ML20137M532
| ML20137M532 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/19/1996 |
| From: | Norris C, Voorhees J FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137M095 | List: |
| References | |
| FOIA-96-485 QSL-SEC-96-03, QSL-SEC-96-3, NUDOCS 9704080116 | |
| Download: ML20137M532 (42) | |
Text
.- _. _-
- e FPL. FPL Nuclear Division Quality Assurance Audit Report Functional Area Audit of Security And Safeguards Information Control QSL-SEC-96-03 l
Audit Team:
C. E. Norris 1 B. J. Lowery J. H. Ardizzoni L. W. Bladow QA PSL 9704080116 970401 PDR FOIA BINDER 96-485 PDR
Of:PL AUDIT REPORT QSL-SEC-96-03 Page 1 of17 1
Executive Summary This audit examined activities associated with the implementation of Security and Safeguards Information l Control Programs at the St. Lucie Plant (PSL). The objective of this audit was to evaluate the effectiveness j of PSL's implementation of the Security Contingency Plan, Physical Security Plan, Security Training and !
Qualification Plan, and Safeguards Information Control. The audit included implementing procedure and I record reviews, interviews with appropriate site and contractor personnel, and observation of field activities 1 associated with the audit scope. The Corrective Action Program and controls in place for identifying, !
resolving and preventing issues that could degrade the physical security of plant were also reviewed. l 1
l The addition of a security specialist from Texas Utilities (TU) participating through the technical exchange program enhanced the audit team's knowledge of security programs in the nuclear industry. The ,
specialist's primary focus was in the security operations area as requested by security managemem. I The audit verified that site procedures are developed and maintained that delineate and implement the programmatic requirements for control of safeguards information and security plans at PSL. Corrective action processes provide adequate documentation, control, evaluation, and disposition of nonconforming conditions. l Strengths .
l
. Unannounced crucial tasking activities are used by Security Training to verify that Security Officers remain knowledgeable of and qualified to perform crucial tasks applicable to the security post the officer is assigned. (Page 8)
. Security has implemented an aggressive self assessment program. (Page 10)
. Security supervision makes extensive use ofindicators to trend personnel and hardware performance. (Page 10)
Recommendations
- Revise Land Utilization procedure LU-Ol-11.0-20 to include dimensional verification of selected panels during safe-net inspections. (Page 13)
. Document an equivalency evaluation of the probability testing being performed by I & C procedure 3400027 to that described in Regulatory Guide 5.44. Submit the equivalency evaluation and 3400027 for NRC approval as required by Regulatory Guide 5.44. (Page 14)
I i .
l
. AUDIT REPORT QSL-SEC-96-43 FPL Page 2 of17 Weaknesses _
a In general personnel using and purchasing computer software did not fully understand the scoping sections of software control procedures SQM 2.9 and QI 2-PR/PSL-3. Interviews with security personnel on site and Juno based engineering personnel revealed that software control requirements were considered to be applicable only to safety related or quality related process classifications. Other applications of software used to meet license commitments were seldom considered. This miss-understanding resulted in an incomplete software documentation package for the video capture system.
Corrective action for the software contol finding of PMON 95-039 finding is still in progress. The weakness identified in thisaudit is an aJ litional example of software control issues previously identified in that PMON which further demonstrates this generic weakness in implementation of software procurement requirements. Corrective action for the finding of PMON 95-039 is still in progress, therefore an additional finding is not being issuedfor the Security software controls area. (Page 8)
Based on the activities and objective evidence audited, it was determined that the requirements for safeguards information control and the Security Plan requirements for Site Physical Protection, Security Personnel Training and Qualification, and Security Contingencies are adequately addressed by procedures ;
and the implementation of those procedures is effective. The technical recommendations in this report l l
identify areas where improvements in intrusion detection system testing and inspection may be realized.
l i
l l
l
O
- O. AUDIT REPORT QSL-SEC-96-03 Page 3 of17 Location of Audit St. Lucie Plant Date of Audit February 6 -April 19,1996 Audit Scone This Functional Area Audit evaluated the program adequacy and assessed the implementation of activities at the St. Lucie Plant associated with:
e The PSL Site Physical Security Plan,
- The PSL Site Contingency Plan, e The Security Training and Qualification Plan, and
- Nuclear Safeguards Information Controls This audit is intended to satisfy the requirements of PSL Technical Specification 6.2.5.8.a and 10 CFR 73.55 (g) (4). This audit included the expanded scope of Functional Area Audits to provide an evaluation of program adequacy and access the implementation of activities associated with physical security at the St.
Lucie Plant (PSL). The audit reviewed and evaluated the following:
- The necessary ; rograms and procedures exist and comply with applicable baseline standards, TQAR requirements, Technical Specifications. FSAR, and 10 CFR 50, Appendix B.
e LER's, Problem Reports, in-House Events, and NRC Inspection Reports associated with Security's activities to identify trends or performance issues within the Security organization.
- The effectiveness of self assessment activities.
- The effectiveness of Security's corrective action activities and programs as applied to corrective action taken in response to intemally and extemally identified issues as well as known performance issues or management challenges.
e Utilization of industry information for lessons leamed to preclude the occurrence of in-house problems.
Audit Details The audit team's knowledge of security programs in the nuclear industry was enhanced by the utilization of a security specialist from Texas Utilities (TU). The specialist's primary focus was in the security operations area as requested by security management. This addition to the team improved the evaluation capability in the area of security operations by providing a comparison of
, AUDIT REPORT QSL-SEC-96-03 Page 4 of 17 St. Lucie Plant's security methods and practices with those utilized by other utilities in the industry.
Several opportunities for improvement in the areas of administrative controls and Intrusion Detection System surveillance testing were identified that could enhance the records of security activities to p; ovide more convenient record traceability to verify that the Physical Protection Program at St. Lucie is in compliance with regulatory requirements.
This audit examined the areas detailed in the Audit Scope to provide an overall evaluation of the effectiveness of Security and Safeguards Information Control Programs at St. Lucie Plant (PSL). The audit included reviews of Security program implementing procedures and records documenting security activities.
These activities included Intrusion Detection System (IDS) testing, response to equipment failure, lighting checks, vehicle access control, shift operations, personnel access control, security training, and operations at the Central and Secondary Alarm Stations (CAS and SAS). Field observations of activities associated with the above scope were conducted to evaluate personnel and equipment performance. Activities observed included: firearms training, contingency response training, personnel and vehicle access controls, CAS and SAS ope' rations. IDS testing, installation of the vehicle barrier system, and a coordination exercise of the site security force with the primary Local Law Enforcement Agency (LLEA.). Field walkdowns of Protected Area (PA), Vital Area (VA), and vehicle barriers were conducted to determine their physical condition and to verify compliance with the configuration requirements contained in the Physical Security Plan (PSP). A summary of each area examined is included in the report.
Security Department personnel were supportive and responsive during the audit. Personnel demonstrated excellent knowledge of security processes and procedural requirements during the interviews performed and field activities observed as part of this audit process.
Physical Security Plan and Contineenev Plan This area of the audit evaluated Physical Security Plan implementation at the PSL site. Implementation of ;
the PSL Site Contingency Plan including implementing procedures: response team manning, training, and '
equipment; and Local Law Enforcement Agency response coordination were also reviewed / observed and evaluated.
This evaluation was facilitated by review of security records and logs documenting access authorization and the completion of surveillance and inspections required by the Physical Security Plan (PSP) to periodically 1 verify the integrity and operability of physical protection equipment and systems. Security Force Instructions (SFI's), Security Procedures (SP's). Site Administrative, Security Instrument & Control (1 &
C) and Land Utilization procedures were reviewed and evaluated against the requirements of the PSP,10 CFR 73 and Regulatory Guide 5.44. Observation of security activities in the fie'd included: Personnel and vehicle access control: Entiance building operations including hands on searches, key controls visitor control. vehicle gate operation, key card control, search equipment operation and communications:
Secondaiy and Central Alarm Station Operations including CCTV coverage, alarm response, communications and computer operation; and Badging computer terminal operation. Walk downs of the Protected Area. Vital Area, and Vehicle barriers verified compliance to the PSP required configuration for the barriers. Interviews conducted with appropriate site security personnel verified that personnel were 1
OE AUDIT REPORT QSL-SEC-96-03 Page 5 of 17 knowledgeable of procedural requirements and responsibilities. The comprehension of procedural requirements demonstrated during inter iews indicates that Security personnel have adequate understanding of the requirements and responsibilities delineated in the procedures to effectively implement them during contingency and response situations.
Review of badging office activities. records. and badging terminal capabilities verified that safeguards system and software capabilities required by the PSP for the badging office and terminal have been effectively implemented. During a previous audit of badging activities (QSL-OPS-95-05) it was found that in some instances badging records for personnel granted unescorted access did not contain current addresses as required by the PSP and the Code of Federal Regulations.10 CFR 73.70. A review of randomly selected badging records for personnel granted unesconed access to the protected area verified that corrective action taken in response to this item has been completed and is effective. This review revealed that adequate records including addresses are being maintained to verify that PSP requirements are being met for badged personnel including annual reorientation of plant personnel in security matters.
The security events log, trending log, surveillance procedures and records documenting the performance of routine surveillance activities delineated in the PSL Physical Security Plan were reviewed. This review verified compliance to the Plan's requirements for periodicity, acceptance levels, applicable equipment, and testing methodology with the following exceptions:
The Physical Security Plan. section 3.2 limits individual " safe net" sections to less than a 96 square inch aperture. Land Utilization procedure LU-Ql-11.0-20," Inspection of Ocean Intake Underwater Intrusion Detection System (UIDS)" provides the specifics of a periodic inspection including the acceptance criteria. This procedure does not provide for a check ofindividual section dimensions to verify that the PSP limit is met. It is possible that stretching may occur that could cause an individual section to exceed the PSP area limit. This concern is the subject of Technical Recommendation No. I of this report.
The Physical Security Plan (PSP). section 3.14 states that the intrusion Detection System (IDS) is designed to meet the detection probability and confidence level requirements of Regulatory Guide 5.44. Section 12.2 of the PSP requires the system to be tested against design specification for proper detection probability each quarter. after each repair or after each inoperative state. Microwave testing is conducted in accordance with the ins: ructions contained in Instrument and Control procedure IC 3400027. " Security System Microwave Quarterly Calibration" which does not appear to meet the intent of Reg. Guide 5.44. This concern is the subject of Technical Recommendation No. 2 of this report.
The Intrusion Detection System hardware failure rate increased during 1995 to rate that could be considered to be mi unacceptable level. This increase in the failure rate has also resulted in additional compensatory posting hours. The number of hardware failures increased significantly in 1995 as compared to those in 1994 (e.g.,234 hardware failures in 1995 and 154 in 1994). The majority of the hardware failure increasa has been
AUDIT REPORT QSL-SEC-96 43 Page 6 of17 I 1
due to repetitive failures of two microwave zones. This item was discussed with Security Operations supervision who indicated that Security is aware that the failure rate has been increasing and is subsequently I keeping plant management informed of the trend. The auditor was also able to verify through interviews !
with Security and I & C supervisory personnel that, due to the urging by Security, work is in progress to determine the corrective action necessary to retum the failure rate to an acceptable level. This item indicates that security supervision is actively monitoring the performance and effectiveness of their security systems and acting in a responsible pro-active manner to effect improvements when indicators deem it necessary.
A review of the Site Contingency Plan and its implementing procedures verified that the contingency plan contains the requirements of the Code of Federal Regu'ations and the implementing procedures effectively l execute the plan contents. Reviews of shift manning records, training documentation, and equipment !
availability verified that the security force teams for each shift contain at least the number of trained response officers required by the PSP and that the security equipment required to support response activities is available.
l Observation of Central and Secondary Alarm Station activities during normal operation and observation of )
the Seemity Force during daily activities and a Local Law Enforcement Agency response drill verified that actions taken to respond to intrusion detection alarms or intrusion detection equipment failures are in
)
accordance with approved procedures and that routine security fptce operations are conducted in accordance with approved instructions.
The following security program elements specified in the PSL Physical Security and Contingency Plans were physically verified by walk-downs or observation of security activities during the audit to be in compliance with the requirements of these plans.
Response team manning, ;
Weapons and ammunition requirements and controls. l Badging operations.
Security Post manning, Vehicle control and access requirements.
Vital and Protected area barrier configuration.
l Compliance to 20 ft isolation zone and PSP exceptions, Intrusion Detection System (IDS) Alarm indications.
Communications equipment availability and testing, Closed circuit television camera coverage, Lock and key controls.
Visitor control.
IDS testing,
- Lighting surs eys.
Central and Secondary Alarm Station Operation.
Security Equipment inspections.
Response team perfonnance, and Contingency procedure implementation for failed security equipment.
AUDIT REPORT ,
QSL-SEC-96-03 Page 7 of17 The results of the above reviews, observations, and walk downs indicate that the Physical Security and '
Contingency Plans are effectively implemented at PSL.
Security Traintne and Oualifications This section of the audit was conducted to examine and evaluate the adequacy of activities associated with Security Training and Qualification at the St. Lucie Plant. This section of the audit satisfied the requirements of 10 CFR 73.55(b)(4),10 CFR 75.55(g)(4) and PSL Technical Specification 6.8.4.9.
The Security Training and Qualification Plan (TQP) for the St. Lucie Plant delineates specific qualification and requalification processes by which Security Force personnel assigned to PSL are selected and qualified.
This section of the audit examined the adequacy and implementation of the Security Training Program in assuring that Security Force personnel are capable of performing crucial tasks required to implement the l physical security and contingency plans for their post assignments. The audit included a review of the following TQP areas: Employment suitability and qualification, Training and Qualification, Crucial Tasking, :
Firearms Training, and training documentation, records and reports. I l
I Security personnel records and training records of all security personnel hired during 1995 were reviewed to verify documentation of personnel training and qualification. Personnel file reviews verified that the documents maintained by Security provide sufficient evidence to verify compliance with the suitability and qualification requirements of the Training and Qualification Plan. This section of the audit also verified that record retention and training reports were being maintained and submitted as specified in the TQP.
The auditor verified that unannounced crucial tasking activities are still conducted as part of the Security Department's Performance Audit Program. This tasking method was demonstrated to be an effective tool to ensure that Security Officers remain knowledgeable of and qualified to perform crucial tasks applicable to the security post the officer is assigned to. Critical tasking is considered a strength in the continuing training program of the security force.
A sample of training records for current security force members was chosen randomly for each duty position. The assignmern of security personnel only to the positions they are qualified to perform was i verified by comparing their assigned post on the effective shift assignment roaster to their requalification training status as verified by review of their training records. No deviations to qualification requirements were identified during this review.
The auditor also observed an LLEA response coordination drill during a performance monitoring activity earlier this year. The drill demonstrated effective planning, coordination. initiation. control and evaluation by the Training staff, Security supervision was present throughout the drill and its follow up critique.
Supervision actively participated in observation of the activity to determine where changes to response tactics. plant physical characteristics and drill control could be made to improve physical security at the St.
Lucie Plant. This involvement by Security supervision is an excellent example of the effective use of a self assessment opportunity and is considered a strength.
I o AUDIT REPORT QSL-SEC-96-03 f:PL Page 8 of 17 The reviews, observations and interviews conducted during tnis part of the audit verified that the Security Training and Qualification Plan is being effectively implemented by the Security Training Staff and that sufficient documentation is maintained ta verify qualification and requalification status of assigned personnel.
Nuclear Safecuards Information (SGI) Prrcram This section of the audit reviewed practices associated with safeguards information (SGI) identification, transmittal, destruction, declassification, storage, and receipt by various organizations at St. Lucie Plant.
The review verified these practices to be in compliance with 10 CFR 73.21 and AP 0006127 requirements for the control of safeguards information.
The Security Department conducts an annual review of the safeguards information program as required by Administrative Procedure 0006127. The audit verified completion of this activity by review of the annual Safeguards Assessment Report completed by the Site Safeguards Information Coordinator in March 1996.
The overall results of the Assessment were satisfactory. One area for improvement was identified concerning SGI marking of Plant Change / Modification cover letters. This result reflects an improvement in SGI awareness since the 1995 review during which four (4) areas for improvement were identified. This result also indicates that the Site Safeguards Information Coordinator's proactive approach is influencing general improvements in the safeguards information control program at PSL.
An independent review of safeguards information repositories, contents, and inventories during this audit verified compliance with SGI requirements for: marking, repository identification, control of open repositories, repository locking mechanisms. SGI inventories, control of SGI during usage, SGI transmittals, and determination of"Need to Know."
10 CFR 50. Apnendix it Criterion Ill Desien Control (comnuter software control)
The computer software control program implemented by the security department, was evaluated for compliance with the Site Quality Manual. SQM2.9 and Quality Instruction. QI 2-PR/PSL-3. This evaluation included interviews with cognizant personnel, and a review of program documentation and procurement practices.
The I&C department maintains all hardware and software associated with the security system. Personnel in I&C are familiar with the software control process and have a well developed program. Documentation including verification' validation packages. change requests, and approvals for use are on file. A review of the purechase documentation for the new Video Capture System was conducted. This review determined that 1&C had not been involved in the purchase process. This resu:ted in a bypass of the expertise necessary to properly indentify QA software control requirements relating to initial procurement.
SQM 2.9 and QI 2-PR/PSL-3 state:" Computer software supporting plant license reqmrements, when used to document compliance with regulatory requirements will be controlled." Specific items of this control process are required to be accomplished during the initial acquisition. QI 2-PR/PSL-3, section5.11 requires
l O AUDIT REPORT QSL-SEC-96-03 Page 9 of17 saimmmmmmmmmmmmmmmme a purchase order for acquisition of computer software to contain the following:
- 1. Software Requirements Specification (SRS)
- 2. Software Design Description (SDD) 3, Licence Requirements
- 4. Software Verification and Validation Plan (SVVP)
- 5. Software verification and Validation Report (SVVR)
- 6. Programmer or User Manuals
- 7. Independent Verification and Validation ,
Q12-3 also states that based on vendor license agreements, code application, and historical use of the code in industry, the above requirements may be deleted if an alternative method is established and approved by management to accept the computer software. The method of acceptance will be documented in the documentation package. With the exception ofitem six above, the purchase order for the site security video ,
capture system did not address software control. This generic problem was previously identified in audit report QSL-OPS-95-18. PMON 95-039, in which software programs associated with the Votes 100 System and the Leading Edge Flow Meter were purchased via PC4 purchase orders with no identifiable software control measures. Corrective action for software contol finding of PMON 95-039 finding is still in progress.
This audit of Security provided an additional example which indicated a generic weakness in the implementation of software procurement requirements. It is noted that the video capture system was purchased before the program and procedure updates were initiated in response to PMON 95-039.
Interviews with security personnel on site and Juno based engineering personnel revealed that software control requirements were considered to be applicable only to safety related or quality related process classifications. Other applications of software used to meet license commitments are seldom considered.
Security Department personnel involved in the use, purchase or development of computer software have recently attended a Software Control Awareness program developed by 1&C for the maintenance dept. This training should irnprove compliance to software control requirements if additional software purchases or changes to existing systems are made.
Licensee Event Renorts and NRC Insnection Renorts l l
Two 1.icensee Event Reports were submitted in the security area in 1995 and one NRC inspection was conducted. These reports were reviewed as part of this audit to idemify trends or perfomiance issues within the Security organization. No adverse performance trends or issues were identified during this review.
Corrective Action Effectiveness Corrective actions developed and taken in response to QSL-OPS-95-05, finding 1 and Licensee Event l Reports 95 S01 and 95 502 were reviewed to determine the effectiveness of the actions taken in correcting i
the identified problem and preventing recurrence of the same or similar events. Corrective actions developed in response to the items was found to be acceptable to the auditor based on the immediate corrective actions taken and the generic implications considered.
. AUDIT REPORT .
QSL-SEC-96-03 FPL Page 10 of 17 The security Department uses an extensive array ofindicators to determine trends in performance of both hardware and personnel. Security supervision made effective use of this data in identifying a decreasing trend in hardware failure rates caused by newly installed microwave equipment that had been installed to effect correction to two deficient detection zones. Based on the results derived from review of these indicators Security supervision urged corrective action to be developed and implemented to reduce failure rates. Due to Security's urging, I&C is currently evaluating the equipment effected to determine the corrective actions necessary to reduce the failure rate to an acceptable level. These actions indicate that security supervision takes a proactive role in identification and correction ofinternal problems and follows up the action taken to verify that the corrective action is effective. This persistent questioning attitude is considered a strength of the corrective action program.
Self Assessment and Use ofIndustry Information Security uses a multi-facetted approach to self assessment. A self assessment specialist has been appointed and has formalized Security's self assessment program through an approved administrative procedure to govern the process and scope of routine self assessments. Security presently performs the following types of assessments and reviews:
- Training record reviews to ensure that the records accurately reflect security officer qualification and i requalification status.
- Unannounced tasking to ensure that Security Officers remain knowledgeable of and qualified to perform crucial tasks applicable to the security post they are assigned.
- Review of the " Safeguards Information Program" to evaluate PSL's control of safeguards information.
- Tracking and trending ofloggable security events. equipment failures. compesatory post hours. NRC 1 violations. and NRC reportable events.
- Security event reviews. and
- Self assessments of those areas scoped in Security Administrative Procedure 3. "Self Assessment Program" This program requires a self assessment of all major functional areas for which the Security Department has implementation responsibility.
Review of Security self assessment activities revealed that routine self assessment activities are conducted in addition to those driven oy security events (i.e.. as the result ofidentified problems. increases in loggable events, etc.L Examination of everal Security self assessments: including: the Safeguards infuiTnation Program annual review, evaluation ofloggable events, trending logs. LLEA coordination response drill, and reviews of reportable security events indicated that the results of these activities have been efTective in identifying areas for improvement and that Security has acted on these self assessment observations to strengthen its processes at PSL. Examples of changes made to improve performance in identified areas include: equipment addition and modification to the firing range to enhance weapons qualification;
O f:PL AUDIT REPORT QSL-SEC-96-03 Page 11 of 17
. . j stencilling of water box manways to reduce the posibility of creating an unprotected pathway to the protected area during maintenance; and establishing a security shift operations center in the secondary alarm station to improve shift communications, response and coordination of shift activities. These self assessment activities are viewed as a strength that provides an effective tool to further improve security program implementation.
An examination of Security's Staff Routing log book by the auditor indicates that the Security Department places importance on keeping their personnel informed of events occurring in the industry. The Security Supervisor past.es information obtained from the industry to the security staff. When this information may have an effect PSL's security program, review and evaluation by the security staffis requested. This type ofinformation transfer enhances the level of knowledge of security force personnel and provides for direct input ofindustry related information.
I Conclusintt Based on the activities and objective evidence audited, it was determined that the requirements for safeguards information control and the requirements of Security Plans for Site Physical Protection, Security ,
Personnel Training and Qualification, and Security Contingencies are adequately addressed by procedures !
and the implementation of those procedures is effective. The technical recommendations in this report identify areas where improvements in intrusion detection system testing and inspection may be realized.
Satisfactory Areas l
j
- Security Force Composition Perimeter barrier isolation zone j
- Protected area and vital area barriers and inspections Key and lock control
- Personnel badge and key card control Penmeter and vital area illumination Vehicle control
- Persomiel and material access controls
- Post manning Operational readiness of the security force
- Compensatory measures established for reduction in the effectiveness of physical barriers and intrusion detection alarm failures
- Camera coverage
- Weapons requirements 4 l
- Alarm station activities and responses
- Contincency plan implementing procedures l
- Records and documentation of security activities including training
- MontUy training reports
- Firearms training and qualification
- Secunty training scheduling and proficiency testing of crucial tasks
i AUDIT REPORT QSL-SEC-96-03 l 1: P L Page 12 of17 j Satisfactory Areas (cont'd) l
- Safeguards inforrnation controls including ;
Identification of repositories !
Document identification Document storage j Identification of safeguards information controllers i Repository Logs and inventories l Findings None i i 1
a l
l t
f I
l
4 G.
AUDIT REPORT QSL-SEC-96-03 FPL Page 13 of17 Technical Recommendation No.1 Inspection of Ge Ocean Intake Underwater intrusion Detection System (UIDS)
Observation The St. Lucie Plant Physical Security Plan (F:P) describes the Safe-net as an interconnected net of panels with apertures not exceeding 96 square inches and also requires a physical inspection of the safe-net underwater bmier every twelve months. The physical attributes and acceptance criteria of this inspection are not included in the PSP. Land Utilization procedure LU-QI-l1.0-20. " Inspection of Ocean Intake Underwater intrusion Detection System (UIDS)." provides the specifics for this periodic inspection including the acceptance criteria. However, this procedure does not provide for a check ofindividual section dimensions to verify that the PSP 96 square inch limit is met.
Discussion The auditor discussed this observation with Security stgervision. During the discussion Security personnel indicated that stretching would cause a break in the wire coating causing an alarm indicating a breach of the net. Calculations demonstrated that a relatively small dimensional change (i.e., a strain of 0.085 in/in ) in individual panel wires could cause the PSP aperture limit to be exceeded. It is possible that dimensional creep over years of use may cause an individual section to exceed the aperture limit without causing a break in the outer coating.
Recommendation Revise Land Utilization procedure LU-Ql-11.0 20 o include dimensional verification of selected panels during safe-net inspections.
- AUDIT REPORT QSL-SEC-96-03 Page 14 of17 Technical Recommendation No. 2 Intrusion Detection System Detection Probability Testing Obsenation The St. Lucie Phnt Physical Security Plan (PSP) requires the perimeter intrusion detection system to provide a high assurar.ce of detecting all attempted penetrations of the Protected Area perimeter. The system is designed te raeet the requirements of Regulatory Guide 5.44 to include 90% probability of detection with 95% confidence level. The PSP requires testing Perimeter and Vital Intrusion alarms against design specifications for proper detection probability each quarter. I & C Maintenance Procedure No. 3400027.
" Security System Microwave Quarterly Calibration." tests the system using nmning, walking, and simulated crawl approach methods. The simulated crawl approached is facilitated with a . sled and aluminum ball.
Discussion Regulatory Guide 5.44,1980. Appendix "A," provides an example of a method to determine the detection capability of perimeter alarm systems that does not use simulated crawl techniquesand also says that other testing methods for determining compliance with detection probability levels may be used if fully ;
documented and approved by the NRC. Section "D" of the regulatory guide provides information to j applicants and licensees regarding the NRC staff s plans for using the regulatory guide. The guide indicates that the staff will use the methods described in the the guide to evaluate capability for performance in complying with specified portions of the Commissions regulations after April 1.1980. This wording i indicates that the staff will also evaluate PSL testing against the methods described in the 1980 version of the guide, in the regulatory guide testing method the most vulnerable area of each segment is tested using the method of approach most likely to penetrate the segment. i.e.. walking. running. jumping, crawling.
rolling. or climbing. All segments are tested using the applicable penetration approaches at the most s ulnerable area. The guide does not discuss the use of approach simulation techniques.
1 & C Maintenance Procedure No. 3400027.
- Security System Microwave Quarterly Calibration," tests the system using running. walking. and simulated crawl approach methods. Based on a determination that a crawling approach is the most likely method of approach to penetrate the zone, the I & C procedure does not include jumping. rolling. or climbing. The simulated crawl. conducted by pulling a sled with an aluminum ball through the area is described in Intrusion Detection Systems Handbook. SAND 76-0554 not Regulatory Guide 5.44. The technique may require further evaluation. documentation and NRC staff approval to meet the intent of the regulatory guide. Documentation that the NRC has approved the current crawl approach testing technique could not be located.
Recommendation Perfomi and document an equivalency evaluation of the simulated crawl testing being performed by I & C procedure 3400027 to the testing described in Regulatory Guide 5.44. Submit the equivalency evaluation for NRC approval.
O b AUDIT REPORT QSL-SEC-96-03 Page 15 of 17 Audit
Participants:
Name Department /Groun A B C C. Burton Services Manager x x L. Bladow Quality Manager x J. Ardizzoni Texas Utilities Security x N. Miller SBI-Project Manager x x G. Varnes SBI x x x J. Allen Security Access Specialist x x x M. Bryan Catalytic x x W. White Security-Supervisor x x R. Czarnecki Security-Operations Supervisor x x x K. Basquez Security x N. Radak I&C x G. Patel I&C x J.Toebe Land Utilization x B. Lowery Nuclear Assurance x C. Norris Nuclear Assurance x x x J. Voorhees Nuclear Assurance x x P. Plantz Security x x J. Talley Security x N. Miller SBl x x
- 11. Fagley Construction Services x K. Sanderlin SBl x L. Grif6n SBl x J. Becker SBI x K. Czarnecki SBl x N. Newman-Yates Nuclear Records x J. Bailey Nuclear Records x l
K. Butler Maintenance QC x L. Petrie Technical Staff x j J. Potter Technical Staff x j G. Madden Licensing x Kgn A - Pre-Audit Conference / Noti 6 cation i B - Contacted during the Audit !
C - Attended Post-Audit Conference l 1
References:
- 10 CFR 73.55(g)(4) 10 CFR 73.21
- 10 CFR 73 appendix B & C I
AUDIT REPORT QSL-SEC-96-03 pp Page 16 of 17 4
References:
- 10 CFR 50.54 Continued - 10 CFR 50.7
- Safeguard Contingency Plan Physical Security Plan
- Training and Qualification plan SP 0006022 SP 0006027
- SP 0006024
- SP 0006025 SP 0006028 SP 0006123
- AP 0010721
!&C 3400020 I&C 3400021
- I&C 3400026 1&C 3400027 I&C 3400029 IP-402 SF1-1
- SF1-3 SF1-4 SF1-5 SF1-6
- SF1-7 SF1-8
- SF1-9 Pre-Audit Conference:
Location: St. Lucie Plant Date: February 6.1996 Post-Audit Conference:
Location: St. Lucie Plant Date: April 19,1996
O 4
k
@FPL AUDIT REPORT QSL-SEC-96-03 Page 17 of 17 Summary of Post-Audit Conference:
The audit was presented and discussed. Personnel were thanked for their cooperation and support during the audit process.
Location of Audit: St. Lucie Plant Accomnanvine Auditors: B. J. Lowery J.11. Ardizzoni, Technical Specialist from Texas Utilities i
f ILrincipal 6' 9 /P 6 Auditor: C. E. Norris Date Senior QA Engineer - PSL n
iteviewed by: "
t% IIS [o
). T. corhees Date Supenisor - PSL
QUALITY ASSURANCE DEPARTMENT AUDIT REPORT DISTRIBUTION AUDIT REPORT: OSL-SEC-96-03 PLANT / DEPARTMENT: St. Lucie Plant NUMBER OF FINDINGS: None fERB Stsndard Distribution R. J. Acosta - JNA/JB T. F. Plunkett - JEX/JB :
W. H. Bohlke - VP/PSL T. V. Abbatiello - JNA/PTN R. J. Hovey - VP/PTN L. W. Bladow - JNA/PSL G. J. Boissy - JPN/JB R. A. Symes - JNA/JB Dr. K. R. Craig - JPN/JB D. A. Culpepper - JPN/JB I
H. N. Paduano - JPN/JB QAD Files w/ Checklist & Audit Plan Cheryl Robinson - JNA/JB i Dr. W. R. Corcoran (CNRB)
Health Physics & Chemistry Related Audits
]
D. B. Miller (CNRB) Manager Nuclear Health Physics / Chemistry i
- CNRB Files: K. E. Gutowski Emereener Prenaredness Related Audits ;
Manager - Nuclear Emergency Preparedness j Audit Snecific Distribution NuclearTrainine Related Audits M. Miller - NRC/PSL-1 Manager Nuclear Trammg :
E. Weinkam W. White Plant Snecific Security Audits l J. Scarola Security Supervisor- A. Cummings Services Manager l
Fire Protection Audits S. Martin. Risk Management
'Only distribution outside the plant for Security Audits containing Safeguards information.
\%
N M 4/22/96 r\%
, Inter-Office Correspondence W L.
JQQ-96-054 To: C. L. Burton Date: April 22,1996 From: L. W. Bladow Department: JNA/PSL
Subject:
Quality Assurance Audit QSL-SEC-96-03 Security and Safeenards Information - Functional Area Audit j l
Attached is the final repon of an audit conducted to evaluate site compliance with the applicable requirements of the FPL QA Program, The Code of Federal Regulations 10 CFR 73.21,
" Requirements for the protection of safeguards information," St. Lucie Plant Security Plans including: Physical Security, Training and Qualification, and Contingency, and implementation of the security and safeguards information control programs delineated in plant and security procedures.
This audit is conducted to satisfy the requirements of St. Lucie Plant Technical Specification 6.5.2.8.a.10 CFR 73.55 (g)(4) and the Quality Assurance Department Annual Audit Program Plan i to review the security program at least once per 12 months. l The following technical recommendations are documented within this report and were discussed at the Post-Audit Conference. l l
Technical Recommendation No.1 Land Utilization Procedure LU-QI-l1.0-20, " Inspection of Ocean Intake Uri.ierwater Intrusion ,
Detection System (UIDS)," provides the specifics for this periodic inspection including the acceptance criteria. This procedure does not provide for a check ofindividual section dimensions to verify that the PSP 96 square inch limit is met. It is reconunended that Land Utilization procedure LU-Ql-11,0 20 be revised to include dimensional verification of selected panels during safe-net inspections.
Technical Recommendation No. 2 I&C Maintenance Procedure No. 3400027," Security System Microwave Quarterly Calibration,"
tests the system using nmning, walking, and simulated crawl approach methods. The simulated crawl, co::.iucted by pu!!ing a sled with an aluminum ball through the area, is not described in
, Regulatory Guide 5.44. The technique may require funher evaluation, documentation and NRC staff approval to meet the intent of the regulatory guide. It is recommended that an equivalency evaluation of the simulated crawl technique being utilized in I&C Procedure 3400027 to the testing described in Regulatory Guide 5.44 be performed and documented. This equivalenev evaluation should be submitted to the NRC for approval.
Security and Safeguards Information - Functional Area Audit Page two 1
i Condition Reports (CR) have been issued for the departments affected by the Technical Recommendations above. In accordance with the FPL Quality Assurance Pro. gram, please ensure that the CRs which address these Technical Recommendations are responded to within 30 calendar days of origination.
We sincerely appreciate the cooperation we received from your staff during the course of the audit. [
Please contact me at extension 7111 or the lead auditor, C. E. Norris at extension 7591 if you have i t
any questions.
h I L. W. Bladow Quality Manager - PSL LWB/CEN/str Attachment I
Copies to: Distribution Attached t
T --
1, L
i O
i W W !
$ e "a p
oJ q $
,lll m o g i i
4M g o g
- a p C bE N $.
! 4 i k i
C
_ gA
,-~ . .. - -
8' s t
[NAME ' DUTY ASSIONMENT i 2SN, DOE: s a
k
() HTIA6 OUAuPCATION l) REQUAUFCATION REQUALFCATON
.r Y EVALUATORI DATE ' DATE DATE SECURITY EVALUATOR / !
DATE DATE DATE - sECURITV INSTRUCTOR _I t CRUCIAL .. . -.. . .... .. OFFCER INSTRUCTOR .. . .. .- + .. OPPCER I - SECURITT - PAger pass PASSI S10 NATURE Of0 NATURE PAS 8/ PA$$r PASSI StGNATURE SIONATURE i t . Tate ( DEtcruPTION. FAL PAL PAL Fasl FAL FAL r
i s
ELECTROMC SECURITY SECT 80N assemy veur O Gems 4Fhuses Sp4W ,
Cupener Tes> -'
8 pummsaw wil Cemen emne h flade i
MAhU 6 l
e I s esessane s assessse
' ab uses Comummer Commess Pupumps !
g huMW g Gunuss mennesme ,
W OButsEF Cof9ed88 to numan vem CcTv eyess C aamer to CCTV 8,seem ,
i sb .ama, s nm.e.m l ca.e.m
. - anw= nsuswee pism P
.g et BAS Denaf GA4 Ps ,
l
.# Omasp ESA F. Asseurue ,
M YIM @M j e I 1
1 ansva s . mas j
n== e s . == -
e cA7 4 ELECTRONIC SECURITY GECT8DN . 7400 Sorse ammy.o senses tunames cuisse 8tunwt Danemov 0 A esammes comese C uhme an cosas gone 0 homen g Canaan as tumuus tone
, taus asen pusungy -
asemes
--m amenos tho79mes Rusase to A*sesuis amasamme 0 alpaur &Ced Susastsee
} t alsmuysse tessumme of tensgo
' I am spicast ammen edes D'et noe hague M !
1
s e - a a .. - .s ..-.m.u.. - ~.=u- - . ,
t V. t NAME DUTY ASSIGNMENT SEN; ' DOE; '
e
.v. 9 (l 88fflAL QUAJCATEDN () REQUALFCATION REQUAuPCATION
.7 DATE DATE OF.CUfWTY ' EVALUATofW DATE DATE DATE SECURffY EVALUATOR /
DATE 4 CRUCIAL . .... ...n .... 0FFCER DiBTRUCTOR + . + . . . . . . . . . . . OppceR DesTRUcfon .
$ MCUINTY PA44f PASS PAS $f $10 NATURE $108eATURE PA$$f PAGGI PAS $/ SIGNATURE SeQNATUM
-4 TA8K DESCpuPfl08s FAIL FAL FR FAIL Fall FAIL
)
SEARCH OPFCEM SECTION } -
s I
m Test Pam EEIIsme Desseur {
?
WLW .[
ensma Res.
i inmusDuGIAB ARAs - t i
i g
amme e
g WWM . . .
Imme 4 (Demu4 venuse Ones puunsee Usage tan PA 95 M E8888 ,.
'i e es an me===m IM
.* m.a.am.me
. .- vesso no PA - ;
m m. m A ,
O leue AmW hemig meses vases esoneser U tem as weh One ;
i y W UU 1muuan DO l
. m. :
la W TJ 4 esses }
hay: Ops [
es i
un O'e8 N t
a *==' = =
, es, m e.
Gussnesse L
Muse FeuemeMICA?46ti t i
eme=se saa.
k SEARCH OFFICER SECTION .7400 Senes ,
l usuouse Dome g Ausen beaumat CB A }
ta==s o===
e-. .
Cassalas Sseper ese
- emme (New Ann 8essemp 35 asamme amme ames passessy
% asemme e use Neues ainaus to A poemeus 5
le edueue # 98 Olaudegue ti num6use taumes(W beammgo if M AD0881M t
W~H '
new test . W
I
.e f
- NAME DUTY A5530NMENT 4
.SSN: DOE:
e f
() Def7ML CAJALIFCATION () REQUAUFCATON REQUALFICATION l DAYS DATE DATE SECURrTY EVALUATQ8V DATE DATE DATE SECU8vfY EVALDATORI e CRUCnAL .~ . .. . .... OFFcER INSTRUCTOR .. . . . .. .... OFFCER HSTRUCTOR 6 GBCUfvTY PAGGI PASS PASGI $1GNATURE $1GNATURE PA86/ ' PAS $( PASS / $80 NATURE 010seATURE TAtet DESC8MPTON Pall fat FAIL FAIL FE FAIL +
ACCESS CONTROL SECTON a
fees Peep Gmusesse
- Demeter cumsomww Tes>
8 #wass sans pasases g Charesann 198>
A Asp tigaste
$M 8
Seew4r Aesses'Penst Chapeaney Test PWesWIF EIS . *
- r Operess esse asseum hade Curses A Pousse mees 5 8
N hIhs O M p Ossets the Aer# Duos tesess 3- ,
usansa Rose unen semp 0 loansne pensent Cremas home upp PeseW I
teses Conseer W Camees The B. Rey $Wesse -
Cantees A seems eens asses }
'it Demeuw couhan a seeneone names Demone 6asse
. -Ae e J
, -A-esse i
e, voarime aussen, os em 4840se 6 e e .
ACCESS CONTROL SECTON (CONT.)
l h . s., - .e s caem pseenw oorum Casuhat payeus Mouse i g i Saee Coeu6 A yeuses Furne
-womaso Assess heem A Destes
% Asas g, pesens tusene ses FA l
last beewsp % l i
esmh emesenses O &aas teamit hamig I
& *u'e's'p"y '**"
D G a*s"Ce'nWowd i
g h M $45 @A Ike humas4, haang Pueuses vouses suMIEter 8'
r E01150& Aus weh Guls 9 Gesesus fases Causes i s =s e esey3 Lan
, emase am -
wasums GS %ERWIf Io Asutes g, tsaus "
Meeum a ftsus " l O ***k'_****'
- A
% -, l
. ...-. - -++.
, e NAh8E - DUTY AS$KiNMENT SSM: DOE;
+
3
() perYAL QUALFCAYcal () REQUALFCATION flEQUALFCATION T .. DATE DATE DATE SECURrTY EVALUATORf DATE DATE DATE stCURsYY EVALuATORf A .... . ... ** . . . . .. a. .. ... OFFCER pf6TRUCTOR I CRUCIAL ' ...... OFFCER DESTRUCTOR
-8 MCURrry Pats / PASS PAges geopeATURE ge0 NATURE. pat
- PASSF PASSI WIONATURE SIONATURE l N TAAst DESCRPTON FAL fu FAIL FAL FAL FM ,
ACCESS CONTROL SECTION (CONT.)
, caos asume te ses 4eas Se e
CelWIs W AW PufWS
'" Omeg A Coenysgr i Psepa* MesmW Puese
" O ushname tenus hemmemcAT
- .p W 136 hasse-este ensema.eans Acctes costra arcTions44ao e 1
emmense o.ame Ausst lueesey G A tassauge Deene .
C s. .e e,
- haee Ceuksi An $sesus huse O tessup Ces Asse pennieser 9 Aneuses g assess asas peauespy AAesang P Uno DC 9enge 31summas to a #semanas Duartenes 9 etoep AComelasesmense ce 3 anm dethen aanse hasse - Seat bee Giesse 307 RESPONSE PERSONNEL SECTION on.e. Asees menn Ree.
Ogoem 5 #winnas Ciene Cesusume senas mese camp IS Deces W I
3 Cem.mo a sinne on seem c a, e as-e 3
tumuh e togtsey AusmomeOmege '
.t Cases spessued assies Cenessmovemytensee '
empo -
- tmap timme Puesmas. At )
stb j l
'" einseeesse 09ARMem tukinsin assume i
I 4 Cassassiunas tues j 1
g easesteamhunesos team -
m + r
L NAM 2- DUTY A5$1GNMENT CN: DOE;
}
() INITIAL QUAUFCATION () REQUAAIFCATION REQUAUFCATION -
DATE DATE DATE MCURITY EVALUATORI DATE DATE DATE SECURfTY EVALUATORf CRUCIAL . 4... . .... .. . OFFCER NesTRUCTOR = * *. ** +- * **
- OFFCER INSTRUCTOR 86CUnffY PA44I PAS $ PA$$f SIGNATURE StuNATURE ' se ASG/ PAS 8f PAS $/ 38GNATURE SIGetATURE TA$st DESC94PflDN FAL PAL ' FAIL FAL Fall FAlt ,
C RESPONSE PERSONNEL SECTION (CONT.)
lseen Asman -
'[vesem a 5
eTWe: Apea reew
' esseson e.Laesow causkPues Aree seuse Posenseer Fouse 8
he vue mes ne=e a inessend $2 asense . -
Dome femessed Anna L.
Ceseel Assene to A Ved see e Ce=W *eism Game Ace =esar some.
- cueseenneo Du** ACoogeur nessensmu senscieser esses oresses to Desmene9Heessaf einiese Geseneses>
i nesm %esencAt RESPONSE PERSONNEL SECTION . 7A00 Senes i nesserme **uase lesseen Demos Aseen Desseev 01 A Geeseos Desse Commes An eesus hele SeeJe O Connel As Gospur Aspe Rene e
l g case see evenemer l
1
, -e w l annan
- um e sees Quesesame 8 teones & Cas Owenesus
+1 mensgnes testeue 9 5etence e esones,an omen anse benent Ste$
nas Esopel M I
RESPONSE PERSONNEL SECTION 7500 Sense i
se neew r - reveems
, Denne m se*
iPg teereissesses g sessessee-88epe i e sees tee meses. Teen Ihenseele & saamen O Diwesse sissamme 9p W 4 messeen
,e esse ne e-mew eeusse -
. . . _ . _ . _ _ .m . _ - _ - . . 4._ .. _ ..
I e !
+
pdAh2 ' DUTY AS$10NMENT SON: DOE; _ j j i
8 i) DNTML QUAUFCATION - -() REOuAUFCATION REOUAUFCATION .j DATE DATE DATE eECURffY EVALUAtorv DATE DATE DATE SECunfTY EVALUATOR /
CRUCIAL . ...... ..... ,.... OpFCen DESTRUCTOR . . .e * ** ...u . OFFCER De8TRUCTOR )
MCURITY - PASSI PASS PAGE/ possATURE SIGNATURE PAS $/ PAS $/ PAS $/ SsONATL#tt SIGNATURE TAdst DESCRIPTIOls FAA FAL PAA, FAIL Fan past ?
RESPONSE PERSONNEL $2CTION . 71100 Senes MONT.)
, assue fa 8ess 01Gessnes -t aumAemory a sesess4menism < ;
.t g %sse empt '
an=== h, {
amas unas ans caen ey 8
Amesm9
{
i amamme tesumsassunsee .
h emas poo passen essumsmay s W Pee Dessee 3
s
- M M Ine Insasans puesmasas C.a.ese s= ameans tenusmos f Commeilmansas fame O tapenne use pne .
Essess nueue uum eseyan '
W- i e aaem aa tusamme f
- s. ;
- t A,.u4 m e. ,
e tesse as Asesus Deusess 85CURffY BHIFT SUPERVISORS r
M IGN-4mm89 W e
N
.i
.i Opson A Pwenne pines Deen assie aussues i fama put See ones n i re===,u a .
1
- ase, I
- a. ase Regund D' 4.emmser poemme Poesse le Hesus n==m == me. enum i
- fans to to Rummt 1 eduus & Smasey amen tusassitunug i
tamanne caos nem se Asummun amtIts Etussee #ses Raummes 4& seemen '
Enamen eased h Sauseum lutamsm poem 8teump 8euse 8"E' ,
l a m. '..
]
i seampuseums mat l
_- - , . . - - . . - . - . - - . = , .- - ,. .- - . - - . > - .. .. nn~ -
&_ j r
i
~ \
1 NAME ~ DUTY A55IGNMENT
]
DOE:
. [,SN: ]
() INfflAL QUALMCAtl0N () REQuAtiPCATON REQUALFICATION 1
t f5 DATE DATE DaTE afCURITV EVALUATOfv DATE DATE DATE SECURrry EvAtuATORI -
4 ,
' CRUCIAL ' ,
...... .... . . . . osvlCER pi4TRUCTOR . . .. . . ... OPPCER . MSTRUCTOR S MCUfWTY Patel PAtt PA8AI saoNATURE gioseATURE ' PA88/ PAS $/ Pa$$/ th0 NATURE $8GNATURE 4 TAes pracMurforms saa - pas saa saa see see SECURffy SHIFT SUPERVISOR $ + 7400 genes er game asses D W Dudus i
]
nom an==, e a tange Sean [
e Om.m.es
. a. son = es=e
- Cunt.as e .
as hapur tue Omur Aus passauer . i 8 meusse aman nes I
- punemy asmew .
e unsgrams - i almanus ba puesumi i g
IBesumes t 0 teuer Athe teasennus k l
,.g 81mepu.s Snapus 0F I e.
3 aussir am oven deman hemme 9ses nas -
i 1
e,emme. smet Om n.se=a.
~b v =, n. sv a
." An bene em us ter *= === ease Cesent84 penas Oweg a j
.
- l l l l I I I I ;
Range Quahncatons !
e sA . l Cenese emmenamnes ca a van I
Vm hoped Atessepaa i g fee A aesnape ano tesem ma Poe a 6msnge m ames l
e An = ;
mas j
=
ca . on A emmaissiums seen ag tusesa M ^ f Imp
- fesiammammuner Nas name basem
- A~ g IIIe 4 sere i
et sus A menpa !
- Cesens - OnA '
i Buumm
[
. Ae 8
= .he.19.ugs
. amu
- aus 8est emuse m any, i s TfWNING AND QUAUFICATON RECORD CARD AUDIT I
' Daft l FEUNTastest l OIENATURE l DATt l PIIsfTeamedE l SIGNATURE l DATE l PUNNTNAAfE f GeGueATURE l DATE l PastrfenesE l 810 FEATURE l
4
.-- , , .x> .-.n , a a - , u an.-
+-
0 %
- -- g e e . . c , .
. 5 ** ?
3 1 f #
I I !
e l
f f
5 i
l a E
l l a .
I E
. . . . . , . .
N E
- m
' k 4
O 5 i j
,i . ;
i ;
i a 1 i
e I l I
e I
. I ;
i 4 !
1 f
a e
L .
N s
\_ . . -
Operations Department St. Lucie Nuclear Power Plant Night Order DISTRIBUTION: Unit 1 Control Room Unit 2 Control Room OPS Support (D-13) Work Control Group System Specialists Training ,
Simulator ;
From: Operations Supervisor's Office Date: January 26,1996 To: All Operations Personnel
- 1. During the weekend a NRC team will be onsite for the purpose investigating aspects of the dilution event that occurred.
They will be requesting to interview operators. These interviews should occur in the control room of the affected unit. The interviews should only last 30 minutes. Please ensure that two SRO's and RO are in the Control Room when an RCO is being interviewed. Jeff West is on call Saturday for any help that the NRC requires and I am on call for Sunday. l l
- 2. Now for some good news. There is a possibility that the shift-splitting agreement may come back soon. More information will follow when i hear about it.
..( -.
, j l -
f g D~ .
Vlh h; /
L ' j !
p
i Operations Department l St. Lucie Nuclear Power Plant Night Order DISTRIBUTION: Unit 1 Control Room Unit 2 Control Room OPS Support (D-13) Work Control Group System Specialists Training Simulator From: Operations Supervisor's Office Date: 01/25/96 To: All Operations Personnel
- 1. On 1/22/96 during the midnight shift a loss of reactivity control occurred when the board operator initiated a dilution to the RCS and then forgot that the dilution was in progress. This is a extremely serious event! Attached are copies of T.C.'s to the conduct of operations procedure that will put in place process changes to prevent reoccurrence. Also attached is a synopsis of the event with lessons learned and corrective actions. It is my expectation that an event of this type will never occur at this facility again.
One of the lessons learned from this event is the need for prompt notification of management of significant events at the plant. The Conduct of Operations has specific guidance in this area. Appendix E, Notification of Operations Supervisor /FPL management, has specific criteria of when I am to be immediately notified. Specifically step 1.F., " unexplained or unplanned reactivity changes" and step 1.J., "any operational event which generates an In House Event report and causes heightened awareness to EPl. sources offsite". Both of these criteria were clearly met. However I was not immediately notified. This is unacceptable. This accountability rests with the NPS,although I expect the shift team to cack up the NPS should he fail to recognize that the notification criteria has been met. To ensure that we are clear on the criteria all NPS's are to get with me at the earliest cpportunity to review the criteria for notification.
The ANPS's are to review, with their crew, this night order and all of the attacnments at their next shift turnover meetinc. A signed attendance roster is to be forwarded to me. The NPS's are accountable to ensure that all shift members receive the brief.
A three person NRC team will be on site investigating this event starting tomorrow and going through the weekend. I anticipate that they will be wanting to interview operations personnel. Ensure you are cooperative and professional.
W
j) .
[ .
g, ,
m
~1 -
1 .
i i
ST. LUCIE PLANT PLAN TO IMPROVE OPERATIONAL PERFORMANCE i l
i l kp o
n, visas
i y..
4 F
PROBLEM #1: Acceptance Of Long Standing, Repetitive Protrienes By Plant Managensent ITEM . ACTION DESCRIPTION DATE B
I. Immediate work ' stoppage on Unit I to Employee meetings were held with President-Nuclear Division, impress on all personnel the need for Site Vice President, and Plant General Manager. Meetings change.
g focused on the need to reduce equipment deficiencies which .
impact operations.
2 Kept Unit I shut down to correct Existing deficiencies (including work orders, jumpers, St. Lucie' deficiencies which could impact operations. Action Reports and caution tags) were reviewed by senior plant management. 83 deficiencies which could adversely impact .
Complete operations were added to tne work scope of the Unit I shutdown. These deficiencies were corrected prior to retuming -
the Unit to service.
- 3 Unit 2 was maintained at reduced power to A plant nwdification was implemented to eliminate a history of resolve a long-standing IIcater Drain pump low flow trips on the IIcater Drain pumps. The same . Complete problem. modification was made on Unit 1.
I 4 Improve the process for ensuring timely Administrative Procedure, " Assessment of Abnormal Plant
- resolution of existing deficiencies. Configurations or SigniGcant Material Deficient Conditions on Plant Operation," was developed to enhance outage scope Complete .
review and ensure that equipment deficiencies are restored in a timely manner.
i-5 Identify and correct deficiencies on Unit 2 A review of the scope for the upcoming Unit 2 refueling outage was completed, which utilizes the new outage scope review process. Appropriate activities were added to the scope of the outage.
Complete To accommodate the increased scope of the Unit 2 refueling outage, and to provide an added measure of conservatism, the planned fuel shuffle has been rep!xed by a complete core ' .
offload / reload sequence.
Rev. 4.1/18/96 '
. J
-._-.-_u.-
PROBLEM #2: Equipment Performance Is Not Satisfactory ITEM ACTION DESCRIPTION DATE I Strengthened technical leadership. Vacant Technical Manager position was filled on September I, 1995.
Consolidated plant engineers from the Maintenance Department gg ,
into the Technical Department.
Suspended SRO certification course. The Engineering Manager ,
and Maintenance Manager retumed to nonnal duty on August 19,1995. They will complete this course at a later date. b 2 Strengthen root cause analysis. Reinforced and carried out the Division policy that defines the threshold of when a formal root cause should be executed and standardizes root cause format.
Complete Refresh key personnel in the new technical group and engineering in root cause analysis methods.
3 Improve management visibility of Established weekly management review of appropriate equipment performance problems. performance indicators and work backlog status, including magnitude of work, age of oper. items, and operator workarounds*.
Complete 1
- Operator workarounds at St. Lucie include operator -
inconveniences which are categorized separately by other utilities.
4 Expedite reduction of operator 83 operator workarounds exist. These will be reduced by half. 6/1/96 workarounds.
Rev. 4,1/18/96
s.
c ..
PRO LEM #2: Equipment Performance Is Not Satisfactory (conthmed)
ITEM ACTION DESCRII" TION DATE 5 Improve the post maintenance test program Test groups were consolidated under a single manager reporting to ensure equipment will operate as to the Operations Manager on September I,1995. Complete designed in the plant.
Review Unit 2 outage scope test procedures to ensure critical component functions are addressed. Complete i
Revise process for post maintenance testing to improve coordination between Outage Management, Operations and 2/29/96 Maintenance.
-6 Improve quality of work performed by Re-emphasized that each FPL contract administrator contractor personnel. is accountable for the quality of work performed by contractors.
Provided training /re-training for FPL persennel assigned duties as contract administrators.
Technical and engineering personnel will review procedures Complete used by contractors during the upcoming Unit 2 refueling outage to identify quality control attributes / processes and -
desired confidence Icvels deemed critical for equipment performance. +
t Inspection plans to meet these specifications will be developed.
i
-i Rev. 4.1/19/96 ,
I i
PROBLEM #3: Personnel Performance Ilas Not Been Adequate -
ITEM ACTION DESCRIPTION DATE I Reinforcement of management expectations Plant meetings have been held to reinforce high standards of at Manager / Supervisor level. personal accountability and will continue periodically. Nuclear Complete -
Plant Supervisors met with their crews to se: clear expectations for error-free performance.
2 Increase management oversight. Managers and supervisors have been directed to spend significantly more time monitoring work areas under their Complete oversight.
3 Push high standards of accountability Reinforced expectations of recognition for superior performance g throughout the organization. and discipline for substandard performance.
4 Assess ability of first line supervision to The independent assessment team identified this as an area for address personnel perfonnance issues further evaluation. Conduct an assessment of this issue and g develop an action plan to resolve.
The assessment was completed. The action plan follows:
Develop and issue guidelines to Foremen and Supervisory personnel on assessing employee performance. Complete R4 Evaluate and modify if necessary, the accountabilities of our supervisors to ensure handling employee performance is identified as a Key Responsibility. Ensure these accountabilities 3/15M6 are clearly identified in the Supervisor selection process. ,
Rev 4,1/18/96
-t .
PROBLEM #3: Personnel Performance Has Not Been Adequate <cond ee ITEM ACTION DESCRIPTION DATE 4 Assess ability of first line supervision to Evaluate and modify, if necessary, the accountabilities of the address personnel performance issues. Foreman level positions to ensure handling employee .
(continued) performance issues is a Key Responsibility. Ensure these '1/15N6 -
accountabilities are clearly identified in the Foreman selection process.
Interview all incumbents in the Foreman and First Level of supervision positions to ensure they are willing to meet the expectations of the position in handling employee performance 6/30N6 issues.
For those employees who are not willing to meet the new R4
^
expectations, identify replacements and replace incumbents. 8/31N6 For those employees who are willing to meet the new expectations, review performance to determine if expectations 12/3 tN6 are being met. If not meeting expectations, replace incumbent.
Rev. 4,1/18/96 l
_ _ _ _ _ _ _ _ _ . __ _ _ _ _ _ _ _ _ _ = _ . _ _ _ _ _ _ _ ._ -_ __ _. . ._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
PROBLEM #4: Procedures Have Been Approved With Technical Deficiencies ITEM ACI' ION I)ESCRIPTION t) ATE I Improve the technical review for first-time Plant Policy 105 has been revised to include a technical review use of procedures under conditions for first-time use of procedures under conditions different than different than originally intended. originally intended.
Complete This policy has been integrated into the infrequently Performed Test and Evolutions procedure.
2 Improve the Facility Review Group (FRG) FRG administrative accountabilities have been transferred from process so that senior plant management the Plant General Manager to the Services Manager. Complete can apply greater focus to safety-significant procedures. Reduce the volume of material reviewed by the FRG.
Maintenance has redefined the review criteria for work orders Comp Icte that reference Technical Manual instructions.
Establish a screening subcommittee [similar to the Company Nuclear Review Board (.CNRB) subcommittee] to allow the FRG to perform a more detailed critical review of safety- Complete significant procedures.
Convert those procedures which are not required by Technical gg Specification 6.8.I to departmental guidelines.
3 Expedite correction of errors in plant Place emphasis on accountability of a procedure user to ensure procedures that the procedure is correct and adequate.
l l Adopt the definition of procedure compliance consistent with Turkey Point Plant.
Cosgh l
Establish a schedule for upgrade of Operations Normal and Off-Normal procedures. _.
Rev. 4.1/18/96
J O PROBLEM #5: Management Is Not Provided With Sufficient Information To Assess / Trend Plant Events ITEM ACTION DESCRIFrlON DATE I improve assessment of deficient conditions. Combined the personnel from St. Lucie Action p .apst l (STARS), in-flouse Event (IllEs), and lin e. s'erformance Complete Enhancement System (IIPES) into a smgle group.
l Include all data bases in trending effort, including Operations, Security, Ilealth Physics and Safety. Complete R4 Policy established to integrate personnel from outside St. I.ucie Plant into benchmarking efforts and plant assessments. "*P'#'"
2 Better utilize Quality Assurance (QA) Management will review QA audit inspection plans quarterly.
capabilities. The initial meeting reviewed the fourth quarter inspection plan. Complete Delegate QA management oversight of Steam Generator Replacement Project from QA Manager to QA Supervisor.
Complete Establish a rotation plan to line/QA employees for 1996.
Complete R4 Participate in QA Audit Technical Specialist Exchange Program Complete with other utilities.
3 Upgrade log-keeping activities to supply Management is conducting a daily review of Control Room logs better communications to management. to reinforce the expectations for detail and completeness. Complete Computerize Control Room logs and make logs available to management electronically. Complete -
4 Improve focus of the Shift Technical Reassign the STA's to the Operations Department. Complete Advisor (STA) during normal plant
! operation. Reduce STA administrative duties that do not relate to Control Room activities.
Rev. 4.1/18/96
. - _ _ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ - _ . ___ . .~ - _ .. . - - .
ST. LUCIE PLANT PLAN TO IMPROVE OPERATIONAL PERFORMANCE ITDi PROBLEM ACTION DESCRIPTION DATE STATUS DUE 2-Sb Equipment Performance Improve the post Review Unit 2 outage scope 11/9/95 CI4 SED Is Not Satisfactory maintenance test test procedures to ensure program to ensure critical component equipment will operate functions as designed in the are addressed plant 3-4a Personnel Performance Assess ability of The independent assessment 12/10/95 CLOSED Has Not Been Adequate first line supervision team identified this as an to address personnel area for further performance issues evaluation. Conduct an assessment of this issue and develop an action plan to resolve 5-4b Management Is Not Improve focus of the Reduce STA administrative 12/31/95 CIASED Provided With Shift Technical duties Sufficient Information Advisor (STA) during To Assess / Trend Plant normal plant operation Events 5-lb Management Is Not Improve assessment of Include all data bases in 12/31/95 CIASED Provided With deficient conditions trending. effort, including Sufficient Information Operations, Security, To Assess / Trend Plant Health Physics and Safety Events 5-2c Management Is Not Better utilize Quality Establish a rotation plan 12/31/95 CIASED Provided With Assurance (QA) to line/QA employees for Sufficient Information capabilities 1996 To Assess / Trend Plant Events 5-2d Management Is Not Better utilize Quality Participate in QA Audit 12/31/95 CI4 SED Provided With Assurance (QA) Technical Specialist Sufficient Information capabilities Exchange Program with To Assess / Trend Plant other utilities Events erum/czosso erarve meersomse 11/t/es at.t. zruses rarem to 11/1/ps ama crosso
~ , a.n
_ . .- _ _- _ .~ ~ _ _. _. ._. _ _ _ __ . _ . . . . ._.
- - -e
- -
1 i
ST. LUCIE PLANT "
PLAN TO IMPROVE OPERATIONAL-PERIORMANCE !
ITEBf PROBLEM ACTION' DESCRIPTmet- DATEi STATUS DUE-3-4b Personnel Performance Assess ability of Develop and issue 1/15/96 CLOSED ;
Has Not Been Adequate first line supervision guidelines to Foreman and '
, to address personnel Supervisory personnel on i performance issues assessing employee t'
, performance
, 2-Sc Equipment Performance Improve the ost Review process for post 2/29/96; OPEN Is Not Satisfactory maintenance est maintenance-testing to . .
1 program to ensure improve coordination '
equipment will operate between Outage Management, as designed in the Operations and Maintenance plant 4-2d Procedures Have Been Improve the Facility Convert those procedures 2/29/96 OPEN Approved With Review Group (FRG) which are not required-by Technical Deficiencies process so that senior Technical Specification ,
plant management can 6.8.1 to departmental l apoly greater focus to guidelines salety significant-procedures -
Personnel Performance 3-4c Assess ability of Has Not Been Adequate first line supervision Evaluate and modify if 3/15/96 OPEN !
the '
to address personnel necessary,lities accountabi of our F performance issues supervisors to ensure handling. employee performance is identified ,
as a Key Responsibility.
Ensure these accountabilities are clearly identified in the -
Supervisor selection '
process !
l l
4 oems/cumous staTwo amerserrise 11/1/es 4 aza ITunes onzen To 11/1/95 ans essene '
. - - - - - c- -r ,---es, ,- -- e- - g..m n- m-* . . . - - = rm. .w-e- - w - c --se ,w--e,s ,
ST. LUCIE PLANT PLAN TO IMPROVE OPERATIONAL PEREVRMANCE ITEM PROBLEM ACTION DESCRIPTION DATE STATUS DUE 3-4d Personnel Performance Assess ability of Evaluate and modify if 3/15/96 OPEN Has Not Been Adequate first line supervision the to address personnel necessary,lities accountabi of the performance issues Foreman level positions to ensure handling employee performance issues is a Key Responsibility. Ensure these accountabilities are clearly identified in the Foreman selection process 2-4 Equipment Performance Expedite reduction of 83 operator work-arounds 6/1/96 OPEN Is Not Satisfactory operator work-arounds exist. These will be reduced by half i 3-4e Personnel Performance Assess ability of Interview all incumbents 6/30/96 OPEN Has Not Been Adequate first line supervision in the Foreman and first to address personnel level of supervision performance issues positions to ensure they are willing to meet the expectations of the position in handling employee performance issues 3-4f Personnel Performance Assess ability of For those employees who 8/31/96 OPEN Has Not Been Adequate first line supervision are not willing to meet to address personnel the new expectations, performance issues identify replacements and replace incumbents 3-4g Personnel Performance Assess ability of For those employees who 12/31/96 OPEN Has Not Been Adequate first line supervision are willing to meet the to address personnel new expectations, review performance issues performance to determine if expectations are met.
If not meetin expectations,greplace incumbent .
orms/ctosas starve amerzwtwa 11/1/9s alt. zrmMs enroe To 11/1/95 ama exosmo
NF-TR-95-01 FPL i
i i
i i
NUCLEAR PHYSICS
! METHODOLOGY FOR RELOAD DESIGN OF : ,
TURKEY POINT & ST. LUCIE i
- NUCLEAR PLANTS
- JANUARY 1995 b/
- 4. ,,,,
, t-
- 4remong pw w,,j>\gg bb !
NUCLEAR PHYSICS METHODOLOGY FOR RELOAD DESIGN OF TURKEY POINT & ST. LUCIE NUCLEAR PLANTS NF-TR-95-01 JANUARY 1995 l
i J
h FLORIDA POWER & LIGHT COMPANY NUCLEAR FUEL SECTION JUNO BEACH, FLORIDA
4 ABSTRACT l This document describes the nuclear design methodology employed by Florida
! Power & Light Company (FPL) to analyze the core design characteristics
- necessary to support a fuel reload for Turkey Point Units 3 and 4 and St. Lucie Units 1 and 2. This methodology, including all computer programs used, was obtained from Westinghouse Electric Corporation. Calculations were performed
- using this methodology and the results compared to operating data from Turkey Point and St. Lucie. The quality of the comparisons demonstrates FPL's ability to perform reload core design for FPL's nuclear units.
a i
l I
i 1
i i
l l
l 1
L___-___- -
a TABLE OF CONTENTS SECTION PAGE
?
1.0 INTRODUCTION
AND CONCLUSIONS .......................... 1
- 1.1 OBJ ECTIVE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i
1.2 BACKGROUND
..................................... 1
- t
- 1.3 SC OPE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1.4 CONCLUSION
S ..................................... 4 2.0 PHYSICS METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 i i 2.1 CROSS SECTION LIBRARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 i 6 !
- 2.2 LATTICE MODELING IN PHOENIX-P .....................
2.2.1 FUEL CELL MODEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
! 2.2.2 DISCRETE ABSORBER MODEL . . . . . . . . . . . . . . . . . . . . 7 2.2.3 STRUCTURAL CELL MODEL . . . . . . . . . . . . . . . . . . . . . . 9
]
2.3 BAFFLE-REFLECTOR MODELING . . . . . . . . . . . . . . . . . . . . . . . 9 2.4 THREE-DIMENSIONAL NODAL MODEL . . . . . . . . . . . . . . . . . . . 9
! 2.5 ONE-DIMENSIONAL DIFFUSION THEORY MODEL ......... 10 3.0 PHYSICS MODEL APPLICATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 j 3.1 CORE POWER DISTRIBUTIONS AT STEADY STATE C O N DITI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 3.1.1 POWER DISTRIBUTIONS ........................ 11 l 12 3.1.2 POWER PEAKING .............................
! 3.1.3 FUEL DEPLETION ............................. 12 3.2 AXIAL POWER DISTRIBUTION CONTROL LIMITS . . . . . . . . . . 13 3.3 CORE REACTIVITY PARAMETERS ..................... 14 l 15 3.3.1 MODERATOR TEMPERATURE COEFFICIENT ........
3.3.2 DOPPLER COEFFICIENTS ...................... 15
! 3.3.3 TOTAL POWER COEFFICIENT . . . . . . . . . . . . . . . . . . . . 16 3.3.4 ISOTHERMAL TEMPERATURE COEFFICIENT . . . . . . . . 17 3.3.5 BORON REACTIVITY COEFFICIENT . . . . . . . . . . . . . . . . 17 3.3.6 XENON AND SAMARIUM WORTH ................. 18 i 3.3.7 CONTROL ROD WORTH ........................ 18 3.3.8 NEUTRON KINETICS PARAMETERS ............... 19 j 3.4 CORE PHYSICS PARAMETERS FOR TRANSIENT ANALYSIS i INPUT ........................................... 20 J
4.0 PHYSICS MODEL VERIFICATION TURKEY POINT UNITS . . . . . . . . . . 21 i
4.1 CYCLE DESCRIPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
- 4.2' ZERO POWER PHYSICS TESTS ....................... 23 4.2.1 CRITICAL BORON CONCENTRATIONS ............. 24 4.2.2 TEMPERATURE COEFFICIENTS .................. 24 e 4.2.3 CONTROL ROD WORTH ........................ 24 4.2.4 DIFFERENTIAL BORON WORTH .................. 25 i
i i
t
i TABf F OF CONTENTS (CONTINUED) 1 SECTION PAGE-1 i
4.3 POWER OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 i 4.3.1 BORON LETDOWN CURVES ..................... 26 !
. 4.3.2 POWER PEAKING FACTORS . . . . . . . . . . . . . . . . . . . . . 27 l 4.3.3 RADIAL POWER DISTRIBUTIONS ................. 28 4.3.4 AXIAL POWER DISTRIBUTIONS AND AXIAL-OFFSETS . 28 4.4 SUM MARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 5.0 PHYSICS MODEL VERIFICATION ST. LUCIE UNITS . . . . . . . . . . . . . . 73 5.1 CYC LE DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 5.2 ZERO POWER PHYSICS TESTS ....................... 74 5.2.1 CRITICAL BORON CONCENTRATION .............. 75 5.2.2 MODERATOR TEMPERATURE COEFFICIENT ........ 75 5.2.3 CONTROL ROD WORTH ........................ 75 5.2.4 DIFFERENTIAL BORON WORTH .................. 75 5.3 POWER OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 1 5.3.1 BORON LETDOWN CURVES ..................... 76 !
5.3.2 AXIAL POWER DISTRIBUTIONS . . . . . . . . . . . . . . . . . . . 76 5.4 S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76
6.0 REFERENCES
.......................................... 102 APPENDIX A WESTINGHOUSE COMPUTER CODES . . . . . . . . . . 104 A.1 FIGHTH .................................... 104 l
A.2 P HO ENIX-P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 l A.3 ANC........................................ 106 A.4 A PO LLO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 i
l l
ii
l
!4 l '
i
- l. l LIST OF TABLES TABLE PAGE !
1 4.1-1 Turkey Point Unit 4 Fuel Specification . . . . . . . . . . . . . . . . . . . . . . . 30 4.2-1 Turkey Point Unit 4 HZP Physics Test Review C rite ria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 l
4.2-2 Turkey Point Unit 4 Critical Boron !
Concentration Comparison Between '
Measurement and Predictions for Cycles 12, 13, a n d 14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.2-3 Turkey Point Unit 4 Moderator and isothermal Temperature Coefficient Comparison Between Measurement and i Prediction for Cycles 12,13, and 14 . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.2-4 Turkey Point Unit 4 Control Rod Worth !
Comparison Between Measurement and I Prediction for Cycles 12,13, and 14 . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.2 5 Turkey Point Unit 4 HZP Differential Boron Worth Comparison Between Measurement and Prediction for Cycles 12, 13, a n d 14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 4.3-1 Turkey Point Unit 4 Cycles 12,13, and 14 Boron Letdown Comparison Between Measurement and Prediction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 4.3-2 Turkey Point Unit 4 Cycles 12,13, and 14 Power Peaking Factor (F,) Comparison Between Measurement and Prediction . . . . . . . . . . . . . . . . . . . . . . . 37 i
4.3-3 Turkey Point Unit 4 Cycles 12,13, and 14 Power Peaking Factor (Fo) Comparison Between Measurement and Prediction . . . . . . . . . . . . . . . . . . . . . . . 38 4.3-4 Turkey Point Unit 4 Cycles 12,13, and 14 Axial Offset Comparison Between Measurement and Prediction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 iii
r LIST OF TAB FS (CONTINUED)
TABE PAGE 5.2-1 St. Lucie Unit 1 Critical Boron Concentration Comparison Between Measurement and Predictions for Cycles 10, 1 1, a nd 12 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 5.2-2 St. Lucie Unit 1 Moderator Temperature Coefficient Comparison Between Measurement and Prediction for Cycles 10,11; and 12 . . . . . . . . . . . . . . . . . . . . . . . . . 79 5.2-3 St. Lucie Unit 1 Control Rod Worth Comparison Between Measurement and Prediction for Cycles 10,11, and 12 . . . . . . . . . . , . . . . . . . . . . . . . . 80 5.2-4 St. Lucie Unit 1 HZP Differential Boron Worth Comparison Between Measurement and Prediction for Cycles 10, 1 1, a n d 1 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 5.3-1 St. Lucie Unit 1 Cycles 10,11, and 12 Boron Letdown Comparison Between Measurement and Prediction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 iv
)
. 1 LIST OF FIGURES i l i
FIGURE PAGE i i
4.1-1 Turkey. Point Unit 4 Cycle 12 Core Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 4.1-2 Turkey Point Unit 4 Cycle 13 Core I Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . r . . . . . . . . . 41 l J
4 4.1-3 Turkey Point Unit 4 Cycle 14 Core
. Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 l
- (
l 4.2-1 Turkey Point Unit 4 Cycle 12 Measured i versus Predicted Control Bank C Integral Rod 1 Worth............................................... 43 i
4.2-2 Turkey Point Unit 4 Cycle 13 Measured !
versus Predicted Control Bank A integral Rod l Worth............................................... 44 1 4.2-3 Turkey Point Unit 4 Cycle 14 Measured
. versus Predicted Shutdown Bank B Integral Rod 2 Worth............................................... 45 '
4.3-1 Turkey , Point Unit 4 Cycle 12 Boron '
Letdown Comparison Between Measurement
> a n d Pred ictio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46
- 4.3-2 Turkey Point Unit 4 Cycle 13 Boron Letdown Comparison Between Measurement a nd Pred iction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 4.3-3 Turkey Point Unit 4 Cycle 14 Boron
- Letdown Comparison Between Measurement 4
and Pred ictio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 s
4.3-4 Turkey Point Unit 4 Cycle 12 F.
Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . 49 4.3-5 Turkey Point Unit 4 Cycle 13 F, 4
Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . 50 i
4.3-6 Turkey Point Unit 4 Cycle 14 F, Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . 51 v
LIST OF FIGURES (CONTINUED)
FIGURE PAGE t
< 4.3-7 Turkey Point Unit 4 Cycle 12 Fa Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 L 4.3-8 Turkey Point Unit 4 Cycle 13 Fo Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 l
4.3-9 Turkey Point Unit 4 Cycle 14 Fa Comparison Between INCORE and ANC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 4.3-10 ' Turkey Point Unit 4 Cycle 12 Radial Power Distribution Comparison Between INCORE and ANC - 2320 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 4.3-11 Turkey Point Unit 4 Cycle 12 Radial Power
- Distribution Comparison Between INCORE and ANC - 6975 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 4.3-12 Turkey Point Unit 4 Cycle 12 Radial Power i Distribution Comparison Between INCORE and ANC - 11812 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57
. 4.3-13 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and A NC - 2440 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.3-14 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and l ANC - 6678 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 4.3-15 Turkey Point Unit 4 Cycle 13 Radial Power Distribution Comparison Between INCORE and ANC . - 12316 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 60 4.3-16 Turkey Point Unit 4 Cycle 14 Radial Power Distribution Comparison Between INCORE and ANC - 600 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 i
4.3-17 Turkey Point Unit 4 Cycle 14 Radial Power l Distribution Comparison Between INCORE and ANC - 68 36 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 I
vi l l
l 1
M 4
UST OF FIGURES 1 (CONTINUED) .
FIGURE PAGE ,
4.3-18 Turkey Point Unit 4 Cycle 14 Radial Power :
Distribution Comparison Between INCORE and !'
ANC - 10704 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 4.3-19 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 7620 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 4.3-20 Turkey Point Unit 4 Cycle 12 Axial Power Distribution Comparison Between INCORE and ANC - 9468 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 1
4.3-21 Turkey Point Unit 4 Cycle 12 Axial Power i Distribution Comparison Between INCORE and
. ANC - 11812 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 '
4.3-22 Turkey Point Unit 4 Cycle 13 Axial Power l
Distribution Comparison Between INCORE and AN C - 2440 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 l 4.3-23 Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE and
- ANC - 6678 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 i 4.3-24 ' Turkey Point Unit 4 Cycle 13 Axial Power Distribution Comparison Between INCORE and
- ANC - 12316 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69
! 4.3-25 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE and ANC - 600 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 l 4.3-26 Turkey Point Unit 4 Cycle 14 Axial Power i i Distribution Comparison Between INCORE and i ANC - 6836 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 l 4.3-27 Turkey Point Unit 4 Cycle 14 Axial Power Distribution Comparison Between INCORE and ;
AN C - 10704 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 1
i vii. <
--r- -- . . , _ . . - . . _ . _ _ _ _ _ _ _ _ . __ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l.- .
I
- LIST OF FIGURES (CONTINUED) i FIGURE PAGE
! 5.1-1 St. Lucie Unit 1 Cycle 10 Core j Loading Pattom . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 l l
5.1-2 St Lucie Unit 1 Cycle 11 Core Loading Pattom . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 ,
5.1-3 St. Lucie Unit 1 Cycle 12 Core Loading Pattern . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 5.2-1 St. Lucie Unit 1 Cycle 10 Measured versus Predicted Reference Bank integral Rod Worth............................................... 87 5.2-2 St. Lucie Unit 1 Cycle 11 Measured ;
versus Predicted Reference Bank integral Rod '
Worth............................................... 88 l
5.2-3 St. Lucie Unit 1 Cycle 12 Measured i versus Predicted Reference Bank Integral Rod Worth............................................... 89 5.3-1 St. Lucie Unit 1 Cycle 10 Boron Letdowr. Comparison Between Measurement e d Prod ;tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 5.3-2 St. Lucie Unit 1 Cycle 11 Boron Letdown Comparison Between Measurement a nd Prediction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 91 5.3-3 St. Lucie Unit 1 Cycle 12 Boron Letdown Comparison Between Measurement a n d Prediction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 i
5.3-4 St. Lucie Unit 1 Cycle 10 Axial Power Distribution Comparison Between INPAX and ANC - 372 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 l 5.3-5 St. Lucie Unit 1 Cycle 10 Axial Power Distribution Comparison Between INPAX and i ANC - 6904 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 vtLL i
~~
LIST.OF FIGURES (CONTINUED)
FIGURE PAGE j j
5.3-6 St. Lucie Unit 1 Cycle 10 Axial Power l Distribution Comparison Between INPAX and l 1
ANC - 15718 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 5.3-7 St. Lucie Unit 1 Cycle 11 Axial Power Distribution Comparison Between INPAX and ANC - 185 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 i I
5.3-8 St. Lucie Unit 1 Cycle 11 Axial Power Distribution Comparison Between INPAX and ANC - 6721 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97
(
5.3-9 St. Lucie Unit 1 Cycle 11 Axial Power l '
Distribution Comparison Between INPAX and j ANC - 12118 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 l
. 5.3-10 St. Lucie Unit 1 Cycle 12 Axial Power 4
Distribution Comparison Between INPAX and AN C - 625 MWD /MTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99 i
5.3-11 St. Lucie Unit 1 Cycle 12 Axial Power Distribution Comparison Between INPAX and ANC - 66 20 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 5.3-12 St. Lucie Unit 1 Cycle 12 Axial Power Distribution Comparison Between INPAX and ANC - 13320 MWDIMTU . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101 1
1 ix
a
' ~
1.0 INTRODUCTION
AND CONCLUSIONS This report describes the physics methods used by Florida Power & Light 4
Company (FPL) to analyze the core characteristics for our four Pressurized Water 3 Reactors (PWR). It includes a summary description of the Westinghouse ,
computer programs and methodology as applied by FPL to model the Turkey Point and St. Lucie Nuclear Power Station cores. Comparisons between predictions and operating data are provided as a demonstration of FPL's qualifications to use the Westinghouse methodology to perform reload design
, calculations for the Turkey Point and St. Lucie nuclear units.
1.1 OBJECTIVE The objective of this report is to demonstrate FPL's competence to perform reload design analyses for our four nuclear power plants. To this i end, extensive design calculations have been performed for Cycles 12,13
- and 14 of Turkey Point Unit 4 and the results are compared to actual plant operating data herein. Unit 4 was chosen for its wide variety of assembly and poison types, its transition to axial blanketed fuel, its large number of reinserted fuel assemblies, vessel flux reduction features (e.g., Hafnium inserts at the periphery), and its low leakage fuel management. Design calculations have also been performed for St. Lucie Unit 1, Cycles 10,11, i l
and 12 and a limited set of results have been compared to actual plant l operating data. Unit 1 was chosen for comparison because of its use of i Gidolinium bumable poisons, axial blankets and vessel flux reduction features in the core design.
1.2 BACKGROUND
FPL has determined that in-house capability to design reload cores for our units would provide the following benefits:
i 1
.. l
~~
- - Improved control over the design, yielding more control of the decision process, i
- Improved optimization of the design, allowing better fuel utilization and economics, and j
- A better understanding of the design, leading to more comprehensive !
evaluations of core safety.
l Various physics methodologies were renfewed to determine which best satisfied FPL's needs. FPL decided to use the West 39ouse approach, one of our NSSS vendors and present fuel supplier for Turkey Point. The Westinghouse methodology provided four important advantages:
1 1
i
- A physics methodology which included extensive written procedures (METCOM) which documented in step by step fashion core design j
calculational practices.
- A training program which provided hands on experience by utilizing
]
4 METCOM and performing actual calculations on the computer workstation to ensure that the FPL engineers understood the Westinghouse methodology.
- A physics methodology previously reviewed and generically i
approved by the NRC for all PWR applications, and
- An agreed upon process under which FPL engineers would perform the calculations related to the reload physics analysis process independently of Westinghouse for Turkey Point Unit 3, Cycle 14 with Westinghouse providing Quality Assurance of all calculations.
The purpose of this effort was to demonstrate the ability of FPL to perform the required analysis and to use lessons loamed to improve the implementation prior to operating independently from Westinghouse.
9 2
- - ., _ _l
j Implementation of the above decision required entering into a technology :
l exchange agreement with Westinghouse Electric Corporation. This agreement also provides FPL the abilhy to upgrade codes and methods !
I to be consistent with any revisions developed by Westinghouse. The !
relevant computer programs and associated methodology of I Westinghouse's Commercial Nuclear Fuel Division have been transferred j to FPL. A description of the applicable physics models is provided in the !
j next chapter while the computer programs themselves are discussed in Appendix A. The computer programs and procedures (METCOM) are l incorporated into the FPL Quality Assurance Program. !
i Training of FPL personnel in the Westinghouse methods was performed during 1993 utilizing the Nuclear Core Design Training Center approach provided by Westinghouse. FPL Individuals were trained in areas ranging from Loading Pattom Scoping, Cross-Section Development, Loading Pattom Generation, Safety Analysis Models and Analysis, Nuclear Design l Models and Analysis, to the development of Core Follow Analysis. In all, 14 FPL individuals were trained by Westinghouse in these areas representing well over 5500 manhours of training. Ongoing training by i Westinghouse has also been provided, a recent two day training session i
reviewed modifications to METCOM and provided technical interactions j between FPL perso'nnel and Westinghouse designers.
l l
j 1.3 3QQPE j l FPL has performed in-house core design calculations and core follow analysis for Turkey Point for many cycles. Core follow results obtained during Unit 4 Cycles 12,13, and 14 provide ample data with which to compare ppdicted power distributions, predicted boron letdown curves, ;
and fuel depletion calculations. In addition, the startup physics
~
measurements conducted during the startup of each cycle provide an l 3
. - - - . - . . - - . - . - - - . . - - . - - - - .=- .
r . !
i additional source of valid data for evaluating the physics model j predictions of critical boron concentrations, control rod worth, and j temperature coefficients. Detailed comparisons of the predictions and measurements are presented in Section 4. -
i FPL has also performed in-house core design calculations and core follow analysis forthe St. Lucie Units. Comparisons between measurements and 4
predictions for St Lucie Unit 1 Cycles 10,11, and 12 are presented in l Section 5 using Westirighouse methodology.
{
1 !
All methods used to generate the results detailed in this report (computer l' programs and model development) are standard licensed methods used by :
the Westinghouse Commercial Nuclear Fuel Division. Therefore, the !
calculational uncertainties (e.g., see Reference 1) associated with the methods are unchanged and do not require re-quantification. In addition, the methods utilized to process measured data (e.g., see Reference 2) for !
Turkey Point are also standard to Westinghouse such that measurement ,
uncertainties do not require re-determination by FPL.
{
1.4 CONCLUSION
S This report describes the use of the Westinghouse methodology as applied by FPL to model the Turkey Point Unit 4 and St. Lucie Unit 1 cores.
Calculations were performed for Cycles 12,13, and 14 for Turkey Point Unit 4 and the results were compared to actual operating data. Assemblies from Turkey Point Unit 4, Cycles 9,10, and 11 were also modeled to '
l establish the appropriate axial bumup distributions. Calculations were performed for Cycles 10,11, and 12 for St. Lucie Unit 1 as described in '
,_ Section 5. The results from these comparisons demonstrate FPL's understanding of the methodology and show that FPL can apply the METCOM procedures and computer codes during the performance of future reload design analyses for FPL nuclear units.
--- - .c
- r
. t i
).
2.0 PHYSICS METHODOLOGY
- This section describes the Westinghouse codes and methodology used by FPL to perform design calculations for reload cores. The major features associated with each model are discussed, as is the interaction between models. This methodology was also used to obtain the results presented in Section 4 and Section 5. Descriptions of the individual computer codes used are provided in Appendix A.
- Lattice physics parameters for unitassemblies and baffle-reflector cross sections are calculated with PHOENIX-P (Reference 3 and 11), a two-dimensional multi- !
, group transport theory code. Fuel and clad temperatures are generated with the 1
FIGHTH (Reference 9 and 10) code. The three-dimensional advanced nodal code )
ANC (Reference 8) is used to predict reactivity, power distributions, and other relevant core characteristics. In addition, APOLLO (Reference 12), a one-dimensional diffusion theory code is available to calculate differential control rod 4
worth and axial power distributions for the heat flux hot channel factor (Fo) synthesis to establish operational limits. The cross section library, as well as PHOENIX-P, nodal, and diffusion theory models are discussed in the following i sect *cas.
The models described here are representative of current Westinghouse practices.
FPL's calculational capabilities are anticipated to evolve in parallel with W6stinghouse's through planned implementation of the technology exchange !
agreement between the two corporations.
2.1 CROSS SECTION LIBRARY The PHOENIX-P computer program's nuclear cross section library contains microscopic cross section data based on a 42 energy group structure derived from ENDF/B-V files. This cross section library was designed to I property capture integral properties of the multigroup data during the ;
I 5
l
_f.
F .
'~
group collapse in order to accurately model important resonance ;
parameters, and to provide the overall accuracy of reactivity predictions necessary for core design. In addition, this library has been developed in l a manner consistent with current Westinghouse methodologies and f accumulated core design experience. The development and benchmarking of the PHOENIX-P library are described in Reference 3.
For gadolinium, the cross-sections are obtained from the Criticality Safety CSRL-V 227 group ENDF/B-V library. Resonance effects are added by the NITAWL-S code using Nordheim treatment.
The 227 groups are subsequently collapsed to the' PHOENIX-P 42 group structure using the XSDRN-PM transport theory cell code.
2.2 LATTICE MODELING IN PHOENIX-P in PHOENIX-P, the fuel, discrete absorbers, and structural components within a single fuel assembly are represented in their exact lattice configuration. Discontinuity factors, pin factors, and homogenized two-group microscopic cross sections are generated as a function of bumup forinput to ANC. Forisotopes and materials represented explicitly in ANC, microscopic cross sections are generated, including xenon, samarium, soluble boron, water density, and bumable absorbers. To obtain constants for rodded assemblies, branch calculations are performed at selected bumups.
A thme region cylindrical cell description for each cell within the lattice is 4 allowed in PHOENIX-P. Principles of material preservation are employed )
to construct three region cell representatiors, since most lattice cells l 1
consist of more than three subregions. The outer region (third region) of I each cell, defined by the fuel pin pitch, has a common composition in all cells in a given lattice configuration. Grids are modeled by smearing the l i grid material unifonnly over this common outer region. Grids are only )
l
.s. j l
^
l g,p smeared in the active fuel region. The sections following describe the various types of cell models.
2.2.1 FUEL CELL B00 DEL The fuel pellet outer radius defines the innermost region of a fuel rod cell. l
- The middle region is defined by the clad outer diameter and incorporates the pellet clad gap. Appropriate number densities are specified for the uranium isotopes and oxygen for fresh fuel. Isotopic information for bumed fuel, including decay chains, is obtained from previous depletion
]
j calculations of fresh fuel. For fuel pellets with integral fuel bumable absorber (IFBA) zirconium diboride coating, the coating material is smeared into the clad region rather than being explicitly installed as a j coating on the surface of the pellet. PHOENIX # corrects for the reactivity I effect of modeling the absorber as smeared into the clad instead of on the
! pellet.
. 2.2.2 DISCRETE ABSORBER MODEL A. BURNABLE ABSORBER RODS I
Turkey Point has used two types of discrete bumable absorber (BA) rods: Wet Annular Bumable Absorbers (WABAs) and Pyrex glass.
The cell representation for the two BA types is significantly different.
The WABA contains moderator material in the central region, while the Pyrex BA is voided in the central region. The surface area of the absorber material must be preserved in addition to the quantity of j 4 material.
Since a fast neutron can pass through the absorber region of a
, WABA, become thermalized in the inner region, and be absorbed, both the inner and outer surfaces of the absorber are important.
Region 1 of the cellis therefore defined as moderator material with I an outer radius equivalent to the BA pellet inner radius. Region 2 is i 7
t
- - - - ~_ -
I .
' ~'
defined as pure pellet material with an outer radius equal to the outer radius of the pellet. The inner WABA cladding, inner pellet-clad gap, outer pellet-clad gap, outer cladding, guide tube, and sleeve materials are all smeared into the moderator region in order to preserve material quantities.
For Pyrex absorbers, the inner gap, inner clad and pellet absorber material are smeared into the first region with a radius equivalent to
~
the pellet outer radius. Region 2 is made up of the absorber outer clad, moderator, guide tube and sleeve volumes, and materials. The small volume of moderator between the outer clad and the guide :
i tube is modeled as if it were outside the guide tube. This is a minor
- approximation, since the zircaloy guide tube material is nearly transparent to neutrons.
For gadolinium, PHOENIX-P uses 42 group microscopic cross sections for the gadolinium isotopes as a function of Gd-155 and 1- Gd-157 depletion along with lattice and other geometry specific
- aspects to produce appropriately weighted two group, homogenized cross-sections for ANC.
B. CONTROL RODS Control rod cells are modeled in a manner similar to Pyrex BA cells, except that the dimensions and material in the pellet region are different. Resonance calculations are performed by PHOENIX-P for the Ag-In-Cd control rod material. For St. Lucie, control rods are modeled as five regions consisting of B,C absorber, clad, moderator, guide tube, and moderator.
3
y- r
'k i C. HAFNIUM . ABSORBERS S
. Hafnium absorber rod cells are modeled in a manner similar to Pyrex BA cells, except that the dimensions and material in the pellet region i l are different. Hafnium rods decrease the power and thereby the fast j
fluence in core locations close to the reactor pressure vessel weld.
This reduction is required for pressurized thermal shock (PTS) considerations.
. 2.2.3 STRUCTURAL CELL MODELS
! Certain cells, known as structural cells, contain neither a strong absorber i or material that is depletable. Examples of these include guide tubes, instrument tubes, water displacer rods, and stainless steel rods. These
! can typically be represented with three regions or less and do not require
> special neutronic considerations. Sleeve volume is preserved by .j l
calculating an effective guide tube thickness that equates to the total sleeve volume.
! l l
) 2.3 BAFFLEREFLECTOR MODELING
' Bame-reflector cross sections are generated by performing a one-i dimensional slab calculation with PHOENIX-P. Such a model is developed by using a series of fuel cells approximating two fuel assemblies, the assemblylbame gap, bame, reflector, core barrel, thermal pad (on the flats),
and moderator. A homogenized set of cross sections for ANC is obtained, i representing the spectrum variations existing between the fuel assemblies, j, bame, and reflector.
- 1 2.4 THREE t"4ENSIONAL NODAL MODEL l Homogenized cross sections, discontinuity factors, and pin factors are generated on a cycle specific basis using PHOENIX-P depletion calculations. These parameters are then used to model the three-dimensional core in ANC. A fuel assembly consists of four radial nodes. ;
{
l a , .
r ,
l in order to obtain an accurate pin power recovery solution, the bumup gradient within each node is represented in ANC. A bumup gradient '
algorithm matches nodal corner and surface average bumups.
Explicit representations of axially heterogeneous features such as axial i
blankets and bumable absorbers are made using the variable axial mesh capability in ANC. Typically, 24 axial mesh intervals produce accurate axial power distributions. To account for spectrum effects induced by variable length bumable absorbers and fuel bumup gradients, axial zoning of the bumup dependent cross sections is employed. Bumable absorber historv rufects are also taken into account by using appropriate sets of fuel 4
cross sections.
The three-dimensional ANC calculational results can be used to predict
{ peaking factors, critical boron concentrations, core power distributions,
' l control rod worth, and reactivity coefficients. This model can also be collapsed to two dimensions for those calculations (e.g., determination of l
the highest worth stuck rod) where a three-dimensional representation is l not required.
2.5 ONE-DIMENSIONAL DIFFUSION THEORY MODEL A three-dimensional ANC model can be collapsed radially to generate a one-dimensional APOLLO model. The cross sections are flux and volume weighted, and a bumup and elevation dependent radial buckling search is j performed to normalize the APOLLO model to ANC. The one-dimensional diffusion theory model is used for calculations where additional detail is
, desired in the axial direction. To this end, the axial mesh is redefined to comprise 40 or more axial intervals. APOLLO can be used to generate integral and differential control rod worth curves, determine control rod insertion limits, and analyze axial power distributions in order to establish Ilmits on axial offset during power operation.
1 10
.P 3.0 PHYSICS MODEL APPLICATIONS The physics methodology discussed in Section 2 was developed in order to j provide reliable analytical predictions in the following four major areas:
- Core power distributions at steady state conditions,
- Axial power distribution ;ontrol limits,
- Core reactivity parameters, and
- Core physics parameters for transient analysis input l Often more than one model may be used to perform a specific analysis. The preferred model depends upon a number of considerations including the degree
$ of accuracy desired and the specific applications.
3.1 CORE POWER DISTR:DUTIONS AT STEADY STATE CONDITIONS The prediction of steady-state core power distributions is fundamental to the design, analysis, and surveillance of nuclear reactor cores. Accurate prediction of core power distributions leads to confidence in developing and optimizing core loading pattems, ensuring compliance with Technical I i
Specification limits, and determining fuel assembly burnups and isotopic l l
inventories.
3.1.1 POWER DISTRIBUTIONS i l
Global coie power distributions are obtained as a function of burnup from )
three-dimensional ANC depletion calculations. Calculations are also performed at selected burnups for various power levels and control rod configurations. Peak rod powers and hot channel factors are generated by pin power reconstruction within ANC using rod-by-rod power distributions from single assembly two-dimensional PHOENIX-P fine mesh spectrum calculations. ,
I I
. is .
s l
3.1.2 POWER PEAKING i i
j Local power peaking is monitored to ensure that the peak pellet power and i
> the total energy content within each coolant channel remain within
. Technical Specification and/or fuel design limits. The factors used to l
measure local power peaking include:
q i l
- the heat flux hot channel factor, Fo, defined as the maximum local I heat flux on the surface of a fuel rod divided by the average fuel rod
)
heat flux, l
- the nuclear enthalpy rise hot channel factor, F , defined as the ratio
! of the integral of linear power along the rod with the highest j intog m_.:' power to the average rod power, and l
1
- the planar radial power peaking factor, Fn(Z), defined as the ratio of }
.i l
- the peak power density to the average power density in the horizontal f
! plane at elevation z.
For steady state conditions, these are obtained from three-dimensional 1 ANC calculations using pin power reconstruction. For maneuvering and translent xenon conditions, a three-dimensional, one-dimensional,
, synthesis technique (see Section 3.2) may be used.
3.1.3 FUEL DEPLETION Three-dimensional fuel depletion calculations are performed with ANC.
Rod-by-rod bumup distributions are obtained from the ANC depletions.
Specific fuel nuclide inventories are obtained from two-dimensional single assembly PHOENIX-P depletion calculations.
12
~'
3.2 AX1AL POWER DISTRIBUTION CONTROL LlHITS The axial power distribution is primarily affected by control rod position, xenon, burnup, and temperature distributions. Axial power distribution control limits are used to ensure that thermal limits are not violated during power level changes, control rod motion, acd the resulting xenon f redistributions. This is accomplished by maintaining the axial flux difference within acceptable boundaries. Axial flux difference, AI, is defined as the difference between the upper and lower excore detector
- signals.
Axial power distribution control limits for Turkey Point are determined using Westinghouse's Relaxed Axial Offset Control (RAOC) calculational procedure (Reference 4). The RAOC calculational procedure begins by defining " provisional" Allimits which are wider than the expected LOCA limits (or, attemately, the RAOC Allimits frorn the previous cycle may be used if it is desired only to verify their acceptability). Xenon transient simulations are performed with the one-dimensional APOLLO code at various burnups and for different power levels, constrained by the provisional Allimits and power dependent rod insertion limits. A library of axial xenon shapes is constructed at each bumup. Next, axial power shapes are generated with APOLLO for all possible combinations of xenon shapes, power levels, and rod insertions. These axial shapes are synthesized with height dependent planar radial power distributions from three dimensional ANC calculations. Imposition of the LOCA F o limits for normal operation then defines the allowable Allimits (or verifies that the previous cycle's limits are acceptable) for the cycle. The axial power shapes corresponding to cases within the Allimits are checked against thermal hydraulic constraints from Loss of Flow Accident simulations and the peak power and DNB limits for accident conditions.
13
3 For normal operations, more restrictive Allimits are developed if either the
[* l Fa limits or thermal hydraulic constraints are exceeded. For accident conditions, analyses are performed to verify that all design limits are met.
j if necessary. trip setpoints may be revised and/or the RAOC Al limits
! tightened. Therefore, the RAOC procedure provides axial power shape
!~ information which is used to verify that all design limits are met. The t
1 RAOC Allimits are placed in the Turkey Point Core Operating Limit Report t i and apply during plant operation. ,
- 3.3 CORE REACTMTY PARAMETERS The core reactivity is affected by changes in the reactor which occur i j
during operation as the result of fuel depletion and abnormal or accident l i-conditions. Reactivity coefficients quantify the rate of reactivity change to be expected in response to changes in power moderator or fuel
! temperatures, and soluble boron concentration. Reactivity defects refer to l the integral of the corresponding reactivity coefficient between two reactor statopoints with all other variables remaining constant. Xenon, samarium, and control rod worth are also typically required to fully define the change in reactivity between two core configurations. In addition, neutron kinetics pammsters are needed to describe the time dependent behavior of the l core.
Quantification of the'se effects are needed: (a) to provide input to safety l analyses, (b) to provide guidance to the reactor operators, and (c) to l' ensure compliance with Technical Specifications. Therefore, the physics l models described in Section 2 are used to calculate reactivity coefficients, i
reactivity worth, and kinetics parameters as a function of core bumup, il L
moderator temperature, and power level.
I I 1 l
. u.
I l w-e,- ,,ww .-
. ,(
l l
L .,
- l" 3.3.1 MODERATOR TEMPERATURE COEFFICIENT The moderator temperature coefficient (MTC) is defined as the change in reactivity per degree change in moderator temperature. The effect of
- concomitant changes in moderator and soluble boron densities are included. The MTC is sensitive to the values of the moderator density, i
moderator temperature, soluble boron concentration, fuel bumup, and the l presence of control rods and/or bumable absorbers which reduce the I required soluble boron concentration and increase the leakage of the core.
l The MTC may be positive or negative depending on the magnitude of change of the individual components of this coefficient.
i
- The MTC is calculated using the ANC core model described in Section 2.4 by varying the inlet temperature around a reference temperature. The moderator temperature coefficient is analyzed for various reactor
! conditions, from hot zero power (HZP) to hot full power (HFP), for various i
{ boron concentrations and control rod positions, and at various cycle
, bumups. The moderator temperature defect is also obtained using data from the ANC core model.
f l 3.3.2 DOPPLER COEFFICIENTS The Doppler temperature coefficient is defined as the change in reactivity per degree change in effective fuel temperature. The effective fuel i temperature accounts for the spatial variation in fuel temperature )
! throughout the core. The Doppler power coefficient represents the corresponding change in reactivity per percent change in reactor power.
These coefficients are primarily a consequence of the Doppler broadening i of U-238 and Pu-240. resonance absorption peaks which increases the effective resonance absorption cross section of the fuel with increasing fuel temperature.
1 I
15 l
)
,g -
~~
The Doppler power coefficient is normally calculated using the'ANC core !
l model by varying the reactor power level about a reference power (which in tum varies the fuel temperature) while holding the product of the power ,
- level and the enthalpy rise constant. The FIGHTH code provides effective I
- fuel temperatures, which account for spatial variations in temperature i within the pellet, as a function of power level and bumup. The Doppler
- coefficient is analyzed at different power levels and for various cycle bumups. Doppler reactivity defects can also be obtained using the ANC
! model by varying the reactor power at various times in life, while holding $
the product of the power level and the enthalpy rise constant.
! i l At hot zero power, the Doppler temperature coefficient may be calculated by subtracting the moderator temperature coefficient from the isothermal
- temperature coefficient (lTC), provided ITC is explicitly calculated (see Section 3.3.4).
"] !
3.3.3 TOTAL POWER COEFFICIENT The total power coefficient is defined as the change in reactivity per percent change in core power level. This coefficient represents the combined effect of moderator temperature and fuel temperature changes I for an associated change in core power level. I
~
The total power coefficient is calculated using the ANC core model by varying the core power level around a reference value while allowing the inlet temperature to change in accordance with the inlet program for the plant. The power coefficient is analyzed at different power levels and at various times in core life. The power defect is also obtained using the .
ANC model by varying the reactor power.
16
} [
- t. t ;
l' 2 - -
3.3A NBON NTURE COEFFICIENT '
)
The isothermal temperature ' coefficient (lTC) is defined as the change in ;
i reactivity per uniform degree change in core temperature. The ITC is the temperature coefficient directly measured during startup physics testing.
The ITC can be calculated by summing the moderator temperature i coefficient and the Doppler temperature coefficient. Altemately, the ITC may be calculated explicitly using the ANC core model by varying both the i moderator temperature and the fuel temperature about a uniform reference 1
l temperature.
I tl '
l
! [ The isothermal temperature defect (ITD) refers to the change in reactivity between hot zero power temperatures and temperatures below hot zero power. ITDs are needed as a function of temperature and bumup for various rod pattems to establish shutdown boron concentration 4
requirements. ITDs are calculated with the ANC model using cross sections generated with PHOENIX-P at specific temperatures between hot
- zero power and 68'F.
h 4
, 3.3.5 BORON REACTIVITY COEFFICIENT The boron reactivity coefficient, also referred to as the differential boron ,
worth, is defined as the change in reactivity per ppm change in the soluble I 4
boron concentration. The inverse of the boron reactivity coefficient is l
referred to as the inverse boron worth, it provides a means of determining 1
the change in soluble boron concentration necessary to compensate for a given reactivity change. The magnitude of the boron reactivity l
- coefficient depends primarily on the soluble boron concentration, the moderator temperature, control rod insertion, and the presence of bumable i absort>ers.
L The boron reactivity coefficient is calculated using the ANC core model by i perturbing the boron concentration in both directions about a reference L
l
- .u.
value and computing the reactivity change. Boron worths are calculated as a function of boron concentration, power level, temperature, burnup, and control rod configuration. i l
3.3.6 XEMON AND SAMARIUM WORTH I The fission products Xe-135 and Sm-149 possess large thermal absorption
, cross sections. Knowledge of the concentrations and reactivity worth of these isotopes as well as the changes which occur in response to plant
. maneuvers is crucial to reactor control. Since Xe-135 is also produced by
- iodine decay, it initially builds up and then decays following a reduction in
' I power or shutdown. Sm-149 is a stable isotope produced by promethium
!i decay. Following a reactor shutdown, its concentration increases. Upon l l restart it gradually returns to its equilibrium value. )
i l
Equilibrium xenon and samarium worth are calculated with the ANC core model at various power levels and core bumups. Changes in their worth and axial fluctuations in isotopic concentrations during transient operation are obtained using the ANC and/or APOLLO models.
3.3.7 CONTROL ROD WORTH Control rod worth refers to the reactivity difference between two control rod configurations. The total control rod worth, trip reactivity shape (i.e.,
the inserted rod worth versus rod position), integral and differential worth ofindividual banks, and worth of individual rod cluster control assemblies (e.g., stuck, ejected, and dropped rods) are determined as required for startup physics testing, plant operations, and input to safety analyses.
Control rod worths are analyzed for all normal and many abnormal control rod configurations as a function of burnup, power level, and moderator temperature. Total rod worth and the integral worth of individual rod banks and rod clusters are calculated using the ANC core model.
-18 e
~
Differential rod worths are obtained with the ANC and/or APC' LO models.
3.3.8 NEUTRON KINETICS PARAMETERS Neutron kinetics parameters, which include delayed neutron fractions, decay constants, and the prompt neutron lifetime, are required as input to
- the plant reactivity computer and to various safety analyses. These parameters are also input to the inhour equation to generate core reactivity as a function of startup rate and period. The kinetics parameters are i evaluated at hot full power and hot zero power conditions for various cycle l bumups and control rod configurations.
l The PHOENIX-P cross section library contains delayed neutron fractions i
and decay constants for fissionable nuclides for each of the six delayed neutron energy groups. The core averaged delayed neutron fractions are i obtained by weighting the delayed neutron fractions for each group by the I
regionwise fraction of fissions in each isotope and the regionwise power and volume weighting in the core. The core average decay constants are calculated in a similar manner. The fraction of fissions in each isotope are obtained from single assembly PHOENIX-P calculations. Regionwise
. power sharings for various core conditions are obtained using the ANC core model. A delayed neutron importance factor (to account for spectrum l differences between delayed and prompt neutrons) is used to calculate an j effective core average delayed neutron fraction.
i The prompt neutron lifetime also depends upon the core composition (fuel
,I
- enrichment, bumup, absorbers, etc.). Single assembly PHOENIX-P calculations provide the neutron lifetime for the fuel in each core region.
The core average value is determined through a power and volume weighting process.
-19
m 1
l
\
3A QORE PHYSICS PARAMETERS FOR. TRANSIENT ANALYSIS INPUT The physics modals described in Section 2 are used to generate key input parameters for various safety analyses. These key safety parameters !
. Include reactivity coefficients, control rod worth, and limiting power l distributions duritig both normal operations and accidental transients. l I
Reference 5 provides a detailed description of how these parameters are calculated for Turkey Point.
P l
4 l
l 1
1 I
i I
1 l
l l
l
. 20 ,
l 1
i
9Mm- r i
4.0 PHYSICS MODEL_ VERIFICATION TURKEY POINT _ UNITS t
Core physics model verification typically includes comparisons of predictions to ,
plant startup and operating data. Turkey Point Units 3 & 4 are currently in their fourteenth and fifteenth cycles of operation, respectively. In this section, predictions made using the physics methodology described in Section 2 are compared to zero power physics test measurements and at power operating data for Turkey Point. For St. Lucie, this data is presented in Section 5.
I 4 As stated in Section 1, the methods employed to generate the predictions l reported in this section are standard licensed and NRC approved methods used l by Westinghouse's Commercial Nuclear Fuel Division. The comparisons reported herein provide additional verification of the predictive capabilities of this methodology; however, their primary purpose is to demonstrate FPL's ability to
! perform design calculations for the Turkey Point Units 3 & 4.
E i Turkey Point Units 3 & 4 are similar in design. Each reactor is a closed cycle
! pressurized light water moderated and cooled system, which uses slightly enriched uranium dioxide fuel. Each unit is currently designed to produce 2200 MWt core power. The reactor core consists of 157 fuel assemblies. Turkey Point Units 3 and 4 core and fuel assembly designs are essentially identical, both utilizing a low leakage core design. Each fuel assembly consists of a 15x15 array 4
of 204 fuel rods, 20 guide thimbles, and one instrument thimble. The Turkey j Point Unit 4 Cycles 12,13, and 14 were selected for core physics model verification, since each of these cyclas has different design attributes which i
provide an opportunity to model different design features.
4.1 CYCLE DESCRIPTIONS Turkey Point Unit 4 Cycle 12 began operation on June 11,1989, and shutdown on November 24,1990 after 406 effective full power days (EFPD),
corresponding to a cycle bumup of 12441 megawatt days per metric ton f
21 t
7 , , - , . . , - - - - , . - - .
1 d
(MWD /MTU). Turkey Point Unit 4 Cycle 12 was fueled with two different fuel designs. The bumed fuel of Regions 9B,11B,12C,13A,13B, and 13C
- are the familiar Low Parasitic Fuel (LOPAR) design. Regions 12A,12B,
- 13D, and 13E and the fresh regions 14A,148,14C, and 14D are of the Westinghouse Optimized Fuel Assembly (OFA) design. The core loading pattom for Cycle 12, including the assembly locations, the number of l Integral Fuel Bumable Absorbers (IFBAs), the number of Wet Annular Bumable Absorbers (WABAs), and the locations of control banks are i shown in Figure 4.1-1. The core also contains part-length hafnium rods.
l These rods decreas e the power and thereby the fast fluence in core
{ locations close to the reactor pressure vessel weld. This reduction is required for pressurized thermal shock (PTS) considerations. There are
- 240 hafnium rods in the core. They are 36 inches long and positioned slightly below the core midplane. Figure 4.1-1 gives the core locations for the hafnium rods. A quarter core representation is used since the core is symmetric.
i Turkey Point Unit 4 Cycle 13 began operation on October 27,1991 and shutdown on April 10,1993 after 441 EFPD, corresponding to a cycle i
bumup of 13433 MWDIMTU. Turkey Point Unit 4 Cycle 13 was also fueled I
with both LOPAR and OFA fuel assembly designs. In Cycle 13, sixteen assemblies from earlier cycles were re-inserted. Special modeling of these re-inserted assemblies was necessary to account for their loss in reactivity due to the excessive time that the re-inserted assemblies resided in the spent fuel pool. The core loading pattom for Cycle 13, including the i assembly locations, the number of IFBAs, the number of WABAs, and the
, locations of control banks are shown in Figure 4.1-2. The Cycle 13 core i
also contains the part-length hafnium rods as in Cycle 12. !
t Turkey Point Unit 4 Cycle 14 began operation on May 26,1993 and was 4
- shutdown.on October 2,1994 after 454 EFPD, corresponding to a cycle ;
i j 22
. Y burnup of 13793 MWD /MTU. Turkey Point Unit 4 Cycle 14 was fueled entirely with assemblies of OFA design, and the fresh fuel of Region 16 introduced axial blankets into Unit 4. Axial blankets consist of a nominal six inches of natural UO, pellets at the top and bottom of the fuel pellet stack to reduce neutron leakage and to improve uranium utilization. The core loading pattern for Cycle 14, including the assembly locations, the number of IFBAs, the number of WABAs, and the locations of control i banks are shown in Figure 4.1-3. The Cycle 14 core also contains the part-length hafnium rods as in Cycles 12 and 13. Fuel batch characteristics for l Cycles 12,13, and 14 are summarized in Table 4.11.
4.2 ZERO POWER PHYSICS TESTS After each refueling at the Turkey Point units, startup physics tests are conducted to verify that the nuclear characteristics of the core are l
consistent with design predictions. While the reactor is maintained at hot zero power (HZP) conditions, the following physics parameters are measured:
i I
- Criticci boron concentrations,
- lsothermal temperature coefficient,
- Control rod worth, and
- Differential boron worth Table 4.2-1 contains the zero power physics test review criteria, which represent the maximum expected deviation between predicted and measured values for each parameter.
The following sections briefly describe the measurement and calculational techniques and summarize the results of the zero power physics tests for Turkey Point Unit 4, Cycles 12,13, and 14. Small changes in core reactivity were measured by feeding the signal from a power range neutron 23
3 -
l g ' detector ~ into a reactivity computer which solves the point kinetics j equation. The computer output was plotted on a strip chart recorder. All
- . predictions were made with the three-dimensional ANC model described
! in Section 2.4.
i
) 4.2.1 CRITICAL BORON CoenWTRATrms c Critical boron concentrations were measured by acid-based titration of reactor coolant samples taken under equilibrium conditions. Samples were taken with all rods essentially out and with the reference bank (see .
- Section 4.2.3) inserted. Critical boron searches wors performed with the 4
~
- , three-dimensional ANC model for these core configurations to obtain the predicted concentrations. The measured and predicted critical boron
- concentrations are compared in Table 4.2-2. All differences are within the t50 ppm review criteria.
i
, 4.2.2 TEMPERATURE COEFFICIENTS lacthermal temperature coefficients (lTCs) were measured by making small .
changes in the reactor coolant system temperature and determining the corresponding change in reactivity with the reactivity computer. ITCs were predicted by uniformly varying the core temperature by t5'F about the HZP temperature in the ANC model. The moderator temperature is varied l directly; Doppler effects on reactivity are determined using fitting i
coefficients obtained from FIGHTH calculations. The measured and .
predicted ITCs and Moderator Temperature Coefficients (MTCs) are
[
'I compared in Table 4.2-3. All differences are well within the review criteria of $2 pcmi'F. The measured Moderator Temperature Coefficient is ,
obtained by subtracting the Doppler Coefficient from the measured ITC.
4.2.3 CONTROL ROD WORTH Control rod worths were measured by the Rod Swap Technique. First, the '
worth of the reference bank (the bank of highest worth) was measured by 9
24
i 4
boron dilution. Stepwise bank insertion was used to maintain criticality )
. and differential worth were obtained from the reactivity computer
~
response. The differential worths were summed to provide the integral
! worth of the reference bank. Then, maintaining the boron concentration l
at a constant value, critical configurations were established with each
- remaining bank fully inserted and the reference bank partially withdrawn. 4 L The integral worth of each inserted bank was determined from the critical ,
position of the reference bank after the exchange by applying analytical I corrections to account for the effect of the inserted bank on the partial j integral worth of the reference bank. This procedure is described in l l Reference 6.
l 4 The ANC model was used to predict the individual control rod bank worth
, as well as to generate the corrections used to infer the measured worth.
4 The measured and predicted worth are compared in Table 4.2-4; all differences are within the review criteria listed in Table 4.2-1. Measured and predicted reference bank integral rod worth shapes are compared in i
l Figures 4.2-1 through 4.2-3.
2 4.2.4 DIFFERENTIAL BORON WORTH l Measured differential boron worths were obtained by dividing the i measured reference bank worth (see Section 4.2.3) by the difference I
{ between the critical boron concentrations measured with all rods out and with the reference bank inserted. The differential boron worth does not i l change significantly over this range of boron concentration. Boron worths i were predicted by varying the boron concentration by *25 ppm about the !
HZP all rods out critical boron concentration in the ANC model. The measured and predicted boron worth are compared in Table 4.2-5. All
! differences are well within the 115% review criteria.
2s . l
- _.~
N pm, f .
' 4.3 POWER OPERATION in support of the Turkey Point Technical Specification requirements, the core power distribution is measured at least once every 31 EFPD using the in core instrumentation system. Neutron flux measurements made by movable in-core fission chambers are combined with analytically determined power to reaction rate ratios using the computer program INCORE (Reference 2) to infer (i.e., " measure"), a three-dimensional power distribution. The power to reaction rate ratios are generated 'with the three-dimensional ANC model using cross sections derived from PHOENIX.
INCORE is a data analysis code written to process information obtained by in-core instrumentation. INCORE synthesizes measured axial flux shapes
, and theoretical elevation dependent X-Y power distributions to obtain a power distribution throughout the core.
In this section, measured data obtained from INCORE is compared to predictions made with the threewfimensional ANC Model. Included are:
Power peaking factors, F o and F,,
- Average assembly radial power distributions,
- Core average axial power distributions, and Axial offset Also, measurea ano pr Gicted bcron fetdown curves are compared. Boron letdown refers to the reduction of the all rods out (ARO), hot full power (HFP) critical boron concentration as a function of core bumup.
4.3.1 BORON LETDOWN CURVES :
Reactor coolant system boron concentrations are measured daily regardless of power level or control rod bank insertion. Critical boron ,
concentrations measured at or very close to hot full power all rods out equilibrium xenon and samarium conditions are compared to the predicted
.n. >
r
-- boron letdown curves for Cycles 12,13, and 14 in Figures 4.3-1 through 4.3.3. The predicted curves were obtained from design depletions with the three-dimensional ANC model. !
Table 4.3-1 compares measured and predicted critical boron .
concentrations at the time of INCORE power distribution measurements.
The measured concentrations were corrected to hot full power all rods out j equilibrium xenon and samarium conditions in accordance with the Turkey i Point units surveillance procedures. The predicted concentrations were obtained by performing critical boron searches with the ANC model at the f specified burnups of the measurements. The mean difference between measured and predicted critical boron concentrations for all three cycles is 9 ppm with a standard deviation of 13 ppm.
4.3.2 POWER PEAKING FACTORS l I
The nuclear enthalpy rise hot channel factor (F ) and the heat flux hot i channel factor (Fo) were measured using the INCORE code, as discussed above. Predicted peaking factors were obtained from three-dimensional i ANC calculations performed for core conditions similar to those at the time i of the measurements. Power peaking factors measured during Cycles 12, !
13, and 14 are compared to predicted values in Figures 4.3-4 through 4.3-9 and in Tables 4.3-2 and 4.3-3. For Fe, the mean difference between the :
measured and predicted values for the three cycles is 2.02% with a standard deviation of 1.27%; for Fo the mean difference is 3.33% with a standard deviation of 1.86%. Regarding the Fo comparisons, it is noted that spacer grid effects are inherent in the measured values but the grids are not explicitly modeled in ANC. The magnitude of this effect can be seen from Figures 4.3-19 through 4.3-27.
\ ^r .
i 4.3,3 ftADIAL POWWER DISTRIBUTIOND i Core power distributions were measuroa with the INCORE code,~ as discussed above. The measured power distributions are typically referred I to as flux maps. INCORE also produces predicted power distributions at the bumup of the flux map by interpolating between power distributions generated using the three-dimensional ANC model at specific burnups
'during a. depletion calculation. Since the core is loaded symmetrically, ANC depletion calculations are performed assuming quarter-core reflective symmetry for Cycles 12 and 13, and rotational symmetry for Cycle 14. The predicted power distributions are expanded to full core for comparison to
' l the measured distributions.
Figures 4.3-10 through 4.3-18 compare measured and predicted assembly relative power distributions at selected bumups for Cycles 12,13, and 14.
All comparisons are for the hot full power all rods out condition since this is the normal mode of operation for the Turkey Point units. The mean absolute difference between measured and predicted assembly relative powers is less than .021 and the standard deviation is less than .023 for these comparisons.
4.3.4 AXIAL POWER DISTRIBUTIONS AND AXIAL OFFSETS Measured core average axial power distributions from each of the flux maps discussed in the previous section are compared to predicted axial distributions in Figures 4.3-19 through 4.3-27. The predicted distributions were obtained from three-dimensional ANC calculations performed forcore conditions similar to those.at the time of the flux maps. Note that since the grid straps are not modeled explicitly in the ANC model, no depressions are seen at the grid locations in the predicted distributions.
This difference coupled with the normalization of both measuroa and
- predicted axial power distributions to unity causes the measured relative power to appear slightly higher between grid locations.
24 1
Axial offset refers to the percent difference between the relative power in ,
I the top half of the core and that in the bottom half of the core divided by the sum of these two relativo powers. Axial offsets measured using the INCORE code are compared to predicted values from ANC calculations for core conditions similar to those at the time of the measurements in Table 4.3-4. The mean difference between measured and predicted values for Cycles 12,13, and 14 is 0.66% with a standard deviation of 1.54%.
4 1
4.4
SUMMARY
- in this section, predictions made using Westinghouse's reload core design methodology are compared to zero power physics test measurements and f
at-power operating data from Turkey Point Unit 4, Cycles 12,13, and 14.
i in all cases, the predictions agree well with the measurements. All startup test predictions are within the review criteria listed in Table 4.2-1.
- Predicted critical boron concentrations at power are within 50 ppm of the measured values, and the predicted power distributions are close to the measured values, as evidenced by Figures 4.3-10 through 4.3-27. The l
- excellent agreement between the predictions and the measurements
- reported here demonstrates FPL's capability to apply the Westinghouse
' licensed methodology to reload core design for Turkey Point Units 3
! and 4.
4 1
h l
. 29
- OH 1^Me sq ,,,,
TABLE 4.1-1 TURKEY POINT UNIT 4 CTCLE 12,13 AND 14 FUEL SPECIFICATION CYCLE BATCg gman OF 23fITIAL BOC ASSEMBLIES ENRICBMENT BURNUP i w/o U-235 MND/NTU 12 9B 1 3.40 16080 11B 8 3.40 29279 12A 8 2.60 22862
, 12B 12 3.45 28354
- 12C 8 3.00 24975 13A 4 3.00 18285 13B 4 3.10 17489 i 13C 4 3.10 17592 13D 24 3.20 18769 13E 32 3.40 15354 14A 28 3.40 0 14B 4 3.40 0 14C 8 3.80 0 14D 12 3.80 0 13 9 8 3.30 27419 9B 1 3.40 29845 11A 8 3.10 26325 13B 4 3.10 28073
' 13C 4 3.10 31944 13D 8 3.20 32654 13E 24 3.40 27752 14A 28 3.40 16549 14B 4 3.40 16232 14C 8 3.80 13179 14D 12 3.80 13952 15A 16 3.60 0 15B 12 3.60 0 15C 16 4.00 0 15D 4 4.00 0 14 13E 8 3.40 32717 14A 25 3.40 31222 \
14B 4 3.40 31768 it.
14C 8 3.80 24027 14D 12 3.80 30454 15A 16 3.60 17883 ISB 12 3.60 17489 15C 16 4.00 15004 15D 4 4.00 15055 16A 16 3.60 0 16B 16 3.60 0 16C 4 3.60 0 16D 8 4.00 0 16E 8 4.00 0 1
.M.
. Mr -
TABLE 4.2-1 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 EZP PHYSICS TEST REVIEW CRITERIA PARA 1 STER REVIEW CRITERIA Critical Boron Concentration 150 ppm Temperature Coefficients: 12. 0 pcm/*F Moderator Temperature Coefficient Is3 thermal Temperature coefficient Control . Rod Bank Worths:
Reference Bank Worth 10%
" Swap" Worths 115% or 100 pcm whichever is greater Differential Boron Worth 115%
31
'I '
TABLE 4.2-2 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 CRITICAL BORON CONCENTRATION COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CRITICAL BORON CONCENTRATION (PPM)
CONFIGURATION ,
MEASURED PREDICTED DIFFERENCE M P (M-P) 12 ARO 1538 1584 -46 ;
12 BANK C in 1399 1428 -29 13 ARO 1554 1560 -6 '
13 BANK A in 1401 1408 -7 14 ARO 1698 1691 7 l 14 BANK SB in 1552 1544 8 :
l i
Acceptance Criteria is ISO ppm i
h 32 1
T
(
- TABLE 4.2-3 i
i TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 ,
MODERATOR AND ISOTHERMAL TEMPERATURE COEFFICIENT COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK MODERATOR TEMPERATURE COEFFICIENT (PCM/*F)
CONFIGURATION MEASURED PREDICTED DIFFERENCE M r (M-P) 12 ARO 0.92 0.58 0.34 13 ARO 0.24 0.06 0.18 14 ARO 0.26 1.13 -0.87 CYCLE BANK ISOTHERMAL TEMPERATURE COEFFICIENT (PCM/ F)
CONFIGURATION MEASURED PREDICTED DIFFERENCE M P (M-P) 12 ARO -0.88 -0.63 -0.25 13 ARO -1.66 -1.65 -0.01 14 ARO -1.44 -0.57 -0.88 Acceptance Criteria is 2 pcm/*F I
33
~
l TABLE 4.2-4 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 CONTROL ROD WORTH COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CONTROL ROD WORTH (PCM)
CONFIGITRATION MEAStTRED PREDICTED DIFFERENCE (%)
M P ( (M-P) /P) *100 12 BANK D 691 718 -3.76 BANK C(1) 1314 1365 -3.74 BANK B 375 380 -1.31 BANK A 1177 1204 -2.24 BANK SB 1180 1202 -1.83 BANK SA 1000 1017 -1.67 TOTAL (2) 5737 5886 -2.53 13 BANK D 641 682 -6.09 BANK C 1022 992 2.97 BANK B 435 457 -4.88 BANK A(1) 1232 1275 -3.41 BANK SB 1183 1233 -4.03 BANK SA 826 836 -1.17 TOTAL (2 ) 5338 5475 -2.51 14 BANK D 636 661 -3.78 BANK C 1093 1172 -6.74 BANK B 435 480 -9.37 I BANK A 1086 1102 -1.45 BANK SB(1) 1173 1195 -1.84 BANK SA 1052 1094 -3.84 TOTAL (2) 5475 5704 -4.01 Acceptance Criteria is 15% or 100 pcm which ever is greater (1) Reference Bank - Acceptance Criteria is 10%
(2) Sum of all measured banks within 7%
.u.
9
1
( .
TABLE 4.2-5 i
TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 i HZP DIFFERENTIAL BORON WORTE COMPARISON BETWEEN MEASUREMENT AND' PREDICTION ; i I
CYCLE BANK DIFFERENTIAL BORON WORTH (PCN/ PPM)
CONFIGURATION l MEASURED PREDICTED DIFFERENCE (%)
M P ( (M-P) /P) *100 ;
12 Average Over Bank C insertion 9.45 8.78 7.63 13 Average Over Bank A insertion 8.05 8.34 -3.47 14 Average Over Bank SB insertion 8.56 8.13 5.29 I
l 1
l
- 35
TABLE 4.3-1 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE CYCLE BtHtNDP CRITICAL BORON CONCENTRATION (PPM)
MND/MTU MEASURED PREDICTED DIFFERENCE M P (M-P) 12 0 1437 1426 11 150 1111 1124 -13 2000 1020 1006 14 2320 986 989 -3 3000 954 950 4 4020 867 882 -15 4940 815 812 3 5890 744 733 11 6975 657 637 20 8000 567 546 21 10000 379 361 18 11184 277 250 27 12000 200 174 26 12441 161 133 28 13 150 1082 1104 -22 1000 1014 1027 -13 2000 938 953 -15 2440 924 926 -2 3224 862 869 -7 4888 750 739 11 6678 610 591 19 8265 473 458 15 8754 436 418 18 '
10608 276 258 18 12316 121 109 12 ,
l 14 150 1213 1212 1 600 1189 1166 23 1000 1135 1142 -7 j 1830 1103 1098 5 i 2521 1056 1043 13 j 3428 991 980 11 !
5000 869 858 11 )
5986 792 775 17 j 8148 601 587 14 i 8995 520 510 10 I 9871 453 428 25 10704 362 349 13 12000 218 226 -8
~
l
-M.
l l
- r. . %
l i
l TABLE 4.3-2 1
TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 l POWER PEAKING FACTOR (F ,) COMPARISON
. BETWEEN MEASUREMENT AND PREDICTION 1
l l
CYCLE CYCLE BURNUP F,(MAZ)
Mwo/MTU MEASURED PREDICTED DIFFERENCE N P ( (M-P) /P) *100 12 150 1.475 1.416 4.17 4945 1.492 1.456 2.47 5860 1.499 1.459 2.74 1 6890 1.498 1.456 2.88 7620 1.493 1.471 1.51 l 8363 1.502 1.481 1.41 l 9082 1.511 1.486 1.70 9458 1.509 1.489 1.35 10323 1.511 1.487 1.60 11121 1.504 1.484- 1.35 11812 1.508 1.479 2.01 13 150 1.557 1.486 4.77 2440 1.438 1.440 -0.14 3224 1.442 1.437 0.35 4888 1.447 1.435 0.84 6678 1.519 1.455 4.39 8265 1.509 1.477 2.16 8754 1.511 1.491 1.34 10608 1.541 1.507 2.26 12316 1.545 1.508 2.45 14 600 1.468 1.420 3.38 ,
1830 1.462 1.421 2.89 2521 1.485 1.427 4.06 3428 1.465 1.427 2.66 5986 1.471 1.440 2.15 6836 1.472 ..452 1.38 7659 1.478 1.460 1.23 8143 1.476 1.464 0.82 8995 1.482 1.464 1.23 9871 1.494 1.461 2.26 10704 1.496 1.462 2.33
. sr .
3 TABLE 4.3-3 TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 POWER PEAKING FACTOR (F,) COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE CYCLE BURNUP Fo (MAZ)
MND/MTU MEASURED PREDICTED DIFFERENCE M P ((M-P)/P)*100 12 150 2.920 1.874 2.39 4945 1.692 1.632 3.67 5860 1.709 1.649 3.63 6890 1.718 1.644 4.50 7620 1.722 1.657 3.92 8363 1.709 1.672 2.21 9082 1.699 1.672 1.61 9458 1.724 1.671 3.17 10323 1.713 1.661 3.13 I 11121 1.692 1.652 2.42 j 11812 1.688 1.644 2.67 13 150 1.773 1.756 0.97 i 2440 1.642 1.634 0.49 3224 1.614 1.621 -0.43
)
4888 1.664 1.596 4.26 6678 1.719 1.617 6.31 8265 1.726 1.650 4.60 8754 1.732 1.669 3.77 10608 1.762 1.681 4.82 12316 1.755 1.671 5.02 14 600 1.815 1.682 7.91 1830 1.745 1.677 4.06 2521 1.825 1.686 8.24 1 3428 1.730 1.672 3.42 !
5986 1.725 1.681 2.62 j 6836 1.742 1.688 3.20 7659 1.741 1.699 2.47 I 8143 1.741 1.702 2.29 8995 1.735 1.701 2.00 i 9871 1.729 1.701 1.65 l 10704 1.732 1.696 2.12 1
l
.n.
Dl TJOBIJR 4.3-4 l
TURKEY POINT UNIT 4 CYCLE 12,13 AND 14 f AXIAL OFFSET COMPARISON BETWEEN MEASUREMENT AND PREDICTION
\
l
. CYCLE CYCLE BURNUP AIIAL OFFSET (%)
l MND/MTU MEASURED PREDICTED DIFFERENCE M P M-P 12 150 -2.46 -2.57 -0.11 4945 -2.55 -1.85 -0.70 5860 -2.99 -1.96 1.03 6890 -3,05 -2.13 -0.92 !
7620 -1.60 -2.28 0.68 l 8363 -1.42 -2.24 0.82 i 9082 -0.87 -2.11 1.24 l 9458 -1.66 -2.06 0.40 l 10323 -3.09 -1.96 -1.13 l 11121 -2.04 -1.93 -0.11 l 11812 -2.23 -1.93 -0.30 l l
13 150 7.00 0.34 6.66 :
2440 2.34 -1.42 3.76 3224 0.93 -1.79 2.72 4888 -0.41 -2.25 1.84 6678 -0.66 -2.45 1.79 8265 -1.51 -2.59 1.08 8754 -1.99 -2.50 0.51 10608 -1.99 -2.23 0.24 '
12316 -1.61 -1.86 0.25 14 600 3.62 1.85 1.77 1830 1.19 0.56 0.63 2521 -1.30 0.17 -1.47 3428 -0.13 -0.60 0.47 5986 -1.36 -1.88 0.52 6836 -1.64 -2.08 0.44 7659 -1.93 -2.20 0.27 8143 -2.08 -2.17 0.09 8995 -2.10 -2.09 -0.01 9871 -1.70 -1.93 0.23 10704 -2.12 -1.82 -0.30 39
l f .
FIGURE 4.1-1 t
TURKEY POINT UNIT 4, CYCLE 12 LOADING PATTERN 52 ASSEMBLY FEED 8 7 6 5 4 3 2. 1 H 9B 13A 13E 13E 128 14A 13B 12B D SB D 20 HF G 13A 14B 13E 13D 14A 13D 14C 12A 4 WABA A 8 WABA SA 20 HF F 13E 13E 128 14A 13D 14A 13E SB 8 WABA C B E 13E 13D 14A 13C 13E 14D 11B A 8 WABA SB D 12B 14A 13D 13E 14D 12C WOEW D 8 WABA C /
C 14A 13D 14A 14D 12C L""
SA w.rtsneenius Ameerter B 13B 14C 13E 118 14A - 3.4 w/o CO IFBA B 14B - 3.4 w/o 116 IFBA 14C - 3.8 w/o No IFBA A 128 12A 14D - 3.8 w/o 60 IFBA 20 HF 20 HF 32 Assemblies @ 3.4 Wt.%
20 Assemblies @ 3.8 W1.%
40 l
1 FIGURE 4.1-2 i
TURKEY POINT l UNIT 4, CYCLE 13 l 1
LOADING PATTERN '
l l 48 ASSEMBLY FEED 8 7 6 5 4- 3 2. 1
~
i H 9B 15B 13E 15A 13B 15A 14A 13C D 8 WABA SB 20 WABA D 8 WABA 20 HF R
G 15B 13E 14D 14A 15A 9 15C 13D
- 8 WABA A 16 WABA SA 20 HF l R F 13E 14D 14B 14A 11A 15B 14C l
SB C 4 WABA B l
l E 15A 14A 14A 14D 14A 15C 13E l 20 WABA A SB 8 WABA R
D 13B 15A 11A 14A 15D 13E MGEM D 16 WABA C R '"'""
C 15A 9 15B 15C 13E L*'*;
4 8 WABA SA 4 WABA 8 WABA w Ms Het*= Anesmer n . - ase r
- B 14A 15C 14C 15A - 3.6 w/o No IFBA 13E B 158 - 3.6 w/o 32 IFBA j 15C - 4.0 w/o No IFBA A 13C 13D 15D - 4.0 w/o 88 IFBA 20 HF 20 HF i
28 Assemblies @ 3.6 Wt.%
20 Assemblies @ 4.0 Wt.%
.c.
i FIGURE 4.1-3 TURKEY POINT
! UNrr 4, CYCLE 14 LOADING PATTERN 52 ASSEMBLY FEED 8 7 6 5. 4 3 2. 1 H 14A 15A 15B 16B 14A 16B 14D 148 D SB eWABA D 4 WABA 20 HF G 15A 16C 15A 14C 16A 158 16D 14A 8 WABA A 16 WABA SA 20 HF F 15B 15A 14A 16A 14D 168 15C SB 16 WABA C 4 WABA B E 163 14C 16A 15A 15C 16E 13E 8 WABA A 16 WABA S9 D 14A 16A 14D 15C 15D 14A "*
D 16 WABA C neesh e C 16B 15B 16B 16E 14A """
4WABA SA 4 WABA HFamNuhhan Ahearter B 14D 16D 15C 13E 16A - 3.6 w/o No IFBA B 16B - 3.6 w/o 32 IFBA 16C - 3.6 w/o 64 IFBA A 14B 14A 16D - 4.0 w/o 16 IFBA 20 HF 20 HF 16E - 4.0 w/o 48 IFBA 36 Assemblies @ 3.6 Wt.%
16 Assemblies @ 4.0 Wt.%
42
. N FIGURE 4.2-1 TCHUGT POINT UNIT 4 CYCLE 12 MEASURED VERSUS PREDICTED i BA3DC C INTEGRAL ROD WORTH 1400 .
i 130d
\
\^
~
,,, mEasomED PREDICTED 1100 \, a 1 i I
^ 1000 \ '
l 900 1 l!! \ .
.: x h 500
\
- 8 a; 700 N e
'\ 1
$ .00 N
- l s x
! s00 A
. x ,
400 .
300
\
200 l
\
100 ^, \,
k 0 \
0 40 80 120 160 200 - 240 ROD POSITION (STEPS WITEDRAWN) 43
h.
a FIGURE 4.2-2 TURIGnf POINT UNIT 4 CYCLE 13 MEASURED VERSUS PREDICTED BANK A INTEGRAL ROD WORTH 1400 1300 N
1200 '
t X ,
, PREDICTED 1100
\,
^
_ 1000 \
l \
900
^
gi \
" \
- h 800 8
g 700 \
l
^
\ '
l g"
\
N
^
~
$300 \ ^ !
x .
400 ,
N l 300
\
ma 100
\ ,
N O
0 40
\
80 120 160 200 - 240 ROD POSITION (STEPS WITEDRAWN) l l
l 44 .
I i
FIGURE 4.2-3 TURKEY POINT THEIT 4 CYCLE 14 l NEASURED VERSUS PREDICTED BANK SB INTEGRAL ROD WORTH !
1400 !
1300
^
N Nx ., PREDICTED 1100
\
'N
^ 1000 :
900 x E4 k
h 800 8
g 700
\j\.
8 k -
o l w 1 b
H 500 k 3\
400
\
300 l \ '
l 200 '
\
100 N -
l 0 \
0 40 80 120 160 200 - 240 ROD POSITION (STEPS WITEDRAWN)
.u.
v-FIGURE 4.3-1 TURKEY POINT UNIT 4 CYCLE 12 BORON LETDOWN COBEPARISON BETWEEN MEASUREBRENT AND PREDICTION 1600 1500 a
1400 1300 E
- ==orcT="
31200 3
m 0 1100 s
1000 E
0 900 A
\
g 800 s ,
~
so 700 600 500 400 N
300 hp 200 .
A 100 0
0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP, IOfD/MTU 46
'N FIGURE 4.3-2 TUREEY POINT IntIT 4 CYCLE 13 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION 1600 1500 1400 1300 k- PREDICTED 3 1200 1100 ,
T 1000 900 1
800 0 700 -
a N 600 500 4
400 l 300 \
N 200 \
\
100 h I
O I O 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNDP, IOfD/3tTU 47
FIGURE 4.3-3 TURKEY POINT UNIT 4 CYCLE 14 BORON LETDOWN COBEPARISON BET 95'EN MEASUREMENT AND PREDICTION 1600 )
1500 1400 N
n 1300 l PREDICTED
$ 1200 b
" 1100 ,
L N -
j I1000 N 900 X
i O \ l E 800 g x 0 700 \\ -
2 \
U 600 -
5 \
5 500 t
\.
400 k
\
300 \
\
l' l 200 100 I l I
0 O 2000 4000 6000 8000 10000 12000 14000 1 CORE AVERAGE BtHUEUP, 30fD/3tTU 4s .
3 FIGURE 4.3-4 TURKET POINT UNIT 4 CYCLE 12 F DELTA B COBEPARISON BETWEEN INCORE AND ANC
. 2.0 1.9 .
1.8 ,
ANC 1.7 1.6 se h
^^ ^ ^
d 1.5 Q
. 6
^
6
) j - -
I #
- p A 1.4 1.3 1.2 1.1 1.0 0 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUP, MWD /MTU 49
L _
FIGURE 4.3-5
- TURKEY POINT UNIT 4 CYCLE 13 F DELTA B COBEPARISON BETNEEN INCORE AND ANC 2.0 1.9 1.8 ANC 1.7 1.6 tc 4 a a N
d 1.5 ^^
t s/
1.4 i
1.3
- 1.2 1.1 1.0 0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP, NWD/MTU
. sc .
N i
FIGURE 4.3-6 TUR13Y POINT UNIT 4 CYCLE 14 F DELTA B COMPARISON BETWEEN INCORE AND ANC 2.0 1.9 y,, INCORE ANC 1.7 1.6
- c h
a g 1.5 ,
, o a ^ ' #
g a A ,
l s f l
1.4 1.3 1.2 1.1
. * # wt !
1.0 0 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUP, MWD /MTU
. si .
3 FIGURE 4.3-7 TURKEY POINT UNIT 4 CYCLE 12 FQ COMPARISON BETWEEN INCORE AND ANC 2.50 INCORE 2.25 _
ANC 2.00 A
N E1.75 N ^ ^ '
^ ^
- N A i
f 1.50 1.25 1
1.00 0 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNUP, NWD/rfU 52
FIGURE 4.3-8 TURIWY FOINT UNIT 4 CYCLE 13 FQ COMPARISON BETWEEN INCORE AND ANC 2.50 2.25 ANC 1
2.00 l l
1 I
E1.75 *
\ 6 A '
\ W A
s f
~
I h . we j 1.50 l
1.25 l i
l 1.00 0 2000 4000 6000 8000 10000 12000 14000 '
CORE AVERAGE BURNUP, NND/MTU 53
FIGURE 4.3-9 TURKEY POINT UNIT 4 CYCLE 14 FQ COBEPARISON BETWEEN INCORE AND ANC 2.50 R
N 2.25 ANC 2.00 -
A 6
E1.75 o A a, a *
- a a 4
- -e, l
1.50 l
l 1.25 1.00 I O 2000 4000 6000 8000 10000 12000 CORE AVERAGE BURNETP, 30fD/3tTU
. s4 I
1
. 8j 1
FIGURE 4.3-10 TURKEY POINT UNIT 4, CYCLE 12 RADIAL POWER Dl51RIBUTION COMPARISON BETWEENINCORE AND ANC U 14 13 12 11 10 9 8 7 6 6 4 3 2 1 l l 1 1 I I a240 n2s3 n240 I I I I I I l l l l l 1 n241 a286 a241 1 1 1 I I I R I i 1 I 0Al% 1m5% -0Als I l I I I l l l 1 a380 a704 1.102 a858 1483 a685 a387 1 1 I I l l l l l 0364 a673 1488 a859 1.088 a673 a364 l l l 1 -P l l l 4A0% 441% 129% 412% 446% 1.78 % 6.32% i l I i l I I aM8 1.101 1.315 1.14 6 1.341 1.117 1280 1.122 a456 I I I l l l a445 157 1290 1.130 1.362 1.130 1290 1.097 aMS l l l N I l a67% a36% 1.94% IA2% -1.54% -1.15% 0.78% 228% 2A7% 1 i l l QMO l.112 1207 1.170 1.313 1.060 1.311 1.175 1231 1.130 0454 l l l 1 QM4 1.122 1.205 1.174 1.337 ID66 1.337 1.174 1.205 1.122 0A44 I l -M l 4 90% 489% 0.17% 4 34% 1.80% 4 56% -1.94% n09% 2.16 % a71% 225% l 1 l a358 1.039 1.204 1.156 1288 1.139 1.207 1.166 1.324 1.181 1.206 1.092 0.367 l )
I a364 1.096 1.204 1.164 1.330 1.147 1207 1.147 1.330 1.164 1.204 1D96 0.364 1 1. l l 1.65% -6.20% a00% 049% 3.16% 4 70% a00% 1.57% -0AS% 1A6% 0.17% .O.36% G82% l l QM8 1.261 1.174 1.306 1.062 1.184 1.209 1.207 1.087 1.318 1.164 1.277 0470 l 1 0.672 1.288 1.172 1.327 1.060 1.177 1.193 1.177 1.060 1.327 1.172 1.288 a672 1 -K 3.57 % -2.10% 0.17% 1.66% Q19% a59% 1.34% 2.55 % 2.55 % 048% 468% 485% 4 30% i 0232 1.094 1.137 1.317 1.156 1.193 1257 1486 1.264 1.196 1.114 1270 1.100 1483 n240 J 0241 1.086 1.127 1.334 1.1M 1.174 1249 1.047 1249 1.174 1.1M 1.334 1.127 1.086 a241 -J i l
-3.73% a74% G89% 127% 1.05% 1.62 % a64% 3.72% 120% 1.87% -2.62% 4.80% 2A0% -028% DAl%
G290 n867 1.359 1475 1226 1224 1.087 1.146 1.081 1.220 1.182 1.037 1.318 G859 G291 0.205 a&58 1.358 1.063 1207 1.191 1.046 1.087 1.046 1.191 1.206 1.063 1.358 0.858 n285 -H 1.75% IIPL On7% 1.13 % 1.57% 2.77% 3.92% 5.43% 3.35% 243% 1.99% -2AS% -2.95% Q12% 2.11 %
0245 1.0 94 1.125 1.326 1.17 l 1.212 1274 1.087 1259 1.191 1.122 1273 1.103 1.099 a246 0241 1.086 1.127 1.334 1.1M 1.174 1249 1.047 1249 1.174 1.1M , 1.334 1.127 1.086 0241 G 1.66% G74% 418% 460% 2.36% 324% 2.00% 3.82% 0.80% 1AS% 1.92% -4.57% -2.13% 120% 2.07%
a686 1.336 1.179 1.312 1.086 1209 1228 1.196 1466 1278 1.141 1259 0.674 l a672 1288 1.172 1.327 1.060 1.177 1.193 1.177 1.050 1.327 1.172 1288 a672 F 2 08 % 3 73% ORI% 1.13% 2AS% 2.72 % 293% 1.61% G57% -3.69% -2.65% -225% 0.30%
G384 1.115 1.222 1.1/2 1.316 1.160 1224 1.14 9 1292 1.148 1.191 1.080 a372 a364 1.096 1.204 1.164 1.330 1.147 1207 1.147 1.330 1.164 1204 1.096 0.364 E 549% 1.73% 1.50% 0.69% -1.05% 1.13% 1Al% 0.17% -2.86% -1.37% -1.08% -1 A6% 220%
E464 1.127 1.206 1.145 1.313 1.073 1.313 1.160 1.1% l.115 a457 a444 1.122 1205 1.174 1.337 1.066 1.337 1.174 1.205 1.122 a444 D 4 50 % ,0AS% a08% 2A7% 1.80% 0.66% 1.80% 1.19% 4 75% 462% 2.93%
G465 1.00 1239 1.102 1.342 1.121 1263 1.059 a441 INCoef a445 1D97 1290 1.130 1.362 1.130 1290 1.097 aMS C 4A9% 464% 3.95% 2A8% -1A7% 4 80% -2 m % 3A6% 4 90%
ANC 0.367 a670 ID72 a857 1.084 a662 0.351 a364 a673 1D87 0.859 1.087 a673 a364 8
% DiFFERFNCE n82% 445% 1.38% 023% 028% 1.63% -3.57%
0238 0285 n241 0241 0286 0241 A Moon Absolute Difference = 0.016 124% -0.35% 0.00%
sianoord Devkstion - a014 BURNUP = 2320 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS ss -
v FIGUltE 43-11 l TURKEY POINT UNIT 4, CYCLE 12 ,
l RADIAL PCWVER Dl51RIBU110N COMPARISON !
BETWEEN INCORE AND ANC i i
9 8 7 6 5 4 3 2 1 O 14 13 12 11 10 I I I G245 a290 0245 I l i l i I I l l 1 I I G244 G290 a244 I I l l I l -R l -1 I I l OAl% E00% G41% i l I l l l l l I , 1 I i R393 a686 1.049 a824 125 a669 a389 I I I I I l a383 0.675 1D45 0.832 1.045 a675 0.383 I 1 l 1 -P I I 1 1 I 2.61% 1.63% R38% 4 96% 4 96% 4 89% 1.57 % i l i a466 1.118 1.294 ID87 1.3G2 lose 1.260 1.107 a462 I I I i i I N
1293 1.093 1.341 1A93 1.293 1.128 G467 l l l 1 l l OA67 1.127 I 1 0 Ass .c30% 0.08% -0.56 s -2.91% -3.11% -2.56% -1.86% -1D7% i l I. a458 1.163 1.133 1.362 la39 1.363 1.133 1.172 1.133 0469 I I l 1.117 1.150 1.381 1D44 1.381 1.150 1.171 1.147 OA67 l l M i 1 R467 1.147 1.171 1A8% 1.38 % 0AS% -1.30% -1AB% a09% 1.22% 0A3% I i -1.93% -2.62% 1.54%
I a376 1.119 1.346 1.111 1.156 1.12 8 1.368 1.131 1.152 1.109 R384 l 1.068 1.194 l 0.384 1.127 1.171 1.142 1.385 1.124 1.161 1.124 1.385 1.142 1.171 1.128 0.384 1 -L l -2D8% -524% 1.96% 2D1% -2.82% -1.16% 0A3% 0.36% -1.23% 4 96% 1.62 5 -1.68% 06 l l a666 1267 1.160 1.384 1.043 1.149 1.173 1.179 ID71 1.360 1.134 1.283 a660 l l a675 1293 1.150 1.383 1.058 1.161 1.159 1.161 1DS9 1.383 1.1M IJe93 0.675 l -K
-1.33% -2.01% OA7% OD7% -IA2% -1.03% 1.04% 1.55% 1.13% -1.66% 1.39% 4 77 % -2.22%
0246 1.062 1.096 1.368 1.118 1.143 1.310 1.111 1.377 1.174 1.122 1.378 1.115 1.069 G247 0243 1.044 1.092 1.380 1.122 1.158 1.344 ID66 1.344 1.158 1.122 1.380 1.092 1.045 R243 J 123% 0.77% a37% 4 87% 436% -1.30% -2.53% 4.22% 2A6% 1.38 % 0.00% 4 14 % 2.11% '2.30% 1.66%
0292 a837 1.345 1460 1.162 1.182 1.119 1.169 1.101 1.186 1.169 1.059 1.370 a850 0293 0290 G833 1.339 1.0 43 1.161 1.159 1.066 1.104 1.066 1.159 1.162 1D43 1.339 a833 0290 H 0 60% 048% 045% 0.67% 0.09% 1.98 % 4.97% 5.89% 328% 2.33 % a60% 1.53 % 2.32% 2.04% 1.03% !
0246 1.061 1.096 1.372 1.117 1.175 1.386 1.107 1.363 1.174 1.116 1.371 1.084 1.062 G244 a243 1.044 1.092 1.380 1.122 1.158 1.344 1.066 1.344 1.158 1.122 1.380 1.092 1.045 G243 G l 123% 1.63% 0.55% -0.58% 445% IA7% 3.12 % 3.85% IAl% 1.38 % 453% '4 66% 4 73% E67% OAl%
G701 1.320 1.151 1.366 1.081 1.197 1.204 1.178 1.082 1.356 1.129 1270 a666 0.675 1293 1.150 1.383 1.069 1.161 1.159 1.161 1.059 1.383 1.150 1293 a675 F 3.85% 2.09% 0.09% 123% 2.08% 3.10% 3.88 % lA6% 2.17% -1.95% 1.83% -1.78% -1.04% 1 0 408 1.149 1.181 1.130 1.370 1.121 1.185 1.125 1.3c8 1.125 1.14 9 1.108 0.386 {
i 0.384 1.128 1.171 1.142 1.385 1.124 1.161 1.124 1.385 1.142 1.171 1.128 a384 E 625% 1.86% 0.85% 1.05% 1.08% 027% 2.07% n09% 123% -1A9% 138% -1.77% 0.52 %
a4al 1.152 1.165 1.127 1.364 1.068 1.390 1.145 1.156 1.125 a462 OA67 1.147 1.171 1.150 1.381 1.044 1.381 1.150 1.171 1.147 G467 D 3.00% 044% 451% -2.00% -123% 2.30% 0.66% 0A3% 128% -1.92% -1D7%
OA82 n.121 1267 1.081 1.355 1.102 1.280 1.100 QA64 INCORE 0A67 1.128 1294 1.093 1.341 1.093 1294 1.128 OA67 C 3.21% 4 62% 2.09% -1.10% 1.04% 0.82% ID8% 2A8% 4 64%
ANC Q.389 a679 1.060 a842 1.068 a668 a380 0.383 a675 1.045 0333 1.045 a675 a383 8
% OtFFERENCE 1.57% 0.59 % OA8% 1.2% 124% 1.04% 4 78 %
03A5 0293 0246 0244 0290 0244 A Moon Absolute Dttforonce - a015 OAl% 1D3% 0.82%
Standord Devk2 tion = E012 BURNUP = 6975 MWD / Mill POWER LEVEL = 100% D BANK AT 228 STEPS l
l l
! )
l 56 -
l 1
l
w o
FIGUltE 4.3-12 TURKEY POINT UNIT 4, CYCLE 12 RADIAL POWER DISTRIBU110N COMPARISON BETWEEN INCORE AND ANC 15 14 13 12 11 10 9 8 7 6 5 4 ~3 2 1 l l 1 1 I i 0267 a325 0267 1 1 I I I I I I I I i 1 0266 a319 0266 1 1 I I I I R I l l 1 0.38 % 1.88% 0.38% _
l i I l l i I I OA10 a704 1.058 0.847 1.038 a687 0.393 I l i I i 1 l l OA10 a697 1.049 a849 1.049 a697 a410 l 1 l l P l l l 0.00% 1.00% a86% -024% -1.05% -1A3% 4.15% 1 1 I I I l a492 1.158 1.301 1.084 1296 1D75 1.263 1.111 OA74 l I I I I l 0A% 1.139 1275 1.075 1.316 1.075 1275 1.139 OA95 I I l N I l 461% 1.67% 2.04% 0.84% -1.52% 0.00% 0.94% -2A6% 424% 1 i i l OA92 1.175 1.170 1.14 0 1.370 1.048 1.369 1.138 1.165 1.135 a475 I I i l OAM 1.158 1.150 1.126 1.377 1.033 1.377 1.126 1.150 1.158 OAM i l M l 4 61% IA7% 1.74 % 1.24 % 0.51% 1AS% -0.58% 1D7% 1.30% -1.99% 4.04% l l 0.397 1.140 1.162 1.108 1.383 1.115 1.147 1.112 1.374 1.107 1.151 1.128 0.398 l l a411 1.139 1.150 1.123 1.385 1.109 1.142 1.108 1.384 1.123 1.150 1.139 OA10 i L l 3Al% 0.09% 1.04% -1.34% 4 14 % 0.54% OAd% 0.36% 0.72 % 1A2% a09% -0.97% -2.93% l l 0.683 1261 1.117 l.366 1.073 1.163 1.159 1.152 1.058 1.347 1.126 1277 0.691 l l a697 1.275 1.126 1.383 1.056 1.153 1.145 1.152 1.056 1.383 1.126 1275 0.697 l K
-2.01% -1.10% 480% -123% 1.61% 0.87% 1.22% 0.00% 0.19% -2.60% 0.00% 0.16% 0.86%
0265 1.038 1.079 1.373 1.120 1.158 1.383 1.019 1.347 1.14 3 1.092 1.361 1.085 1.046 0267 0266 1.049 1.075 1.377 1.108 1.151 1.361 1.071 1.361 1.151 1.107 1.377 1.075 1.049 0266 J
-0.38% 1.05% 0.37% -029% 1.08% 0.61% 1.62 % 0.75% -1.03% -0.70% -1.36% -1.16% 0.93% -029% G38%
G327 0.879 1.324 1.060 1.156 1.156 1.089 1.147 1.098 1.153 1.146 1.043 1.325 0.846 0.324 a319 0.850 1.315 1.033 1.143 1.145 1.071 1.109 1.071 1.145 1.143 1.032 1.315 0.849 a319 H 2.51% 3 41% 068% 2.61% 1.14 % 0.87% 1.68% 3 43% 2.52 % 0.70% 0.26% 1.07% 0.76% -0.35% 1.57%
0273 1.077 1.101 1.378 1.11/ l.168 1.397 1.091 1.390 1.168 1.118 1.362 1.054 1.036 0267 0266 1.049 1.075 1.377 1.107 1.151 1.361 1.071 1.361 1.151 1.107 1.377 1.075 1.049 0266 G 2.63% 2 67% 2 42% 0.07% 0.90% 1A8% 2.65% 1.87% 2.13% 1A8% 0.99% ~-1.0?% -1.95% -124% 0.38 %
0.095 1281 1.129 1.365 1.065 1.167 1.158 1.174 1.075 1.388 1.130 1270 0.687 0 697 1275 1.126 1.383 1.056 1.152 1.145 1.152 1.056 1.383 1.126 1275 0.697 F 0 29% 0 47% 0.27% 1.30% 0.85% 1.30% 1.14 % 1.91% 1.80% a36% 0.36% 4 39% -1A3%
0.392 1.120 1.155 1.105 1.353 1.107 1.160 1.122 1.359 1.104 1.161 1.156 OA09 0411 1.139 1.150 1.123 1.384 1.108 1.141 1.108 1.384 1.123 1.150 1.139 OA10 E 4 62% 1.67% 0 43% 1.60% -2.24% 4 09% 1.67% 126% 1.81% -1.69% 0%% 1A9% -024%
0 475 1.137 1.142 1.114 1.365 1.073 1.361 1.113 1.145 1.146 0.484 G495 1.158 1.150 1.126 1.377 1.032 1.376 1.126 1.150 1.157 OA95 D 4.04% 1.81% 470% 1.07% 4 87% 3.97% 1.09% 1.15% 0A3% 0.95% -2.22%
0471 1.113 1254 1.083 1.342 1.055 1252 1.119 0.475 OA95 1.139 1.274 1.075 1.315 1.075 1274 1.138 OA95 C INCORE 485% -228% 1.57% 0 74 % 2.05% 1.86% 1.73% -1.67% 4 04%
0.391 0.679 1.061 0.853 1.0 44 0.674 0.394 ANC 0410 0.697 1.048 0.849 1.048 0.697 OA10 B
-4 63% 2.58% 124% OA7% 4 38 % -3.30% -3.90%
% DtFFERENCE 0269 0.327 0269 0266 E319 0266 A Meon Absoute Dttference = 0.013 1.13 % 2.51% 1.13%
Stoncord Devioton - a009 BURNUP = 11812 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS 37
, N ,
l l
> NGURE 4.3-13 -
TURKEY PolNT UNIT 4, CYCLE 13 !
RADIAL POWER DIS 1RIBUTION COMPARISON BETWEENINCORE AND ANC J
9 8 7 6 5 4 3 2 1 G W 13 12 11 10 l l l 1 I n251 n27s nasi l I I I i I l
I n244 n272 nada I I I i l I #
i 1- 1 I I I l
, I l l l 2m n37% 2A7% i i I t 04a4 Da23 1.119 n907 1.lui Os06 0431 I I I i l l 1 n416 I l P l l l l 0416 AsM 1.le n928 1.M6 0 s14 I I 1 I i 1.925 1.11% -2.36% -2.26% dals 4 98% 3Als I l I nei t015 1.326 n971 1.252 1.109 M63 I I l I i l IDS 9 1277 '
N n447 1.106 1.232 1D10 1.342 late 1.292 1.106 0447 l l l ,
I I I I I 2.24% 1.545 1.1 5 n50% -1.19% -3.865 -3.10% n27% 3.88% I i i l I 04m laid 1.104 IDl4 1.282 n991 1.a65 Im 1.us Im6 0455 I I
'l -l RAS 1480 1.106 E989 1.288 n998 1288 n9R9 1.105 1480 n448 I l -M l
OA9% 4 93% 0.09% 2.53% 047% 4 70 % -1.79% 1A2% 0.90% 0.56% 1.56% I i l l n409 1D51 1.139 1270 1.205 1.223 1.310 1233 1.197 1275 1.106 1D82 n418 l ,
l Q417 1.107 1.106 1.268 1.187 1214 1.292 12M 1.Hl7 1.288 1.106 1.107 0417 I L l 1 1.925 6D6% 2.90 % 0.98% 1.82% G75 1.39% 1.57% DA4% 1.38% nMI5 -226% 024% i 1 0,799 1.263 ID34 1216 1.236 1.315 ID72 IJ04 1.231 1.203 128 1.257 E804 I i n815 1.293 n992 1.188 1.222 1.300 1D61 1.300 1.222 1.1PS n992 1293 n815 l K i
-1.96% 2.32% 4.23T 2.36% 1.06% 1.15% 2.005 R31% E73 1.265 1.61 % -2.78% -1.38%
n257 1.133 1.031 I ?M l.237 IJ01 ID66 1.316 ID44 1.294 1.228 1.261 0.994 1.117 G250 0244 1.M7 ID11 -1 4 1215 1.298 1D51 1.369 ID61 1298 !2 15 1.290 tall 1.M7 0244 -J 1Al% 0.235 1.33% -3.87% 0A7% 0.31% 1.07% -2295 -1 A8% -242% 2A6% :
5.33% -1.22% 1.98% 0.78 %
0.200 0.932 1.338 ID11 1.303 ID67 1.324 1.000 1.319 1.062 1.266 E966 1.343 n919 Q272 :
0272 E929 1.343 E999 1292 1.048 1.363 1.063 1.363 1.048 1.292 n999 1.343 0.929 n272 -H 120% E85% 1Al% 2.86% 123% -323% 1.34% -2D1% -3.30% 0.00% -1 5 % nmp%
- 2 94% a32% 4 37 % '
n257 1.153 ID26 1.208 1.238 1.309 1.046 1.335 1.048 1291 1.202 1.254 127 1.125 0255 0244 1.147 -IDil 1290 1215 1.298 1D61 1.369 1.051 1297 1215 1.290 1D11 1.M7 Q244 -G s I
5.33% 0.52 % 1A8% 4 16 % 1.89% 0.86% -0A8% -2A8% 029% 0A6% -1.07% ' 2.79% 0ADE -1.92% 4.51%
n833 1292 154 1214 12C 1.322 ID70 1295 1.229 1.198 n989 1.262 0521 G815 1293 0.992 1.188 1.222 1.300 1.061 1.351 1.222 1.188 0.992 1293 OA15 F 221% 0.08% 323% 2.19% 1A7% 149% 1Al% 0.38% a57% 0.85 4 30% -2ADE O.73 <
a433 1.112 1.128 1.284 1.203 1214 1291 1.226 1.207 1.206 1.105 1D85 a434 )
0 417 1.107 1.106 1258 1.187 1214 1.292 1214 1.187 1.258 1.106 1.107 a417 E 3 84% 045% 1.99% 2.07% 1.35% OIDE 4 08% 0.99% 1.68% 223% 0.09% -1.99% 148% !
0472 1.103 1.111 0.977 1251 125 1297 ID18 1.103 Im7 a459 j 0 448 1.080 1.106 0.989 1.288 0.998 1.288 0.989 1.106 1D79 0448 D -
5.36% 2.13% 0.54% 121% 2A7% 0.70% n?0% 2.93% 4 18 % 0.74 % 2A6% l 0 471 1.084 1243 121 1.332 ID20 1272 ID77 0458 l INCORE W 47 1.106 1292 1D10 1.342 1D10 1291 1.106 0A47 C 5:O % 1.99% 3.79 % 089% 4 75 % 0.995 -1A7% -2.62% 2A6% )
ANC OA20 ' n807 1.116 0.921 1.14 9 QAl4 0415 i OA16 E814 1.M6 0.928 1.146 0.814 0416 8 )
% DNSSBNCE 0 96% 4 86% -2.62% -0.75% 026% E00% 024%
0252 n281 0261 0244 0.272 0244 A Hoon Abscute Difference = n015 3.28 % 3.31% 6.97%
8tondord Deviction = n012 BURNUP = 2440 MWD /MTU POWER LEVEL = 99.4% D BANK AT215 STEPS 88 -
i
_ _ . _ - _. __. . . . _ . _ . _.__._-._..__ _ __.._ .__ _ ___.. _ _ _ - _ _ _ _ . _ . _ . - _ _ - . _ . _ . . _ ~ . _ . .
- neues4.s-w
, )
i TURKEY POINT UNIT 4, CYCLE 13 i RADIAL POWER DISTRIBUTION COWARISON IETWEEN NCORE AND ANC l; ,
i j 16 M 13 12 11 4) 9 a .7 6 5 4 3 2 1 !
! ! I I I I I name om name i I 'I I I I I
- 1 I I I i- 1 0280 Q2s0 a2eu I I I I i 1 -R ,
l i i i l i i 1A01. -Imm 1ADI I I I
, I I I I nee 0 Das I.xm E901 1.isl naiv n445 l i I 1 i i I' l 1 QAss Osl3 1.115 0.911 1.116 Gals Q43s i I I I -P i l i I 2.74 % 1As% en3% -1285 1.17% C74L 1A0!b I I I i
i i i i nar ude 1.30s uOs 1.356 la m 1.29s u3s n491 1 I i I I I Q480 1.M6 1J10 1A06 1.352 1A05 1210 1.M6 Q4eo l l l -N .
- I- l 3.64% 0.3e5 43e5 QODI, Q22% OADE -1.37% 4 41% 2290 l I i i i i ns30 1.167 Imps a9si 1.3ss Ims IJ64 0.9se Im7 1.le GAso l I t
l l 0A00 1.156 1.100 Q986 1.368 1A36 1.354 Q906 1.100 1.156 0A10 l l M i 10A2% Q170 0.18 % 0.10 5 OIDE QSSE 0440 02TL -1.180 -IDet 1.281, 1 I E451 1.229 1.17D 1.19ll 1.127 1.178 IA09 1.1EIi 1.131 1.19H IA7d 1.121i 0A5 l l- QA3C 1.M6 1.101 1210 1.M1 1.196 IJ77 1.196 1.M1 1210 1.101 1.M6 0438 l - 1.
l 9A2% 7.24% 6.54% -1 A35 -123% 1.51% 2.32% R78% 0AB% -1.325 -2ASL 133% QA6L l 1 E852 1.396 Q954 1.13 1.141 1.340 1435 1.236 1.136 1.12 2 G976 1.275 E51P l l G813 1.311 - Q908 1.142 1.164 1.237 1D19 1.237 1.164 1.142 R908 1.311 G813 l -K
.' 4.80E M8% 4 405 -1.23% -1.13% 0.24% 1.86% 02% -1.56% -1.79% 121% -2.825 0205
- 0261 1.117 1.G36 1.364 I.191 1.221 ID12 1.33B E998 1211 1.MB 1.334 lalo 1.132 0269 0290 1.116 1.006 1.389 1.197 1.236 1.002 1.360 1202 1236 1.197 1.309 1206 1.116 R280 -J 4 405 0.09% 0.10E 0.37% 0.505 -1.13% 1.005 2.365 0ADE -1.945 -2A2% -2.58% REN, IA3% 7A0%
0286 Q914 1.348 1.030 1.380 1.GIS 1.333 127 1.336 Im2 1.333 Q95' IX E926 0295 02B1 0.912 1.362 126 1.377 1417 1.356 122 1.366 1D17 1.377 126 1.382 R912 Q281 H 1.78% 022% 4 30% a39% 0.22% 1.08% 1.62 % DJ9% -1As% OA95 -3.20% 4.39% 4005 1.54% 4.98%
0261 1.118 121 1.348 1.193 1.225 0.999 1.347 IDl4 1.236 1.171 1.323 0.984 1.115 0261 4
0250 1.116 1.005 1.369 1.197 1.236 IM 1.360 1M 1.236 1.197 1.389 1A06 1.116 Q200 -G
} 440% 0.18 % -0 4G% 0.81% -0.33% -0.57% -0.305 496% 120% 473% -2.17% 22A5% -2.095 0.090 4ADE
! G825 1.3CD 0.990 1.131 1.14 6 1235 129 1.230 1.157 1.135 Q957 1275 QBl:
i 0.813 1.311 Q988 1.142 1.154 1237 ID19 1237 1.154 1.M2 Q988 1.311 0813 F 148% 4 84 % 0.20% 4 96% 469% 0.08% 0.98% 457% 026% 0.35% 0.10 5 -2.62% 0.29%
I DA48 1.143 1.096 1212 1.139 1.187 1.372 1.197 1.142 1.198 Im8 1.123 G440 l 0 438 1.146 1.101 1210 1.141 1.1% 1.377 1.196 1.M1 1.209 1.101 1.146 Q438 E l 228% 026% 0 45% 0.17% 4 18 % 4 75% 4 36 % 0.05% 0.09% 0 91% 1.18% -2D1% Q46%
- OA93 1.14 9 1.098 0.999 1.367 1.043 1.378 Q997 ID97 1.14 7 0484 0 480 1.156 1.100 0.986 1.368 1.026 1.368 Q986 1.100 1.166 0400 D
! 2.71% 452% -0.18% 1.32% 046% 1.76% 1A7% 1.12 % 027% 469% 0.83%
.' Q493 1.132 1275 Q994 1.366 150 1.309 1.116 OAB3
! INCORE R400 1.145 1.310 1.004 1.352 124 IJ10 1.M6 OABO C l 2.71% -1. W% -247% -1.005 0 96% 1.59% 4 08% 2.53% 043%
l ANC 0.439 G80B 1.113 0.905 1.173 OA23 0434 3
0.438 Q813 1.116 a911 1.116 0A13 0438 B j 466%
% DIFFERENCE O23% 4 62 % 0.18 % 520% 123% 3205 j 0260 0297 Q271 0260 0200 0250 A
- u on Atmoiute ometence - a013 4 mot 6mm 840E !
seanoord Deviation. 0.0 14 BURNUP = 6678 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS ;
)
1 80
. . _ - . _ _ _ _ _ . . ~ _ . _ _ _ _ _ _ _ . _ - . _ _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
,N ;
MIDME 4.3-15 ., [
- TURKEY POINT UNIT 4, CYCLE 13 t 4
! SETWEENINCORE AND ANC i
G W 13 12 - 11 10 9 -
8 7 6 5 A 3 2 1
- i i
i i i l i I an6 nais asis l l I I I I i l 1 1 I i 1 naes naos 026s i 1 1 1 I I R
! I I I i 672 % A20s, 3.7 m i l l I I 4
i l I i neu used 1.u7 avni 1.im usu nae 9 I I I I
- I l ! I 0467 as21 1.103 n916 1.10s nati 0467 I i i 1 -P 1WE M 1.27% notm 0890 E125 0A30 l 1 I i I I ;
. I I I nsas Li77 lau: 14 - Lami aw lau 1. iso GAzi l I i
! I I l ns13 1.176 inns 1ADI 1.346 1AD1 1.305 1.177 4 513 l l 1 -N I I 2.la OD90 2L7a nas 0d% -2.000 -2.180 -141% 1.56L I l .
1 I l namu L174 1m n959 1.398 L 380 Lai n95 1.ury 1.170 0421 l l ,
4 I 1 as13 1.las 1m92 n9es 1. ass ime6 1. ass n9es im 1.18s n5W l i M I 1.30% -1.18% 4 885 Q105 0A3% 0.39% 4WL 4 71% -1.19 5 1200 1.36% i
,' 1 0A65 1.145 1.1 11i 1.157 ID95 1.156 1A50 1.lE 1.10L 1.164 lath 1.156 0A69 l
-l' 0467 1.177 1A93 1.176 1.112 1.171 1AM 1.171 1.112 1.176 1A93 1.177 0467 l -L l 043% -2.72% 101% -1.53% -1.835 -1.28% 2.68% Q985 0.995 49s% -2A7% -1.78% GA30 l
! I n513 1.299 ID14 1.114 ID96 1.1% 1R25 1.195 ID95 ID9d n975 L273 aslU l
~
l G821 1.306 0.999 1.112 1.110 1.192 1. 5 12 1.192 1.110 1.112 n999 1.306 n821 1 - IC 4 97% 0.54% 2.53% E18% -1.26% 0.34% 2.30% E255 1.30 % -1A2% -1A2% -2.635 4 37% ,
, - Q276 . 1.12 0 ID17 1A12 1.182 1.180 E992 1.3 33 E953 1.172 L152 1.359 0.975 IR97 0275 l l 0.268 1.103 1.01r 1.389 1.172 1.190 n976 1.336 0.976 1.190 1.172 1.388 1432 1.103 0268 -J
, 2.99% 1.54% 1.505 146% RSSE 0.84% 144% -1.12% 0.72 % 1.51% -1.71% -2.095 -2A9% 4 64% 3.73 %
d Q.319 0.929 1.353 1.056 1443 1.021 1.322 IJIl6 1.318 1AIB 1.376 1ADI l.355 0.910 n3CB
!. E306 E915 1.346 1RS6 1A16 1A00 1.333 0.995 1.333 1AIII 1A15 1D36 1.346 0.916 0.305 -H I 4.59% 1.53% 0.52 % 1.93% 1.98 % 2.105 4 83% 1.11% -1.13% 0.805 2.765 3.38% E67% 0.56 % Q90%
0.286 1.117 1. CDP 1.399 1.1% 1.194 0.953 1.333 n985 1.181 1.M7 1.362 E992 1.109 Q22 O.268 L103 1.002 1.309 1.172 1.191 Q976 1.336 E976 1.190 1.172 1.389 1532 1.103 0268 -G j 6 72 % 1.27% 0 70 % 0.72 % 105% 029% 0.72 % 4 19% 0.92% 0.76 % -2.13% "1.985 -1 ADE E64% 522%
0.845 1.313 0.994 1.105 1. lCD 1.197 1R23 1.185 1.105 ID99 0.975 1.296 na43 l 0.821 1.306 0.999 1.112 1.110 1.192 1.002 1.192 1.110 1.112 0.990 1.306 0221 F !
l 2.92% 0.54 % 0.51% 463% 4 905 0 42% 2.10 5 4 995 0A5% -1.17% -1.52% 4 77 % 2.68%
i Q483 1.186 ID93 1.167 1.096 1.157 IA26 1.181 1.110 1.164 ID79 1.175 0479
! 0 467 1.177 ID93 1.175 1.112 1.171 1A15 1.171 1.112 1.175 ID93 1.177 0467 E i 3 43% G76% ECIls 4 68% 1Ad% -1.205 0.78 % 0.85% 0.18 % -0.94% -1.28% 0.17% 2.57%
2 0.533 1.193 ID79 0.961 1.340 ID75 1A31 0.996 1471 1.182 0.529 O.514 1.188 1.092 0.902 1.388 1436 1.308 0.988 1.092 1.188 0.5W D 3.70% OA2% 1.19 % 2.73 % -346% 3.76% 3.10 5 0.81% -1.92% 4 51% 2.92%
l 0.532 1.168 1.250 E988 1.338 ID16 1.302 1.182 a522
{ #eCOtt 0.513 1.177 1.305 1.002 1.346 1AII2 1.305 1.177 0.5W C l .
3.705 4 76% -1.92% 1405 4 59 % 1A35 023% -2.12% 1.56%
i ' ANC 0476 0.833 IDUS 0.916 1.119 0829 G463 l Q467 E821 1.103 E915 1.103 OA21 Q467 8 F % DIFFERENCE 1.93% 1A6% 0 45% 0.11 % IAS% 0.97% 0.86%
, 0278 E313 02B5
! 0268 0.305 0268 A
! Moon AbschJte DNforence = 0.015 3.73 % 2.62 % 6.34%
! 31onoord Deviation - E023 l BURNUP = 12316 MWD /MTU POWER I.EVEL = 100% D BANK AT 228 STEPS L
i i- . so .
j 1 l
wi 1
EON TURKEY POINT UNTT4, CYCLE 14 RADIAL POWER DISTRIBlmON COMPNHSON BETWEEN INCORE AND ANC U 14 13 12 11 10 9 8 7 6 5 4 3 2 1 I I I I I I 0227 a242 0228 I I I l i I I I I i l I a236 a251 n237 I I I I I I -R l l l 1 -3.81% -3.59% -3.80% I i i I i i a349 a798 1.085 a789 1.085 a797 a351 1 I I I i I I I I l a352 n801 1D80 a795 IDB6 G803 a353 l l l l -P l i I 4a5% 4 37% OA6% 0.75% 0D9% 0.75 % 0.57 % I i I ,
I l l a390 1.050 1.252 1.279 1.373 1.2 14 1.254 ID59 adol l I I I l l a391 1.057 1.259 1.230 1.315 1.234 1.262 1D69 a392 1 l 1 -N l l 0.26 % 466% -0.56% 3.98% 4Al% 1.62 % 463% RODE 2.30% l l l l a390 a947 1242 ID90 1290 IDil 1.283 ID72 1207 a936 adil I I I l a392 a945 1.240 1.081 1270 1.004 1.273 1D84 1.241 n945 0.391 l l -M 1 a795 -a95%
1 4 51 % 021% a16% a83% 1.57 % a70% 1.11% 2.74 % 5<12% l I 0.338 1477 1.256 1.261 1.286 1.178 1297 1.154 1274 1.238 1.154 1D50 a368 I i 0.363 1.069 1.241 1246 1267 1.166 1.310 1.169 1271 1.246 1240 1.057 OL352 l L l 4 25% 1.705 1.13% 120% 1.50% 1.03% 4 995 -1.28% 024% -0.64% -6 94% -0.66% 4.55% l l a773 1.313 1.110 1295 ID26 1218 1257 1219 ID19 1243 ID40 1.181 a750 l l l G803 1262 1DB4 1271 ID18 1225 1.263 1227 ID18 1.266 1.081 1.250 a801 1 -K l 3.74% 4D4% 2AO% 1.89% a79% 4 57 % 0A8% 4 65% a105 122% -3.79% 420% 4.37% 1 0229 1.067 1254 1.329 1.191 1219 "1.282 1.236 1293 1.225 1.168 1.220 1.152 ID45 a234 0.237 1.056 1234 1273 1.169 1228 1285 1233 1.285 1.225 1.166 1.270 1230 1.000 0236 J
-3.38% -1.75% 1.62 % 4A0% 1.88% 473% 023% 024% a62% 0.00% 0.17% -3.94% 4.34% -324% 0.85%
0253 a793 1.313 1.023 1.368 1285 1261 0.937 1238 1.266 1.288 a941 1236 a764 a249 0251 a795 1.315 1.004 1.310 1263 1233 a%1 1233 1.263 1.310 1D04 1.315 a795 0251 -H G80% -025% 4 15 % 1.89% 4A3% 1.74 % 227% 2.71% a41% a16% -1.68% 4 27% 4.01% 3305 4 80%
1 0246 1.132 1256 1287 1.172 1248 1.353 1267 1.321 1270 1.181 1252 1.206 1469 0233 0236 1.080 1230 1270 1.166 1225 1285 1233 1285 1227 1.169 , 1273 1234 1.086 0237 G
_508% 4 81% 2.11 % 1.34% 0.51% 1.8e% 529% 2.76% 2.80% 3.50% 1.03% -1.65% 2.35% -1.57% 1.69%
G828 1294 1.091 1273 1.038 1237 1290 1279 1.086 1.324 1.106 1261 a780 0.801 1259 1.081 1267 1D18 1227 1263 1225 1D18 1271 IDB4 1262 G803 F 3.37% 2.78 % 0.93% 0A7% 1.96% G81% 2.14 % 4A1% 6.68 % 4.17% 1.94% 4 08% -2.86%
G360 1.085 1249 1239 1260 1.170 1.323 1.197 1.304 1278 1239 1.041 a338 a352 1D57 1240 1246 1271 1.169 1.310 1.166 1266 1246 1241 1.059 a353 E 227% 2.65% 0.73 % 456% 0.87% 0.09% 0 99% 2.66% 3.00% 2.57 % 0.16% 1.70% 425%
04Q2 a932 1.183 1.035 1279 1.021 1254 1.083 1.272 E933 a374 l a391 a945 1241 1.084 1273 1.004 1270 1.081 1240 a945 0.392 D 2.81% 4 55% 4.67% 4.52 % OA7% 1.69% -126% 0.19% 2.58 % -127% 4.59%
E386 ID31 1.230 1216 1.374 1221 1.231 ID25 0.377 INCORE G392 1.059 1262 1234 1.314 1.230 1250 1.057 0.391 C 1.53% -2.64% -2.54% -146% 457% 0.73% -2.22% -3.03% -3.58%
ANC a346 G792 1.082 a753 1.032 G777 a340 0353 0.803 1.086 a794 1D80 0.801 a352 8
% DIFFERENCE -1.98% -1.37% 437% -5.16% 444% -3.00% -3 Al%
0235 0241 0226 0237 0251 Q236 -
A Moon Absolute Difference = aQ21 4 84% 3.98% -424%
Stondord Deviation - a019 BURNUP = 600 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS 61
l FIGURE 4.3-17 TURKEY POINT UNIT 4, CYCL.E 14 i RADIAL. POWER DISTRIBUTION COMPARISON ,
! l BETWEENINCORE AND ANC i i l
i 13 12 11 10 9 8 7 6 5 4 3 2 1 15 14 l l l l l 0248 0268 0249 I I I I I l l
a245 G265 a246 l l l l l l R l l l l l 1 I I i 122% 1.13% 1.22% I I I I I
l l l l 0.371 0.797 1.069 0.784 1.067 a792 a370 l l l l !
0.370 l l I l P l l 1 I a370 a790 1.055 a781 1.058 0.791 i l l a27% as9% 1.33% a38% a85% Q13% ODa I i i OA09 1.092 1290 1.1% l 355 1.179 1.282 1.091 OA03 I I I i i l l l i l a403 1.084 1281 1.188 1.336 1.189 1.282 1D84 OA03 l l ! N 1 l uo% Qi4% G70% a67% u2% 024% a00% a65% a00% I i
- 1. I OA04 0.918 1.177 ID69 1.370 1.G21 1.371 1.066 1.168 0.901 G398 l l ,
I DA03 a910 1.170 1.067 1.359 1.011 1.360 1.069 1.170 0.910 OA03 1 1 M i 1
1 025% Q88% 0.60% 0.19% 0.81% 0.99% a81% -028% -0.17% -0.99% -124% I l l a362 1.067 1.172 1.188 1.346 1.158 1.364 1.167 1.369 1.177 1.148 1.066 0.EI l j l 0.370 1.084 1.170 1.188 1.339 1.1 52 1.375 1.154 1.342 1.180 1.170 1.084 a370 l L ;
I -2.16% -1.57% 0.17% 0.0% 0.52 % 0.52% a65% 1.13% 2D1% -0.93% -1.88% -1.66% -2A3% l l a790 1275 1.067 1.356 0.999 1.176 1.209 1.197 1.020 1.339 1.054 1269 0.767 I l 0.791 1282 1.069 1.342 IDIO 1.183 1.204 1.184 1D10 1.339 1.067 1281 0.790 l K 0.13% -0.55% -0.19% 1.04% 1.09% -0.59% 042% 1.10% 0.99% 0.00% 1.22% -0.94% -0.38%
0249 1.068 1.188 1.368 1.153 1.166 1.346 1.199 1.377 1.180 1.155 1.357 1.178 1.067 0248 0246 1.058 1.189 1.360 1.154 1.184 1.363 1.187 1.363 1.183 1.152 1.359 1.187 1.055 0245 J l 1.22% 0%% 0.08% 0.59% 0 09% -1.52% 125% IDl% 1.03% -025% 026% 0.15% -0.76% 1.14 % 1.22% )
a266 0 783 1.336 1.022 1.393 1201 1.183 0.949 1.185 1.182 1.361 0.996 1.363 a786 0268 0265 0.781 1.136 1.011 1.375 1204 1.187 0.937 1.187 1204 1.375 1.011 1.336 0.780 0265 H 1.13% 026% 0.00% 10?% 1.31% 025% 0.34 % 128% -0.17% 1.83% -ID2% -1A8% 2.02% 0.77% 1.13%
0247 1063 1.185 1.367 1.165 1.189 1.373 1.183 1.363 1.178 1.138 1.351 1.180 ID70 0248 0245 1.055 1.188 1.350 1.152 1.183 1.363 1.187 1.363 1.184 1.154 1.360 1.189 1058 0246 -G 082% 076% 025% 0.59% 1.13% 0.51% 0.73 % -0.3d% 0.00% -0.51% -1.39%' "-0.66% -0.76% 1.13 % 0.81% 1 l
0.792 1280 1.063 1.357 Idio 1.182 1.178 1.175 IDl? 1.329 1.043 1269 a788 0 790 1281 1D67 1.339 1.010 1.184 1204 1.183 IDIO 1.342 1.068 1282 a791 F 025% 0 08% 0.37% 1.34% 0.59% -0.17% -2.16% -0.68% 0.69% 0.97% -2.34% 1.01% -0.38%.
0 367 1.064 1.170 1.178 1.345 1.142 1.377 1.153 1.339 1.168 1.153 1.074 0.366 0370 1.084 1.170 1.188 1.342 1.154 1.375 1.152 1.339 1.188 1.170 1.084 0.370 E 70 81% 0 00% 0.00% 0.84 % 022% .' 04% 0.15% 0.09% 0.00% -1.68% -1AS% 0.92 % -1.08%
0404 0.906 1.162 1.058 382 1.034 1.387 1.060 1.158 0.901 0 402 0 403 0.910 1.170 1.068 i.360 1.011 1.359 1.067 1.170 0.910 OA03 D 025% 044% 0 68 % -0 94 % 1.62% 227% 2.06% -0.66% -1.03% -0.99% -025%
0.392 1.052 1243 1.178 1.360 1206 1293 1.089 0A05 INCORE 0403 1.084 1282 1.189 1.336 1.187 1281 1.084 OA03 C 2 73 % -2.95% 3.04% 093% 1.80% 1.60% 0 94 % 046% 0.50%
ANC 0 356 0.767 1.054 a792 ID76 0.805 0.37(.
0.370 0.791 1.058 Q7PO 1.055 Q790 0.370 B
% DIFFERENCE 3.78 % -3.03% 0.38 % 1.54 % 1.99% 1.90% 1.08%
0239 0271 0251 0246 0265 0245 A Moon Absolute D!rference = 0.00? 2.85% 226% 2A5%
Stoncord Devloton = 0.CD7 BURNUP = 6836 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS I
82 -
1 L. nouns 4.us r TURKEY POINT UNIT 4, CYCLE 14 L RADIALPOWER D5mBUTION COMPARISON l BETWEEN INCORE AND ANC .
t 9 8 7 6 5 4 3 2 1 U 14 13 12 11 10 I I I I I i l i i l I i na65 navi 026 I R
ones n261 I I ! I l I i 1 I I l l n260 2.11% 1.92 % I i l I I I I i 1.92%
I I I I I l n391 naco 1.057 n773 IDe0 ns00 n387 I J I I I I -P
- I I I l a3e8 n797 1sso n?9s iss2 n?97 nae 8 nast 426% 1 I I n7m n38% nem 4 13% n76%
i l
i 1
I ac7 1.104 1274 1.159 1.310 1.167 1.275 IDD8 am 1- 1 I 1.330 1.170 1.272 1D90 0434 I I I N i j i l i nest. 1890 1272 1.169 n24% 037% 42d% i l i i som 128% 0.16 % 486% 47s 0.26%
IDe0 1.369 IDis IJ79 Im71 1.170 ns23 9e i I l I naa8 n9se 1.172 l M '
129 1.189 E921 0434 I 1 I naas n921 1.159 128 1.372 1m08 1.373
)
47m 4 225 n69% Oda n19% n96% a22% 424 I I i n9d% 2.80 % 1.12 %
1.154 1.16l 327 a" I 1.143 1.375 1.156 1.391 l l a379 ID73 1.167 1.184 1.361 n308 l -L 1.143 1.369 1.144 1.361 1.179 1.159 1D90 l n308 1.000 1.189 1.179 1.359 1485 2.EN. 0A2% 0.17% 4 28% -1295 i
. 1 -2.32% -1.66% 0495 n42% GIS% E00% 044% ""
ID69 I.369 1221 1.166 1.176 1.176 3.GB IJ71 l_" L" _
l I 0405 1.269 1A68 1272 n?97 l K 1D12 1.166 1.177 1.166 1A12 1.368 I n?97 1272 1D69 1.361 1Da esa n005 asa n89% nom 048% nas 1.58% n96% Asm ness 1.00%
j IJ75 1.173 ID72 n-1.166 1.362 1.163 1.141 R270 1.069 1.165 IJ87 1.146 1.167 1.361 0260 J i
1.373 1.144 1.166 1.364 - 1.162 1.364 1.166 1.143 IJ72 1.169 1250 n261 1462 1.170
< 3A$% 1.62% 4 43% 1.02% 0.17% OD95 0.52% 0.34% ES9% 4 17 % 4 17% 04 4 0.34 2.10% 3A6%
1.366 ID15 1.339 n" R295 1 G209 G793 1.310 1417 - 1.374 1.173 1.161 0.937 1.159 1.166 0.929 1.162 1.177 :1.369 1A08 1.320 n?93 n286 H 4
! 0285 0 794 1.320 1.008 1.369 1.177 1.162 026% 4 93 % 0.00% E69% 1Ad% 1.39% 3.51% l i
1A0% 4 13% 4 76% ne9% 927% 0 34 % 4 09% 0.86%
1.354 1.14 6 1.351 1.161 1.136 1.372 1.163 I" nw l 0258 1.041 1.147 1.369 1.155 1.165 1.364 1.162 1.36d 1.166 1.14d 1.373 1.170 1462 G261' G 4
0260 1.060 1.169 1.372 1.143 1.166 477% 4 86% -1A8% 022% 1.06% 0.CD% 0.00% 1.38% 0.22% 4 43% 470% "047% 0.60% 1.52% 1.53%
1.388 IDl6 1.149 1.134 1.148 ID18 1.352 1.048 l _'m 0.791 f 0.785 1243 1.052
' 1.166 1D12 1.361 1.069 1.272 0.797 F O.797 1272 1D66 1.359 1.012 1.166 1.177 1.51% -2 28% 1.50% 2.13% OAO% -1A6% 3.66% -146% 0.59% 466% -1.96% 49d% 4 75%
l 1.149 1.151 1.355 1.115 1.363 1.138 1.359 1.171 1.151 1.Gl7 xm l G377 1.006 2 a388 1.090 1.159 1.179 1.361 1.14d 1.369 1.143 1.368 1.179 1.169 1.090 d.388 E 4
-2.84% -2.20% 486% -2.37% 0A4% -2.53% 4 44 % 4 44% n07% 468% 4 69% 0.28 %' 0.52%
DA16 0.904 1 135 1.039 1.389 1436 IA09 ID61 1.139 Q919 0430 l 1.069 1.373 1408 1.371 1.068 1.158 E921 0A23 D
{ 0424 0.921 1.159 4 1.89% 1.85% 2.07% -2.81% 1.17% 2.78% 2.77 % 4 66% -1.6d% -022% 1.65%
adl7 1.067 1245 1.160 1.339 1.187 1.280 1075 0A25 1.090 1272 1.170 1.320 1.169 1272 1.090 DA23 C INCORE DA24 1.66% -2.11% -2.12% 4 85% IAd% 1.5 3 0.63 % -1.3 n Q47%
ANC O.377 G781 IDEO 0.800 1.066 0.820 0.380
' 0.797 1D52 n?93 1.050 0.796 0.388 8 O.388
+
% DIF9EWENCE -2.84% -2.01% 0.19% OAB% 1.52% 3.02% -2.06%
0256 0.290 0265 Q261 G286 0260 A
. Moon AbschJte Difference = n009 -1.92% 1.75 % 1.92 %
Siondord Devlotion = 0.52 BURNUP = 10704 MWD /MTU POWER LEVEL = 100% D BANK AT 228 STEPS
. s3
. ~- -_ , . _ . - _ .
t FIGURE 4.3-19 TURERY POINT UNIT.4 CYCLE 12 JLYTAT. POMER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC ]
2.00 l INCORE 1.75 l
JueC l 1.50 i
I 1.25 )
H I
" ^
k a g ^n a a a u _
01.00 . 'J '
n { M J $ 1 I
\a l
@0.75 ,
l t I \ \
'0.50 o l 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTON TOP AZIAL EEIGET, INCHES BUR 3fUP=7620 IntD/3tTU POWER LEVEL =1004 D REEK AT 228 STEPS l
I
.H.
W FIFJRE 4.3-20 TURKEY POE9T UNIT 4 CYCLE 12 AIIAL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC 2.00 INCORE 1.75 l
AMC !
1.50 l
i 1.25 N
H a A a 1
,,f c
^ - -
+
,1.00 ,
3 g" / \
$0.75 $ I
) \s A \
/
0.50
\ "
0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL BEIGHT, INCHES !
BURNUP=9458 MND/3tTU P000'*. LEVELa100% D BAEEK AT 228 STEPS l l
l I
65 I l
N FIGURE 4.3-21 TtrREET POINT UNIT 4 CYCLE 12 AZIAL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC 2.00
~
RE .
1.75 AMC 1.50
" 1.25 N
H A G L '
a u1.00 g
/ ,
s3
/
l@0.75 '
/ -
i l
\"
0.50" t
0.25 1
0.00 '
l 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTOM TOP AXIAL EEIGHT, INCHES BURNUP=11812 MND/MTU POWER N 1004 D BANK AT 228 STEPS i
. ss .
l l
1 FIGURE.4.3-22 TURKEY' POINT UNIT 4 CYCLE 13 AXIAL POWER DICTRIBUTION COMPARISON BETWEEN INCORE AND ANC 2.00
~
MOM 1.75 ANC 1.50 K
" 1.25 k r. - -
. i 6 6 o i a **A "o
^
O1.00 [ "
I
- / ^
X '
l /* !
@0.75 j l
( l
^
l l
0.50 -- = n l
l 0.25 l 1
0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 l 30TTON TOP AXIAL EEIGHT, INCHES stnufDP=2440 MWD /MTU POWER IJMELa100% D RAntK AT 228 STEPS l
67 l
1 l
l
/
FIGURE 4.3-23 TURKEY POINT UNIT 4 CYCLE 13 arrar POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC -
2.00 1.75 AMC 1.50 11.25 N
H
(
co 3 o _
^
- a -
A 6 a ,
-4 A a ,/ , q ,
g1.00 ,
. y H
A l / -
@0.75 / 3 a
0.50 l l
I 0.25 l
0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 noTTOM TOP j AXIAL BEIGHT, INCHES 1 StDUMP=6678 MWD /MTU POWER LEVEL =100% D BMEK AT 228 STEPS '
l 1
. ss .
I 1
s FIGURE 4.3-24' TtnUGar POINT UNIT 4 CYCLE 13
~
AZIAL POWER DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC
. 2.00 1.75 AMC 1.50 j 1.25 k
H
[ 2-A a A ,
A A ;
~
O1.00 ' '
- / ' '
-h
{ s l A E0.75 i o
0.50 0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTON TOP AXIAL BEIGRT, INCHES BURNUP=12316 800D/MTU POWFR LEVEL =100% D RANK AT 225 STEPS 69
r l
)
FIGURE 4.3-25 TERUCEY POINT UNIT 4 CYCLE 14 arvar. POlmR DISTRIBUTION COMPARISON BET 9 MEN INCORE AND ANC 2.00 l 1.75 ANC r
1.50 11.25 ,
^%
."-- ~ '
n l g 6 4 4 d, o
u 1. 00 ,
/ \',
H {
H n b [
\~
@0.75 f g a
0.50 1 \. .
0.2s o
0.00 0 12 24 36 48 60 72 34 96 108 120 132 144
. m o r tcar top AZIAL BEIGHT, INCHES striuttrP=600 MIfD/MTU POIIER LEVEL-100% D BAEK AT 228 8"EPS 70 ,
. P 4
' FIGURE 4.3-26 TUREEY POINT UNIT 4 CYCLE 14 AZIAL POWER-DISTRIBUTION COMPARISON BETWEEN INCORE AND ANC 2.00 1.75 ANC 1.50
- 11.25 N
w . * * * , .
7 n A O1.00 '
/ N l
@0.75 j 3 ,
I \ .
0.50 l \ Eb 0.25" 0.00 O 12 24 36 48 60 72 84 96 108 120 132 144 N TOP l
.., art 17- EEIGHT, INCHES i EtnufUP 6836 IntD/MTU POWER LEVEL =100% D maMK AT 228 STEPS 71 l
f- @
FIGURE 4.3-27 TUREEY POINT UEFIT 4 CYCLE 14 AZIAL POWER DISTRIBUTICBI COBEPARISOBE BETWEEN INCORE AND AMC 2.00 1.75 ANC 1.50 i" 1.25 N
H 2 .
U C *'- - . .,, i , i<
" ^
o is1.00
/
~
. s-
,t .
l _.l \
@0.75
. I \
0.50 l \
I \, .
0.25 0.00 0 12 24 36 48 60 72 84 96 108 120 132 144 BOTTON yop AZIAL REIGHT, DICEES BtHUWUP=10704 anED/MTU POWER Z2 VEL =100% D BAEEK AT 228 STEPS 72
4 d
5.0 PHYSICS MODEL VERIFICATION ST.- LUCIE UNITS
+
1 I Core physics model verification for St. Lucie will include comparisons between measurement and predictions for St. Lucie Unit 1. St. Lucie Unit 1 is i
- currently in its thirteenth cycle'of operation. In this section, predictions made l using the physics methodology described in Section 2 are compared to zero l I
power phpics test measurements and at power operating data. As stated in Section 1, the methods employed to generate the predictions reported in this section an :.tandard licensed methods used by Westinghouse's Commercial !
- Nuclear Fuel Division. The purpose of these comparisons is to demonstrate FPL's competence to use these methods to analyze the core configurations j found at the St. Lucie Units.
i St. Lucie Units 1 & 2 are similar in design. St. Lucie Unit 1 is a Combustion Engineering (CE) reactor with a thermal rating of 2700 MW. The core consists of 217 assemblies of the CE 14x14 design. St. Lucie Unit 2 is also a CE reactor with a thermal rating of 2700 MW. The core for St. Lucie Unit 2 consists of 217 assemblies of the CE 16x16 design. The St. Lucie Unit 1 Cycles 10,11, and 12 were selected for the core physics model vedfication due to the greater complexity in modelling the design features utilized in St.
l Lucie Unit 1. These design features include axial blankets, Gadolinium bumable absorbers, and Vessel Fluence Reduct!on Assemblies (initiated in f.
Cycle 11) which contain uranium tails and Hafnium absorbers placed in the j guide tubes.
l 5.1 CYCLE DESciur iCNS_
St. Lucie Unit 1 Cycle 10 began operation in April 1990 and shutdown in October 1991 after a 477 Effective Full Power Days (EFPD) cycle. Cycle
! 10 consisted of debris resistant fuel (long and cap design) with an active fuel length of 134.06 inches. All fuel utilized axial blankets. The 73 -
D 4
- core loading pattern for Cycle 10,' including a description of the fresh l fuel and the locations of control rods are shown in Figure 5.1-1. A quarter core representation is used since the core is symmetric.
St Lucie Unit 1 Cycle 11 began operation in December 1991 and shutdown in March 1993 after a 442 EFPD cycle. Cycle 11 consisted of
, debris resistant fuel with an active fuel length of 136.7 inches. Vessel Fluence Reduction Assemblies (VFRA) on the core periphery were t t
introduced in Cycle 11. The VFRA assemblies utilized uranium tails and Hafnium absorbers to reduce peripheral power. All fuel with the .
exception of VFRA utilized axial blankets. The core loading pattern for Cycle 11, including a description of the fresh fuel and the locations of control rods are shown in Figure 5.1-2. ,
1 St. Lucie Unit 1 Cycle 12 began operation in June 1993 and shutdown in October 1994 after a 463 EFPD cycle. Cycle 12 consisted of debris resistant fuel with an active fuel length of 136.7 inches. All fuel utilized axial blankets with the exception of the VFRA . The core loading pattern for Cycle 12, including a description of the fresh fuel and the locations :
of control rods are shown in Figure 5.1-3.
I l 5.2 ZERO POWER PHYSICS TESTS l After each refueling at the St. Lucie Units, startup physics tests are conducted to verify that the nuclear characteristics of the core are consistent with design predictions. While the reactor is maintained at hot zero power (HZP) conditions, the following physics parameters are measured;
- Critical Boron Concentrations,
- Moderator Temperature Coefficient,
- Control Rod Worth, and
- Differential boron worth 74
?
4 5,2.1 CRmCAL BnRON CL* CENTRATION i Table 5.2-1 provides the comparisons between HZP critical boron concentrations measurements and predictions for Cycles 10,11, and 12.
j - The values represent all rods out (ARO) and reference bank in conditions. As shown, excellent agreement is demonstrated for each case with all differences well within the 150 ppm review criteria.
4 5.2.2 MODERATOR - ---MTURE COEFFICIENT Table 5.2-2 provides the comparisons between HZP Moderator l Temperature Coefficient measurements and predictions for Cycles 10, 11, and 12. Again, excellent agreement is demonstrated with all
- differences being well within the review criteria of 2 pcmi'F.
j 5.2.3 CONTROL ROD WORTH Table 5.2-3 provides the Control Rod Worth comparisons between i
measurement and prediction for Cycles 10,11, and 12. In all cases, the agreement is within criteria with exceptional agreement being achieved 5 for Cycles 11 and 12. Figures 5.2-1, 5.2-2 and 5.2-3 show the integral l rod worth comparisons for the Reference Bank. The predicted rod worth and integral worth were calculated at the exact conditions which 4
)
l were present during the measurement. Excellent agreement is observed
- between measured and predicted integral worth.
5.2.4 DIFFERENTIAL BORON WORTH Table 5.2-4 provides the Differential boron worth comparisons between measurement and predictions for Cycles 10,11, and 12. Both the measured and predicted values are obtained using the worth of the Reference Bank in pcm divided by the change in boron concentration from ARO to Reference Bank inserted. All differences are well within the expected performance.
75
5.3 POWER OPERATION 5.3.1 BORON LETDOWN CURVES Reactor coolant system boron concentrations are measured daily at the plant. Critical boron concentrations measured at or very close to hot full power all rods out equilibrium xenon and samarium conditions are compared to the predicted boron letdown curves for Cycles 10,11, and 12 in Figures 5.3-1, 5.3-2 and 5.3-3. The predicted curves were obtained from design depletions with the three-dimensional ANC model. Table 5.3-1 shows the difference in ppm between measurement and ANC at various cycle exposures. The mean difference between measured and j predicted critical boron concentration for all three cycles is 3 ppm with a standard deviation of 15 ppm.
! 5.3.2 AXIAL POWER DISTRIBUTIONS Measured core average axial power distributions from Beginning-of- !
! Cycle (BOC), Middle-of-Cycle (MOC) and End-of-Cycle (EOC) obtained i with the incore monitoring code INPAX (Reference 13) using incore detector " snapshots" were compared to predicted axial distributions in Figures 5.3-4 through 5.3-12. The predicted distributions were obtained
- from three-dimensional ANC calculations performed for core conditions I i similar to those at the time of the " snapshots". Overall, the !
! comparisons show excellent agreement between measured and l
j predicted axial power distributions.
. 5.4
SUMMARY
in this section, predictions made using Westinghouse's reload core design methodology are compared to zero power physics test measurements and at power operating data from St. Lucie Unit 1, Cycles 10,11, and 12. In all cases, the predictions agree very well with the measurements. The excellent agreement between the predictions and
-76 k
l
t i
'~
the measurements reported here demonstrates FPL's capability to apply the Westinghouse licensed methodology to perform reload core design i
for the St. Lucie Units.
1 i
J 6
+
I f l
l I
l l
l I
.n..
p t
TABLE 5.2-1 i
ST. LUCIE UNIT 1 CYCLE 10,11 AND 12 '
HZP CRITICAL BORON CONCENTRATION COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CRITICAL BORON CONCENTRATION (PPM)
I CONFIGURATION MEASURED PREDICTED DIFFERENCE I M P (M-P) 10 ARO 1598 1609 -11 10 BANK A in 1477 1496 -19 11 ARO 1393 1396 -3
. 11 BANK A in 1279 1281 -2 12 ARO 1419 1427 -8 12 BANK A in 1303 1307 -4 J
4 Acceptance Criteria is 150 ppm i
t 73
~ ~ ~ ' ' ' " - ~ m.._w a u- - .A e.a .- - - - - - - - - . - - - - - , . - - - - - - - - - - -
TABLE 5.2-2 i
l ST. LUCIE UNIT 1 CYCLE 10,11 AND 12 HZP MODERATOR TEMPERATURE COEFFICIENT COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK MODERATOR TEMPERATURE COEFFICIENT (PCM/'F)
CONFIGURATION '
MEASURED PREDICTED DIFFERENCE N P (M-P) 10 ARO 4.41 5.70 -1.29 11 ARO 2.56 2.57 -0.01 12 ARO 1.54 2.18 -0.64 Acceptance Criteria is 2 pcm/*F f
f
/
J k
4 73
E l-TABLE 5.2-3 ST. LUCIE UNIT 1 CYCLE 10,11 AND 12
' CONTROL ROD WORTE COMPARISON BETWEEN MEASUREMENT AND PREDICTION CYCLE BANK CONTROL ROD WORTH (PCM)
CONFIGURATION MEASURED PREDICTED DIFFERENCE (%)
M P ((M-P)/P)*100 10 BANK 7 516 480 7.50 BANK 6 367 404 -9.16 BANK 5 374 430 -13.02 BANK 4 584 631 -7.45 BANK 3 368 420 -12.38 BANK 2 789 848 -6.96 BANK 1 746 822 -9.25 BANK B 543 584 -7.02 BANK A(1) 1015 1014 0.10 TOTAL (2) 5302 5635 -5.91 11 BANK 7 590 522 13.03 BANK 6 & B 715 774 -7.62 BANK 5 & 3 850 882 -3.63 BANK 4 738 69.9 5.58 BANK 2 791 795 -0.50 BANK 1 808 806 0.25 BANK A(1) 1136 1106 2.71 TOTAL (2) 5628 5583 0.81 12 BANK 7 654 573 14.14 BANK 6 & 3 929 915 1.53 BANK 5 & B 509 550 -7.45 BANK 4 824 809 1.85 BANK 2 699 740 -5.54 BANK 1 759 771 -1.56 BANK A(1) 1099 1136 -3.26 TOTAL (2) 5473 5495 -0.40 Acceptance Criteria is t15% or 100 pcm which ever is greater (1) Reference Bank - Acceptance Criteria is t10%
(2) Sum of all measured banks within t10%
00
j l
TABLE 5.2-4 i ST. LUCIE UNIT 1 CYCLE 10,11 AND 12 HZP DIFFERENTIAL BORON WORTE COMPARISON BETWEEN MEASUREMENT AND PREDICTION 5
CONFIGURATION
~
- MEASURED PREDICTED DIFFERENCE (%)
M P ((M-P)/P)*100 1
{. 10 Average Over Bank A insertion 8.39 8.97 -6.50 i 11 Average Over l Bank A insertion 9.96 9.62 3.53 12 Average Over Bank A insertion 9.47 9.47 0.00 t
p G
f a
J I
. s1 t
p . .
TABLE 5.3-1 ST. LUCIE UNIT 1 CYCLE 10,11.AND 12 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION l i
CYCLE' CYCLE BtTRNUP CRITICAL BORON CONCENTRATION (PPM) mwd /MTU MEASURED PREDICTED DIFFERENCE M P (M-P) 10 139 1160 1178 -18 278 1130 1162 -32 696 1090 1117 -27 1392 1050 1074 -24 '
2784 990 987 3 4176 915 910 5 5568 845 840 5 6960 780 772 8 8352 720 709 11 9744 660 633 27 ,
11136 560 545 15 12528 440 438 2 13920 320 314 6 15312 190 187 3 15947 129 129 0 11 136 952 968 -12 278 939 953 -14 679 903 909 -6 1359 857 862 -3 2718 787 768 19 4078 687 681 6 5437 611 599 12 6796 528 517 11 8155 455 440 15 19514 372 359 13 10874 291 276 15 12233 195 176 19 13592 82 63 19 14404 13 -7 20 Acceptance Criteria is 150 ppm
. s2
, TABLE 5.3-1 (CONTINUED)
ST. LUCIE UNIT 1 CYCLE 10,11 AND 12 BORON LETDOWN COMPARISON BETWEEN MEASUREMENT AND PREDICTION 4
CYCLE CYCLE BURNUP CRITICAL BORON CONCENTRATION (PPM) ,
mwd /MTU l MEASURED PREDICTED DIFFERENCE M P (M-P) !
12 132 961 991 -30 265 940 976 .-36 661 920 934 -14 1324 892 892 0 2648 805 807 2 3972 743 729 14 5296 672 658 14 i i 6614 596 589 7 1 7944 537 525 12 9268 443 453 10 10592 390 369 21 :
11916 287 273 14 l 13240 175 164 11
- 13723 131 122 9 I
l Acceptance Criteria is 50 ppm i
l 33
PIGURE 5.1-1 ST LUCIE UNIT 1 CYCLE 10 -
LOADING PATTERN J2 L3 M4 L4 M4 L5 M3 L4 7 _5 7 J1 L3 M3 L3 M4 L3 M4 L2 M1 s 1 3 H1 M4 L3 L2 L3 M4 L5 M3 L4 .
5 B 2 4 L4 M4 L3 M4 L4 M4 L1 M1 6 A M4 L3 M4 L4 M5 L2 M2 K1 2 A L5 M4 L5- M4 L2 M2 K1 1 A [
M3 L2 M3 L1 M2 K1 5 b b
! L4 M1 L4 M1 K1 LEGEND 3
J1 H1 BATCH ID CEA GROUPO FRESH FUEL INVENTORY:
M1 - 4.0 W/O U-235 l
M2 - 4.0 W/O U-235,4 RODS @ 4 W/O GD203 M3 - 4.0 W/O U-235,12 RODS @ 6 W/O GD203 M4 - 4.0 W/O U-235,12 RODS @ 8 W/O GD203 l M4 - 4.0 W/O U-235,12 RODS @ 6 W/O GD203 & 4 RODS @ 8 W/O GD203 l -u.
i .
7
FIGURE Sai-2 l ST LUCIE UNIT 1 CYCLE 11 )
LOADING PATTERN L2 M4 P5 M3 M3 M4 P3 M5 l 7 1 7 L4 i
M4 M4 M3 M2 M4 P3 M4 P1 1 P6 P5 M3 P4 L1 P3 L2 P3 M1 l 5 B 2 4 j M3 M2 L1 M2 M4 P3 M1 P1 l6 A M3 M4 P3 M4 P3 M4 P2 K1 2_ [
l M4 P3 L2 P3 M4 P2 L2 l 1
A _7 ;
P3 M4 P3 M1 P2 L2 l 5 5 M5 P1 M1 P1 K1 l W LEGEND L4 P6 BATCHID can enourm FRESH FUEL INVENTORY:
P1 - 3.75 W/O U-235 P2 - 3.75 W/O U-235,4 RODS @ 4 W/O GD203 P3 - 3.75 W/O U-235,12 RODS @ 8 W/O GD203 P4 - 3.75 W/O U 235,16 RODS @ 6 W/O GD203 PS - 3.75 W/O U-235,12 RODS @ 8 W/O GD203 & 4 RODS @ 4 W/O GD203 P6 - 0.30 W/O U-235,4 HAFNIUM FLUX REDUCTION INSERTS '
56 4
MGURE 5.1-3 ST LUCIE UNIT 1 CYCLE 12 - .
LOADING PATTERN M4 P3 P3 M2 R3 P5 R3 P2 ;
7 5_ 7 M2 P3 R5 P2 R3 M3 R4 M1 R1 B 1 3 P6 i
P3 P2 P4 P3 R4 P3 R4 P1 5 B 2 4 M2 R3 P3 M5 P3 R5 P1 R1 i l _
6 A
, I R3 M3 R4 P3 R3 P3 R2 M1 2 -
A P5 R4 P3 R5 P3 R2 M4 '
1 A 7 R3 M1 R4 P1 R2 M4 R 4 A P2 R1 P1 R1 M1 !
- 3 LEGEND M2 P6 BATCH ID i cEA I onoue m j
i FRESH FUEL INVENTORY:
R1 - 3.90 W/O U-235 i
R2 - 3.90 W/O U 235,4 RODS O 4 W/O GD203 R3 - 3.90 W/O U-235,12 RODS O 6 W/O GD203 R4 - 3.90 W/O U-235,12 RODS O 8 W/O GD203 i R5 - 3.90 W/O U-235,16 RODS O 8 W/O GD203 l 1
I
.___________________________I
i 9
FIGURE 5.2-1 i ST LUCIE UNIT 1 CYCLE 10 4 MEASURED VERSUS PREDICTED REFERENCE BANK INTEGRAL ROD WORTH 1100 1000'
. h A
=
,00 \ -
\. -DIc-o a 800 \
700 \
8 600 %
^
0500 e t M
" 400
\.
300 \ .
s
\
200
\.
100 \
0 % -
0 40 30 120 ROD POSITION (INCHES WITHDRAWN) 37
w, + , a -
- - , , A a
, P -
0 PIGURE 5.2-2 ST LUCIE UNIT 1 CYCLE 11 MEASURED VERSUS PREDICTED '
REFERENCE BANK INTEGRAL ROD WORTE 1200 t ts 1100
\. u 1000 '
\ -
- ,00 \ =====
- s' Pazozerzo g \
00
\
R g 700 8 \
600 s
g 500 g \ ,
400 i
I N 300
\ I 200 \ l x
100 K
a 0 40 80 K
120 ROD POSITION (INCHES WITEDRAWN) ,
l
f ,
i 1
, 1 1
l FIGURE 5.2-3 i ST LUCIE UNIT 1 CYCLE 12 l MEASURED VERSUS PREDICTED j REFERENCE BABDC INTEGRAL ROD WORTH j 1200 l
I 1100 .
N- '
1000 \
\a .
l- -
1 900 l PIMDICTED .
' E o 800 ^\
g
= 700 \.\
i l
8 g 600 N
^
\
0 500 *\
g *\
H 1 400 ^ '
\
, s 4 JL 300
- \ i 200
\
100
- \
0 0 40 80 120 ROD POSITION (INCHES WITEDRAWN)
FIGURE 5.3-1 BT LUCIE UNIT 1 CTCLE 10 BORON LETDOWN CCMEPARISON BETWEEN MEASUREMENT AMD PREDICTION 1400 1300 1200
- A 1100 *\,
g1000 z H \
> N*
^
900 MEASURED s
g PREDICTED W 800 A 5
o Nk 1
g 700 \
" \a 600 I
\ -
l l N h a l g
500 i l l l
l
\ l l N 400 l \l l l l i
\;
300 h I
N 200 I
l \
l N 100 l I 1 i l l l ,
I O I l I l
0 2000 _ 4000 6000 8000 CORE AVERAGE BURNUP, atWD/3tTU 10000 12000 14000 16000 90 I
y, .
FIGURE 5.3-2 ST LUCIE UNIT 1 CYCLE 11 BORON LETDO90T COMPARISON BET 9 FEEN MEASUREMENT AND PREDICTION 1200 1100 1000
- A l 900 m g PREDICTED
\
E 800 I '
5 700 N 0 %
l E
a 500
! l N'
$ i M 400
$ I l l N N l l \
300 l Ms 200 i I X
! I N(
100 0 \a 0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP, MWD /MTU 91
FIGURE 5.3-3 ST LUCIE UNIT 1 CYCLE 12 BORON LETDOWN COB 8PARISON BETWEEN NEASUREMENT AND PREDICTION 1200 1100 1000 '
900 M PREDIC"ND 8
800 \ 3 4
\
V 700 o Y u N\
g 600
- x 0 l l N- -
500 l \
a b \
C l l 'N !
y 400 U l \,
300 I I Ns 200
\
100 I s 0 i l 0 2000 4000 6000 8000 10000 12000 14000 CORE AVERAGE BURNUP, MWD /3tTU 82
a FIGURE 5.3-4 ST LUCIE UNIT 1 . CYCLE 10 Arran POgga DIgTRIBUTION CONPARISON BET 9EEN INPAX AND ANC
. 2.00 1.75 N AI ANC 1.50
\
l 1.25 - l Q
M / ( d, L ^
- 1h
' N
% V a
as1.00 V' '\ \
i s
( l J
@0.75 ,
/"
'l b ,
0.50 I
, 1 0.25 i i \
l 0.00 0 12 24 36 48 60 72 84 96 30TT006 108 120 132 144 1 TOP AZIAL BEIGHT, INCEES I BURNUP=372 MND/MTU PolGR LEVEL =100%
4
- 93 I
FIGURE 5.3-5 ~
gT LUCIE UNIT 1 crCLE 10 arTAL POWER DISTRIBUTION COBEPARISON BETWEEN INPAX AND ANC 2.00
~ 1 1.75 INPAX -
ANC 1.50 1.25
, yr s a a u o = - r w
4.
01.00 \.
n iO0.75 j
[ '\ ;
\
0.50 f 7
g 1
0.25 0.00
{
0 12 24 36 48 60 72 96 BOTToet 84 108 120 132 144 TOP AXIAL REIGHT, INCHES BURNUP=6,904 MWD /MTU POWER LEVEL =100%
l l
. e4
FIGURE 5.3-6 -
ST LUCIE UNIT 1 CYCLE 10 AXIAL POWER DISTRIBUTION COMPARISON BETWEEN INPAX AND ANC 2.00 1.75 -- NM ANC 1.50 i
" 1.25
/ k n es \
u r x % '-
a 1.00 g.:
, . , i. s <
lE0.75 f j
l f I 0.50 f
0.25 i
0.00 l I 0 12 24 36 48 60 BOTTOM 72 84 96 108 120 132 144 TOP AZIAL EEIGHT, INCHES BURNUP=15,718 NWD/MTU POWER LEVEL =100%
. es .
'f' FIGURE 5.3-7 '
ST LUCIE UNIT 1 CYCLE 11 H m POWER DISTRIBUTION COBEPARISON BETWEEN INPAX AND ANC 3.00 '1
't 1.75 NE '
ANC 1.50 .
I 1.25 g5 ' '
- A h
1
% V r
$1.00 ^'
Y .
i t 7 \
i/ (
@0.75
/ -
\
Y $
0.50 I 0.25
/ k O.00 -
0 12 24 36 48 60 72 84 96 108 120 132 144 BOTT006 TOP AEIAL BEIGHT, INCHES BURNUP=185 antD/MTU POWER LEVEL =100%
. os .
9 FIGURE 5.3-8 ST LUCIE UNIT 1 CYCLE 11 m m t. POWER DISTRIBUTION COMPARISON BETNEEN INPAX AND ANC 2.00 1.75 NAI .t ANC 1.50 11925 N
M g rs a m .s a m , _
4 f '%
M O1.00 -
l \
b O0.75 7 1 0.50 I
O.25 I
\
7 0.00 0 12 24 3E 48 60 72 84 96 108 120 132 144 moTToet Top AXIAL HEIGHT, INCHES BURNUP=6,721 30fD/MTU POWER LEVEL =100%
97
I r
4 FIGURE 5.3-9 ST LUCIE UNIT 1 CYCLE 11 AZIAL POIGR DISTRIBUTION COMPARISON BETWEEN INPAZ AND ANC 2.00 1.75 INPAX e
ANC
. 1.50 1
1.25 H WA a
- a, 4, a a a a l g1.00 t ._
j A 5
1
\'
@0.75 !
l j 0.50 0.25 l \ l
\ !
l ;
0.00
.0 12 24 36 48 -60 72 84 BOTTOM 96 108 120 132 144 TOP 1rm EEIGHT, INCEES BURNUP=12,188 MWD /MTU POWER LEVEL =100%
. se .
1
FIGURE 5.3-10 1 ST LUCIE UNIT 1 CYCLE 12 i AZIAL POIMR DISTRIBUTION COMPARISON BET 9 MEN IMPAZ AND ANC 2.00 i 1.75 NAI E l l
1.50 l
11.25 !
(
f rr * "
u < ,
O1.00 [
/ T" r
@0.75
/ i 0.50 I 7
)
0.25 I
4 a
0.00 0 12 24 36 48 60 72 84 96 BOTT006 108 120 132 144 TOP AZIAL BEIGHT, INCHES BURNUP=625 3GtD/MTU POWER LEVEL =100%
l
.N.
1
. FIGURE 5.3-11 '
ST LUCIE UNIT 1 CYCLE 12 17T17 POlmR DISTRIBUTION COMPARISON BET 9 MEN IMPAZ AND ANC 2.00 1.75 ANC 1.50 11.25 N
- 1 e - . . . ,_
[ K O1.00
u
)Y l l
@0.75
/ \
( \
0.50 ,
l :
0.25 .
, f 0.00 0 12 24 36 48 60 72 84 96 30TT001 108 120 132 144 TOP AZIAL REIGHT, INCHES BURNDP=6,620 POfD/MTU POWER LEVEL =100%
100 l
I.
1 FIGURE 5.3-12 ST LUCIE UNIT 1 CYCLE 12 AZIAL POIER DISTRIBUTION COBEPARISON BETIGEN INPAZ AND ANC 2.00 I i
)
1.75 INPAZ 1
ANC 1.50 )
i
" 1.25 N
H *- % 4 2 a %;,,.f u 1.00 .. .. I l n )
j H l i
h E
@0.75 f
I !
i 0.50 l l
.l 4
0.25 6- l 1
0.00 - I 1
O 12 24 36 48 60 72 84 96 108 120 132 144 BOTTON !
TOP AZIAL EEIGHT, INCHES BURNUP=13,320 MWD /MTU POWER LEVEL =100%
I 1
. ,oi .
l
e i
6.0 REFERENCES
4
- 1. Langford, F.L. and Nath, R.J., " Evaluation of Nuclear Hot Channel Factor Uncertainties," WCAP-7308-L, April 1969, and Spier, E.M. and Nguyen, T.G., " Update to WCAP-7308-L-P-A (Proprietary), Evaluation of Nuclear
! Hot Channel Factor Uncertainties,". June 1988.
- 2. Meyer, C.E. and Stover, R.L, "lNCORE Power Distribution Determination in Westinghouse Pressurized Water Reactors," WCAP-8498, July 1975.
- 3. Nguyen, T.Q., et al, " Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," WCAP-11596-P-A (Proprietary), June 1988.
i
- 4. Miller, R.W., et al, " Relaxation of Constant Axial Offset Control /FQ 1 Surveillance Technical Specification," WCAP-10216-P-A (Proprietary),
June 1983. l i
- 5. Bordelon, F.M., et al, " Westinghouse Reload Safety Evaluation l Methodology," WCAP-9272-P-A (Proprietary), July 1985.
- 6. Camden, T.M., et al, " Rod Bank Worth Measurements Utilizing Bank Exchange," WCAP-9863-A (Proprietary), May 1982.
- 7. Camden, T.M., et al, "PALADON-Westinghouse Nodal Computer Program," WCAP-9485 (Proprietary) and WCAP 9486, December 1978 and Supplement 1, WCAP-9485-A (Proprietary) and WCAP-9486-A (Non-Proprietary), September 1981.
- b. Liu, Y.S., et al, "ANC: A Westinghouse Advanced Nodal Computer Code," WCAP-10965-P-A (Proprietary), December 1985.
102
- i. _ . . __
- 9. Poncelet, C.G., et al, " LASER - A Dapletion Program for Lattice Calculations Based on MUFT and THERMOS," WCAP-6073, April 1966.
- 10. Olheeft, J.E., "The Doppler Effect for a Non-Uniform Temperature Distribution in Reactor Fuel Elements," WCAP-2048, July 1962.
- 11. Harris, A.J., et al, "A Description of the Nuclear Design Analysis Programs for Boiling Water Reactors," WCAP-10106-P-A (Proprietary),
June 1982.
- 12. Barry, R.F., et. al, "The PANDA Code," WCAP-7048-P-A (Proprietary) and WCAP-7757-A, January 1975.
- 13. Correll, G.R., et al, "lNPAX-II: A Reactor Power Distribution Monitoring Code," Exxon Nuclear Company, XN-NF-83-09(p), Maich 1983.
14 . Morita, T., et al, " Power Distribution Control and Load Following i Procedures - Topical Report," WCAP-8385, September 1974.
Y I
i a
103 f
. ~ '
APPENDIX A This section describes the primary Westinghouse computer programs used by FPL to perform the required reload core design calculations for Turkey Point and
- St. Lucie. These codes are used in a manner similar to that outlined in Section j 3 of Westinghouse's licensed reload methodology topical report (Reference 5).
4 Although the codes described in this appendix are not specifically addressed in the topical, two of the codes, FIGHTH and APOLLO (Reference 12), contain the same basic methodology as the licensed versions. The updated code versions 4.
include engineering enhancements (e.g., editing improvements, minor modeling
' improvements, and larger problem size capabilities) relative to the original code versions. The updated code versions were described at a meeting between the
! NRC Core Performance Branch and Westinghouse's Nuclear Fuel Division at the
! October 1984, at which time the differences between the original and updated code versions were discussed. The NRC concurred that the updated code versions were essentially the same as the original versions, employing the same fundamental solution algorithms as the original versions.
1 The two major remaining codes, PHOENIX-P and ANC incorporate significant i improvements to the methodologies discussed at the 1984 Westinghouse /NRC meeting. PHOENIX-P is a two-dimensional multigroup lattice code which does l not rely on the spatial / spectral interaction assumptions inherent in the previous
! methodology. ANC is an advanced version of the PALADON code (Reference 7) incorporating nonlinear nodal expansion, equivalence theory (for cross section l homogenization), and a pin power recovery model. The topical reports I (References 3 and 8) qualifying PHOENIX-P and ANC for use in reload core i design have been approved by the NRC.
i A.1 FIGHTH The FIGHTH code computes effective temperatures in low enriched, ,
sintered UO, fuel rods for specified values of bumup, linear heat 104 i
4
,-a-v ,- -.-. ,,
O
~
j generation rate, moderator temperature, and flow rate. Resulting fuel and
! clad temperatures are used as input for the PHOENIX-P code. FIGHTH accounts for the radial variation of the heat generation rate, thermal conductivity, and thermal expansion in the fuel pellet; elastic deflection in the cladding; and pellet clad gap conductance. The pellet-gap conductance is dependent upon the type of initial fill gas, the hot open i gap dimensions, and the fraction of the pellet circumference over which
- ' the gap is effectively closed due to pellet cracking. References 9 and 10 provide a description of the basis of the FIGHTH program.
i A.2 PHOENIX #
PHOENIX-P is a two-dimensional multigroup transport theory code used to calculate lattice physics parameters for PWR core modeling. In f PHOENIX-P, the detailed spatial flux and energy distribution solution is divided into two major steps. In step one, a two-dimensional fine energy .
group nodal solution which couples individual subcell regions (pellet, clad, and moderator) as well as surrounding pins,is obtained. PHOENIX-P uses a Carivik's collision probability approach and heterogeneous response fluxes to preserve the heterogeneity of the pin cells and their surroundings. The nodal solution provides a detailed and accurate local flux distribution. This distribution is then used to spatially homogenize the pin cells into fewer groups.
In the second step of the solution process, PHOENIX-P solves for the angular flux distribution using a standard S* discrete ordinates calculation.
'This technique utilizes group-collapsed and homogenized cross sections obtained from the first step of the solution. The Sd fluxes are then utilized to normalize the' detailed spatial and energy nodal fluxes. These normalized nodal fluxes are used to compute the reaction rates and power distributions use:I to deplete the fuel and bumable absorbers. A standard B1 calculation is used to evaluate the critical spectrum of the fundamental
- sos .
1
.\
O
~
mode and to provide an improved fast diffusion coefficient for the core l spatial codes.
1 PHOENIX-P employs a 42 er, orgy group library which has been derived
! primarily_ from ENDF/B-V files. The PHOENIX-P cross section library was
) designed to correctly capture integral properties of the multi-group data j during the group collapse, in order to properly model significant resonance J parameters. The library contains all the neutronic data necessary for modeling fuel, fission products, cladding and structural, coolant, and control /bumable absorber materials present in most PWRs.
a A detailed discussion of the methodology and models incorporated in PHOENIX-P may be found in References 3 and 11.
A.3 AE i
. ANC is an advanced multidimensional nodal methods program used to j predict core reactivity parameters, power distributions, detector thimble fluxes, and other important core characteristics. ANC uses the nodal I expansion method to solve the two-group diffusion equations. Partial
- currents and average neutron fluxes for the nodes are determined from l continuous homogeneous neutron flux profiles by employing fourth order j polynomial expansions for each of the x, y, and z directions across the
~
node. Discontinuity factors are used to adjust the homogeneous cross-sections in order to preserve the nodal surface fluxes and currents that f would be obtained from an equivalent heterogeneous model. In addition, ANC contains a pin-power recovery algorithm which couples the analytic
~
solution of the two-group diffusion equations with the pin power information from PHOENIX-P. ANC is able to accurately reconstruct the results of fine mesh models using these methods. A detailed description ;
of the methodology employed in ANC is contained in Reference 8. 1 f
s
- 106
'i I
~ '
ANC is capable of performing either two or three-dimensional calculations with a wide variety of options. The code can handle geometries ranging from octant to full core and supports various symmetries. Feedback
! mechanisms make adjustments to the macroscopic cross sections to i account for any changes in fuel temperature or moderator density. Xenon and samarium buildup and decay are modeled in addition to fuel and bumable absorber depletion. Typical applications of ANC include:
$
- Differential and integral control rod worth,
- Axial and radial power distributions,
- Reactivity coefficients, l
- Critical core configurations, j
- Shutdown margins, and l
- Fuel and bumable absorber loading pattoms.
A.4 APOLLO
! APOLLO is based on a one-dimensional two-group algorithm utilizing l steady state diffusion theory solved via the finite difference method.
Normally, an APOLLO model is generated by radially homogenizing a three-dimensional ANC model. APOLLO is an advanced version of the PANDA code, described in Reference 12. Cross sections are flux and
!1 volume weighted over each mesh interval and a bumup and elevation dependent radial buckling search is performed to normalize the APOLLO model to ANC. APOLLO is used for applications which require a finer
- mesh in the axial direction than ANC, as a relatively high number of mesh points are available. Applications typically include
- Axial power distributions, including Fa synthesis,
- Differential and integral control rod worth,
- Trip reactivity curves,
- Load follow evaluations, and
- Control rod insertion limits.
1 107 a
--y w -er-,-i,--e,, ,.-.,y , -e- - - , ,
6 1
~ ~
I The algorithms used in APOLLO account for space dependent feedback effects due to xenon, samarium, rod position, boron, fuel temperature, and 1 i
water density, j l
k l
t t
t 108-i
, v ,
MAINTENANCE RULE EXPERT PANEL MEETING MINUTES DATE: 8 July,1996 CHAIRMAN: ' Mike Snyder ', M M MEMBERS: Joe Price (SCE) CONSULTANT: RJ Davis (OM)
Brien Vincent (PSA) .
OBSERVER: Charlie Rossie OBSERVER: Dave Lowens !
Bill Hagar (SRO)
CONSULTANT: Kelly Korth (OST) (
i I
(NOTE At least one of the members shall be from the PSA Group, at least one have held a SRO license'or certificate at PSL)
TOPICS REVIEWED ,
- 1. The purpose of the meeting was twofold: a rgview ofmpas.ed_ revision to the unavailability criteria for the SSC logic actuation systems at St. Lucie, and to review the risk determination methodology for the 13 week schedule.
- 2. Mike Snyder provided a handout to panel members of the NUMARC guideline section pertaining to selection of performance criteria to start the discussion. The reason for revising the previously approved unavailability criteria for RPS, AFAS and ESFAS would be to refocus unavailability trending from plant level to a train based unavailability criterion. A written description and example of how this would be trended was handed out. The example was the U2 RPS, which included a period of unavailability in Modes 1-3 when TCB 5 failed to open during a surveillance test.
- 3. After a discussion, panel members agreed to the definition of trains for ESFAS and AFAS, and agreed that a clear identification of trains in the RPS was not possible. Therefore, performance monitoring would be appropriate at the system level. However, panel members B. Hagar and J. Price disagreed with the unavailability limit proposed for the RPS, citing that it was unduly overconservative. Therefore, the change to all logic system unavailability critena was rejected until further review was performed
- 4. The last item of discussion w;as regarding how risk assessment of equipment removal from service was to be performed. The panel members briefly reviewed the draft guideline for 13 week schedule planning. Brien Vincent, RJ Davis, and Bill Hagar described how risk assessment during the planning stages of equipment removal from service would be accomplished using the PREMRAS prepared by JPN. Kelly Korth observed that there was a weakness in real time risk assessment in removmg equipment from service, on the shift that started work in the plant.s The panel could not agree on how to resolve this concern.
3 ACTION ITEMS ,
- 1. Mike Snyder to benchmark more thoroughly with other utilities on how they define
- unavailability for their logic actuation systems. (Completed 7/12'96, SONGS, Pali sades, PTN, PECO)
- 2. Mike Snyder to discuss with I&C Engineering the proposed change to Performance Criteria for logic actuation systems.- (Completed 7/13/96 Rod Filapeck) e as 4
E I
l 1
i l
l 1
I l
{
~
l
MAINTENANCE RULE EXPERT PANEL MEETLNG MLNUTES DATE: 9 July,1996 CHAIRMAN: Mike Snyder ,
MEMBERS: Joe Price (SCE) l Brien Vincent (PSA)
Bill Hagar (SRO) ,
t (NOTE - A.t least one of the members shall be from the PS A Group, at least one have held a SRO license or certificate at PSL)
TOPICS REVIEWED
- 1. The purpose of the meeting was to review the second draft proposed revision to the unavailability criteria for the SSC logic actuation systems at St. Lucie.
- 2. Mike Snyder provided a handout to panel members of the revised draft performance criteria to start the discussion. The reason for revising the previously approved unavailability i
criteria for RPS, AFAS and ESFAS would be to refocus unavailability trending from plant level to a train based unavailability criterion. A wntten description and example of how this ;
would be trended was handed out. The example was the U2 RPS, which included a period of l
unavailability in Modes 13 when TCB 5 failed to open during a surveillance test.
)
- 3. Mike Snyder shared with panel members that prior to the meeting, Rod Filapeck of JPN I&C engineering was contacted for an opinion on the best method to monitor the health of the -
logic actuation systems. He agreed with the concept of unavailability monitoring on a train level for the AFAS and ESFAS systems, and agreed that it was not possible to identify more than one train for the RPS so system monitoring was satisfactory He did raise the concern
- that more specific monitoring was also necessary for the logic actuation systems, and agreed that reliability tracking for the measurement parameiirs would satisfy that concern. (Rod Filapeck was contacted in lieu of the system owner, since the system owner responsibility was l
about to change due to reorganization) l
- 4. A peer review of SONGS, PTN, Palisades and Limerick for how perfomance criteria was ;
developed for logic actuation systems was shared with the panel members. These plant sites l e
generally monitored unavailability on a 2 train level with the exception of Palisades, which ,
used Plant Level Performance Criteria. Ti.e two sites with Eagle 21 RPS systems monitored on a 2 train level as well. The limit for unavailability was generally more than 1%, appearing
' to be arbitrarily selected. The proposed PSL unavailability would be 0.6%, and based on Tech Spec LCO shutdown times since there is no corresponding limit in the PSA.
- 5. The panel members agreed to accept the revised draft performance criteria for the logic I actuation systems. The revision to the first draft was that the RPS unavailability limit was
' ' raised from less than or equal to 6 up to 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br />, to be consistent with AFAS and ESFAS.
~ f B
4 .m w-- -
- . -. .~ - .. .. -
. .. - . -. . . --- - . .. ..... ..~ . .,
- I i
ACTION ITEMS
- 1. Bob Walcheski to write a PCR to ADM 17.08 to reflect the change in performance criteria ;
for the logic actuation systems. .
i 4e
(
I h
1 e
1 I
I s
I;
,' 1
.1 l
f l
i 4
4 1
1 6
4 i
9 Geen 1
.*y --m =.-. m ..*4 -, - e s,-r r . -,w-- -,..
WAINTENANCE RUIZ EXPERT PANEL MEETING MINUTES
- DATE
- -9 August.1996 I :
CHAIRMAN: Mike Snyder WEMBERS:: Ted Dillard ~ .
SYSTEM OWNER. Roger Kulavich
~
Brien Vincent (PSA)- OBSERVER: Dave Wolf-i Roger Weller (SRO) CONSULTANT: Doug Weeks
'(NOTE - Atleait one of th'e members should be from the PSA Croup. at least one have held a SR0 license or certificate at PSL)
. TOPICS REVIEWED i
- 1. The purpose of the meeting was to confirm. (using the upgraded Mrule Procedure which ,
now requires an expert panel approval) keeping the EDCs governor controls in a(l) .
i classification due to reliability problems. Mike .inyder handed out a copy of the pertinent NEl 93-01 guidance.to panel members. and briefly tughlighted to panel members why the EDG governors were :n atl) status .
- 2. Doug Weeks (ERT leader) handed out to the panei members the response to CR 96-1703 for the cause of the governor failure of the 2A2 EDC on 7/10/96. and discussed the .
circumstances. potential causes, and corrective actions.
- 3. The panel then discussed if this condition was a functional fanure. and with Roger Kulavich's technical perspective agreed that it was a funcuonai failure. A discussion of if
.the failure was maintenance preventable followed Doug Weeks ,ndicated that the cause of
~
the failure could be attributed to either site or vendor mamtenance practices. Therefore, the panel and system owner agreed to conservatively determme this to be an MPff. The i panel then compared this MPff with the previous F,VPFF for gmernor failures, and determined that the 7 '10/96 was not an repeat because the tauses were different (lack of PM versus a procedural deficiency during a performance of a PMI
- 4. Roger Kulavich pomted out to the panel that EDC reliabihty could be significantly improved by the upgraded of the governor controls from the r,01 to 2301A model. This would have elimmated the failure mode on 7'10% and possibis all other previous governor control failures except for one. An industry survet mdicated that three other sites had performed the upgrade. and that all sites with the 2301 A modei ure satisfied with its performance. The REA for the upgrade has been rubmitted for management review and
- approval.
- 15. Ted Dillard expressed a concern relating to the nigh vioranon of the2 A2 Diese lcompared to.the other seven diesels. questionmg if the :;oor ;osernor n r5rmance may have been ,
. . causally related to the governor problems This me e bem acKeii ua PMAI for lo g term j i
( :
resolution. Specifically. pdm and Doug Week are working with mig ES! to ascertain the cause l of the high vibration.
- l
- a. In determining the Maintenance Rule classification of the EDCs. Roger Kulavich requested ;
that the Unit 2 EDGs remain in the a(1) classification for governor performance monitoring 4 until 6/1/97. The goals would remain otherwise unchanged because discussion indicated that monitoring methods and trends were effective to predic future failures of a similar nature. assuming the approval for installation of the 2301A governor controls. The Unit 1 EDG goals would remain unchanged. The Expert Panel unanimously agreed to this :
i recommendation. .
~
- 7. The last itim that the panel discussed was the inability to perform a complete historical review of information during root cause analysis Earlier conversation in the rr.eeting showed
- that failure data for the EDGs was retrieved only back to 1988 The reasons cited for'this :
condition wefe that PW0s were previously writte'n by hand. and that NPRDS does not provide l l
data back to 1976. 4
! ACTION ITEMS t
j 1. Doug Weeks will provide a written copy of the F?! failure anar,m and procedurai .
j upgrades to ESI so that they may incorporate ie.<.<ons ! earned fmm this event This action is I
j important from the standpoint that the maintenante organizanon responsible for the shear pin overload could not be determined to be either FPL or ESI- l
! 2. Ted Dillard will find out who is responsible for equipment hi. tory to potentially aid in the i development of a process for retrieving all relesant historicalinformation f
I a
i i
l
-' ~ . - ,, ,_
l
..I t
- MAINTENANCE RlilI EXPERT PANEL MEETING WINUTES DATE: 27 August,'!996 4
CHAIRMAN: Jeff West (SR0) .
MEMBERS: Ted Dillard Brien Vincent (PSA) ,
, Mike Snyder ,
~
(NOTE . Allea'ts one of the members should be from the PSA croup at least one have held a SR0 license or certificate at PSL) ;
3 t
T6PlCS PSVIEWED
- 1. The purpose of the meeting was to review a proposal to inemde the Operations Support )
l 4 Building as a structure within the scope of the rule The potential need for inclusion was i
identified during the July QA audit. smee metal facia from that building blew off during an unusual onsite micro burst and contacted two phases of a Main Transformer located directly.
across the roadway. The resulting Generator Lockout caused an automatic reactor trip. '
Mike Snyder handed out a copy of the Structures : coping docunient as prepared by Ed Hollowell (structures owner) to panel members. and briefly.highhghted to panel the options of: Include only the Ops support building smee it has proximate location to the Unit 1 Main Transformers and has resulted in a unit trip. Include all onsite buildings as structures, and finally to not include the Ops support building which could be construed as a deviation from the NEl 93-01 guidelines.
- 2. Although the system owner was not present at the meeting. he was nearby should questions have arose on the proposal to include the Ops support building mto the scope of the rule.
3 The panel members agreed that although a similar esent was unlikely to recur. the conservative approach would be to include the Ops support ounding m the scope of the maintenance rule considering its proximate location to the transformers .This would require a baseline inspection and subsequent periodic mspections of Ihe ups . support building by the :
structures owner-ACT10N ITEMS
- 1. Mike Snyder will distribute the revised structures scoping <iocumentation to Ed Hollowell and Expert Panel members for updatint! their Maintenance Rule nandbooks.
i l
__________J
- 1. There is no." updated"-risk ranking lists. The. original PSA risk ranking input'is provided by the attached calculatio~ns.
A PSA update has not yet been-performed and, therefore, no change in PSA input has been provided.
- 2. A PSA update has not yet been completed (presently in progress). A response to an NRC RAI for the PSL RAI has been submitted to the NRC (I assumed that they should know this).
- 4. This question is somewhat confusing. Since we scoped risk significance by system, and only in few cases considered only certain components within a system. risk significant, I would assume that the issue would be the MR systems included in the PSA but not considered risk significant. They are as follows (the MR scoping document lists all non-risk
- significant systems (only around 30 total)):
= Turbine Cooling Water (backup cooling for "A" and "B" instrument air compressors in system 13 but are considered part of IA system for MR)
- Some HVAC (Unit 2 ICW pump enclosure)
- Some 480V swgr/bkrs (non-vital swgr)
+ 6.9kV swgr' 1
5 6
4 I
I FPL Nuclear Division R G-u dy Quality Assurance Audit Report l
l l
MAINTENANCE RULE COMPLIANCE QSL-MR-96-14 t
i I
I 9
Audit Team:
C.V. Rossi D.C. Lowens L. W. Bladow QA PSL n ,.. n
O AUDIT REPORT
- QSL-MR-96-14 I Page1of17 ;
Executive Summary The Maintenance Rule (10CFR 50.65) became effective on July 10.1996. In conjunction with the effective date of the Rule, a joint PSL/ PTN audit team performed an evaluation of Plant St. Lucie (PSL) Maintenance Rule (MR) activities. The purpose of the audit was to verify that the PSL program is in compliance with the regulation. and is in conformance with industry standards. An NRC inspection of the PSL Maintenance Rule Program is scheduled for September 1996.
The following audit results were obtained:
Signincant improvement has occurred since the last Maintenance Rule Independent Technical Review was performed in December 1995. l l
l The PSL Maintenance Rule Program is currentlyin compliance with 10 CFR 50.65, l but contains a number of significant weaknesses. !
The most significant weakness is the lack of program elements to evaluate the overall effect on safety functions when removing equipment from service. This weakness has the potential to be interpreted as regulatory non-compliance (See Finding 1).
The speed with which the PSL Program has been implemented has resulted ina deficiency in personnel training, deficiencies in completeness of documentatim, and a deliciency in the integration of Maintenance Rule processes into the plant daily routine. This latter observation is particularly true with respect to activities of System Owners.
One Quality Assurance Finding and one composite Independent Technical Recommendation were generated as a result of the audit. The Finding addresses a deficiency in implementation of 10 CFR50.65 (ax 3) as mentioned above. The Independent Technical Recommendadon (CR 96-1827) contains several parts. and provides suggestions. both for improvement in documentation of Maintenance Rule acuvities and for the assimilation of these activities into plant processes.
The PSL Maintenance Rule Program is currently in marginal condition although it meets regulatory requirements. Greater emphasis must be placed upon effective implementation of the Program in order to achieve the desired level of excellence.
O
- @ FPL AUDIT REPORT QSL-MR-96-14 Page 2 of 17 Audit Summary Audit criteria were based upon regulations contained in 10 CFR 50.65, industry guidance contained in NUMARC 93-01 " Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants". and NRC expectations as contained in Inspection Procedure 62706 " Maintenance Rule Inspection Procedure". The discussion below is divided into sections that reflect the major portions of'he PSL Maintenance Rule Program. Recommendations are incorporated with the applicable discussion and are documented in CR 96-1827.
Maintenance Rule Scone The Maintenance Rule requires nuclear plant Structures Systems and Components (SSC) that exist within four categories be included within the scope of the Rule:
i
- 1. Safety-related structures, system and components (as defined by 10CFR 100). l
- 2. Non safety-related SSCs that are relied upon to mitigate accidents or transients or are used j in the plant Emergers/ Doerating Procedures 1
- 3. Non safety-related SW. whose failure could prevent safety-related SSCs from fulfilling their l safety related function. i
related system.
At PSL the initial determination of SSCs that fell within the scope of the Ru'e was made by Nuclear Engineering. The results of this determination were provided in draft calculation JPN-PSL-93-006.
This calculation outlines the plant specific methodology for necessary determinations and includes 1 tables that list Unit I and Unit 2 SSCs and applicable status under the Rule. The tables provided by this calculation have subsequently undergone several revisions and were ultimately approved by the PSL Expert Panel on June 27.1996.
The methodology contained in the draft calculation is conservative and well documented. The output tables provide an explicit and accurate listing of SSC status. Subsequent revisions to the tables have incorporated refinements to scoping philosophy that have been obtained from industry )
apenence. The basis for the subsequent revisions is not well documented. The draft status of the calculation, and the detachment of the status of the output tables from the calculation itself, represents a weakness in the continuity of the PSL Maintenance Rule Program. 1 RECOMMENDATION: The methodology used to arrive at the current list of SSCs that are within the scope of the Rule should be documented and approved by the Expert Panel. The
. , administrative mechanism for transition from the resuit of the Engineering calculation to plant control of the scoping process on an ongoing basis should be claritied. iCR 96-1827) ;
O O AUDIT REPORT QSL-MR-96-14 Page 3 of 17 A
In summary, the current status of the document titled " Maintenance Rule Scope Determination St.
Lucie Unit I and 2", Revision 0 dated 6/19/96 and its review by the Expert Panel dated June 27, 1996 constitute acceptable compliance with 10 CFR 50.65(b). Additional improvement in the depth of documentation supponing this decision is desirable.
Exnert Panel NUMARC 93-01 provides a process for division of plant SSCs into risk-signincant and non risk-signincant categories. The industry-standard approach towards accomplishment of this differentiation has been through the use of Probabilistic Safety Analysis (PSA) tempered by the judgement of an Expen Panel. The Expert Panel used in this capacity is composed ofindividuals experienced in disciplines of PSA. operations and maintenance.
PSL has utilized the industry-standard approach. The Expert Panel' chairman is the Manager of Systems & Component Engineering (SCE), and Panel contains representation from other plant disciplines. Meeting minutes exist to document Expen Panel review of system scoping, determination of risk significance. assignment of systems to at 1) status and goal setting activities.
During the audit records of Expert Panel training and meetings were reviewed, a meeting of the Panel was observed, and Expen Panel members were given an examination on topics associated with the Rule. The results of this evaluation indicate that Panel activities have been sustained largely by the strong operations background of the Maintenance Rule Administrator and consistent panicipation by the plant PSA Engineer. With a few exceptions, the training and knowledge level of a majority of other Expert Panel members is in need ofimprovement. At the meeting that was observed. discussion wandered signi6cantly from Maintenance Rule related topics. and seemed to be impeded by the fact that several of the Panel members did not have a clear focus on the purpose of the meeting as related to the Rule. It is noted that the current charter and operating instructions for the Expen Panel are contained in a single paragraph in ADM 17.08 and a one page memorandum from the SCE Manager.
RECOMMENDATION: Explicit operating instructions should be provided for the Expen Panel. These should include criteria for evaluation of risk-significance, criteria for evaluation of repetitive maintenance preventible functional failures, criteria for the use ofindustry exp rience.
trending as a pan of a(1) goal setting and monitoring, and other topics related to implementation of the Rule. The Expert Panel Notebooks should be updated to contain the most current approved PSL MR documentation iCR 96-1827L RECOMMENDATION: Detailed re-training should be conducted for members of the Expert Panel. The training should encompass the body of knowledge necessarv for informed decision-making under the provisions of the Rule. Expert panel quali6 cation should be con 6rmed by the admimstration of an examination (CR 96-1827L in summarv. proper documentation exists of necessarv Expert Panel activities. The personal efforts l
A).
il AUDIT REPORT QSL-MR-96-14 PPL Page 4 of 17 of a few individuals have ensured that the necessay quality is present. Additional training is necessary for a majority of the Panel members. In this connection the following information is noted: 1) The Maintenance Rule Administrator has had to compete with other plant priorities for the attention of necessan personnel 2) Current plant reorganization activities have the potential to damage the existing functionality of the Expert Panel.
Risk Sienincance NUh1 ARC 93-01 provides criteria for differentiation of SSCs into risk-significant and non-risk significant categones. These two categories are permitted to oe treated differently under the provisions of the Rule.
St. Lucie Procedure ADh1 17.08 " Implementation of 10 CFR 50.65 - The Maintenance Rule" adequately addresses the three PSA based calculational methods from the for assessing risk. The procedure uses both alternative forms of the Risk Reduction Worth method. Consistent with NUMARC 93-01. the results of all calculational methods were provided to the Expert Panel for consideration. ADM 17.08 also addresses the NUMARC 93-01 guidelines for addressing a containment failure or bypass situation that could result in an unacceptable release as a risk significant consideration. The following items were noted.
- Approximately 40% of the PSL SSCs are risk significant (RS). This is slightly higher than the peer plants (approximately 25m but acceptable.
- Overall. SSCs were nominally classi6ed as RS if they met any of the risk significant calculational methods screening criteria. This is conservative.
- The Expert Panel used judgement as a basis to include several systems as RS that were either not modeled or not significant contributors to risk in the PSA (such as instrument air or containment sprayL Conversely Expert Paneljudcement re-classi6ed several SSCs as non-RS (such as turoine switchcear room HVAC) w hen the PSA screening critena would have them as RS.
l
- \*eri6 cation was performed via interview that the SSC performance criteria are consistent with the PSA assumptions or it have been evaluated as having minimal impact on risk.
- The need to tlac Maintenance Rule impact as a consideration dunng the modification process is addressed in plant procedures.
NUM ARC 93-01 guidelines also address factoring of actual plant availability / reliability data into the PSA model for updates. The utility has broad latitude for implementine this guideline. If and when the PSA is updated the revised information should be factored into Maintenance Rule requirements. !
l l
2
}
/
I j AUDIT REPORT
\#
FPL QFL-MR-96-14 -
Page 5 of17 RECOMMENDATION: Add an Expert Panel responsibility to ADM 17.08 to review and approve PSA updates (for consideration on the effects on performance criteria) (CR 96-1827).
In summary, differentiation between risk-significant and non-risk significant SSCs meets the requirements of the Rule as amplified by the guidance contained in NUMARC 93-01.
Performance Criteria /Monitorine - Systems NUMARC 93-01 establishes the expectation that SSCs demonstrating acceptable levels of performance will be maintained using normal methods of preventative maintenance. SSCs that fail to meet plant established performance criteria are required to receive increased attention via specific goal setting and monitoring. Risk-significant and standby systems are required to have performance criteria established at the train level. Non-risk significant systems may be monitored by more general plant-level performance criteria.
l The performance criteria used to monitor PSL SSCs are now in conformance with NUMARC 93-01 !
guidelines and industry standards. This is a significant improvement over the condition that '
existed at the time of the previous evaluation. The current performance criteria adequately monitor the designated Maintenance Rule functions of risk-significant and standby SSCs. Plant level performance criteria are intelligently applied and correlate to the Maintenance Rule functions of non-risk significant systems. Performance criteria pertaining to unavailability correlate with equipment unavailability assumed by the PSA in accordance with NUMARC 93-01 expectations. i During the review of performance criteria one situation was identified that requires mention. On September 30.1995 the PSL EDGs were placed under the provisions of the Maintenance Rule in coniunction with deletion of Technical Specification requirements for accelerated testing and special reporting of EDG failures. On May 9.1996 the EDG reliability performance criteria (PC) were modified using the administrative process prescribed by ADM 17.08 for PC changes. The '
performance entena were changed from the trigger values specified by NUMARC 87-00, Rev 1.
" Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors" to a defined failure rate of <=2 maintenance preventible functional failures per ;
operating cycle. '
PSL established a commitment to Regulatory Guide 1.160 " Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", at the time of early implementation of the Rule for EDGs.
RG 1.160 states the following with respect to performance criteria for emergency diesel generators I iEDG)
" Performance enteria for emergency diesel generators (and the support systems vital to their functioning) would be met by the absence of a repetitive maintenance-preventible functional failure (MPFF), by not exceeding the established reliability or unavailability criterion, or by the l l
7
)
k/
FPL b 'N %
l
' Lt te -
AUDIT REPORT QSL-MR-96-14 Page 6 of 17 occurrence of a single maintenance-preventible failure followed by appropriate cause determination and correctiva action."
The Regulatory Guide requirement implies that only an single MPFF is acceptable for EDGs but does not specify a time period over which this failure rate is measured. During the course of the audit. the EDG perfonnance criteria were re modified by the Expert Panel to change the limit to <=1 MPFF per year. Subsequently, a conversation with the Technical Specialist who performed the December 1995 evaluation. and has broad indusuy experience. indicated that the apparent requirement for a single MPFF per cycle may not be an actual NRC intention. The response of the Expert Panel to this condition is considered to be conservative and in the interest of safety. The area of EDG reliability was addressed by NRC Integrated Inspection Report 96-08. and will likely be re-examined in the future.
Following establishment of the performance criteria. SSC performance history during the past two operating cycles was examined to establish baseline status under the provisions of the Rule. This represented a large scope of work. He methodology used to accomplish this task was reviewed during the audit and a spot check of documentation was performed, ne limited scope of this audit (two man weeks) precluded detailed verification of the results of SSC status determinations. Within this constraint, the following observations are relevant:
- The tasks of calculation of plant level performance. collection of out-of-service hours and detection of equipment failures were able to be performed with predictable accuracy on a historical basis.
- The tasks of determination of functional failures and assignment of functional failures as maintenance preventible were able to be performed with lower accuracy, particularly in cases where System Owners were not functioning in their present capacity during the time period under review.
- Due to the speed with which the historical review was performed. improvement is possible in the organization of the data reviewed and intermediate results obtained. 1 It is likely that this arca may be the focus of significant attention during the NRC Inspection.
Additional review of this area by PSL Quality Assurance is warranted.
RECOMMENDATION: A detailed examination of the historical review should be performed by Quality Assurance. t PMAl to be issued)
On an ongoing basis, out-of-service hours are obtained, and unavailabilities are calculated by an SCE )
technician. Equipment failures are detected through the work order process, in-house Events and !
the Condition Report process and are evaluated for status as functional failures by the Maintenance ]
l i
OI:PL AUDIT REPORT QSL-MR-96-14 Page 7 of 17 b
Rule Administrator and Condition Report Oversight Group (CROG). In cases where functional failures are identified, Condition Repons are written to the System Owners to determine whether or not the causes were mamtenance preventible. In the event where performance criteria are exceeded, the System Owners are tasked by the Maintenance Rule Administrator to prepare goal setting and monitoring activities. This system is_ adequate for basic cornphance-with4he4ule_but-ha.tseyeral
~
disadvantages: .-
- NRC expectations are that all repetitive component failures, whether or not MPFFs and whether or not in the une system. be analyzed for generic concerns. The present process does not clearly account for this task being performed.
- NRC expects that failure cause determuiations establish whether there are generic implications to the failures by performing a review to determine if the component is used in other applications in the plant. The present process does not clearly account for this task being performed.
- NRC expectations lepeatedly emphasize the use of trending for systems in (a)(2) status 'and~
require criteria for the initiation of a cause determination in the case of clearly declining trends. Presently, trending is performed by the plant Predictive Maintenance Group, but is almost completely detached from the Maintenance Rule activities of the System Owners.
1 The problems above derive from the need to accomplish Maintenance Rule activities in the absence l of a developea system engineer program. Problems with the system engineer program were I previously identified in independent Technical Review t ITR)96-008 " Emergency Diesel Generator 1 Maintenance Program". The need for corrective action in this area is currently tracked by PMAI PM96-05-057. Plant management is aware of problems in this area. but these problems will represent an impediment to proper presentation of the plant program in the upcoming Maintenance Rule inspection.
In summary. the following comments are applicable: performance criteria are properly established. l the historical review is complete but documentation is in need ofimprovement and additional review l for accuracy is desirable. ongoing monitoring activities are sufficient to meet the requirements of the ;
Rule, but lack retinements that will be expected by the NRC.
Performance Criteria /Monitorine - Structures Detinitive industry guidance on the nature of an effective structural monitonng program was not available to the nuclear industry until March 1996. This audit evaluated the PSL procedure and scoping determination for structures against the recently provided NEI guidelines. On an overall basis. ADM 17.08 adequately implements the applicable reqmrements. The following items were noted.
x, '
AUDIT REPORT
. v QSL-MR-96-14 Page 8 of 17
- Baseline / periodic evaluations are required to be performed and evaluated by qualified individuals. Ownership of structural monitoring at PSL will be by Civil Engineers.
- Corrective action requirements in the PSL program are defined and consistent with NEI guidelines.
- An evaluation was performed as to whether condition monitoring criteria is predictive in nature, specifically where a failure of a structure would result in a loss of safety function.
Via inteniew it was confirmed that. criteria are conservatively set to be preaictive.
- The baseline monitoring schedule for structures is not yet set. but the goal is to complete the baseline surveys prior to year end. Blis appears adequate to meet the Rule as the three year historical reviews have been completed. The NEI guidelines do not give a timetable for completion of the baseline surveys.
- PSL Maintenance Rule scoping documents acknowledge existing programs that monitor structures. This is a positive feature of the PSL Program which addresses an area that was found to be a weakness during pilot plant Maintenance Rule inspections.
One minor documentation issue was noted during the review of the program for structures. Expert Panel meeting minutes do not clearly state that structures are initially categorized in (aK2). ADM 17.08 step 7.2.1.4c requires Expen Panel review and approval of the initial disposition category for SSCs. On a global scale. this can easily be accomplished by having the Expert Panel review and approve the most recent Maintenance Rule Quanerly Report.
RECOMMENDATION: The Expen Panel should review and approve the initial categorization of structures as being in ia)(2) status t CR 96-1827).
In summary. the PSL Maintenance Rule program for structures is in compliance wnh applicable requirements and consistent with industry guidance.
Svstem On ners Panicipation of System Owners in the Maintenance Rule process is neither required by the Rule. nor explicitly mentioned in either industry guidance or the NRC inspection procedure. At the same time. the System Owmer as the focal point of Maintenance Rule activities appears to be an expectation of every organization associated with evaluation of activities under the Rule.
At PSL System Owners are divided between the SCE Department and 1&C Maintenance Group.
\1ost sy stem owners have attended a singie onentation session on the Maintenance Rule.
participated in the histoncal review of system reliability and panicipate in the ongoing analysis of
~
- uncuonal failures as described above. Continuity of activities under the Rule has been hampered
. I AUDIT REPORT QSL-MR-%14 FPL '
Page 9 of 17 l l
by reorganization and personnel tumover in the SCE Department. As an example. the System Owner for the HPSI, LPSI, Safety Injection Tanks and Contamment Spray Systems has changed two times in the last 18 months. Rese systems are current'y unassigned following the departure of the most recent individual to hold this assignment. The overall number of System Owners at PSL is lower than the industry average, c.s mentioned above. other problems with the functionality of the System Engineering program are described in ITR 96-008. The effect of the upcoming SCE/ Engineering re-organization upon current System Owner assignments and Maintenance Rule functionality is of concern with respect to the continued effectiveness of the MR Program.
In summary, the System Engineer program does not make a contribution to activities under the Rule that is in line with industry standard practices. Organizational instability and lack of functionality of the System Engineer position are the primary reasons for this deficiency. Plant management is aware of these problems.
Goal Settine/Monitorine in cases where SSC perfo mance does not meet applicable performance criteria or a repetit(ve maintenance preventible functional failure occurs, goal setting and monitoring is specified by ParagrMYTT6TmtK ADM 17.08 adequately addresses necessary aspects regarding consideration ofindustry events, goal setting emnmensurate with safety, goals that are predictive in nature and conditions under which goals are required to be established. On an overall basis. the Maintenance Rule Administrator has properly implemented the requirements for (a)(1) dispositions. A weak area in goal setting exists in the level of documentation that supports the conclusions asociated with (a)(1) dispositions.
ADM 17.08. Figure 4 is used to document goal setting. This Figure does not lend itself to easily addressing procedural requirements, and decision points _are not clearly documented. The information contained on these forms will be presented to every individual / organization that reviews activities performed under the Rule.
Several SSCs that are currently being dispositioned under Paragraph (ah 1) were reviewed for the following attributes: goals commensurate with safety; indusur experience addressed, frequency and method of monitoring established. goal is measurable. trendable. and predictive in nature as required.
The following items were noted.
e I'l PORVs - Figure 4 does not retlect the method by which unavailability is affected/ monitored. Ahhough unavailability is adequately discussed in the Quarterly Report. Figure 4 is the proper location in which to discuss both the reliability and unavailability criteria.
e tJ1 "C" AFW - Good consideration ofgeneric/ cross system implicationsfor MOV
=
i n% AUDIT REPORT i; )
e ' QSL-MR-96-14 i FPL. Page 10 of 17 m
problems. Goal as written may keep AFW in (a)(1) longer than the muumum period necessary while generic implications are addressed. The " measurable" criteria to ensure MOV functionality have not yet established in Figure 4. Once it is established they should be reviewed by the Expert Panel.
Similar to U1/U2 EDGs, goals say "No failures."
e UI/U2 RCP Seals. Conservative entry into (a)(1) for a risk significant SSC. The RCP seals provided a large contribution to exceeding the plant level unavailability criteria. although in themselves did not lead directly to the criteria being exceeded.(Measurable _ criteria for
~
monitoring seal pressure breakdowns was have not yet been established in FIFtre7."
~
RECOMMENDATION: a) The format of ADM 17.08 should be revised or used in a manner that is more commensurate with an expanded discussion of necessary goal setting and monitoring activities. b) Figure 4 for Ul PORVs should be revised to reliect unavailability criteria c)
Measurable goals should be established for the Unit 1 "C" AFW pump and Unit I and Unit 2 RCP seals (CR 96-1827).
As an enhancement to the process, if an unavailability or reliability criteria have not been met for i a risk significant SSC, the (a)(1) goal setting process should provide for an assessment of the balance between availability and reliability. This information would then be available for the balancing i reliability and unavailability ponion of the periodic assessment.
In summary, goal setting and monitoring activities meet the requirements of the Rule. Improvement is possible in the documentation associated with goal setting and monitoring. In several cases specific actions described above associated with goal setting and monitoring will help to clarify present activities in the area.
Periodic Assessment Paragraph (a)(3) of the Rule requires that perhance and condition monitoring activities and associated goals and predictive maintenance activities be evaluated every refueling cycle provided that the interval between evaluations does not exceed 24 months.lhe purpose of the assessment is to adjust the program for activities under the Rule to ensure that the ooiective of preventing failures of SSCs through maintenance is appropriately balanced against the objective of minimizing unavailability of SSCs due to monitoring or preventative maintenance. l ADM 17.08 implements the maiority of the requirements specified by the Rule, including a review of goals and performance criteria. frequency of the periodic assessments. balancing availability and reliability. and actions taken as a result of cause determinations. Recommendations for procedure I ennancement have been provided to the Maintenance Rule Administrator via separate corresponaence.
l
e l AUDIT REPORT g QSL-MR-96-14 ,
Page 12 of 17 l
.___ /
expectations in this area. Usjng this guidance, existing plant procedures were reviewed and found AoTe7eficient for bo on-line and shutdown conditions. Existing informal plant informatierrarid j psoposed pmeedures may be used as a base to build upon in constmction of the ~~ey program.
i (See Finding 1).
N _ . - - - -
In sum 7naVthTPSL program elements in this area do n'otmeet the industry standard. Due to varying interpretations of existing regulatory guidance, it is not clear that this deficiency is a clear-cut failure to comply with the of the Rule but certainly does represent a critical deficiency in the PSL Maintenance Rule Program and should be corrected immediately.
Overall Conclusion In December 1995 an independent review of the PSL Maintenance Rule Program identified potentially significant problems in achieving compliance with the Rule. Since that time. a significant amount of work has been directed towards bringing the PSL Program into compliance with the applicable regulation and into conformance with industry standards. During this period the following significant results have been achieved:
- Performance criteria have been brought into agreement with industry norms e Historical review has been completed for the revised performance criteria
- (a)('l) systems have been identified and a majority of the necessary corrective actions, goals and monitoring established e ADM 17.08. the governing procedure. has been revised and brought into agreement with industry standards.
As a result. on July 10.1995, when the Rule became effective, the PSL Program had achieved the objective of compliance with applicable regulatory requirements. One area exists in which compliance is marginal, assessment of e of service on safety functions. His
~ '
area is addressed in.the Eindilig To'nTained in this report. -
Due to the speed with which the revised PSL Program was established, other deficiencies exist in training, documentation and integration of Maintenance Rule processes into the plant daily routine.
This latter observation is particularly true with respect to activities of the System Owners. The PSL Maintenance Rule Program is currently in marginal condition although it meet regulatory requirements. Greater emphasis must be placed upon effective implementation of the Program in the future in order to achieve the desired level of excellence.
(-
i, l AUDIT REPORT j
/
QSleMR-96-14
'k Page 13 of 17 m ,. . .
Findings Finding 1: Formal program elements necessary to assess of the overall effect on performance of safety functions of plant equipment that is out of service have not been implemented Criteria: 10CFR50.65(a)(3)
"In performing monitoring and preventive maintenance activities, an assessment of
~
the total plarit equipment that is out of service should be taken into account to determine the overall effect on performance of safety functions."
Regulatory Guide 1.160. Rev.1.
Section C Reculatorv Position "NUMARC 93-01 provides methods acceptable to the NRC staff for complying with the provisions of 10CFR50.65."
NUMARC 93-01, Rev. 2 Section 11.2.3.
"During the planning and scheduling phas: and prior to removal of SSCs from service, each planned maintenance activity that results in the removal of an SSC identified in l Section 11.2.2 from service should be assessed for its impact. j i
Discussion: Although plant personnel exhibit a sensitivity towards risk. the auditor concluded that the procedures in place on the effective date of the Rule (July 10.1996) did not meet l the industry guidelines. The guidelines refer to evaluation of risk ifor both on-line I and shutdown conditions) during both the planning stages and when removing components from service.
On-Line Conditions The maintenance rule provides broad latitude on the method by which necessary measures may can be implemented. but there are certain key elements that must be addressed. The following are examples of aspects specified by industry guidance that are not included in present plant procedures. )
- ' .. assessment applies to all modes of plant operation and should take l l
into account... expected changes to plant contiguration." (NUMARC 93-01. section i1.2.3) e "The decision to take equipment out of service for maintenance during power operation should take into consideration the likelihood and possible consequences of an event occurring while the equipment is out
f
'('j
% AUDIT REPORT F4DL QSL-MR-96-14 Page 14 of 17 of service." (NUMARC 93-01 section i1.2.3) e " ..the licensee has established and implemented an ongoing, documented process for assessing the overall effect on the performance of safety functions before SSCs are taken out of service... The inspector should verify that the licensee maintains a current status of all SSCs within the scope of the maintenance rule and that the licensee updates this status to indicate when SSCs are in or out of service."
(NRC Inspection Procedure 62706. section 03.03c)
From a planning perspective, the NRC has determined that a 12-week rolling window cycle to schedule preventive maintenance activities is acceptable. Although a draft procedure has been proposed it was not implemented at the time that this issue was reviewed during the audit. It is noted that prior to completion of the audit, the draft procedure was issued.
The Maintenance Rule permits a qualitative assessment of safety when removing i equipment from service, but au SSCs unponant to safety must be adtud The i documented basis by which this accomplisherTrTniid~e'q'ubaMent d PSL procedures. The plant procedure goveming Maintenance Rule activities ( ADM 17.08) l simply states "an assessment of the total plant equipment that is out of service should i be taken into account..." Nominal industry practice is to use either a simple matrix l (for combinations of equipmat out of service) or an on-line living PSA/ risk meter.
Although these may be considered quantitative approaches and not required, even from a qualitative perspective, a current list of all SSCs within the scope of the rule and their relative risk importance is not available in the control rooms at PSL for use j by Operations personnel. '
PSL procedure AP 0010460. " Critical Maintenance Management". together with I engineering prepared Pre-Evaluated Maintenance Risk Assessments (PREMRAs)is currently used to evaluate safety impact but only for certain T/S components (ir, not all maintenance rule SSCs) and only under certain conditions (i.e., excludes ;
equipment removed from service for surveillance testing.)
1 Procedure OP 0010129. Equipment Out of Service. is currently only required to be l used for Tech Spec related items and not for all maintenance rule SSCs.
chutdown Conditions Similar conditions exist for shutdown conditions. AP 0010526 " Outage Schedule
' Safety Review", has mixed results for up-front outage plannmg. The procedure .
l l
I
/%
. (d1 AUDIT REPORT QSL-MR-96-14 Page 15 of17 contains well thought-out questions for preserving safety functions. including potential challenges to those safety functions, but ais adent on the mechanism by which risk is evaluated einsungtina,=bainandan outage. Presently the ORAM (Otarge Risk' Assessment andManapmmt) enmputer program is used but not mentioned.
The crocess breaks down when viewed across the olant orcanization since different croues areusinn different information sources for assessine risk assessine the effect on safety functions. As stated. outage planning / management uses ORAM for up-front planning. Twice daily during an outage the STA uses a form titled " Shutdown Safety AssessmentV (from ADM 17.01) to assess relative risk while Operations uses approved plant procedures to implicitly address minimization of risk. The following aspects of this situation are noteworthy.
e The logic in ORAM for evaluating risk is different from the logic used in the Shutdown Safety Assessment. (CR 96-1196, written 6/1/96 by the Maintenance Rule Administrator to address this concern).
e The Shutdown Safety Assessment is not normally available, or used, in the control room by Operations personnel although the results are prominently displayed in the Outage meeting room.
Recommendations:
The following recommendations are offered to aid you in responding to the finding.
However, additional or altemate actions may be necessary based upon your investigation of the finding and the causal factors and generic implications identified.
Your response must address each of the tive elements in the audit cover letter.
- 1. Establish a coherent plant method to meet the industrv standard of risk associated with all modes of plant operation.
AUDIT REPORT Fgat., QSL-MR-%14 '
Page 16 or,17 ;
l
~
Audit Participants Name Department / Group A H _C !
R. B all Maintenance x L. Bladow Quality Assurance Manager x D. Brown Ops. Support x J. Cimmino SCE .x L. Croteau Training x B. Czachor , Ops. Support x R. De La Espriella ' QA Supervisor - ITR x D. Denver Engineering x T. Dillard Management x x P. Fulford OST x D. English Maintenance x T. Glenn I&C x C. Guey RRAG x B. Hagar Outage Mgmt x E. Hollowell Nuclear Engineering x D. Howard ICM x H. Johnson Operations Manager x W. Korte Electrical Maintenance x T. Marvin Outage Mgmt. x A. Pell Outage Manager x J. Poner SCE x J. Price SCE x G. Pustover Maintenance x J. Scarola Plant General Manager x x M. Snyder SCE x x x J. Stall Site VP x D. Stewart SCE x B. Vincent Nuclear Engineering x x x B. Walcheski SCE x x x E. Weinkam Licensing x R. Weller Ops. Support x J. West Outage C. Wood SCE x x B. Young SCE x C. Rossi . PTN QA x D. Lowens PSL QA x x
h' ))
( ':: j AUDIT REPORT
'Q QSL-MR-96-14 Page 17 of17 Kat A- Pre Audit Conference B- Interviewed or Contacted During Audit C- Attended Post-Audit Cor.ference 2
References:
10 CFR 50.65 Regulatory Guide 1.160 Rev.I NUMARC 93-01 Rev.2 3 NRC Inspection Procedure 62706 - 9/12/95 NUREG 1526 June 1995 ADM 17.08 Rev.4 ADM 17.01 Rev.15 AP 0010526 Rev.O Pre-Audit Conference: Post-Audit Conference Location: St. Lucie Location: St. Lucie Date: July 8.1996 Date: July 23.1996 i
Summary of Post Audit Conference 3e results of the audit were discussed. Responsibility for the Finding was assigned to the Site Engineering Manager PSL by the Vice President - PSL Accompanying.
.\uditor: C.V. Rossi PTN Quality Assurance Supervisor Prepared by:
D.C. Lowens, Principal Auditor Date,
b T
1 t
O i
e b
.i I
l I
1 1
i i
1 I
I 1
1 J
t i
1
ww
- l. Maintenance Rule Category (c)(1) Status l 8/27/96 System, Structure, Reason for (a)(1) status Action Plan . Due Responsible Goal (s)
Cornponent (SSC) Individual U1 & U2 Repetitive MPFFs resutting in 1. Procedure development for U1/U2 Complete R. Kulavich 1. No U1 fadures of gov, actuators
.WDiesel Generator failure to carry load Cause due to require periodic overhaul. due to aging thru 12/96.
governors to inadequate PM on govemors. 2. Perform overhaul for U2 Complete R. Kutavich 1 EM . 2. No U2 failures of gov. actuators 3 Perform overhaal for U1 Complete R. Kufavich 1 EM due to aging thru 6/97.
- 4. No fadures of govemor and R. Kufavich 3. Gov. control stable load while stable load control. operating thru 12N 102 6mergency Desel Unavailability and reliabdity has 1. Individual problems were venfied Complete R. Kufavich 1.12 month roihng total of OOS hrs Generator exceeded performance entena to have addressed root cause to trend to less than 240 by 12/31/96.
of OOS hrs and SBO tngger values 2. For any additional start failures or 12/31/96 R. Kulavich 2. Less than 5 start failures per 100 OOS hrs. an ERT will be initiated. . starts demands by 12/31M U1 & U2 Repetitive MPFFs due to 1. Revise PM procedure for 4.16 kv Complete J. Campbell 1. By 12/31/97, s 2 PMT faifures Ut& 2 4.16 kw AC safety related floor tnpper and latch check switch breakers to improve reliability. units due to fir tripper or latch check.
Breakers fadures, which can be addressed 2. Use PSA to prioritize PM schedule. Complete J. Campbell 2. No demand failures by 12/31197.
by improved PM for breakers. 3. Notify Training to revew floor inpper 12/31/96 R. Raldiris 3. All 4.16 bkrs PM'd for floor inppers and latch check problems w/EM. and latch check switch by 12/31/97.
U1 PORVs A MPFF resu'ted in PORV 1. Improve testing steps in 1-M-0037. - 10/30/96 T. Sanders 1. Two SAT bench PMTs, and ISTs unavailability exceeding following next Ut PORV rebuild.
their performance critena. 2. Improve testag to ensure soleniod 8/30/96 M.Snyder 2. No inservice circuit failures during IST.
& circuitry SAT prior to LTOP. (PMAl to I&C)
U1 'C' AFW train Train Reliability has exceeded 1. Upgrade the SMB-000 actuator PM Complete J. Cook 1. No SMB-000 actuator PMT contactor performance criteria for to include as found/as left resistance failures after new PM impkmiented for Maintenance Preventable check to ensure proper contact cleaning 18 month period.
Functional Faitures. 2. Upgrade the U1 EGR PM to include 9/6/96 M. Wolaver 2. No as found out of spec EGR resistance check for proper inspection. resistance checks for 18 months.
Ut & U2 Reactor Coolant Unplanned U1 unavailability 1. Revise SU seal changeout PM 12/31/96 M. Snyder 1. RCP seat changeouts every other Pump seals exceeded in 1995,in part, due to frequency to no more than 2 cycles- refueling cycle.
faded 1A2 RCP seat 2. SCE will evaluate 2nd cycle use of Ongoing B Kelfy 2. Monthly trend of all RCP seal SU seals at each refueling downpower. stage pressure breakdowns to show
- 3. Submit REA for potential upgrade to Complete B. Kelly improved performance with shorter runs more rugged N9000 seal design. for 2 refuelings, both units.
Categgy_(3J(D - Maintenance Rute Systems, Structures, and Components (SSCs) determined to have UNACCEPTABLE performance based on Maintenance Ruie performance monitoring l (s)(1) refers to the apolicable paragraph in 10CFR50 65.
MEff (Maintenance Preventable Functonal Failure)- Failure of a SSC to perform its intended functon that should have been prevented by the performance of appropriate maintenance actens.
l Structure monitoring is an important part of Maintenance Rule performance monitoring Monitoring of plant structures is done by all employees. It is done during operation rounds, treterial co" %n inspections, system engineer walkdowns, etc. Degraded structural performance -hould be identified on a Work Request and/or Conditon Report.
t
SSCs in a(2) Classification
{
Risk Sia? Sy310m System & Comoonents Trains & Comoonents not in a(2) {.
Y 1a Fuel and CEAs Y 1b Reactor Coolant System RCP seals, Unit 1 PORVs Y 2 CVCS Y 3a HPSI Y 3b LPSI- -
Y 3c Safety injection Tanks Y 7 Containment Spray Y 8 Main Steam Y 9a Main Feedwater Y 9b Auxiliary Feedwater 1C AFWtrain Y 9c AFAS Y 14 CCW Y 17 b EDG Fuel Oil Y -18 b instrument Air ,
Y 21 a ICW Y 25 A HVAC-Risk Significant Y 47 A 480 VAC swgr & bkrs Y 48 120G08 vac Y 49 120 Vital VAC inverters Y 50 125 VDC & chargers Y 52 4.16kv swgr & bkrs vital 4.16kv breakers Y 53 Generation & Distribution Y 59 EDGs 182 EDGs, All EDG govemors Y 63 RPS Y 68 Containment Penetrations Y 69 ESFAS & Annunciators N 4 FuelPool N 6 Waste Managment N 10 Extraction Steam N 11 , Heater Drains and Vents N 12 a Condensate N 13 Turbine Cooling Water N 17 a Turbine Lube Oil N 19 Condensate Polishing N 21 b Circulating Water N 22 Turbine N 23 a SGBD Rad Monitoring N 25 B HVAC - Non Risk N 25 n ECCS Drains N 26 Radiation Monitoring N 27 a Hydrogen Analyzer N 27 b Hydrogen Recombiners N 37 Ultimate Heat Sink Valves N 46 6.9 kv swgr & bkrs N 47 B 480 VAC swgr & bkrs N 60 Station Grounding N 62 Reactor Regulating N 64 Nuclear Instrumentation N 66 CEDM N ' 67 Fuel Handling
. N 70 QSPDS N 73 Structures N 74 Polar Crane N 75 Cathodic Protection