Administrative Procedure.
5t/,.'( r"i-i r i T c~af.,%lA,/s., e"-1 q. "3".* .- @. ".~
- n'-i Other Comments "INote- Recording of Step Nurnibersi is not ir.quired for multiple occurrences of identicIl mnfomrrison c.- %%hennot bcncfici. to rcvice';rs PBF-0026c Revision 6 04/11 &ot1References NP 1.1 3, NP 1 2 3
Point Beach Nuclear Plant 10 CFR 50.59172.48 APPLICABILITY FORM Pace I Brief Activity Title Total rewrite of ONI 4.3.2, EOP/AOP VerificationfValidation Process or
Description:
This form is required to be completed and attached to the applicable activity change forms to document all or portions of an activity that are covered by another regulation other than 10 CFR 50.59 and 10 CFR 72.48 (pre-screening criteria 2). See NP 5.1.8, 10 CFR 50.59/72.4S Applicability, Screening and Evaluation (New Rule).
NOTE: Guidance for searching the FSAR, Technical Specifications, Regulatory Commitments (CLB Commitment Database) 2nd other licensing basis documents can be found in NP 5.1.8, Attachment G.
NOTE: Although 10 CFR 50.59 and 72.48 may not be applikable to the processes listed below, change activities conducted under these processes may require changes to the FSAR. If so, initiate FSAR changes per NP 5.2.6, FSAR Revisions.
Regulatory or Plant Process YES NO
- 1. Does the activity require a change to the Facility Operating License, License Conditions or Technical Specifications? (If the answer is YES, process the applicable changes per EL 0 NP 5.2.7, License Amendment Request Preparation, Review and Approval.)
- 2. NOTE: The Quality Assurance Plan is described in FSAR Section 1.4.
Does the activity require a change to the Quality Assurance Program? If the answer is I [
YES, process the applicable changes per NP 11.1.3, QA Program Revisions.
- 3. NOTE: Implementation of Security Plan changes that require physical changes to the plant, or changes to operator access to the plant require a screening.
NOTE: Security is described in FSAR Section 12.7.
Does the activity require a change to the PBNP Security Plan, a safeguards contingency plan, or security training and qualification plan? If the answer is YES, assess the acceptability of the change per 10 CFR 50.54 (p) using Security procedures.
- 4. NOTE: The Emergency Plan is described in FSAR Section 12.6.
Does the activity require a change to the Emergency Plan? If the answer is YES, process the applicable changes per NP 1.8.1, Emergency Preparedness Procedures.
- 5. NOTE: The Radiation Protection Program is described in FSAR Section 11.4.
Does the activity require a change to the PFaNP Radiation Protection Program described in NP 4.2.9, Radiation Protection, OR is the activity within the scope of NP 4.2.9 and 10 CFR 20, Standards for Protection Against Radiation?
- 6. NOTE: Change.; to the plant or method of evaluation that result in re-analysis of the FSAR loss-of-coolant accident (LOCA) analysis rec'ire a screening.
Does the activity require a change to the FSAR LOCA analysis r ,..lts subject to 10 CFR 50.46, Acceptance Criteria for Emergency Core Coolii.. Systems for Light Water Nuclear Power Reactors? If the answer is YES, process the applicable changes per NP 5.2.12, 10 CFR 50.46 Reporting Requirements, and NP 5.2.6 FSAR Revisions.
- 7. NOTE: Regulatory commitments are found in the CLB Commitment Database.
Does the activity involve a change to a Regulatory Commitment ? If the answer is YES. LI [
process the applicable changes per NP 5.1.7, Regulatory Commitment Changes.
S. Docs the activity involve a change to the Environmental Manual (EM), Radiological Effluent Control Program Manual (RECM), Offsite Dose Calculation Manual (ODCM).
or Process Control Program (PCP), AND does NOT involve changes in use of explosive El Z gases in waste treatment systems? If the answer is YFS, document the applicable changes per the requirements of TS 15.7.8.7.B {ITS 5.5.1}.
PBF-1515i Pi... ,,onn fl 1/2_4/"1N Rcft-rence. NP 5 1 9
Point Beach Nuclcar Plant 10 CFR 50.59/72.48 APPLICABILITY FORM Page 2 YEiO YES INO Regulatory or Plant Process applicability, the NOTE: For purposes of detcrmining 10 CFR 50.59 / 72.48 below takes precedence determination of an administrative procedure procedures or guidelines.
over definitions or classifications in other plant procedure or controlled document 9, Does the activity require a change to an administrative ONLY? [] 1 or controlled documevt ALL of the following statements shall be true for the procedure to be considered administrative.
are operated,
- a. DOES NOT direct how plant structures, systems, or components generically.
maintained, tested or repaired either specifically O.R.
for plant structures,
- b. DOES NOT specify acceptance criteria or operating limits systems, or components.
etc. to be uscd in plant
- c. DOES NOT specify parts, materials, chemicals, lubricants, structures, systems, or components.
structures, systems, or
- d. DOES NOT specify compensatory action(s) to address plant conditions.
components out of service, or to address non-conforming of the plant.
- e. DOES NOT affect operator access to operating areas -A-.'-
10 CFR 50.59172.48 APPLICABILITY CONCLUSION one or more of the processes above, further NOTE: If ANY portion of the activity is NOT controlled by covered by the above processes shall be 10 CFR 50.59172.48 review is required (i.e., portions not prescreened to other criteria or screened).
NO YES NO ALL aspects of the activity are controlled by one or more of the processes above, therefore 0 ['
additional 10 CFR 50.59 and 72.48 review is required.
NOT covered by one or more If the above question is answered NO, briefly describe the portions of the activity of the above processes:
I _ Date , 0 Performed By JamesGGretn 2nat " atured,
Name (.Print Date Reviewed By Name (Print) Signature PBF-1515aRfeec"P5 vF.t;;nnnIS: -iot
OM 4.3.2 EOP/AOP VERIFICATION/VALIDATION PROCESS DOCUMENT TYPE: Administrative 2
REVISION:
May 13, 2002 EFFECTIVE DATE:
Department Manager APPROVAL AUTHORITY:
Group Head PROCEDURE OWNER (title):
Operations OWNER GROUP:
POINT BEACH NUCLEAR PLANT OM 4.3.2 Revision 2 OPERATIONS MANUAL May 13,2002 TOTAL REWRITE EOP'AOP VERIFICATION/VAUIDATION PROCESS 1.0 PURPOSE and validation The purpose of this procedure is to establish the requirements for the verification Operating Procedures processes for the Emergency Operating Procedures (EOP) and Abnormal (AOP).
EOP, The verification and validation processes are applicable to procedures designated with ECA, SEP, CSP, ST, and AOP.
2.0 DISCUSSION that the 2.1 Verification of EOPs and AOPs is the process of independently checking ERG/ARG procedures are technically correct, that any deviations from the corresponding with plant hardware, and that guidance are justified, that the procedures are compatible Guide.
the procedures adhere to the guidance in OM 4.3.1, AOP and EOP Writers' ensure that they 2.2 Validation of EOPs and AOPs is the process of exercising procedures to that the are usable, that the language and level of information is appropriate, and of this procedure are procedures will function as intended. The validation requirements not applicable to revisions made for the correction of typographical errors.
with a frequency of 2.3 The matrix in Attachment D was developed based on initiating events that I E-3. The core damage greater than IE-6 and an initiating event frequency of greater operator would most selected scenarios were then compared to the procedures that the would likely use to prevent core damage. It is expected that procedure validation is risk based only and should consider those scenarios where an X is marked. This matrix procedural validation.
not be used as the sole consideration for determining scenarios for reasons:
2.4 EOPs, AOPs, and supporting documentation are revised for the following
- Plant design changes
- Operator comments or change requests 0 Industry or plant operating experience 0 ERG or ARG revisions
- Corrective action program
- Tech Spec changes Revisions to other related program instructions Page 2 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOPIAOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 2.5 EOP revisions associated with design changes, Tech Spec changes, or other related procedure changes should normally be implemented concurrently with the change. EOP revisions required to correct technical deficiencies in the EOPs shall be completed in a timely manner.
2.6 Operator requalification training on EOPs provides a means of periodically verifying the technical adequacy of emergency procedures. Operators and training personnel are responsible for ensuring tat problems or discrepancies discovered in EOPs during training are documented. Proposed enhancements and suggestions for improvement of the EOPs should also be encouraged.
2.7 Temporary changes to the EOPs and AOPs will be processed and controlled by NP 1.2.3, technical Temporary Procedure Changes. These changes are usually limited to emergent changes and do not require verification or validation per this procedure.
3.0 RESPONSIBILITIES 3.1 Manager's Supervisory Staff (MSS)
The MSS shall have the responsibility of reviewing and approving revisions to the EOPs and AOPs.
3.2 Operations Manager The Operations Manager shall have the overall responsibility for the EOP Verification and Validation processes.
The Operations Manager shall designaze the personnel who will comprise the Verification Team.
3.3 EOP Writer documents The EOP writer shall determine the need for revision of the EOP supporting and develop revisions for those documents as necessary.
3.4 Nuclear Eneineering the Nuclear Engineering should coordinate the nc-cessary changes if a revision to EOPSTPT is required.
3.5 Reactor Eneineering System Reactor Engineering should initiate revisions to the Safety Parameter Display (SPDS) if revision to CSP-ST.0, Critical Safety Function Status Trees, are required.
These revisions shall not be implemented until approval of the CSP-ST.0 revision.
Page 3 of 28 INFORMATION USE
. . . - ~~~~
I Wý . . .. . - - N o -' I.* * * -- l ow--
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE 4.0 PROCEDURE 4.1 General 4.1.1 The Emergency Response Guideline (ERG) or Abnormal Response Guideline (ARG) documents shall be reviewed to evaluate the intent of the corresponding ERG/ARG steps and whether the proposed change constitutes a deviation from the WOG guidelines.
4.1.2 The applicable EOP Deviation Document shall be reviewed to ensure that previous commitments are properly evaluated and to assess the justification for the present version of the step.
4.1.3 Similar or related steps/actions contained in other emergency procedures shall be evaluated for potential impact.
4.1.A When setpoints are involved, the EOP Setpoint Document (EOPSTPT) shall be reviewed to ensure that setpoints are correctly implemented and to determine if revision of the EOPSTPT is required.
4.1.5 Review the applicable portions of OM 4.3.1, AOP and EOP Writers' Guide, to ensure compliance with the writers guide.
4.1.6 All safety related deviations from the WOG guidelines shall be documented and justified in the associated Deviation Document.
4.2 Verification Process NOTE: Technical changes involve any of the following:
- Changing the method of performing a step or the sequential order of steps
- Changing the intent of any step, note, or caution
- Adding, deleting, or changing numerical values, limits, bands, or setpoints 0 Changing instrumentation or controls used in the procedure
- Changing entry/exit conditions or symptorm-s 0 Addition or deletion of steps, notes, cautions, graphs, tables, etc.
0 Any change which deviates from the WOG guidelines 4.2.1 Technical changes to EOPs should be verified by a multi-discipline team (at least three members) to maximize effectiveness of the verification process.
Non-technical (editorial) changes to EOPs and changes to AOPs may be verified by a single i.ndividual provided that the individual is a licensed operator and a qualified reviewer.
Page 4 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATIONJVALIDATION PROCESS TOTAL REWRITE 4.2.2 Technical changes to EOPs and changes to AOPs should be evaluated using Attachment A, Technical Evaluation Guidelines. Changes to Critical Safety Function Status Trees should be evaluated using Attachment B, Status Tree Evaluation Guidelines.
4.2.3 To ensure an independent verification process, personnel who have been involved in the development of the procedures(s) being verified should not be selected as verifier or appointed to the Verification Team.
4.2.4 The Verification Team members shall consist of, as a minimum, a Chairman, a licensed operator (SRO or RO), and a Training representative. Other members should be selected based on the type of change(s) being made to the procedure. For technical changes, a member of the PRA Group should review the changes but does not have to be a part of the Verification Team meeting.
4.2.5 The Verification Team members shall be listed on PBF-2102a, EOP Verification Team Meeting Form.
4.2.6 Verification Team members should obtain source documents as necessary, such as WOG guidelines, Deviation Documents, and Background Documents.
Other documents such as Tech Specs, FSAR, and other supporting procedures may also be applicable.
4.2.7 Review applicable portion(s) of the revised procedure. Depending upon the scope of the revision, it may be necessary to review the entire procedure and other interfacing procedures to adequately verify the revision. If step numbering or sequencing is affected by the revision, then the entire procedure shall be verified for internal step number referencing.
NOTE: Minor discrepancies may be resolved by the Verification Team without the use of PBF-2102b, EOP Verification Discrepancy Form.
4.2.8 Identify and document discrepancies on PBF-2102b, EOP Verification Discrepancy Form.
4.2.9 A safety evaluation, in addition to the screening review, should be I::,pared for changes which involve new deviations from the WOG guidelines.
Page 5 of 28 INFORMATION USE
OM 4.3.2 POINT BEACH NUCLEAR PLANT Revision 2 OPERATIONS MANUAL May 13,2002 TOTAL REWRITE EOP/AOP VERIFICATION/VALIDATION PROCESS be performed NOTE: The required reviews cuatained in the following steps may concurrently with the verification process if the appropriate personnel review are part of the verification team. If performed separately, the should be identified as a Cross-Discipline Review. The Operations procedure writer is responsibIc for ensuring that assigned reviewers understand the scope of the review required.
of the 4.2.10 Engineering shall review EOP/AOP revisions which involve any following:
or scope
- b. Addition, deletion, or changes in setpoints or setpoint usage.
- c. Changes to status trees or other changes affecting SPDS displays.
which could
- d. Additions or changes to actions outside the control room impact radiation dose estimates.
affect compliance
- e. Changes in instrumentation used in EOPs which could with Reg Guide 1.97, Post Accident Monitoring Instrumentation.
- f. Proposed revisions to AOPs that affect Technical Specifications surveillance requirements.
EOPs or AOPs 4.2.11 Reactor Engineering should review proposed revisions to which may affect Reactivity Management.
to EOPs or 4.2.12 The PRA Group should review any proposed major revisions AOPs.
Protection, 4.2.13 Organizations other than Operations (such as Chemistry, Radiation and AOPs which or Maintenance) should review proposed revisions to EOPs affect actions by the affected organization.
4.3 Resolution of Verification Discrepancies 4.3.1 Verification discrepancies are documented using PBF-2102b, will not undo EOP Verification Discrepancy Form, so that future revisions process.
corrections or improvements made as a result of the verification those 4.3.2 The Validation Chairperson shall assign personnel (preferably for each responsible for writing the procedures) to prepare a resolution discrepancy.
Page 6 of 28 INFORMATION USE
OM 4.3.2 POINT BEACH NUCLE',R PLAN.'T Revision 2 OPERATIONS MANUAL May 13,2002 TOTAL REWRITE EOP/AOP VERIFICATIONIVALIDATION PROCESS 4.3.3 The personnel assigned to resolve the discrepancy shall:
- a. Propose.a resolution to correct the discrepancy on PBF-2102b, EOP Verification Discrepancy Form.
as applicable.
- b. Obtain concurrence from the Verification Chairperson, resolution,
- c. If the Verification Chairperson does not concur with the coordinate efforts to assess and resolve the discrepancy.
Verification
- d. Document the final resolution on PBF-2102b, EOP Discrepancy Form.
assigned to 4.3.4 If the discrepancy cannot be resolved between the personnel then the Verification resolve the discrepancy and the Verification Chairperson, approval from the Chairperson shall recommend a corrective action and obtain Operations Manager or designee.
the Verification 4.3.5 After resolution of the discrepancy has been determined, Chairperson shall:
resolution of the
- a. Ensure the procedure is changed to incorporate the discrepancy.
required.
- b. Determine the scope of any additional verification
- c. Document completion of the additional verification.
notify the Training
- d. Determine if additional training is required and, if so, Department.
INFORMATION USE Page 7 of"28
OM 4.3.2 POINT BEACH NUCLEAR PLANT OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE 4.4 Validation Process 4.4.1 The validation method shall be selected using the following guidance:
- a. The simulator method is preferred and should be used, when practical, because this method:
More accurately demonstrztes operator response to a specific scenario.
"* Effectively identifies discrepancies between instructions and Control Room hardware.
"* Effectively identifies discrepancies between instructions and the operators execution of them.
- b. The walkthrough method should be used when:
Use of the simulator method is impractical due to modeling constraints or other limitations.
In combination with the simulator method when the simulator method is partially impractical.
- When the revision affects action taken outside the Control Room.
For changes which do not warrant simulator validation due to the nature or scope of the change.
NOTE: The walkthrough method is more effective than a table-top discussion in ensuring that the instructions contain the necessary level of detail and are compatible with plant hardware and personnel.
- c. The table-top method should be used only when th, simulator and walkthrough methods cannot be used effecti ;el," OR -'-r :rcinor ecitnrI'.. or technical revisions which do not involve pli nt ,,,: .ware and do iiot warrant simulator or walkthrough , aiid,-'ion.
Page 8 of 28 INFORMATION USE q.M. E- - gmna= IN1; 1 .P220NIM
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4A.2 The Validation Team Leader shall be designated based upon the scope of the validation and the vwlidation method(s) to be used. The Validation Team Leader should possess expertise in as many of the following areas as possible:
- a. Supervisory skills
- b. Plant Operations
- c. Operations Training
- d. Technical Bases
- e. Development of EOP/AOPs 4.4.3 The Validation Team members requirements should be based on the following:
- a. Technical changes to an EOP should be validated by a multi-disciplined team consisting of at least three members. Revisions to AOPs and minor changes to EOPs do not require a multi-disciplined team nor do they require a minimum of three team members.
- b. The Validation Team should collectively be knowledgeable in the following areas:
- Plant Operations
- Training/Simulator Instruction
- Technical Bases
- Development of EOP/AOPs
- c. At least one member of the Validation Team shall be a licensed operator.
The operations personnel used as the operating crew for the validation scenarios may be included as part of the Validation Team.
- d. At least one member of the Validation Team shall be a simulator instructor (N/A for walkthrough or tabletop validation methods).
- e. The Validation Team members shall be listed on PBF-2103a, EOP Validation Form.
Page 9 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 My 13, 2002 EOPiAOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.4.4 The Validation Team Leader shall review the PBF-2102a, EOP Verification Team Meeting Form and any PBF-2I02b, EOP Verification Discrepancy Form(s) to determine the validation methods to be used and identify significant changes incorporated into the new procedure revision.
4.4.5 The Validation Team Leader shall outline one or more scenarios encompassing the ;dent;ficd changes in the procedure. Select plant failures that will initiate the desire response. considering the following:
- a. Use both single and multiple failures where practical.
- b. Use concurrent and sequential failures where practical.
- c. Use dual unit failures where practical.
- d. If the simulator is to be used, select simulator malfunctions that closely model the selected failures.
4.4.6 Each validation scenario shall be documented using on PBF-2103b, EOP Validation Scenario Form.
4.5 Simulator Validation Method 4.5.1 The Procedure Writer or Validation Team Leader should prepare for simulator validation as follows:
- a. Schedule licensed uperators and a sirnulator instructor to participate in the simulator validation. Operators selected should be representative of the training level expected of all operators.
- b. Arrange for the needed resources to support the validation such as simulator time, copies of procedures and relate instructions, and copies of tfie scenarios .o be covered.
- c. Review the purpose and objective of the validation with the opcrator(s) involved. Include a discussion of the procedure revision.
- d. Brief the operators on how the validation will be conducted.
- e. Eva!uate any known simulator characteristics which are different from the actual plant responses for impact on the validation.
Page 10 of 28 INFORMATION USE a,
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOPIAOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE
- f. Prior to beginning the scenario, the Validation Team will discuss any differences between units that may come into play during execution of the scenario. The Validation Team Leader should ensure that the operators are aware of these differences and what effect they have on execution of the steps to be validated.
4.5.2 Conduct of the Simulator Method.
- a. The operators will use the procedures in response to the sc,.aario enacted on the simulator. The procedure writer may be present but should not interfere or provide guidance during the scenario.
- b. The Validation Team will assess the procedures by noting any problems or deviations during the simulator run.
- c. At the conclusion of each simulator run, the Validation Team will conduct a debriefing as follows:
a Evaluate the instruction using Attachment C, Validation Guidelines and document all discrepancies on PBF-2103c, EOP Validation Discrepancy Form.
0 Allow operators to present any problems or discrepancies that they identified during the simulator run. Document all discrepancies identified.
- Discuss any deviations noted during the simulator run to identify discrepancies in the procedures.
- d. Any portions of the procedure or other procedures impacted by the revision which cannot be validated on the simulator should be validated separately using the walkthrough or tabletop methods.
Page I11 of"28 INFORMATION USE
=- - 030nw
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOPIAOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.6 Walkthrough Validation Method 4.6.1 The Procedure Writer or Validation Team Leader should prepare for walkthrough validation as follows:
- a. Schedule personnel to participate in the walkthrough. Individuals selected should be representative of the training level expected of all similarly qualified personnel.
- b. Arrange for the needed resources to support the validation such as copies of procedures and relate instructions, and copies of the scenarios to be covered, and related technical documentation.
- c. Review the purpose and objective of the validation with the personnel involved. Include a discussion of the procedure revision.
- d. Brief the personnel on how the validation will be conducted.
- e. Prior to beginning the walkthrough, the Validation Team will discuss any differences between units that may come into play during execution of the walkthrough. The Validation Team Leader should ensure that the personnel are aware of these differences and what effect they have on execution of the steps to be validated.
4.6.2 Conduct of the Walkthrough Validation
- a. Walkthrough validation should be performed at the in-plant location(s) where the procedure would be performed.
- b. If the procedure being validated is written for either unit, then a walkthrough should be performed on both units.
- c. The Validation Team Leader will use the scenario to direct the walkthrough by first providing the plant initial conditions and then providing appropriate cues while the personnel walk through each procedure step.
Page 12 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE
- d. The personnel will use the procedures in accordance with the scenario and walk through or talk through actions they would take in response to each instruction step. Personnel should:
"* Describe actions they are taking.
"* Identify information sources uscd to take actions.
"* Identify controls used to carry out actions expected system response(s), how response(s) are verified, and action(s) to be taken if response(s) did not occur.
- e. At any time during the walkthrough, personnel may stop to identify any problems or discrepancies in the procedures. Validation Team members may ask questions during the validation.
- f. The Validation Team will assess the procedures by noting any performance problems during the walkthrough.
- g. At the conclusion of each walkthrough, the Validation Team will conduct a debriefing as follows:
- Evaluate the instruction using Attachment C, Validation Guidelines.
"* Review comments made during the walkthrough and document all discrepancies identified.
"* Discuss any performance deviations to identify discrepancies in the procedures which resulted in the deviation.
Page 13 of 28 INFORMATION USE MROM
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFCATIONIVALIDATION PROCESS TOTAL REWRITE 4.7 Table-Top Validation Method 4.7.1 The Procedure Writer or Validation Team Leader should prepare for table-top validation as follows:
- a. Schedule personnel to participate in the validation. Individuals selected should be representative of the training level expected of all similarly qualified personnel.
- b. Arrange for the needed resources to support the validation such as copies of procedures and relate instructions, and the scenarios to be covered.
- c. Review the purpose and objective of the validation with the personnel involved. Include a discussion of the procedure revision.
- d. Brief the personnel on how the validation will be conducted.
- e. Prior to beginning the scenario, the Validation Team will discuss any differences between units that may come into play during execution of the scenario. The Validation Team should ensure that the personnel are aware of these differences and what effect they have on execution of the steps to be validated.
4.7.2 Conduct of the Table-Top "a. The Validation Team Leader will use the scenario to direct the table-top discussion by first providing the plant initial conditions and then providing appropriate cues while the performer discusses each procedure step.
- b. The personnel will use the procedures in accoidance with the scenario, discussing the actions taken in response to each instruction step while identifying any problems or discrepanciec, in the procedure(s).
- c. During the table-top, the Validation Team will discuss and evaluate the instructions against Attachment C, Validation Guidelines. All discrepancies from the checklist or from individual comments will be documented on an on PBF-2103c, EOP Validation Discrepancy Form.
- d. The Validation Team will assess the procedures by noting any pcrfoimance problems during the walkthrough.
- e. At the conclusion of the table-top discussion, the Validation Team will discuss any deviations to identify discrepancies in the procedures which resulted in the deviation and document all discrepancies on PBF-2103c, EOP Validation Discrepancy Form.
Page 14 of 28 hNFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE NOTE: EOP/AOP changes to resolve verification/validation discrepancies may require repeating portions of the verification and/or validation process.
4.8 Resolution of Validation Discrepancies 4.8.1 Validation discrepancies are documented using form PBF-2103c, EOP Validation Discrepancy Form, so that future revisions will not undo corrections or improvements made as a result of the validation process.
4.8.2 The Verification Team Leader shall assign personnel (preferably those responsible for writing the procedures) to prepare a resolution for each discrepancy.
4.8.3 Discrepancies involving plant response from simulator validation shall be evaluated to determine if they were caused or aggravated by simulator modeling deficiencies.
4.8.4 The personnel assigned to resolve the discrapancy shall:
- a. Propose a resolution to correct the discrepancy on PBF-2103c, EOP Validation Discrepancy Form.
- b. Obtain concurrence from the Validation Team Leader, as applicable.
- c. If the Validation Team Leader does not concur with the resolution, coordinate efforts to assess and resolve the discrepancy.
- d. Document the final resolution on PBF-2103c, EOP Validation Discrepancy Form.
4.8.5 If the discrepancy cannot be resolved between the personnel assigned to resolve the discrepancy and the Validation Team Leader, then the Validation Team Leader shall recommend a corrective action and obtain approval from the Operations Manager or designee.
Page 15 of 28 INFORMATION USE
OM 4.3.2 POINT BEACH NUCLEAR PLANT Revision 2 OPERATIONS MANUAL May 13,2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.8.6 After resolution of the discrepancy has been determined, the Validation Team Leader shall:
- a. Ensure the procedure is changed to incorporate the resolution of the discrepancy.
- b. Determine the scope of any additional validation required.
- c. Document completion of the additional validation.
- d. Determine if additional training is required and, if so, notify the Training Department.
4.9 Final Approval of EOP/AOP Revisions NOTE: Temporary changes to the EOPs and AOPs can be approved via NP 1.2.3, Temporary Procedure Changes.
4.9.1 Following completion of the verification and validation process, including resolution of all discrepancies, final approval is obtained.
If 4.9.2 MSS review and approval is required for technical revisions to the EOPs.
the revised the basis and step deviation documents are affected by the change, with the EOP for MSS review.
background document should be submitted 4.9.3 All EOP/AOPs and background documents shall be approved by the Opera'ions Manager or his designee.
i I
Pae1o2 NOMTO S Page 16 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE
5.0 REFERENCES
5.1 NUREG 0899, Guidelines for the Preparation of Emergency Operating Procedures 5.2 NRC Generic Letter 82-33, Supplement I to NUREG-0737 - Requirements for Emergency Response Capability 5.3 C. W. Fay letter to H. R. Denton, "Response to Generic Letter No. 82-33 .... " April 15, 1983.
5.4 OM 4.3.1, AOP and EOP Writers' Guide 5.5 Westinghouse Owners' Croup (WOG), Emergency Response Guidelines (ERGs) 5.6 Westinghouse Owners' Group (WOG), Abnormal Response Guidelines (ARGs) 5.7 PBF-2102a, EOP Verification Team Meeting Form 5.8 PBF-2102b, EOP Verification Discrepancy Form 5.9 Institute of Nuclear Power Operations (INPO) Guidelines, Emergency Operating Procedures Verification Guidelines,83-004, March 1983 5.10 PBF-2103a, EOP Validation Form 5.11 PBF-2103b, EOP Validation Scenario Form 5.12 PBF-2103c, EOP Validation Discrepancy Form 6.0 BASES NONE Page 17 of 28 INFORMATION USE r * *J _ * -* ' * * :,r--
- __*.r_*_- _ * *_ *-*-OE M_
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOPIAOP VERIFICATION/VALIDATICN PROCESS - TOTAL REWRITE AIATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 1 of 3 1.0 (EOP)
Are entry conditions consistent with those listed in the Owner's Group guidelines or are deviaz.ionsjustificd in the basis and deviation documents (AOP/SEP)
Are entry conditions logical. (reflective of the expected conditions leading to performance of the instruction). Are the entry conditions observable.
2.0 (EOP)
Is the sequence of steps consistent with that in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOPlSEP)
Are the steps sequenced logically. Does the sequence follow good operations principles.
3.0 (EOP)
Are all steps consistent with the intent of those in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Is the intent of each step understandable. Does the step provide adequate detail.
4.0 (EOP)
Have all applicable Owner's Group Guideline steps been incorporated into the procedure or are deviations adequately justIfied in the basis and deviation documents.
(AOP/SEP)
Are the steps necessary instructions provided to the user.
5.0 (EOP)
Are differences from the Owner's Group Guidelines consistent with the intent of the Owner's Group Guidelines.
6.0 (EOP)
Is documentation adequate to explain the intent of complex steps.
(AOP/SEP)
Is documentation adequate to explain the intent of complex steps.
7.0 (EOP)
Is all Owner's Group Ct, idelines "bracketed" information, pertinent to the plant design, incorporated.
(AOP/SEP)
Is applicable plant design and components clearly addressed by the instruction.
Page 18 of 28 INFORMATION USE
OM 4.3.2 POINT BEACH NUCLEAR PLANT Revision 2 OPERATIONS MANUAL May 13,2002 TOTAL REWRITE EOP/AOP VERIFICATION/VALIDATION PROCESS ATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 2 of 3 8.0 (EOP) that are Have all references to systems or components in the Owner's Group Guidelines applicable to the plant design been included.
(AOP/SEP) and correct.
Are all references to system, component and plant design clear 9.0 (EOP) with Owner's Group Guidelines Are required computations, specified in the procedure. consistent or deviations adequately justified within source documents.
(AOP/SEP) guidance been given and Are all required computations specified in the procedure. Has adequate is space available for working and recording computations.
10.0 (EOP) with the Owner's Group Are the cautions and notes, as specified in the procedure, consistent documents.
Guidelines or are deviations adequately justified in the basis and deviation (AOP/SEP)
Do they provide adequate Are cautions and notes specified in the instruction clear and concise.
information to convey the message.
11.0 (EOP) specified in the Owner's Are the contingency actions in the procedure consistent with those and deviation documents.
Group Guidelines or are deviations adequately justified in the basis (AOP/SEP)
Do they provide adequate If specified/used, are contingency actions clear and easily understood.
detail for implementation.
12.0 (EOP) action steps of the procedure.
Is there a conflict between the foldout page requirements and the (AOP/SEP)
Is there any conflict between steps and required actions.
13.0 (EOP) design.
Are the required steps to be performed cor.si,,tent with the plant (AOP/SEP)
Are the steps consistent with plant design.
14.0 (EOP) with the plant design.
Are the quantitative ranges as specified in the procedure consistent (AOP/SEP) with the plant design.
Are the quantitative ranges as specified in the piocedure consistent Patge 19 of 2S INFORMATION USE
OM 4.3.2 POINT BEACH NUCLEAR PLANT Revision 2 OPERATIONS MANUAL May 13,2002 TOTAL REWRITE EOP/AOP VERIFICATION/VALIDATION PROCESS
. ATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 3 of 3 15.0 (EOP) with those specified in the Owner's Group Are the limits, as specified in the procedure consistent the basis and deviation documents.
Guidelines or are dcviations adequately justified in (AOPISEP)
Are limits clearly specified.
16.0 (EOP) procedure consistent with the Owner's Group Are the charts, tables, and curves presented in the the basis and deviation documents.
Guidelines or are deviations adequately justified in (AOPISEP) the information provided in source documents.
Are the charts, tables, and curves consistent with 17.0 (EOP) in the procedure correspond with the Do parameter values, numerical values, and setpoints in Setpoints Document.
parameter values, numerical values, and setpoints specified (AOP/SEP) in the procedure correspond with the Do parameter values, numerical values, and setpoints specified in supporting technical parameter values, numerical values, and setpoints documentation.
18.0 (EOP) all the procedures affected been revised.
If the revision involves a change to a setpoint, have in the setpoints document.
Verify against the list of affected procedures contained 19.0 (EOP) the procedures affected been revised. Verify If the revision affects a "standard" step, have all of in the standard step document.
against the list of affected procedures contained INFORMATION USE Page 20 of 28 W
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANU,-L Revision 2 May 13,2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTIACHMENT B STATUS TREE EVALUATION GUIDELINE Page I of 3 1.0 WRITERS' GUIDE CONVENTIONS 1.1 Procedure Title 1.1.1 Is the title 10 words or less.
1.1.2 Are the important words placed at or near the beginning of the title.
1.2 Identification Information 1.2.1 Does the procedure number include the required information:
- a. Instruction type
- b. Instruction number 2.0 STATUS TREE FORMAT 2.1 Page Format 2.1.1 Does the Status Tree clearly show the transitions.
2.2 Symbol Coding 2.2.1 Are the symbols used correctly.
2.2.2 Are arrows positioned correctly.
2.3 Function Flow and Branching 2.3.1 Does the flow path move from left-to-right.
2.3.2 Is sufficient spacing allowed between flow paths.
2.3.3 Are the number of arrowheads sufficient to indicate flow.
2.3.4 Does the flow path go down for each favorable response.
INFORMATION USE Page 2 of 28 11MOPRBW 11 mff
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE
- ATTACHMENT B STATUS TREE EVALUATION GUIDELINE Page 2 of 3 3.0 READABILITY 3.1 Text 3.1.1 Is the text in black type against a light background.
3.1.2 Is the text readable at arms length under degraded lighting conditions.
3.1.3 Is the typeface legible and consistent.
3.1.4 Is spacing between letters and words adequate.
3.1.5 Is the correct line spacing used.
4.0 WRITING STYLE
-4.1 Step Construction 4.1.1 Does eaLI. step contain only one statement.
4.1.2 Are the statements simple and precise.
4.1.3 Are double negatives avoided.
4.1.4 Are terms used consistently within and among status trees.
4.1.5 Does each decision step clearly indicate a yes or no answer.
5.0 MECHANICS OF STYLE 5.1 Spelling 5.1.1 Is the spelling correct.
5.2 Abbreviations and Acronyms 5.2.1 Are abbreviations and acronyms used consistently.
5.2.2 Are abbreviations used in accordance with the Writers' Guide.
Page 22 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE ATTACHMENT B STATUS TREE EVALUATION GUIDELINE Page 3 of 3 5.3 Curves and Tables 5.3.1 Are the curves and tables legible, consistent with the instructions, and usable.
5.3.2 Are the safe and unsafe regions of curves labeled.
5.4 Hyphenation 5.4.1 Are hyphens used correctly.
5A.2 Is hyphening at the end of a line avoided.
Page 23 of 28 INFORMATION USE
- POINT BEACH NUCLEAR PLANT OM..4.f..
OPERATIONS MANUAL Revision 2
.,ay 13.2002 EOP/AOP VERIFICATION[VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 1 of 4 1.0 USABILITY 1.1 Level of Detail 1.1.1 Are the introductory sections of the instruction sufficient.
1.1.2 Is there sufficient information to perform the specified act',ons at each step.
1.1.3 Are the alternatives adequately described at eacn decision step.
1.1.4 Are labeling, abbreviations, and nomenclature as provided in the instruction sufficient to enable the operator to find tihe needed equipment.
1.1.5 Does the instruction have all information or instructions needed to manage the emergency condition.
1.1.6 Are the actions sufficient to correct the condition.
1.1.7 Are the titles and numbers sufficiently descriptive to enable the operator to find appropriate instnictions.
1.2 Understandability 1.2.1 Is the instruction's typeface easy to read.
1.2.2 Are the figures and tables easy to read with accuracy.
1.2.3 Can the values on figures and chaits be easily determined.
1.2.4 Are the cautions a'nd note statements readily understandable.
1.2.5 Are the individual instruction steps readily understandable.
1.2.6 Were the step sequences understood.
Page 24 oi"28 INFORMATION USE 11m, g1:0.1 1111 C It II S 11Kil 1=
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 2 of 4 2.0 OPERATIONAL CORRECTNESS 2.1 Plant Compatibility 2.1.1 Can the actions specified in the procedure be performed in the designate sequence.
2.1.2 If alternate success paths exist, does the procedure use the best method to accomplish the task.
2.1.3 Can the information from the plant instrumentation be obtained, as specified, by the instructions.
2.1.4 Are the available Control Room instrumentation and annunciators adequate for the Operator to recognize the entry or prerequisite conditions.
2.1.5 Are the instructions entry or prerequisite conditions appropriate for the plant symptoms displayed to the operator.
2.1.6 Is all the equipment required to accomplish the task specified in the instruction.
2.1.7 Do the plant resources agree with the instruction.
2.1.8 Are the instrument readings and tolerances stated in the instruction consistent with the instrument values displayed on the instruments.
2.1.9 Is the instruction physically compatible with the work situation (e.g., too bulky to hold, binding would not allow them to lie flat in the work space, no place to lay the instruction down to use).
2.1.10 Are the instrument readings and tolerances specified by the instruction for remotely located instruments accurate.
2.1.11 Can plant parameters be maintained within limits or bands specified in the procedure.
Page 25 of 28 INFORMATION USE
- 1111 1911111 ilailligism il 11 1 1 MINIM1111011 I -
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MIANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATION/VALIDATiON PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 3 of 4 2.2 Operator Compatibility 2.2.1 If time intervals are specified, can the instruction action steps be performed on the plant within or at the designated time intervals.
2.2.2 Will environmental conditions permit completing the required actions.
can 2.2.3 If concurrent or sequential steps are required by more than one individual, the required actions be coordinated adequately.
2.2.4 Can personnel follow the designated action step sequences.
2.2.5 Can a particular step, set of steps, or other information be readily located when required.
2.2.6 Can instruction branches be entered at the correct point.
2.2.7 Are place keeping aids utilized as required by the user's guide.
2.2.8 Are instruction exit points adequately specified.
2.2.9 Are the procedures compatible with the operating shift manning.
2.2.10 If steps and instructions are verified with signoffs, are provisions adequate.
2.2.11 Do Operators interfere with each other physically.
make the 2.2.12 Is there adequate Radiation Protection support and/or provisions to required entries into contaminated areas.
2.2.13 Does plant staffing support procedure requirements.
complete the 2.2.14 Is the procedure adequate to allow properly trained personnel to task without errors.
Page 26 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 4 of4 2.3 Additional Guidelines for Validation of Local Operato. Actions 2.3.1 Can the Operator easily locate the component from a combination of the information in the procedure and operator training/knowledge.
2.3.2 Is the component clearly identified by name and/or number.
2.3.3 Is the component easily accessible.
2.3.4 Are special tools needed to operate the component.
2.3.5 Is the environment at the component location suitable to allow the operator to perform desired actions.
2.3.6 Do the local actions require more than one operator.
2.3.7 Are communications available from the remote location.
2.3.8 Is the Operator performing the local actions familiar with the procedure and does he/she understand the objective and/or consequences of his/her actions.
2.3.9 Are the local actions required to be performed in a specific time period. If so, can the actions be completed within this time period.
Page 27 of 28 INFORMATION USE
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13,2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE
. ATTACHMENT D PRA CORE DAMAGE RISK MATRIX Page I of I Procedure EVENT SGTR Turbine Trip LOOP Loss of CCW Steam Line without the Break Condenser ECP 0 X X X X X EOP 0.0 ..........
EOP0.1 -= X X X -
EOP 0.2 - ..... X X -
EOP 0.3 -- X X EOPO.4 -- -- X X -
EOPI X -- X -- X EO P 1.1 ..........
EOP 1.2 " X ... X -
I '-. EOP 1.3 ..... X ....
EOP 2 ........ X EOP 3 X ...... X "EOP3.1- - X ........
EOP 3.2 X ........
EOP 3.3 X ........
- ECA 0.0 - -° .. X ....
ECAO.1-. .... X ....
ECA 0.2 ..........
ECA 1.1 ..........
ECA 1.2 ..........
ECA 2.1 "X ...... X ECA 3.1 X ........
ECA 3.2 X ........
ECA 3.3 X ....
CSP C. X-- X X X CSP H.1 X X -X Page 28 of 28 INFORMATION USE
Point Beach Nuclear Plant DOCUMENT FEEDBACK FEEDBACK REQUEST Document Number OM 4.3.2 Revision 3 Unit 0 Title EOP/AOP verification/validation process Requested Change (attach mark-up as necessary): Step 4.2.4 states "other members should" change to" other members shall". Step 4.2.12 states "The PRA group should" change to "The PRA group shall". Shall is more appropriate for these statements.
Reason for Change: The original intent of these steps was to be a shall.
Suggested Priority ( [] Immediate Action [] Sta= Revision 10 Next Revision ) Date Needed (if applicable)
Requested By (print and sign) J. Pruit I Date 9/18/02 DISPOSITION El APPROVED (0D Immediate Action 0] Start Revision 0] Next Revision)
This issue Ml DOES NOT / [] DOES require an Action Request according to NP 1.1.4 and NP 5.3.1.
AR No.
El REJECTED (include reason below)
Comments:
Document Owner (print and sign) / Date (Forward copy to requestor and original to procedure writer)
PBF-0026p Revision 3 1116102 Reference. NP I 1.3, NP I I 4
Flessner, Richard "V-o0m: Pruitt, Jerry it: Wednesday, September 18, 2002 11:09 AM Flessner, Richard
Subject:
FW: Feedback Additional info for you. Doc
-Original Message-From: Vandenbosch, Terry Sent. Wednesday, September 18, 2002 10:10 AM To: Pruit Jerry
Subject:
RE: Feedback I'll add it. The number is OPS 2002-01364.
Onginal Message From: Pruitt, Jerry Sent: Wednesday, September 18, 2002 10:02 AM To: Vandenbosch, Terry Cc: Pruitt, Jerry
Subject:
RE: Feedback Terry, On step 4.2.12, you may want to also change "major" to "technical". Should be ok without it, but would be a little cleaner. Thanks, Doc Requested Change (attach mark-up as necessary): Step 4.2.4 states "other members should" change to" other members shall". Step 4.2.12 states "The PRA group should" change to "The PRA group shall". Shall is more appropriate for these statements.
Onginal Message From: Vandenbosch, Terry Sent: Wednesday, September 18, 2002 9:42 AM To: Pruitt, Jerry
Subject:
Feedback Attached is a copy of the feedback. I won't have a number until tomorrow.
<< File: OM 431 feedback.doc >>
1
/
Nuclear Management Company Page 1 of 2 STATE CHANGE-- HISTORY /
- ~ -w s -~- -
Conduct Work Review & Quality Initiate Assign Work Assign Work Complete Approval Approved Check 9/3/2002 9/3/2002 9/3/2002 9/3/2002 7.07 24 PM 7.08.14 PM 7:13.12 PM 7:13 41 PM Owner Owner Owner Owner by RICHARD RICHARD by RICHARD by RICHARD RICHARD by RICHARD RICHARD FLESSNER PBNP CAP FLESSNER FLESSNER FLESSNER FLESSNER FLESSNER Admin FLESSNER P.
SECTION 1 Activity Request Id: CA026225 Activity Type: Corrective Action Submit Date: 9/3/2002 7:07:24 PM Site/Unit: Point Beach - Common Activity Requested: CA#19. Modify the AFW recirculation valves to provide a back-up pneumatic supply to allow time for operator actions.
0 CATPR: N Initiator: MASTERLARK, JAMES Initiator Department: EPN Engineering Responsible Group Code: EXC Engineering Programs Nuclear Processes Continuous Safety Analysis PB Improvement PB Z D*
Responsible Department: Engineering Activity Supervisor: RICHARD FLESSNER Activity Performer: RICHARD FLESSNER SECTION 2 Priority: 3 Due Date: 9/3/2002 "0Mode Change Restraint: (None) Management Exception From P1?: N "0QA/Nuclear Oversight?: N 0 Licensing Review?: N NRC Commitment?: Y 0 NRC Commitment Date:
SECTION 3 Activity Completed: 9/3/2002 7:07:24 PM - RICHARD FLESSNER:
This CA is being issed to document a completed corrective action.
9/3/2002 7:13:12 PM - RICHARD FLESSNER:
MR 01-144 was initiated to provide N2 back-up to AFW mini-flow valves AF-4007 and AF 4014. MR 01-144 was accepted on 2/6/02. MR 02-001 was initiated to provide air back-up to AFW mini-flow valves 1/2AF-4002. MR 02-001 was accepted on 411102.
SECTION 4 QA Supervisor: (None) Licensing Supervisor: (None)
SECTION 5 https://nmc.ttrackonline.comltmtrack/tmtrack.dl] ?IssuePage&Tab]eId= 000&RecordId=26:... 9/20/2002
Page 2 of 2 Nuclear Management Company 0 State: Quality Check 0 Active/Inactive: Active Parent 0 Owner: PBNP CAP Admin AR Type:
913/2002 "0Submitter: RICHARD FLESSNER Assigned Date:
"0Last Modified Date: 9/12/2002 10:10:07 AM 0 Last Modifier: RICHARD FLESSNER RICHARD FLESSNER 0 Last State Change Date: 9/3/2002 7:13:41 PM 0 Last State Changer:
" Close Date:
"0One Line
Description:
Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW NUTRK ID: CR 01-3595 Child Number: 0
References:
CR 01-2278 RCE 01-069 GOOD CATCH LER 266/2001-005-00 Update:
Import Memo Field: LER 266/2001-005-00 made the commitment that 'Plant modifications to enhance system reliability, including providing a backup air or nitrogen supply to the minimum recirculation valves, are being evaluated.'
CAP Admin: PBNP CAP Admin Site: Point Beach OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD UNKS E 9-Subtask from CAP001415: Probabilistic Risk Assessment PRA For Auxihary Feedwater System AFD 2 Z2 Linked to ACE000314: Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW https://nmc.ttrackonline.comltmtrackltmtrack.dli?IssuePage&TableId= 1000&Recordld=26.... 9/20/2002
Point Beach Nuclear Plant MODCATI ONM'ORPL.kANT CHANGE NO.: 01-144 PLN wOD PLANT CHANGE INITIATION INITIATION VALVE MODIFICATION
Title:
AFXV MOTOR DRIVEN PUMP MINI RECIRC CONTROL El AQ [I Non-QA 0 SR nl Non-SR Unit 1 El Unit 2 El Common ED Z QA AF EWR: CR:: C___1-Z_7__C,_/_3_'
CHAMPS System Code:
Project Objectives: PROVIDE A F;k11"EzY BACKUP N2 SYSTEM TO THE AFW MlTR DRIVEN PMP ON LOSS OF INST AIR.
MIN'I-FLOW CONTROL VLS , AF-4007 AND AF-401-1, SO THE VALVES FULNCTION MI-INI-FLOW R-ECIRC Proposed Scope: INSTALL JUMPERS FROM THlE AFW MOTOR-DRIVEN PUMP DISCH CONT VLVS, AF-4012 AND AF-4019, WHICH CONTROL VALVES TO THE AFW MTR DRfVEN PMP CURRENTLY HAVE SEPARATE BACKUP NITROGEN SUPPLIES.
T.;;.,. *,.Stewart A. Witole Date. 1.4z/20VI
_u CHANGE CHANGE DETERMINATION YES x
If YES go to NP 7.3.1 Temp Mod Is the change Temporary? x If YES go to NP 7.3.8 Setpoints.
Is this a Setpoint Only change? x If YES go to NP 9.3.3 SPEED Is this an Equivalent change?
X If YES determine if previously Document change only? evaluated X If YES proceed with document Does previous evaluation encompass change? changes X If YES, determine if document Commercial Facihty Change? updates are required.
X If YES contact design supervisor. If For Commercial Facility Change Only: NO proceed outside of Engineering Document Updates? process controls Document below.
If YES perform Minor Plant Change Is this small scope? x If NO, it is a Plant Modification. Go to EAC for review and approval (N'P 7.2.1) document and/or attach justification Also. attach If it is determined that this is not a Plant Change or Modification, document update checklist if necessary.
ENGINEERING CHANGE PROCESS TO USE:
Minor Plant Change Date Enineering Group Leadat Prepared By:
PBF-1605i Page I ot 2
Point Beach Nuclear Plant PLANT MODIFICATION/MINOR PLANT CHANGE NO.:
01-144 PLANT DESIGN CHANGE CHECKLIST VALVE MODIFICATION
Title:
AFW MOTOR DRIVEN PUMP MINI RECIRC CONTROL DESIGN SUPERVISOR for completion of this section.)
Design Controls and Project Controls (Ref. NP 7.2.1, Commentary, Check Applicable Design Controls: Clarifications/Basis.
J* Design Input Checklist (PBF-1584)
[] DUC (PBF-1606)
N Design Verification Notice (PBF-15S3)
J* Calculations
[] Design Documentation (PBF-1585), or equivalent Design Change In Progress DCN's 0 Engineering Change Requests 1.o Specifications ClanficationslBasis:
Check Applicable Project Controls:
[] Modification Team Required (indicate minimum groups to request)
[] Conceptual Design Package Required o/1- *2-oo071 0.
Budget Design Project (Impact) Number
[] Detailed Project Schedule
[] BV? Required
. Stewart Wtetholter Assigned Modification Engineer:
Date: , *I,*:"
e/
Design Supervisor.
PBF-1605 Referencets) NP 7 2 I. PBF-19S3. PBF-15Sa NP , .1 PBF-1585 PSBF-1606 Page I ot 4 Revision 6 10/0:101
Point Beach Nuclear Plant PLANT MODIFICATION/MLINOR PLANT CHANGE NO.: 01-144 PLANT DESIGN CH-ANGE CHECKLIST CONCEPTUAL DESCRIPTION/REFERENCE INFORMATION (IF APPLICABLE)
GROUP HEAD CONCEPTUAL DESIGN REVIEW AND ACCEPTANCE [Check here if not required:
Review conceptual design. Attach comments on NPBU Document Review Comment Sheet (PBF-1622 or equivalent)
GrouD Accentance Signature Date Comments
["_] None 1"' Attached Radiation Protection None " Attached Fire Protection ___]
0'-_ None D Attached Installine Orzanization
[_] None "- Attached
__ None M Attached
___] None M Attached
["] None ["' Attached Design Supervisor tJ I A E" None 1 Attached PBF-1605 Page 2 of 4 Refercnces) NP7 2.1. PBF-1583. PBF-1583 Revision 6 10/02/01 NP7'2.2. PBF-1535. POF-1606
Point Beach Nuclear Plant PLANT MODEFICATION/M-NOR PLANT CHANGE NO.:
01-144 PLANT DESIGN CHANGE CHECKLIST FINAL DESIGN REVIEWS (PBF-1622 or equivalent)
Review final design. Attach comments on Document Review Comment Sheet Date Comments Grouo Acceptance Siýgnature one El Attached Ipp/cbLe DN Radiation Protection h oli- one Attached Fire Protection Engineer .
.~-Z2 one E Attached
/ / 1. 7 Installing Organization ([ aC)
/ ne [-Attached
- Chni, ,a;IAttAA' 1 4one [ Attached
[one D Attached
,one E Attached Oo~oS-. /a. L%4-11Jone -" Attached I.so ~~
Tech. Review all IWP's and WO's "NDEPENDENT REVIEW OF INSTALLATION DOCUMENTS (IWP or Work Order Plan) List used for installation anOl?95qe9&8ents IWPAsIWO#(s) -v JO n7!a nt gt All design and licensing requirements have been incorpo d in the installation and testing, document(s).
RELEASE FOR INSTALLATION has been appropriately reviewed. All necessary documents All design controls have been properly implemented and the project regarding release of this design are noted below:
are approved. This design is released for installation. Comments Date: . /
Design Supervisor:
esig -- -T .
COMMENTS PBF-1605 Reference(s) NP7 Re'vision 6 10/'02101 Page 3 of 4 NP7 22.2.
- 1. PBF-15S3. PBF-1384 PBF-1585. PBF-1606
Point Beach Nuclear Plant PLA*'T MODIFICATIONR'MEiOR PLA*Nr MAN.GE NO.: 01-144 PLANT DESIGN CHANGE CHECKLIST ACCEPTANCE are complete.
Plant modification is installed, tested, and all documents required for acceptance L1. Date: /
Modification Engineer:
CLOSEOUT the Document Update Checklist (PBF-1606).
Plant modification is complete, including submittal of all document updates in FCR numbers, etc.)
Reference change tracking numbers on PBF-1606 where appropriate (DCN numbers, Date:
Modification Engineer:
Date:
Design Supervisor:
NUCLEAR INFORIVlATION MANAGEMENT Microfilm the entire modification package.
L-PBF-1605 PBF-15S3 PBF-1584 2 1. PBF-15S5.
paoe -, Of-4 Reterenceis) NP7 NP, 22 PBF-1606 Revwson 6 10/02101