ML20137K929
| ML20137K929 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 04/01/1996 |
| From: | Lowens D, Voorhees J FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17354B293 | List:
|
| References | |
| FOIA-96-485 QSL-CA-96-05, QSL-CA-96-5, NUDOCS 9704070124 | |
| Download: ML20137K929 (199) | |
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, FPL Nuclear Division FPL Quality Assurance Audit Report t
STA'R/PMAI CONVERSION PROCESS I
- QSL-CA-96-05 l
i I i Auditor: 1 D.C. Lowens
)
e i L. W. Bladow QA PSL l 9704070124 970325 PDR FOIA BINDER 96-485 PDR
I l 'O AUDIT REPORT p QSL-CA-96-05 1 age 2 of 16 i Executive I Smnman: The St. Lucie plant implemented a Plant Manager Action item (PMAI) process in the beginning of 1996. The PMAI process was used to absorb outstanding action items from approximately 1000 open STARS that were ! closed during the period of January - March 1996. The PSL Quality Assurance Department performed an audit of the conversion process in late March. The audit discovered problems with the PMAI procedure,'the implementation of the PMAI process, and with the method being used to close existing St. Lucie Action Reports (STARS). These problems contributed towards the delay cf necessary corrective actions on technical issues in several cases discovered by the audit. Findings Finding 1 discusses procedure content and implementation problems relating to AP 0006129 "PMAl Corrective Action Tracking Program" (See Pg.7). Finding 2 discusses the fact that the Corrective Action Department closed STARS without concurrence of the initiating department in cases where initiating department concurrence was required (See Pg.10). Finding 3 discusses three technical issues requiring immediate attention associated with items that were converted from STARS to PMAls (See Pg.12). Recommendations Prepare quality records system for PMAI entry (Page 5) Reduce backlog of unfilmed STARS (Page 5) , Update CR procedure to reflect approval authority (Page 5) Ensure CR supplement will be generated for PMAls involving extended root cause analysis. (Page 5) Based on the activities and objective evidence examined, it was determined that the requirements of the QA program and technical specifications are not adequately addressed by the procedure reviewed j and that ' implementation of the procedure is not effective. Prompt corrective action, as discussed with the St. Lucie Business Manager and initiated during the audit, is necessary to rectify the identified problems. l l l
i l l t d AUDIT REPORT o - E QSL-CA-96-05 Page 3 of 16 i l Location of Audit St. Lucie Plant Date of Audit March 12 - 29,1996 Audit Scope This audit was conducted to evaluate use of the Plant Manager's Action Item (PMAI) list at St. Lucie, and conversion of outstanding actions associated with St. Lucie Action Requests (STARS) ; to the PMAI list. Audit Details Administrative Procedure 0006129,"PMAl Corrective Action Tracking Program", was issued on 1 January 16, 1996. The procedure is similar to one in use at Turkey Point, and was intended I create a system to track action items on a plant wide basis. The PMAI is designed to track action actions deriving from the completion of evaluations associated with Condition Reports and other requests generated by the Plant General Manager. At the time that the PMAI procedure was introauced there were approximately 1600 open STARS in the St. Lucie system. To facilitate the goal of a smooth transition to a uniform Nuclear Division Condition Report process at the beginning of April, the Corrective Action (CA) l Department was given direction to make use of the PMAI process to reduce the number of open , STARS. I The PMAI process was new to St. Lucie, and St. Lucie personnel were not familiar with its steps. The procedure itselfis brief, and deals primarily with administrative details by which PMAls are processed within the CA Department. Little information is provided in the procedure about interface processes to be used by departments that receive and disposition PMAls. The procedure also contains several internal inconsistencies. Shortly after the issuance of the PMAI procedure, the process was initiated to track corrective actions deriving from the closure of STARS and requests from other sources. No training on the process was provided to plant personnel. The lack of detail in the PMAI procedure, combined with unfamiliarity on the part of personnel responsible for its use, caused confusion and frustration on the part of personnel responsible for responding to the action items issued by the process. Details on the problems with the PMAI procedbre and plant process are provided in Finding #1. The St. Lucie STAR process contains a feature which permits the originator of a STAR to specify that his or her concurrence signature is necessary prior to closure of a STAR. When the effort to convert large numbers of open STARS in to PMAls began, the CA Department adopted a method of operation by which the requirement to obtain initiator concurrence signature, on
l l l AUDIT REPORT QSL-CA-96-05 Page 4 of 16 l STARS where it was required, was waived. Non-compliance with the applicable requirement in QI-16-PR/PSL-2, "St. Lucie Action Request Process", is documented in Finding #2. ! I On March 12, 1995, a quality assurance audit of the STAIUPMAI conversion process was ! initiated. As part of the audit process, personnel in CA Department and other plant departments ! were interviewed on all aspects of the PMAI process. A selected sample of quality records I associated with STARS that had been closed / converted to PMAls was also examined. During this process, several technical issues that require further attention were discovered. These issues are documented in Finding #3. In addition to the problems documented in the Findings, several overall observations were generated by the audit process. These are listed below l
- 1. AP 0006129 requires that all PMAI corrective action records related to STARS become Quality Assurance Records. At the time that the audit was performed, no PMAls had been submitted to the quality records vault for entry into the records system, and the St. Lucie Quality Records Index had not been modified to recognize the creation of l PMAls. A method to distinguish between PMAls related to STARS and those not ;
l related to STARS had not been developed. It is recommended that these steps be I taken before the backlog of PMAls that requires processing into the records system ; becomes too large. When setting up a record category for PMAls in the Quality ' Records Index, it is recommended that both the PMAI number and the associated STAR /CR number be established as indexed fields to facilitate future retrieval efforts.
- 2. At the time that the audit was performed, a backlog of approximately five months of STARS was awaiting entry into the quality records system in the St. Lucie vault. This is a large backlog to be in existence prior to an outage. It is noted that during the audit an effort to reduce the backlog was initiated by records vault personnel.
- 3. During the audit it was noted that both the initiation and closure signature for the Plant General Manager on a majority of recent STARS, had been performed by l Corrective Action Department personnel on a "by direction" basis. If this function is
! to be one which is permanently assigned to the Corrective Action Department, it is ! recommended that AP 0006130, " Condition Reports', the successor to the STAR procedure, be modified to reflect the actual execution of the process.
- 4. During the audit, concern was expressed by several parties about " loop closure" on issues processed by the CR/PMAI process. The concern centered around cases in which PMAI actions disclosed the need for further review and/or another iteration of the investigation / analysis process. In this connection it is noted that at Turkey Point, if a PMAl is issued for an extended action associated with root cause analysis (e.g.
vendor disassembly of a valve), the PMAI includes a requirement for a CR supplement to be issued upon receipt of the applicable information. It is recommended that this
AUDIT llEPORT NL QSL-CA-96-05 Page 5 of 16 operating method be adopted at St. Lucie. On an overall basis, the activities performed by this audit disclosed several signiDeant weaknesses in the St. Luci site corrective action process. The overall evaluation to date is that the audited elements are not effective in performing applicable functions as part of the St. Lucie quality program. Prompt corrective action, as discussed with the St. Lucie Business Manager and initiated during the audit, is necessary to rectify the identified problems. With respect to the items discussed with the St. Lucie Business Manager, the following actions had been taken or were in progress at the time that this report was prepared:
- 1. A draft revision to AP 0006129, incorporating greater detail has been prepared.
- 2. Arrangements have been made with the PSL Training Department to provide training on the uniform Condition Report and PMAI Processes.
- 3. Implementation of the unifomt Condition Report Process will eliminate the requirement for originator concurrence in corrective action.
- 4. A review of STARS that have been converted to PMAls is planned to detect other issues that may require near term corrective action.
Satisfactory Are:15 STAR Da3 base Status Activities of Departmental STAR Coordinators Completeness and Organization of Assembled STAR Record Packages PMAI Database Status Findings
- 1. AP 0006129 "PMAI Corrective Action Tracking Program", lacks instructions necessary for successful use by plant personnel. The procedure was implemented without training of affected personnel.
- 2. The Corrective Action Department has closed STARS without concurrence of the initiating department in cases where concurrence has been required.
- 3. Technical issues requiring near term corrective action are currently documented on PMAls with extended due dates.
. O L AUDIT REPORT QS L-CA-96-05 Page 6 of 16 Finding No.1 Procedure Content and Implementation Problems - AP 0006129 Criteria: FPL TQAR 5.1," Instructions, Procedures and I)rawings", Rev.10 Paragranh 5.1
" Activities affecting quality or nuclear safety-related structures, systems and components shall be prescribed by documented instructions, procedures or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures or drawings.
FPL TQAR 2.0, " Quality. Assurance Program", Rev.13 Paragranh 2.2.5
" Instructions shall also require the head of each department to be responsible for a training plan which assures that persennel performing activities affecting quality are trained in the principles and techniques of the activity being performed.. . This training shall be conducted to reflect significant procedure changes, or plant modifications which significantly affect the operation or the department" ANSI N 18.7- 1976 Quality Assurance for the Operational Phase of Nuclear Power Plante Parafranh 5.215 "Those particeting in any activity shall be made aware of, and use, proper !
and current instructions, procedures, drawings and engineering requirements for I performing the activity" 1 Finding: AP 0006129 "PMAI Corrective Action Tracking Program", lacks 1 instructions necessary for successful use by plant personnel. The I procedure was implemented without training of affected personnel. Discussion: AP 0006129 was implemented at the St. Lucie site on January 16,1996 to provide a process for timely resolution of assigned action items. As part of the transition to a uniform Condition Report process, AP 0006129 has been used to absorb large numbers of open items associated with closure of existing overdue STARS. Finding No.1 (Continued) , The corrective action process created by AP 0006129 is currently tracking i safety and quality related items that affect many plant departments. In its
. O
@b AUDIT REPORT QSL-CA-96-05 Page 7 of 16 current revision, the procedure provides instructions that relate primarily to internal processing of applicable documents (PMAI's) by the Corrective Action Department. Important detail relating to interface activities with other plant departments is missing from the procedure. Examples are provided below:
Individuals /nositions authorized to initiate PMAI's are not designated. In some cases Department Managers been assigned PMAl's of which they had no previous knowledge and with which they have a disagreement in the content , of the depanmental commitment. The source and authority for these PMAI's I is not designated by AP 0006129. - l Tvnes of items that may or may not be tracked by PMAI are not delineated. In some cases PMAl's are being used to track STAR closecut items where a root cause investigation on failure of safety related components is still , outstanding. The PMAI process does not satisfactorily address reportability I issues that may arise from such activities. Process for accentance/ rejection by imolementing department is not snecified. The PMAI form contains a signature blank for the manager of the implementing department that states the following "I understand and accept the responsibility of the above listed action & date due" AP 0006129 does not contain instructions as to the use or sequencing of this signature block (i.e. signed before or after the PMAI is issued, provisions for post is,uance rejection of the PMAI by the implementing department). In some cases Department Managers been assigned PMAl's of which they had no previous knowledge and l with which they have a disagreement in the content of the departmental commitment. Necessarv close-out nrocessing is not snecified. The PMAI form contains a signature to document review by the originator of the PMAl. AP 0006129 does not contain instructions as to the use or sequencing of this signature block (i.e. signed before or after corrective action is complete, individual who is considered the originator). l'inding No.1 (Continued) The PMAI process was implemented without training of affected plant personnel. This was responsible for confusion and frustration on the part of implementing personnel. In this connection it is noted that TQAR 2.0 contains a requirement for each department head to conduct training on significant procedure changes. Research on the implementation of this TQAR requirement
. O NL AUDIT REPORT QSL-CA-96-05 Page 8 of 16 at the St. Lucie site disclosed an absence of site level requirements or programs to provide training on administrative procedures or plant quality instructions.
Additional e:: amination of this area by the Quality Assurance Department is currently in progress In some cases, as documented in Finding #3, STARS closed to PMAls contain technical issues requiring near term action. Lack of clarity in procedural guidance concerning implementation of the PMAI process has removed one
- of the barriers that might have acted to prevent these problems from escalating l
in importance. Recommendation: Your response must address the finding above. The following recommendations are offered for your consideration.
- 1. Revise AP 0006129 to address necessary interface methods with plant departments.
- 2. Provide training to affected personnel on use of the process prescribed by AP 0006129.
e
AUDIT REPORT NL QSL-CA-96-05 Page 9 of 16 Finding No. 2 Pracedural Non-Compliance - QI 16 PIUPSL-2 Criteria: QI 5-PIUPSL-1," Preparation, Revision, Review / Approval of Procedures", Rev.68 Paragrana 5.13 " Adherence to Procedures" "A strict adherence to procedural requirements - Verbatim Compliance is the policy expected and required of S' Lucie Plant personnel." QI 16-PR/PSL-2, "St. Lucie Action Report (STAR) Program, Rev. 4 Paragraph 5.6.1 "Closecut of STARS"-
"Once all corrective measures are complete:
B. Obtain concurrence signature of initiating department if required. FPL TQAR 15.0," Nonconforming Materials Parts or Components", Rev.10
" Items shall be reworked or repaired in accordance with documented procedures and shall be verified by reinspecting the item as originally inspected or by a documented method which is at least equal to the original inspection method."
Finding: The Corrective Action Department has closed STARS without concurrence of the initiating department in cases where concurrence has been required. Discussion: The STAR process contains a provision that permits an initiating department head to require his or her concurrence in corrective action that has been accomplished for a particular STAR prior to closure of the STAR. Early in 1996, when the large scale conversion of STARS to PMAls was first planned, the Corrective Action Department prepared a change to QI 16-PR/PSL-2 that would have eliminated the requirement to obtain. concurrence signatures from initiating departments where it was required. This change was disapproved by the Quality Assurance Department on the grounds that initiating department concurrence was the method currently used to satisfy reinspection requirements contained in the Topical Quality Assurance Report. During the PMON 96-018, it was discovered that the Corrective Action Department has closed STARS being converted to PMAls without obtaining initiating department concurrence, in violation of the requirement contained
AUDIT REPORT I:PL QSL-CA-96-05 Page 10 of 16 Finding No. 2 (Continued) in the existing revision of QI 16 PIUPSL-2. In some cases, as documented in Finding #3, STARS that were closed in this f:shion contained technical issues that necessitate near term action. Closure of thee STARS without initiator concurrence has acted to remove one of the barriers that might have acted to prevent these problems from escalating in importance. Recommendation: Your response must address the findinF above. The following recommendation is offered for your consideration. l Review requirements for procedural compliance with applicable personnel.
l l AUDIT REPORT QSL-CA-96-05 l Page 11 of 16 1 Finding No. 3 Technical Issues - Overdue STAR /PMAI Conversion l i Criteria: FPL TQAR 16.0, " Corrective Action", Rev. 8 j i Paracranh 16.1
" Documented measures shall be used to assure that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment and nonconformances, are promptly identified and ,
corrected as soon as practicable." l AP 0010021, " Administration of 10 CFR21 Notifications", Revision 5 Parafraoh 8.3 l ' "Within 60 days of discovery of a Deviation, Failure to Comply or receipt of notification from a supplier of their inability to evaluate a deviation, one of the following shall be completed: I A. An evaluation to determine whether or not a deviation could create a SSH and is reportable to the NRC 1 B. An interim report submitted by the Site Vice President to the NRC which should include the following:
- 1. a description of the potential defect that is being evaluated, and
- 2. a statement of the date when the evaluation will be completed.
ASME Section XI- 1992, IWA-4150," Verification of Acceptability" j
" Prior to authorizing a repair or installation of an item to be used for ;
replacement, the Owner shall conduct an evaluation of the suitability of the I repair or item to be used for replacement including consideration of the cause of the failure. In addition, when repair or replacement is required because of the failure of an item, the evaluation shall consider the cause of failure of the existing item to ensure that the selected repair or replacement item is suitable." l l Finding: Technical issues requiring near term corrective action are currently : documented on PMAIs with extended due dates. Discussion: " STAR 950428 identified the fact that the upper stem in MV-09-13 (Motor Operated Valve for Crosstie Between lA and IB AFW Pump Discharge) on Unit I had thread damage. Investigation by engineering disclosem ' hat the
. @b AUDIT REPORT QSL-CA.96-05 Page 12 of 16 Finding No. 3 (Continued) upper stem nut in the valve had been fabricated from cast leaded red brass rather than cast manganese bronze alloy as required. Engineering disposition of the STAR called for an inspection of 14 other motor operated valves on both Unit I and Unit 2 in which stem nuts of the same description are installed.
The engineering disposition was issued in May 1995. STAR 950428 was closed to PMAl's PM96-03-099 and PM96-03-100 on March 6,1996. The first of the two PMAl's calls for 14 MOVs on Unit I to be inspected. The second PMAI calls for the results of the inspection to be returned to engineering for a final disposition of the non-conformance issue. Investigation disclosed that PM96-03-099 has not been incorporated into the Unit 1 outage schedule for Spring 1996. On July 24,1995, STAR 950794 was issued to investigate a potential 10CFR21 notification associated with mismarked AMP f/12-10 insulated lugs. The engineering disposition for the STAR required an inspection of 240 volt switchgear located in Unit 2 to detect potentially defective terminals. Numerous lugs fitting the applicable description were discovered, and the STAR was returned to engineering for preparation of an operability sampling piaa. The plan was implemented and required a report to engineering if any loose lugs were found. One loose terminal lug was found, and the required report was made to engineering. STAR 950794 was closed to PMAI PM96 047 on March 1,1996. The PMAI requires engineering to evaluate funher switchgear inspections. The STAR disposition does not document whether or not the deviation represents a substantial safety hazard. STAR l-94110358 was issued on November 6,1994 to document the fact that three Unit 1 pressurizer code safety valves failed high on set pressure testing. An engineering evaluation was performed on November '25,1994, and demonstrated that the observed set pressure deviations would not have resulted in violation of the plant safety limits of any of the analyzed FUSAR Chapter 15 events. The engineering evaluation specified that final disposition of the STAR was to be performed following completion of the root cause evaluation for the failure. STAR l-94110358 was closed to PMAI PM96-02-321 on March 5,1996. The PMAI calls for the root cause evaluation to be performed and has a completion due date of August 29,1996. Unit 1 pressurizer code safety valves are scheduled to be replaced again during the spring 1996 Unit 1 outage. The first two of the STARS above required initiator concurrence prior to closure. All three of the STARS required Plant General Manager l l
1 'c @- NL AUDIT REPORT QSL-CA-96-05 Page 13 or 16 Finding No. 3 (Continued) concurrence prior to closure. In all three cases initiator concurrence was not obtained, and Corrective Action Department personnel signed as designees for the plant General Manager, when these STARS were closed to PMAls. The individuals who were assigned corrective action for these STARS are responsible for the status discovered during the audit. However the fact that these STARS were closed without required initiator concurrence has acted to remove one of the barriers that might have acted to prevent these problems from escalating in importance. ' All three of the issues require prompt attention, both to satisfy applicable regulatory requirements, and to verify that potential technical issues have been properly addressed. This PMON reviewed only a sample of STARS that have been closed to PMAls. Additional review is necessary to verify that similar issues do not exist in the total population of STARS that have undergoine this process. Recommendation: Your response must address the finding above. The following recommendations are offered for your consideration.
- 1. Resolve the identified technical issues.
2. Perform a review of other STARS that have been closed to PMAl's to verify that technical issues have been satisfactorily addressed.
9 NL AUDIT REPORT QSL-CA-96-05 Page 14 of 16 Audit
Participants:
Name Department / Group A Il C W. Bladow Quality Manager x x K. Butler Quality Control x R. Dawson Business Manager x x x A. Locke Corrective Action x x T. Lyons Maintenance x G. Madden Licensing x J. Mann Information Services x R. Mixon Corrective Action x x A. Pell Outage Manager x L. Petrie SCE x J. Poner SCE x C. Rossi Quality Assurance /PTN x D. Sager VP/ Nuclear Assurance x J. Scarola Plant General Manager /PSL x T. Schiffley Corrective Action x J. Schorn Outage Mgmt. x M. Tarascio Nuclear Engineering x J. Valdes Operations x J. Voorhees Quality Assurance x x L. Whitwell Corrective Action x x Key: A - Attended Pre-Audit Conference / Notified of Audit B -Interviewed or Contacted During Audit C - Attended Post-Audit Conference
o @b ' AUDIT REPORT QSL-CA-96-05 Page 15 of 16
References:
ANSI N 18.7-1976 Quality Assurance for the Operational Phase of Nuclear Power Plants AP 0006129, PMAI Corrective Action Tracking Program Rev. 0 , AP 0010021 Administration of 10 CFR21 Notifications Rev. 5 FPL TQAR 16.0 Corrective Action Rev. 8 FPL TQAR 15.0 Nonconforming Materials Parts or
- Components Rev.10 FPL TQAR 5.1, Instructions, Procedures and Drawings Rev.10 FPL TQAR 2., Quality Assurance Program Rev.13 QI 5-PR/PSL-1 Preparation, Revision, Review / Approval of Procedures Rev.68 QI 16-PR/PSL-2 St. Lucie Action Report (STAR) Program Rev. 4 Sect XI-1992 IWA-4150 Verification of Acceptability l 4
, Pre-Audit Conference: Location: St. Lucie Plant ' Date: March 12,1996 > l Post-Audit Conferencet Location: St. Lucie Plant Date: March 29,1996 1 Summary of Post-Audit Conference: Audit results and fmdings were discussed. Personnel contacted during the audit were thanked for their assistance. Location of Audit: St. Lucie Plant
AUDIT REPORT 4 o ppg QSL-CA-96-05 Page 16 of 16 t
, LW hflff( \
D. L. Lowens Date Quality Assurance - PSL i Reviewed By: P sd -/-f l J. T. Voorhees Date QA Supervisor - PSL [ i l 4 I a i l l l 4 1 1 4 5 1 i
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i
4 y, Inter-Office Correspondence a' . FPL JQQ-96-048 To: J. Scarola Date: April 1,1996 From: L. W. Bladow Department: JNA/PSL
Subject:
Quality Assurance Audit OSL-CA-96 STAR /PM Al Conversion Process Attached is an audit of the STAIUPMAI conversion process performed in March 1996 The following findings are documented in this report and have been discussed with appropriate personnel and exited with PSL Plant Management. Findine 1: AP 0006129 "PMAI Corrective Action Tracking Program", lacks instructions necessary for successful use by plant personnel. The procedure was implemented without training of affected personnel. Fmding_h The Corrective Action Department has closed STARS without concurrence of the initiating department in cases where concurrence has been required. Finding _h Technical issues requiring near term corrective action are currently documented on PMAls with extended due dates. Condition Reports (CR) have been generated for the above findings. In accordance with the FPL
. Quality Assurance Program, please ensure that the CRs which address these findings are responded to within 30 days of origination. As noted in AP 0006130 responses to CRs resulting from QA audit findings must include the following:
- 1. The results of review and investigation of the findings including identification of the probable root cause/ causal factors.
- 2. Results of your examination of potential weaknesses in departmental self-assessment programs which may have impeded self-identification of the problem.
- 3. A determination of the generic impact of the finding, i.e., whether it extends to other dreas, systems, drawings, procedures, etc., or whether it is isolated to those examples cited in the audited report.
- 4. Actions taken or planned to correct the findings identified and to prevent recurrence of the deficiency.' Corrective actions should address the causal factors and enhancements to the audited department's self assessment program. .[
h
- Q .:
JQQ-96-048 April 1,1996 Page 2
- 5. Date when full corrective action was or will be achieved.
- 6. Identification of the individual (s) responsible for the corrective action.
For those corrective actions which cannot be completed within 90 days from the audit report transmittal, the response shall (1) include an explanation why the action cannot be completed within 90 days and (2) include both the cognizant Vice President (or Director where the Director is a direct report of the President - Nuclear Division) and the Vice President Nuclear Assurance on distribution. An evaluation should be made of the findings identified.in this audit to determine reportability. We sincerely appreciate the cooperation we received from your staff during the course of the audit. Please contact me at extension 7111 or the respective QA contact if you have any questions.
}f -r ('l L. W. Bladow Quality Manager - PSL LWB/dcl
- Attachment '
Copies to: Dist. Attached
1 Inter-Office Correspondence FPL I I l To: L. W. Bladow Date: April 29,1996 From: W. H. Bohlke Went: St. Lucie Plant
Subject:
QA Audit QSL-CA-96-05 ' l l i Attached is our response to the subject audit. Enclosed you will find individual responses to i the three findings. Four additional observations were offered. Our feeaoack to address these items is also I included in this package. l 1 vpps1017-96 l
- khw'5 A l
l I l
i QS L-CA-96-05 l i FINDING 1: l AP 0006129 "PMAI Corrective Action Tracking Program" lacks instructions necessa y for successful use by plant personnel. The procedure was implemented without training of affected personnel. L RESPONSE: ' 1 1. The probable root cause was that even though the process was relatively simple and in ! use at another FPL plant, we underestimated the effect on plant personnel by changing this plant business process at St. Lucie. The St. Lucie PMAI program was initiated by replicating the successful system in use at Turkey Point. AP0006129 "PMAI Corrective Action Tracking"is a virtual duplicate of the procedure ( ADM-054) in use : at Turkey Point. We duplicated the product of another plant without making sure that the users had a working knowledge and a commitment to make the process work. Due to the simplicity of the process, the Training department was not requested to provide formal training. Briefings were conducted, however, prior to putting the PMAI system into operation. On several occasions, Department Heads and other key personnel were briefed on the procedure and its use at regularly scheduled plant meetings. At those meetings, the procedure was distributed and instructions were given ! as to how to interface with the PMAI database and procedure. 1 Interviews with plant personnel identified two common problems areas. The form in use (produced by the computer) was found to be less descriptive than the one shown in the procedure and there was a section on the form that was not used nor described ; in the procedure. It was agreed that the procedure could be improved.
- 2. Several departments had given feedback that demonstrated some difficulty with the procedure. A revision to the PMAI procedure was planned.
- 3. The other procedures that were imported frorr. Turkey Point (the Condition Report procedure and the Quality Instruction on corrective action) had a detailed front end technical and management review prior to implementation.
Training was conducted on the PMAI and CR procedures during the weeks of 4/8 and 4/15 and was complete on 4/19/96. Training on Administrative procedures was looked at generically. STAR 950820 was written on 8/3/95 to address the generic topic. The response to this STAR initiated actions that would require the FRG to consider training implications while reviewing changes to Administrative procedures. 2
QSL-CA-96-05
- 4. a. The PMAI program was revised to put the necessary detail on the computer generated form.
- b. The PMAI procedure (AP 0006129) will be revised to correct the problems noted.
- 5. The procedure revision will be issued on or before 5/15/96.
- 6. Bob Dawson is responsible for this procedure revision.
l l l l 1 c f 3
QSL-CA-96-05 l FINDING 2: The Corrective Action Department has closed STARS without concurrence of the initiating j depanment in cases where concurrence has been required.
RESPONSE
- 1. We concur with the finding.
1 Conversion of STARS to the CR/PMAI system did not always obtain the concurrence l signatures required by the STAR procedure. To convert from the STAR process to the CR/PMAI process, the decision was made to substitute an independent review of the ; completed STAR in lieu of obtaining originator's concurrence. (In areas of QA findings and Licensing items add Plant Manager concurrence, the concurrence signatures were obtained.) Concurrence by the originator was not a part of the Division standard CR process. The CR relies on an independent review. A procedure change request was stibmitted to formalize this process change. The procedure change initiated to reflect the desired process change was not approved by QA. Due to pereonnel transfers, supervision was not aware that the procedure change was rejected until the conversion process was essentially complete.
- 2. The plant has no notification process for procedure changes other than the issuance of the new controlled copy of the procedure. A controlled copy of the STAR procedure was not assigned to the Corrective Action department. Had there been one, procedure changes would be issued to the Corrective Action Supervisor when the change was approved. It would have been evident that the procedure change requested was not l approved.
- 3. Procedure adherence is not a generic problem in this department. Understanding the status of procedure changes was.
- 4. The Condition Report procedure has replaced the STAR procedure. The CR process does not require the originator's concurrence.
Controlled copies of the procedures governing CR/PMAI processes and the Quality Instruction governing Corrective Action (Q116-3) are now assigned to the Corrective Action Supervisor. All revisions to these procedures will be distributed promptly. Procedural compliance requirements were discussed with section personnel.
- 5. Complete 4
t I '4 QSL-CA-96-05 s
!' FINDING #3:
4 1 Technical issues requiring near term corrective action are currently documented on PMAls with extended due dates.
RESPONSE
i 4
- 1. STARS 950428,950794, and 94110358 were reviewed for unresolved technical issues.
l ) STAR 950428 was converted to PMAls (PM96-03-99 and PM96-03-100) and assigned to EM/Jeff Cook and ENG/ Denver, respectively. The action for EM is to 1 perform hardness testing and measure thread length on the valve stem nuts of seven ' MOVs on Unit I and seven on Unit 2 during the next outages. Initially, these activities had not been scheduled for the upcoming outage on Unit I due to a failure to designate
- the PMAI as an SL1-14 item. This activity has now been scheduled for the SL1-14 3 outage on Unit I and will be scheduled for the SL2-10 outage on Unit 2. All data will then be submitted to Engineering for a Final Engineering Disposition of the technical issues associated with the STAR. Thus, no unresolved technical issues are associated
} with the closure of STAR 950428. STAR 950794 was converted to PMAI 96-03-047 and assigned to ENG/ Denver. The 4 PMAI required Engineering to evaluate further switchgear inspections. However, due 4 to an oversight during the STAR closecut process, no requirement was made for a j disposition on whether or not the deviation associated with this STAR rep- sented a i substantial safety hazard. Subsequently, a Condition Report was written (CR 96-0475) to ENG/Busch to evaluate this condition. This evaluation is scheduled for completion ! by May 3,1996. '! STAR 1-94110358 was converted to PMAI 96-02-321. The action is for Engineering } to perform a root cause analysis on the Unit 1 pressurizer code safety valves after they l-failed high on set pressure testing in November,1994. This work, while scheduled to be completed by 8/29/96 in accordance with the PMAI, should be completed by ! 5/30/96. However, the presently scheduled date is technically acceptable. The valves ! were sent offsite for a detailed analysis. The pressurizer code safeties are scheduled to i be replaced with an upgraded design in the April,1996 refueling outage. Any root i cause analysis results would not necessarily be applicable to the upgraded valves.
- 2. Approximately 1200 STARS were closed between 1/1/96 and 4/1/96. Of these, a
- significant number had remaining activities and/or corrective actions requiring transfer .
- to PMAls. The Corrective Action Group had responsibility for ensuring that STARS I were not closed, nor remaining corrective actions transferred to PMAls, without adequate intemal review to verify the adequacy of the corrective actions proposed by the responsible groups. However, due to the high number of STARS being closed, ;
4 some corrective actions, as noted above, required follow-up to ensure that the work was ! completed in a timely manner. Only one case was identified where the corrective j i active was inadequate. This finding has been reviewed with the Corrective Action I 3 5
QS L-C A-96-05 Group to fu-ther ensure that the importance of proper closure of corrective action documents is stressed and understood.
- 3. A representative sampling of NCR STARS closed during the 1st Quarter of 1996 has been reviewed for closecut adequacy. NCR STARS were chosen since this sampling would have the most significant impact with respect to operability, etc. The total number of NCR STARS closed during this time frame was 112. Of this population, 20% (or 23) were randomly chosen for review. Based upon a successful review of this sampling, this sample size will give a 95% confidence that approximately 90% of the entire population is acceptable. A detailed review of the closecut process used on these STARS found no inadequacies associated with the corrective actions completed or transferred to PMAls. Thus, the problems identified in the subject audit finding appear to be isolated to those examples provided in the report.
- 4. No additional actions are planned. The new Condition Report Tracking System has been implemented in accordance with AP 0006130. This procedure is very specific as to the closecut process to ensure that the initiating conditions are evaluated prior to closure of the Condition Report.
- 5. All corrective actions have been completed.
- 6. Not Applicable.
A h 6
.QSL-CA-96-05 0
OBSERVATIONS
- 1. Prepare Quality Records for PMAI entry The PMAI records are now cataloged in the records vault. Transfers to QA Records are now routine.
- 2. Reduce Backlog of Unfilmed STARS.
The QA Records Vault has initiated efforts to film the backlog of completed STARS.
- 3. The PGM signature was performed'by the Corrective Action Department personnel in a majority of cases.
The PGM assigned the Business Manager responsibility for reviewing each STAR and summarizing them for the management team each moming. Even though the Business Manager was signing the STARS, the PGM was fully involved with the process. The PGM reviewed each one and made changes to the recommended assignments and due dates. Since convening to the Condition Report process, the PGM personally assigns each CR and reviews them for closure.
- 4. Some PMAls require further review or analysis.(ie. vendor disassembly of.a component for root cause)
A requirement to initiate another CR or CR supplement will be placed in the revised PMAI procedure if root cause analysis results demonstrate the need for further review. I l l 7
6 l e AUDIT REPORT ppg. QSL-PM-96-22 Page 9 of 30 Based upon the information reviewed, implementation of the In-plant Radiciodine Monitoring Program is evaluated as satisfactory. I I Performance Monitor : J. J. Walls
)
l Maintenance: PMON 96-064 was initiated to verify site compliance with AP 0005753, Rev.18, section 8.4," Cold Weather Preparations and Precautions" The absence of actual cold weather conditions prevented a walkdown to verify the implementation of the following checklists: i l Checklist 17, " Operations Pre-Cold Weather Checklist" { Checklist 18, " Operations Cold Weather Checklist" ' Checklist 19, " Maintenance Department Pre-Cold Weather Checklist" Checklist 20, "I&C Cold Weather Checklist" In the absence of a walkdown to verify actual implementation, this audit concentrated on verifying the availability of freeze protection equipment, the adequacy of procedural guidance and conective actions for issues identified in the previous examination of this area (PMON 95-076). i l Corrective Action Review j STAR 96-0128 - This STAR identified that Data Sheet 3, " Temporary Hose Installation / l Restoration", of AP 0005753 had not been filled out in accordance with Checklist 19," Maintenance Department Pre-Cold Weather Checklist". When this discrepancy was identified, the hose installation process was a joint effort between Operations and Maintenance personnel. The Maintenance personnel performing the hose installation did not know where Data Sheet 3 was located. Corrective action for the STAR generated a Procedure Change Request (PCR) which combined Data Sheet 3, and Data Sheet 2," Valve Position"into a single Data Sheet. Revision 18 of this procedure contains Data Sheet 2 which incorporates the required char.ges. STAR 96-0326 - This STAR was generated by a walkdown that identified the absence of heat tape cm instrumentation listed in Checklist 20 of AP 0005753. Revision 15 of this procedure specifically instructed personnel to install heat tape and gave no provisions to bypass this step for instruments that are already protected by insulation. A PCR was generated and the checklist was modified to read," Verify insulation is installed or install heat tape for cold weather preparation for each unit." l Revision 18 of this procedure contains the modified checklist. i
l AUDIT REPG J FPL, QSL-PM-96-22 Page 10 of 30 Eauipment Storage. Condition and Availability: Per procedure, eleven kerosene heaters are required for deployment during cold weather conditions. Twelve heaters are stored in the F4 warehouse and, in general, appear to be in good condition. No preventive maintenance is accomplished between usage periods. Problems are identified and corrected when heaters are tumed in or, as an altemative, the heater is tagged out for repair at a later date. All heater fuel (kerosene), is purchased PC3 (Quality Related), because of an alternate potential use ' as a cleaning solvent on safety related equipment. Nuclear Materials Management (NMM) maintains a minimum of 2 drums and a maximum of 4 drums assigned to the Stores Department on site. At the time that this PMON was performed there were three untapped 55 gallon drums in Stores. One other drum is set up and tapped for use. There are 2 additional drums marked NIS, "Not In Stores". These drums are signed out to the Maintenance Department and can consist of kerosene drained out of the heaters prior to their storage in the warehouse. The kerosene in these drums is available for use in the haaters, as is the designated PC3' kerosene. It whs noted that Checklist 19 requires Maintenance to verify heater fuel supply and make arrangements through NMM for adc'itional supplies if necessary. By procedure, this is usually accomplished 16 to 24 hours prior to the onset of cold weather conditions. I&C maintains extension cords, drop lights and heat tape specifically dedicated for use in the freeze protection plan. At the present time they also maintain approximately 70 new 4 foot heat tapes in excess of those required by Checklist 20,"I&C Cold Weather Checklist", in case of additional need J or as potential spares for the stored, previously used material. Interviews revealed that, in the absence of procedural guidance, personnel were operating on their own initiative to maintain an adequate stock ofmaterial to implement the requirements ofChecklist 20. I&C had previous plans in place to pre-identify storage locations for power sources and the lengths of extension cord needed l for each item designated in the Checklist. However, the two personnel who were assigned this task are no longer employed by FPL. This was a very good idea. It would shorten the time required to deploy the freeze protection equipment and considerably maximize personnel time and effort. It is recommended that work should proceed in this effort. l AP 0005753, Rev.18, Checklist 17," Operations Pre-Cold Weather Checklist", directs operations l to station electric space heaters in the C AFW pump area to maintain AFW pump oil temperature at or above 70 degrees F. Interviews with F4 warehouse personnel revealed they were aware of three electric, space heaters, of which one was broken and two were checked out. Personnel indicated there is no procedural guidance specifying a set number of heaters to be dedicated for freeze protection deployment. Personnel also indicated that they were in the process of performing a complete warehouse inventory which would probably identify additional space heaters available for use.
AUDIT REPORT p:PL QSL-PM-96-22 Page 11 of 30
Conclusion:
The lack of procedural guidance delineating required inventory and accountability requirements, limits the plants ability to effectively organize and prepare for extremely cold weather conditions. It also interjects the possibility of limiting conditions, (i.e.., lack of material), that could
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conceivably hinder plant i forts in locating and deploying its freeze protecH.,n equipment. This problem is discussed further in Finding 1 (see below). Performance Monitor: B. Lowery Services / Engineering: PMON 96-044 examined the closure process for a selected sample of Condition Reports and associated PMAl's generated to implement and track corrective action completion. One STAR, four Condition Reports and numerous Change Request Notices,(CRN's), Document Change Requests', (DCR's) and PMAI's were evaluated. Star 960210 was written to review NRC discrepancies idemified during a system walkdown of the Unit 1 Instrument Air System, and to provide corrective actions for the NRC noted items. This STAR generated two CRN's, one DCR, one OPS Procedure Change Request,(PCR) and several PMAl's. An examination of a sampling of these documents and affected procedures and drawings did not identify any problems in the corrective actions completed. Four Condition Reports and twelv: resulting PMAl's were also reviewed. All corrective actions taken were identifiable and traceable. All actions associated with changes to procedures, procedure development or work to be performed were verified to be complete. Condition Report 96-551 identified the following condition:
" Unit 1 RAB pipe penetration room to RCB purge plenum fire barrier was breached by 2, 10 inch unattended holes, which were created to implement PCM 01-196. These holes bring into question the design basis operability and cornpliance of the fire suppression system and the ECCS exhaust ventilation system. Reference IHE 96-034."
i This C,R along with the accompanying engineering evaluation generated 7 PMAl's to track the l necessary corrective actions. PMAI PM96-05-122 was initiated and assigned to Operations to review the circumstances of this breach approval by the ANPS for procedure non-compliance. Attachment I of this PMAl effectively answered the condition addressed. However, the last paragraph of the response stated the following: l
f. AUDIT REPORT FPL QSL-PM-96-22 Page 20 of 30 11 Findings: Finding 1: Site procedures do not adequately address the storage, accountability, and inventory requirements for required freeze protection equipment. Criteria: TQR 5.0," Instructions, Procedures and Drawings", Revision 11 Paragranh 5.1 Activities affecting quality of nuclear safety-related structures, systems, and components shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. These documents shall include appropriate quantitative criteria such as dimensions, tolerances, and operating limits, and qualitative criteria such as comparative workmanship samples, to assure that the quality assurance activity has been satisfactorily accomplished. Discussion: AP 0005753, " Severe Weather Preparations", Revision 18, paragraph 8.4, and associated checklists, delineate steps to be taken in the deployment of freeze proteuion equipment at the St Lucie Plant. The checklists are speci6c in that they contain detailed requirements conceming the type of freeze protection equipment to be used and the locations at which it is to be installed. The procedure states that the freeze protection plan should be implemented 16 to 24 hours in advance of the onset of cold weather conditions. Checklist 19,
" Maintenance Department Pre-Cold Weather Checklist", directs ie Mechanical Maintenance department to ensure that an adequate supply of tarps /herculite, space heaters, and heater fuel is available, and to preposition electric / kerosene heaters in their designated positions. Checklist 20,"I&C Cold Weather Checklist", instructs I&C to install heat tape, if necessary, on equipment identined by the checklist.
Interviews with personnel in the F4 warehouse revealed a lack of procedural direction in the inventory requirements pertinent to equipment required for implementation of the freeze protection plan. At the time of the interview there were no electric space heaters available for check out in the warehouse. Of the three heaters identi6ed, two were checked out and one was awaiting repair. Personnel also indicated that a complete inventory of the F4 warehouse was planned which would probably locate several heaters presently unaccounted for. i I I i
. ~ .
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, AUDIT REPORT FPL QSL-PM-96-22 Page 21 of 30 Additional interviews with I&C department supervision also revealed a lack of procedural direction in inventory requirements. Steps had been taken since the last audit to acquire a significant number of spare heat tapes. Approximately 70 new spares were on hand. I&C also indicated that personnel had been in the process of identifying the power sources and the length of extension cord needed for each transmitter location. Their intent was to proceduralize the process to make it both easier and simpler to implement Checklist 20. This however had not been completed. The two personnel that ~ere assigned this task are no longer with FPL, and due to manpower and time constraints the task was not considered a high priority item.
Nuclear Materials Management, (NMM), supervision also indicated that all kerosene on site was purchased PC3 due to its' availability for use as a cleaning solvent on safety related equipment. Personnel also indicated that there had been problems in the past in the time of delivery due to this PC3 designation. This procedure inherently relies on several individuals to maintain significant quantities of material and equipment on hand in a readiness condition. With the Plants', departments and personnel structure in a constant state of flux, the continuity required for unproceduralized programs does not exist and can substantially degrade the efficiency of the overall program. As was demonstrated in 1989, the onset of extremely cold weather, without having completed the required preparations, has the potential to both cause undesirable operating transients and to affect the operability of safety related equipment. Recommendation: 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 five elements in the audit cover letter. The intent of the suggestions given below is to improve the overall plan efficiency by maximizing personnel time and effort. This can be accomplished if equipment is staged and readily available for use.
j AUDIT REPORT FPL. QSL-FM-96-22 Page 22 of 30
- 1. Complete work started by I&C in identifying power locations and extension cord requirements for all transmitters.
- 2. Proceduralize and dedicate inventory needed for implementation of the Freeze Protection Plan.
2
- 3. Proceduralize personnel or position accountability.
- 4. Consider a threshold for initiating the purchase of PC4 kerosene during cold weather conditions when on site supply falls below minimum.
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. _ . - . - ~ . - . . _ - . - - - . . . _ - . - . - . . -- .. - - . - . . . . - - -
? W r Daily Ouality Report Monday. January 6.1997 A. QC surveilled NPWO 3565 for IC work on HVE-13 HEPA filter D/P Indicating Switch. ; Results were satisfactory. j Performance Monitor: D. Gingras j Tuesday. January 7.1997 ! [ A. Site freeze protection plan: QA Interviewed personnel and performed a partial walk-down ; of stores to verify that freeze protection material was on hand and available for deployment j during potentially freezing temperatures expected late in the week. j I&C indicated that the procedure to identify power scur e locations and extension cord length requirements had not yet been completed. There was however, sufficient material on i hand to support the implementation of checklist # 20,"I&C Cold Weather Checklist" , Personnel in the F4 Warehouse indicated that they were in the process of ordering 4 electric space heaters which will be used exclusively for freeze protection. The heaters are expected < to be here by the end of the week. In addition, approximately 4 weeks ago, there were four ; 55 gal drums of kerosene in stores, which is an adequate supply for kerosene space heaters. , Personnel in the G2 warehouse will re-verify the quantity of kerosene on hand tonight. 1 Five tarps (10 x 12) were on hand and stored in the same location as the kerosene heaters, and one 10 X 12 tarp was in the G2 Warehouse. However, personnel indicated that it takes approximately 15 tarps to adequately cover the areas susceptible to freezing. NMM needs to ensure that the additional tarps are ordered and received by Friday, when temperatures are expected to approach freezing.. Performance Monitor: Joey Lowery B. During review of NPWO 8542, W.O. 96020697, QC noted that the required QI-l1-2 Att. A was not included in the NPWO. QC generated a CR to document that this required . documentation is missing. ! Performance Monitor: R. Miller i C. . As a follow up to the recent contact testing problem in breaker testing, QC suneilled the ) Quarterly PM of U-l TCB-1 and completed the applicable hold points of MP 092071. Results were satisfactory. Performance Monitor: T. Silvio i
Daily _ Quality Repod Tuesday,Janunry 7; 1997 A. AFAR Tecting: QA observed the performance of AFAS Relay Acutation Testing by I&C, with Operations support. There was good preparation for this infrequently-performed, load-threatening surveillance. The tailboard session held by the ANPS emphasized the potential for a plant trip, planned contigencies, and tiie need for good three way communciations. Communications between I&C and Operations personnel were excellent, and the procedures were followed verbatim. Operator log entries were appropriate. QA noted no discrepancies in the completed procedure and work order, and observed a conservative approach to plant operati'ons. Performance Monitor: L. W. Neely _W.ednesday Januaryl._1991 A. Free 7e Protection: Conducted a follow up on the Freeze completed yesterday. Bill White of NMM 20' Locateh_ X 50' sheets of Protection Herculite in Equipm Warehouse 3, location 2f37-123-000. He also verified 3 full in stock 55 gal drums of kerosene and 2 partials. His recommendation is that 4 additional drums of kerosene be ordered to support any ficc:e contingency. He also stated that Art Neuberger would expedite the delhery of addibonal tarps during the day shift today. Tarps can also be l found in the south forty Conex boxes. Points of contact or BR Nichols or Dan Hickory. Andy Flowers informed me this morning that the tarps on hand would be sufficient and that additionallupplies had been identified in different_d_epartments on site. He also stated I that he would insure that the additional kerosene was placed _oJLorder. l Performance Monitor: Joey Lowery l l l
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PSL NUCLEAR ASSURANCE unit: 02 e r 3ef 9 ! QUALITY REPORT DATE: 11/20/96 4 PAGE: 1 OF 02 ; m EVALUATOR: L.W.NEELY/B.M. PARKS SIGNATURE:
SUBJECT:
UNIT 02 CCW HX CRITICAL MAINTENANCE MANAGEMENT REVIEWWITH RESPECT TO l THE UNIT 02 FUSAR l RESULTS: UNSAT DEPARTMENT: ENG PROCEDURE: OP 2-0640020 Rev. 31 WO NUMBER: N/A PCM NUMBER: N/A I ACTIVITY TYPE: REVIEW I PRIMARY FUNCT. AREA: DESIGN /CONFIG. CONTROL PRIMARY PROG. AREA: DESIGN CONTROL i SECONDARY FUNCT. AREA:N/A SECONDARY PROG. AREA:N/A BRIEF DESCRIPTION : Observed cleaning of the 2A CCW HX. Considerable mud / silt was observed being flushed out of the 2A CCW HX . In addition, shells were observed covering approximately 30% of the inlet ' tubesheet. This was reviewed against the Unit 02 FSAR. The Natural Phenomena section of the FSAR indicates that the plant is operated such that silt will not accumulate in the tubes. The observed condition of the i heat exchanger is contrary to the FSAR. ICW/CCW operability curves may not account for the sitt in the tubes. 4 POTENTIAL CAUSE: N/A DISPOSITION: CONDITION REPORT
REFERENCE:
CR 96-2853 INSPECTION PLAN / CRITERIA CONDITIONS NOTED
, PLAN: Observe the Unit 02 CCW Heat Exchanger
- cleaning. Review the as found condition of the heat 1. UNSAT- The 2A CCW HX and 2B CCW HX had silt
! exchangers with respect to the Unit 02 FSAR. Review being flushed out of the tubes. This appears to be the operation of the CCW HX's as described in the inconsistent with the FSAR, which describes that silt FSAR with respect to the plant operating procedures. will not accumulate in the tubes. Additionally, ICW system operability curves do not appear to account for Criteria: the potential loss of heat transfer capability due to the
- 1. Unit 02 F.SAR Sections 9.2.1.3.2 & 9.2.1.3.3 & accumulation of silt in the heat exchanger tubes or silt 10.4.5 in solution. This is further described in Section 4.
,~
- 2. OP 2-0640020 Rev. 31," INTAKE COOLING 2. UNSAT - The strainers for the 2A and 2B CCW Heat WATER SYSTEM OPERATION" Exchangers are not operated as described in the FSAR. It appears that there is an error in the FSAR.
This is further described in Section 5.
- 3. UNSAT - The 2A CCW HX had shells covering approximately 30% of the inlet tubesheet, causing a reduction in flow through the heat exchanger. The 2B CCW HX had shells on the tubesheet as well. The FSAR desenbes operation of the CCW HX (strainer, hypochlorite) is such that shells will not get to the heat exchanger. QA observed seaworms and various bi-valve organisms growing in situ, which is indicative of less than effective hypochlorite injection.
s. C ' ' REVIEWED BY: DATE: y
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/ i cn E.-2ts3 Pay 4.f 9
/ PSL QUALITY REPORT ORNO 96-0007 Continuation Sheet Page 2 of 2
- 4. Silt Accumulation in the 2A & 2B CCW HX Tubes i
During the 2A & 2B CCW HX cleaning on 11/18/96,11/19/96, and 11/20/96, considerable sitt was observed being removed, indicative that it was not being scoured from the tubes during normal operation. Unit 02 FSAR Section 9.2.1.3.3, Natural Phenomena, describes, in part, that *High water velocities in the ICW system scours sitt from the tubes." This section also describes that the heat removal capability of a water / silt composition is less than water alone, and that the reduction in ; heat transfer is not expected to be more than seven to ten percent. QA noted that, although there is no heat exchanger performance testing currently being performed to determine the heat removal capability of the CCW heat exchangers, there are ICW system operability curves that are a function i I of heat exchanger and strainer differential pressure. However, the current ICW operability curves do not appear to take into account any loss of heat transfer capability due to sitting in the l water or silt accumulating in the tubes. I JPN is requested to evaluate wiiether the loss of heat transfer capability due to sitting in the water or silt accumulating in the tubes should be accounted for in the ICW operability curves as part of the , response to Condition Report 96-2BC3. <
- 5. CCW HX Inlet Strainer Ooeration l
Prior to the 2A CCW HX cleaning, QA observed that approximately 30% of the inlet tubesheet was covered by shells. This is indicative of shells bypassing the strainer in their path to the heat exchanger. In addition, various clams and marine organisms were observed growing in the heat exchanger. This is indicative of less than effective HYPOCHLORITE injection to minimize biofouling. Unit 02 UFSAR Section 9.2.1.3.3, Natural Phenomena, describes operation of the ICW system with suspended matter. Paragraph b) states,in part, "The CCW heat exchangers are protected from larger suspended particles that may be present in the intake structure by the use of basket strainers as shown on Figure 9.2-1. The basket strainers are designed to remove all suspended particles greater than 1/4 inch (6.36 mm) and are designed to seismic Category 1, Quality Group C requirements. The strainers are designed for and operated under continuous backwash during normal operation.' This is an apparent error in the FSAR. The strainers are not normally operated under continuous backwash. During reviews of OP-2-0640020, Rev. 31, INTAKE COOLING WATER SYSTEM OPERATION, QA found that the strainer is not normally operated under continuous backtvash. For example, the procedure requires the normal position of SB-21189 and SB-21188, the 2A CCW Heat Exchanger Strainer Flush Supply valves, to be closed. in addition, SB-21334, the 2A CC'N Hx Strainer Drain valve, is required to be closed. Similar valve lineups are required for the *B* side components. j i JPN is requested to evaluate the discrepancy between the description of the CCW HX strainer operation in the FSAR and the normal operating procedure as part of the respose to Condition j Report 96-2853.
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- 3. RITIA R. DE LA ESPRIELLA f R. WAlrIIESKI C. NORRIS J. WALLS K. BUTLER L DEARROR T. gel $ SINGER J. IDWERY R. WEIS S. MEAD ** K. CROSBY S. SANDERS *. R DARRESS W. IIAYES
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. A NATHY T. D. BARNES S. LEWANDOWKSI J. W. GILES C J. WILLIAM J. J. COLANDO . F. BASHWNR M. J. BLADEK M. MACCLELIAN M. W. LANGNES P. F. FARNSWORTH R. R. FRUGGIERO . E BEAKES B, C SMITil K. P. DRAYER M.MACNICHOL - A J. KIMPEL D. S. IfURLBITT P J BO G. E. BRIGGS N, E. MAJOR NUCTURBINE E. L SUMNER R. J. LAUVER T.L LANKOW OPERATOR W. II. TAYLOR T. S. BROWN J. R. MANN B.A NICitOLS D. E. LOWE L M.CIfURCH . W. L BURRELL M. L McCARDELL J. STARKWEATIIER D. MITTENMEYER M. A FOUST J. E. PEREZ DELAGUARDIA B. F. CARROLL, C. E. MOSTACERO
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- 5. A VALDES SUPERVISOR T, V. OSWALD SITE ORAPit!CS SITE TAO /SION DEVELOP SPECIALISTS A PROCUREMENT L M. DONOlllA K. W. KRAM NUCLEAR RECORDS DOCUMENTCONTROL C P. VANDERSCitOT TEC!tN!CIANS TECIINICIANS J. J. BAILEY D.F. ARNE1T i PROCEDURES WORDPROCESSING J.G.MANN J. L DUTCifER PROCEDURES K.1 MANNING C II. OIIISoN PRODUCTION TECIINICIANS N. J. WEST M. J. LANE K. L IIALVERSON L R. DAVIS 1C ANDERSON II. C BARANOWSKY, P. G. IIAMILTON K. A. WOOD M. P. DIMARCO B. 5. GILMOUR FRG SECRETARY iCLAY CONFIGURATION MANAGEMENT i
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og . - i ST.LUCIE NUCLEAR Puuff ORGANIZATION CHART FAOLITIES DEPARTMENTOILGANIEATION SERVICES MANAGER APRE 89e6 , C L BURTON 6 h t FACIUTIES SUPERVISOR ! L C DOUSKA I i y r r FACIUTIES/MAINT. , COORDINATOR COMMUNICATIONS l COORDINATOR f J. L STANTON ! s O. W. EDWARDS l I I CONTRACTOR:R HARDT l I JANITORIALSERE1CE I l I l CDNTRACTOR: AT& T l I Jf I I l 2 ? I l l 1 i ! l CONTRACTOR: TOTAL l l l QUAUTTSERE'!CES l CDNTRACTOR: P& G
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- T.ucs NucuAR PUWT ORGANIZATION CHART RESOURCE CONTROL DEPARTMENTORGANIZA110N APRn.19es SERVICES MANAGER C. L BURTON RESOURCE CONTROL ,
SUPERVISOR W. E. WALKER ACCOUNTS PAYADLE EXPENSE CONTROL & CAPITAL PROJECTS CONTRACTED & SUPERVISOR. O&M JUSTIFICATION CONTROL ENGINEERED PROJECTS A T. ROSSE* E. MESUN J. S. SLIM M J. M. TAXACIIER 7 INVOICE PROCESSING / O&M TECIINICIAN ADMIN CLERKS J. AIAIIEPA C. A DUNCAN C. A KINDER E. S. RIC11 ARDSON J. A TIIOMAS C. A WITEK APRIL 1996 REV02 03/31/96 Page 33 9(20380xsd
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DEPARTMENTORGAMitATION l C L BURTON ANLI"5 l INVENTORY CONTROL i SUPERVISOR , . 5 D. K. QUILTY ! i r l : BUUUCENTRAL . SENIOR PLANT WAREHOUSING RECEIVING ANALYST , A R.NEUBERGER l M. O. ULLMAN J. C EDGAR ; I i SITE ANALYSTS i PROCUREMENT INVENTORY CONTROL AGENTS SPECIALISTS S. D. BARRY C J. IIARRIS W.C. ARMES J. J. DRACK ' R. J. CONKLIN C R.OWENS l T. E. EVANS S. B. PINKSTON ; D. A KORNSTADT E. O. PRICE INVENTORY CONTROL ; O. W. RICIIARDS C D. SUMMERS SPECIALISTS ; D. E. SMITil i W.D. Wit!TE J. D. CltAUVIN [ P. L Wil3ON P. 5. POULOS [ W. M. QUIGLEY , I i MATERIAL MANAGEMENT ! TECilNICIANS ;
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B. A BIRNBAUM ! T. L CIIVIIA ; , D. K. COX l N. A.QUESADA [ i i i i l APRIL 1996 REV02 03/31/96 p,ge 37 960380.vsd [
Inter-Office Correspondence q. JPN-SPSL-96-0089
%Q Date: MAR 2 9 E ;
To S. A. Valdes St. Lucie Plant i j Frcon Nk)W) D. J. Denver Department: JPN/PSL ] Nuclear Engineering Subjects ST. LUCIE PLANT UNIT 1 PC/M #2 144-194 TITLE: RCS Flow Transmittern Damneninc Board j Addition FILE: PCM 144-194 ) I Attached for review, approval and use is the EP for the subject PCM. l This package provides the details necessary to implement the following modifications: Add damnenine boards to the RCS Flow tranemitters to reduce noise. l i If you have any questions, please contact David Howard at x7478. I I l DJD/ p" Copies: H. L. Fagley - DCC-CS/PSL (w/ original) D. M. Stewart - TS/PSL Originator: R. Olson PSL/TCM' L. Miecel PFL/ICM pw ; n- _;;;;m=A - -- -- - PSL INFO SVC
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4-ENGINEERING PACKAGE PC/M NO: 144-194 REV: O SUPPL: O PLANT: St. Lucie UNIT: 1 TITLE: RCS FLOW TRANSMITTER DAMPENING BOARD ADDITION REVISION DESCRIPTION: N/A LEAD DISCIPLINE: I&C EXPIRATION DATE: 6/30/97 REA NO: STAR 1-94100260 DWA NO: N/h PC/M CLASSIFICATION: SR X QR NNS DESIGN ORGANIZATION: _ FPL/JPN DISC CHIEF REVIEW REQD? YES _X_ NO EXTERNAL INTERFACES: None h DISC CHIEF SIGNATURE: 9. SN.O A h ti2acm REVIEW / APPROVAL: GROUP INTERFACE TYPE PREPARED VERIFIED APPROVED FPL APPROVED
- INPUT REVIEW N/A MECH [pd8 [
ELECT X
. - A I '
w ~' !L -_ CIVIL X LIC** X ' CSI X FUELS X a f 1 l
- For pontractor Prepared EPs As Detemined By Projects ** Review Interf ace As A Minimum On All EPs FPL PROJECTS APPROVAL: M '
DATE: ( 4/14 /94
~l I i
i I i
Ol 3-PR/PSL-1 ' R;vi; ion 38 i F bruary,1996 Page 20 of 26 i ATTACHMENT 2 PC/M REVIEW FORM (Page 1 of 2)
- PC/M Number 144 - RM Supplement Number O Expiration Date G / Db / D PC/M TITLE
- RC5 FLOLU~TPASSTAl'TTEF6 IyarAPE.M1OG BO ARD ADDt-TIDW-Safety Classification:
i Safety Related Quality Related Not Nuclear Safety Related Administrative PC/M Division: Normal Eith.er/Or As-Requested-Package As Fail Implementing Documentation: Department ER/NPWO # W/O # WCD# T cm G 3/ 5 2.zF, % 009501-cl + t Areas Affected: Yes or No Descnotion Operator Trasruno . Yes As per T.Ir3.0 Department Requirements Operatina Procedures YES Survedlance Proceduros \AE5 rsa h Ss. I2.
- Maintenance Procoaures V6fS $fC_; hRUIE N Spare Parts U m' "
' 'D Drawinos/Technkal Manuals
'NO SRDs? YN FUSAR Change V65 stPt'#QCJ40TRT- 1 TEDB Chanoe VEN 4TTMurrD:MT- 2.
Hurnan Factors (CREDIT) NG Environmental Co cgre MO in-Serwoe ir.u-- --. AlO Maintenance Rule NO Plant Restnchons VES - btE FM, ID Others HOLD PbiMT- Pern 054-R6 NkET FE COMPLETED FRtR 70 tWITIATicA GF THt5 fren, 4 1 1 9
Qt 3-PR/PSL-1 R:,visi:n 3B February,1996 Page 21 of 26 i i ATTACHMENT 2 i PC/M REVIEW FORM (Page 2 of 2) PC/M Review and Approval: ) JM Date 4 / /f-/ % ulo 60730t SCE % C ^ QL A W
)
maTrjienance OC Date N / O /Y W MY r]S6fM Date YI 0/ V[ Plant.G ral anager
! FRG Number: d' FRG Secretary . M" I
w y[Date Y/ /9/M Comments: i 4 j PC/M Closeout Review: Approved for Closeout: / Date: / : Maintenance QC ' 1 Approved for Closeout: Date: / / SC Procedure Notification Complete: Date: / / SCE l Training Notification Complete . Date: / / SCE 9, . - S T EA Q - Cancellation information: Date: / / SCE 1 1
)
1
. PC/M 144-194 i Revision 0 Page 2 of 12 LIST OF EFFECTIYg_PAGRE PAGE NO. REY 1 0 2 0 3 0 4 0 5 0 6 0 7 0 8 0 9 0 10 0 11 0 12 0 l
4 1 6
i PC/M 144-194 j Revision 0 Page 3 of 12 RCS FLOW TRANSMITTER DAMPENING BOARD ADDITION I j ABSTRACT STAR l-94100260 identified an inherent " noise" in the Reactor Coolant System (RCS) which is monitored by the RCS Low Flow Trip Circuitry. The RCS Low Flow Circuitry is designed to monitor the total RCS flow and provide a reactor trip if the total flow drops below the trip setpoint. The trip setpoint is determined once, at the beginning of each fuel cycle. The RCS 4 Low Flow Trip is designed to prevent exceeding Departure from Nucleate Boiling (DNB) l limits and is presently set at 95% of designed reactor coolant flow with four Reactor Coolant l Pumps running. The work done under STAR 1-94100260 has identified that the noise seen by 3 the RCS Low Flow Trip Circuitry is due to the geometry of the RCS system and is not related j to actual RCS flow. This Engineering Package (EP) provides the details necessary to reduce the undesired noise by adding dampening cards to the RCS flow transmitters (PDT-llll A, B, C, D and PDT-1121 A, B, C, D). I The RCS low Flow Circuitry is a part of the Reactor Protection System which is Safety i Related. Therefore, this EP is classified as Safety Related. A safety evaluation of this modification has been performed in accordance with 10CFR50.59. l This evaluation has shown that implementation of this EP does not have an adverse effect on
- - plant safety, security or operation, does not constitute an unreviewed safety question and does not require a change to plant Technical Specifications. Therefore, prior NRC approval for implementation of this modification is not required.
1 This EP has a Hold Point. This EP can not be implemented until the St. Lucie Unit 1 Cycle 14 refueling PCM (Reference 9) is completed. This PCM will analyze and accept the 1.025 second delay associated with the dampening cards. This Hold Point will be removed via a i supplement after issuance of the refueling PCM. i _ . - _ , . . ~
. . PC/M 144-194 Revision 0 Page 4 of 12 ~
I. DESIGN
- 1. Structure. System, or Component (SSC) Involved:
Reactor Protection System (RPS).
- 2. SSC Purpose / Function / Design Basis The subject pressure transmitters are inputs into the Reactor Protection System and are used to detect a reduction of reactor coolant flow. Per FSAR Section 7.2.1.2, the RCS low How circuitry is designed to protect the core against departure from nucleate boiling (DNB) in the evert of coolant Dow decrease.
- 3. Safety Classification: SR X QR _ NNS The Reactor Protection System is designed to Seismic Category I and IE requirements per Section 7.2 of the FSAR. The RCS low flow portion of the RPS is required to protect the core for DNB considerations. Therefore, this PCM is classified as Safety Related.
- 4. Reason for Design Change STAR l-94100260 was generated due to the inability of 1&C Maintenance department to complete the RCS Imw Flow Setpoint Determination Procedure (Reference 7).
Under STAR 1-94100260, an investigation has been performed and it has been identified that the noise seen by the RCS Imw Flow Trip Circuitry is due to the geometry of the RCS system and is not related to actual RCS Flow. The predominant frequency of the noise 2en is approximately 10.7 hertz and is associated with a i pressure wave, not flow (Reference 10). This noise causes difficulty in the RCS low flow setpoint procedure due to having to compensate for the noise by raising the setpoint accordingly. This PCM will add dampening to reduce the noise seen during the implementation of the RCS low Flow Setpoint Determination Procedure, j
- 5. Design Change Description 4
This Engineering Package (EP) provides the details necessary to add dampent a to the RCS Flow transmitters. The RCS Flow transmitters (PDT-1111 A, B, C, D and PDT-ll21 A, B, C, D) are Rosemount differential pressure transmitters, models 1153 HD6
& 1154HH6RA. These transmitters do not come with the dampening as%tandard feature. The standard electronics cards (amplifier and calibration boards) for each
, transmitter will be replaced with the special boards (option N0037) supplied by Rosemount which have the dampening feature.
. . PC/M 144-194 Revision 0 Page 5 of 12
- 6. Design Change Checklist Does the Design Change involve / impact / require justification of: !
YES NO REFERENCE 4 Internal / External Flooding X Heavy Load Handling X ' Tornado / Internal Missiles X l Single Failure Criteria X l Human Factors X Paging System Audibility X Masonry Block Wall Interaction X Environmental Criteria X 8.1 l Plant Security Capability X HELBA Criteria / Analyses X I Seismic Qualification X 8.2 I Seismic Interaction X l Electrical Separation Criteria X Accessibility /Laydown/ Clearance Requirements X ; Loads Applied to Existing Structures (+ buried) X EDG/ Battery lading / load Sequencing X Hydrogen Generation in Containment X Heat Sinks in Containment X Emergency Plant Operating Procedures (EOPs/ONOPs) X ASME Code X Emergency Lighting Criteria X Snubber Criteria X Material Compatibility / Hazardous Materials X Electrical Equipment Grounding X Cable Tray Seismic Loading X Instrument Setpoints X j Instrument Time Response X 8.3 Hurricane / Tornado Wind Loading X , Thermal / Hydraulic Performance X l Coatings Inside Containment X Emergency Response Data System (ERDS) _ X, Emergency Planning X NML Property Insurance Requirements X Environmental Qualification
- X ATT3 Fire Protection Capability * .
X ATT4 Safe / Alternate Shutdown Capability
- X ATT4 ALARA Exposure Criteria
- X ATT5
- Forms 3E, 3F, and 71 shall be attached to justify the conclusion for these items.
1
. . PC/M 144-194 l Revision 0 l Page 6 of 12 ~
- 7. Design Evaluation / Justification )
This Engineering Package provides justification necessary to allow modification to circuitry of the RCS Flow Transmitters (PDT-1111 A, B, C, D and PDT-1121 A, B, C, D). The circuitry modification consists of adding the dampening feature to the transmitters to reduce . noise. l The modification will consist of replacing the existing amplifier and calibration board in each RCS flow transmitter with boards that contain the dampening feature. The dampening feature consists of a passive RC network. The replacement boards are manufactured by the original transmitter vendor (Rosemount) to the same qualifications as the existing transmitters. The dampening feature is an option that could be purchased standard or retrofitted. Justification of the key design parameters is discussed below in item 8.
- 8. Evaluation of any "YES" responses in Design Change Checklist 8.1 Environmental Criteria The RCS Flow Transmitters are not required for mitigation or monitoring of either a LOCA or MSLB. Even though these transmitters are safety related and are located inside the containment, they are not required to be operable in a post aco 'ent containment environment (Reference 6). The replacement dampening boards are manufactured by the original transmitter vendor (Rosemount) to the same qualifications as the existing transmitters. Therefore, there is no impact on environmental criteria.
8.2 Seismic Qualification The RCS Flow Transmitters are seismically qualified. The replacement dampening boards are manufactured by the original transmitter vendor (Rosemount) to the same 4 qualifications as the existing transmitters and weigh the same as the existing amplifier and calibration boards. There is no net change in the weight of the RCS flow transmitters. Therefore, there is no impact on seismic qualification of the transmitters. 8.3 Instrument Time Response This modification affe::ts the time response portion of the RCS Flow Transmitters. Adding the dampening option on the transmitters adds an acjustable lag of 0.2 to 0.8 seconds to the transmitter response time. Presently, FSAR Sections 7.2.1.6 r.nd 15.1.3 identify that the response of the RCS low flow transmitter (sensor) should be 0.25 seconds. FSAR Section 15.1.3 identifies the following response times:
. The time the RCS low flow transmitters sense the abnormal condition to the TCBs 4
opening should be less than 0.65 seconds. This 650 milliseconds include the 250 millisecond transmitter response time.
PC/M 144-194 Revision 0 Page 7 of 12
=
3.3 seconds for the CEA insertion time, which includes a time delay of 0.5 seconds which is added after the trip breakers open until the CEAs start to insert into the core. This accounts for the decay of the m:gnetic flux of the CEA holding coils. 4 The RCS Low Flow signals are provided by summing the output of two differential pressure transmitters, one across each steam generator. This provides an indication of total coolant flow through the reactor. A review of the RPS time response procedure (Reference 8) shows that the total response time from the transmitter to the Trip Circuit Breaker (TCB) opening is approximately 0.26 seconds. This is measured conservatively by adding the time response from the " transmitter to bistable card" to the longest logic matrix time response. The 0.26 seconds is significantly less than the required 0.650 seconds. This modification will increase the time response of the transmitter from a maximum of O.25 seconds to 0.8 seconds. Adding to this time for the logic matrix to trip the TCB of less than 0.225 seconds produces a total new trip delay time of 1.025 seconds. This new trip delay time has been evaluated in the cycle 14 reload analysis (Reference 9) and it has been determined that with this trip delay time, there is sufficient margin to protect the core against departure from nucleate boiling (DNB) in the event of coolant flow decrease.
- 9. Design Verification Statement The design was reviewed to ensure that the overall design concept meets mplicable FSAR 4
and design basis documents. The Safety Evaluation was verified by review of each conclusion statement required to satisfy 10CFR50.59 criteria for establishing whether an unreviewed safety question does or does not exist. The assumptions required to perform the design were adequately. described and reasonably and appropriately identified for subsequent reverification, if required. The design inputs were correctly selected and incorporated and applicable operating and construction experience was considered. 1 The drawings and specifications used in the design were checked to ensure that the latest . revisions were utilized. The design codes that were used were checked to ensure that the proper revisions were utilized and were applicable to the reference plant and the proposed design. The codes, standards and regulatory requirements were properly identified and the design was shown to meet those requirements in addition to FSAR commitments. Interface requirements were demonstrated to be satisfied. All specified equipment, parts and materials were shown to be suitable for the required application and the environment to which the system is exposed. Environmental qualification, pre-operational and periodic test requirements were verified to be correct. The Safety Analysis has been properly documented and correctly concludes that no unreviewed safety question exists as a result of this modification. The output of this design was reasonable when compared to the inputs and the acceptance criteria for the design was adequately incorporated into the design documents to allow verification that the requirements have been satisfactorily accomplished. l
i PC/M-144-194 Revision 0 Page 8 of 12 j ' l II. S_AFETY l
- 1. Description and Purpose i This Engineering Package (EP) provides the details nt:cessary to add dampening to the RCS Flow transmitters which w .1 r increase the response time of the transmitters from 0.25 seconds to 0.8 seconds. This time response change will minimize the noise seen by the l RPS bistable and ensure that the bistable trips on valid low flow conditions. j
- 2. Analysis of Effects on Safety I This modification is to add dampening to the RCS Plow transmitters which will increase i the response time of the transmitters from 0.25 seconds to 0.8 seconds. As documented in the analyses above, operation of the Reactor Protection System has not been affected, the 1 RCS Low Flow circuitry will still operate to ensure that sufficient margin exisu to protect the core against departure from nucleate boiling (DNB) in the event of coolant flow decrease. A review of the Passport Safety Evaluntion Database and the Drawing Change j Tracking System (DCTS) was conducted. PCM 054-196, "St Lucie Unit 1 Cycle 14 l Reload", justifies the changing of the time response by supporting a low flow trip delay time of 1.025 sec in the safety analysis. No other active documents affecting the scope of I this PCM were identified.
i
- 3. Failure Modes and Effects Analysis -(FMEA) l This PCM adds dampening to the RCS Flow transmitters. This EP replaces the existing amplifier and calibration boards of each transreitter with the special boards (option N0037) that contain an additional passive RC network to provide dampening. There j are no active components added by this package. The design differences between the original design and the design provided herein have been addressed above and have been '
determined to be acceptable. No new failure mo:les have been introduced.
- 4. Effect on Technical Specifications There is no impact on any Technical Specifications or Technical Specification bases.
- 5. Unreviewed Safety Question (US0) Determination in accordance with 10 CFR 50.59, the respor.ses to below listed questions serve to determine whether the proposed design chant,e constitutes an Unreviewed Safety Question:
5.1 Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR7 The proposed change modifies the electronics of the RCS Flow transmitters. The function of the RCS Flow transmitters has not changed. The RCS Flow transmitters are inputs into the Reactor Protective System which is not considered in the initiation of any accidents analyzed in the FSAR. Therefore, the probability of occurrence of an accident previously evaluated in the SAR has not been increased.
I , ,, PC/M 144-194 Revision 0 Page 9 of 12 5.2 Does the proposed activity increase the consequences of an accident previously evaluated in the SAR7 l The proposed change involves a change to the time response characteristics of the flow transmitters only. The transmitters will still function to mitigate the consequences of an accident. Per FSAR section 7.2.1.2 (c), the low reactor coolant flow trip is required to protect the core against DNB. The effects of the increased response time have been analyzed (Reference 9) and determined to be within the existing FSAR and Technical Specification limits. Therefore, the consequences of an accident previously evaluated in the SAR have not been increased. ; 5.3 Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR? The on!';proposed change is to the electronics of the RCS flow transmitters. The change involves replacing the amplifier and calibration board in each RCS flow transmitter . The replacement boards have been fabricated by the original equipment manufacturer to the same standards as the original equipment. No new failure modes have been created. Therefore, the probability of occurrence of malfunction of equipment important to safety previously evaluated in the SAR has not been increased. ' 5.4 Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR? The change described in this package is to add passive dampening to the RCS flow transmitters only. This change does not alter the failure modes of the transmitters or the operation of other safety related component. The consequences of a malfunction of equipment imponant to safety previously evaluated in the SAR have not been increased. 5.5 Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the SAR7. The proposed change in no way degrades the reliability or operability of the Reactor Protection System or any other safety related component. No new failure modes or system interactions are being introduced. Therefore, the possibility of an accident of a different type than evaluated previously in the S AR has not been created. 5.6 Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than any previously evaluated in the SAR? The proposed change affects the response time of the transmitters only. The change has been demonstrated to adequately monitor the RCS flow. No new failure modes or system interactions have been introduced. Therefore, the possibility of a malfunction of a different type than any evaluated previously in the SAR has not been created. 5.7 Does the proposed activity reduce the margin of safety as defined in the basis for any
. technical specification?
The basis for operability of the Reactor Protection System is not impacted by this change. The modification affects only the response time of the RCS flow transmitters which are used to protect the core against DNB. Technical Specification 3.2.5, DNB parameters, provide the limitations for DNB related parameters, including RCS flow. The basis for
PC/M 144-194 l Revision 0 i Page 10 of 12 this specification identifies an acceptance limit for minimum DNBR of 51.22. The effects of the increased response time have been analyzed (Reference 9) and determined to be within the above limits. The proposed activity does not reduce the margin of safety as defined in the basis for any technical specification. 10CFR50.59 allows changes to a facility as described in the SAR if it does r 9 involve an Unreviewed Safety Question or a change in the Technical Specifications is not required. As shown in the preceding sections, the proposed change does not involve an unreviewed i safety question because each concern as posed by 10CFR50.59 that pertains to Unreviewed i Safety Questions can be positively answered and a change to a Technical Specifications is not required.
- 6. Plant Restrictions This EP he.s a Hold Point. This EP can not be implemented until the St. Lucie Unit 1 Cycle 14 refueling PCM (Reference 9) is completed. This PCM will analyze and accept the 1.025 second delay associated with the dampening cards. This Hold Point will be removed via a supplement after issuance of the refueling PCM.
There are no restrictions on plant operation imposed by this EP that are outside the operational restrictions currently provided by the Plant Technical Specifications.
- 7. Conclusions The activity proposed by this safety evaluation is acceptable from the standpoint of nuclear safety, does not involve an unreviewed safety question and does not require a change to the Technical Specifications. Therefore, prior NRC approval is not required.
l 111. CONFIGURATION
- 1. Affected Document Checklist YES N_O Reference FSAR X ATT1 Technical Specifications _ X_,
TEDB , X_ ATT2 Security Plan X DBDs X ATT6 Snubber List X ISI/IST Program Of "Yes'. ESI review is required) X Code Stress Reports _ X,_ l l l
PC/M 144-194 - Revision 0 ; Page 11 of 12
- 2. Affected Drawings !
(1) (2) (3) (4) l PC/M Dwg Rev Description / Title Disc Affected DwasRev Pf EP Rev NONE l i
- 1 (SRD). 2 (non-SRD). or 5 (other) for PSL. or 1 (POD). 2 (CRD), 3 (MD). 4 (ED), or 5 (other) for FrN.
(1) DISCIPLINE: C = CDIL; F = ELECT. I = IAC; M = MECil. N = NUCLEAR (2) REVISION of affected drawinghendor rnanual. Indicate "new" if the drawing / vendor snanual is being created. ~ne engineer is accountable for reserv)ng the aew drawing / vendor manual number. a (3) UPDATE PRIORrrY: 'ndicate 1 (sRD). 2 (non-SRD). or 5 (other) for PsL. Indicate 1 (POD). 2 (CRD), 3 (MD). 4 (ED). or 5
. l i
(other) for l'IN. (4) MEP REVISION under which last drawing change was rnade.
- 3. Affected Vendor Manuals (1) (4) l Vendor Manual #. Rev Description / Title Disc Remarks EP Rev j NONE i
- 4. Attachments 1 FSAR Change Package 6 pgs . j 2 TEDB Change Package 4 pgs !
3 EQ Check List 1 pg l 4 Fire Protection / Safe Shutdown Review 3 pgs 5 Alara Screening 1 pg 6 DBD Change Package 3 pgs
- 5. References
- 1. St. Lucie Unit No.1 FSAR, Amendment 14
- 2. St. Lucie Unit No.1 Technical Specifications. Amendment 141
- 3. Passport Drawing Change Tracking System (DCTS), reviewed 3/19/96
- 4. Passport Safety Evaluation Database, reviewed 3/19/96
- 5. PSL Package Information Tracking System (PITS) reviewed 3/19/96
- 6. Design Basis Document RPS-1 " Reactor Protection System" l
' 7. Reactor Protection System RCS Low Flow Trip Setpoint Determination, Procedure No.
1-0120050.
- 8. Reactor Protective and Engineering Safeguards System Response Time Testing, Procedure 1-1400053.
- 9. PCM 054-196, St Lucie Unit 1 Cycle 14 Reload
- 10. STAR l-94100260 I
~;
PC/M 144-194 Revision 0 Page 12 of 12 IV. MATERIAL
- 1. Equipment and Material 1.1 Equipment and Material Pre-Purchased by Engineering None 1.2 Equipment & Material Field Procured with Engineering Specifications or Requirements Description Rosemount Part Number Amplifier Doard 01154-0021-0002 - PC-1
- Calibration Board 01154-0023-0002
- PC-1 These parts are acceptable for use on the 1153HD6PB transmitter
- This is option N0037 O-rings and lubricant will be required -
4 1
- 2. Spare Parts None l
1 3 V. IMPLEMENTATION 1.0 Implementation Instructions / Specifications This EP has a Hold Point. Tnis EP can not be implemented until the St. Lucie Unit i Cycle 14 refueling j PCM (Reference 9)is completed. This PCM will analyze and accept the 1.025 second delay associated 1 with the dampening cards. This Hold Point will be removed via a supplement after issuance of the refueling
, PCM.
- 1. Coordinate all work with Operations to obtain appropriate work clearances.
- 2. Using applicat;le I&C Maintenance procedures and the vendor technical mnual, install the 1
dampening cards for each transmitter with the special boards (option N0037) supplied by Rosemount. Dampening should start with the adjustment fully counterclockwise giving the minimum time constant. Dampening should be adjusted to ensure that the signal monitored during the RCS Flow Setpoint Determination Procedure (Reference 7) is sufficiently small (approximately 100 millivolts peak-to-peak) to ensure adequate operating margin.
- 3. The plant is to revise any applicable procedures (including time response procedures).
Procedure No. 1-0120050 should be revised to use V- in determining setpoint instead of V.n+ 100 millivolts. 2.0 Post-Mod Acceptance Testing
- 1. Verify correct calibration of the RCS Flow transmitters and loops and successful completion of the RCS Flow Setpoint Determination Procedure (Reference 7).
- 2. Verify that response time of the RCS Flow transmitters is less than 0.8 seconds and the trip delay time is less than 1.025 seconds. Response time should be verified using the applicable methods of Procedure 1-1400053 Reactor Protective and Engineering Safeguards System Response Time Testing, (Reference 8).
3.0 Operations and Maintenance Guidelines
- None
PC/M No.144-194 Rev. 0 Attachment ! ) Pace 1 of 6 1 FSAR CHANGE PACKAGE Plant St. Lucie Unit _1_ 1 FSAR PAGES A1TACHED: j 7.2-11, 7.2-24, 7.2-43 15.1.3-4, 15.1.6-6 , i FSAR FIGURES ATTACHED: 1 1 COMMENTS: 1 Change on page 7.2-11 is correction of typographical error.
.5 N L J 6/* - + me a Prepared by/Date: Verified by/Date:
A 'If Approved by/Date: Notes:
- 1. All affected FSAR pages and tables are to be marked up and attached. If additional space is required, than additional pages should be provided. New information for inclusion in the FSAR shall also be provided.
- 2. The new or marked up pages are to be legible and of sufficient quality to be microfilmed.
- 3. If a figure is provided elsewhere in the design package, then it need not be duplicated in the FCP.
However, a note should be provided in the comment section referring to its location in the package. This also applies to new drawings which should be added to the FSAR. If a figure revision is to be included in the FCP, then a copy of the FSAR figure with a bubble around the affected area is sufficient.
fcu 144-199 GL4 A na w-i v i Pa$t t of b start-up to full power. A reactor trip is initiated if the rate-of-change of reactor power exceeds 2.6 decades per minute over a range of 10" Thepercent trip to 15 percent power, as measured by any two wide-range channels. signal is automatically bypassed below 10-4 above A high rate-of-change of power pretrip alarm is generated from each channel bistable trip unit when the rate-of-change of Thispower exceeds CEA 1.5 decades withdrawal prohibit per minute above a power level of 10" percent, action to prevent the further withdrawal, but not insertion, of any regulating CEA'S. Visual and audible annunciation is also provided. c) Low Reactor Coolant Flow the core against The low reactor coolant flow trip is provided to protect of a conlant flow # departure from nucleate boiling DNB in the event PCr$ l decrease. The functional diagram is shown on Figure. 7.2-910. The flow measurement signals are provided by summing the output of the differential pressure transmitters across each steam generator to provice an indication of the total coolant flow through the reactor. This , measurement ia indicated as a differential pressure (Ap) corresponding to j actual flow. The flow reactor trip is actuated directly by the summed a p l l signal. J A reactor trip is initiated by two-out-of-four coincidence logic from the four independent measurement channels when the flow falls below a preselected value. M -VLU dyr.y h a k(o co< f r.M t u ! l 7.2-11 Am. 7-7/88
Pcs @4- 194 de/ 4 A n44 >N4 e.t i va$c 3 of 4 j l Table 7.2-1 provides sensor response times for instruments providing input to the reactor protection system, and Table 7.3-1 provides similar data for the engineered safety feature actuation system (ESFAS). The total delay time (sensor plus RPS delay) assumed for each RPS trip in the accident analysis of Section 15 is provided in Table 15.1.3-2 l The latter table indicates the canservatism in the accident analysis by providing both expected and assumed delay times. Generally, sensor . delay times are small compared to total RPS trip delay time assumed for accident analysis, e.g., l Sensor Delay Total Delav Low Reactor Flow (Steam Generator AP)
- o. 8 S ec.
-0. 2 5 e c c.
1.015%c
-0.65 -eu. J
( High Pressurizer Pressure 0.032 sec. 0.90 sec. Low Steam Generator Level 0.025 sec. 0.90 sec. With regard to the response. time of the sensor, there are three basic types of sensors to consider, namely, a) Flux - The sensor itself responds to a nuclear event, thus its response is essentially instantaneous. I b) Thermoresistive Elements - The response time of a resistance thermometer element is much less than the response time of the protective well assembly housing the element, and the accident ! analyses utilize a simulation of the entire assembly. Thus , periodic checks of thermometer element response times are of questionable value. It should also be noted that RCS temperature does not provide a reactor trip directly. Hot and cold leg temperatures are one of several inputs to a computed trip function (Thermal Margin / Low Pressure Trip - see Section 7.2.1.2(g). I c) Electromechanical Devices (pressure, level, flow) - Simple mechanical devices convert changes in the measured variable to small displace-ments that upset a balanced electrical circuit. k'ith regard to response ; time, the mechanical portion of the device would be controlling. ! Factors that could conceivable ef fect response time, for example, friction and changes in material mechanical properties also af fect accurace, and thus would be monitored by routine periodic calibrations. The possibility of inferring sensor response time by studying transients resulting from perturbations to the process system itself was considered. For minor system perturbations, sensor response times would generally be much less than the system response time. Perturbations large enough to ; reduce system response time to a point where sensor response time could l conceivably be inferred from transient data would undoubtedly result in l reliance on reactor protective functions, i.e. , unit trip. Pertubations of this natute would not be created intentionally. Even for large pertur- l bations, sensor and system response times are inseparable, and the ability to measure accurately the sum of sensor plus system response times is cuestionable. Nevertheless, testing is performed in accordance with the Technical Specification. 7.2-24 Am. 2-7/84
. TABLE 7.2-1 MONITORED PLANT VARIABLE INSTRUMENTATION RANGES SENSOR FULL POWER CHANNEL RESPONSE MONITORED VARIABLE HINIMUM NOMINAL MAXIMUM ACCURACY TIME 2
-0 Neutron Flux Power, 10 100 200 1.0% <1 mece 3 percent of full power 4.0 sec 4 Cold Leg Temperature, F 465 539.7 615 0.75% 2.5 s,ec 5 Ilot Leg Temperature F 515 595.1 665 0.75% 2.5 sec 5 -
Pressurizer Pressure, peig 1485 2235 2485 0.83% 32 usec , PC Nk 800 Steam Generator AP, psi 0 34.0 50 551.67% 55f99 asec Steam Generator 0 64.7 100 0.90% 25 msee Water Level 2 Y Steam Generator 0 800 985 0.79% 32 usec
$ Pressure, peig Containment Pressure, psig 0 0 45 0.71% 0.5 see Axial Shape Index -1 0 +1 N.A. < 1 usec, EN?
Notes: it t I
- 1. Channel accuracy based on root-mean-square technique 4.
Logarithmic power channel response
*fs T
g' *ii using individual instrument accuracies. at 10-8% power --C.0,sec f
,. 2.
Response time defined in terms of reaching 63% of Logarithmic power channel response p
-n at II power -- 2 msec 1 y final value for a step enange input. om $ 3. Linear power channel response 5. Correlated to 40 feet per second flow rate
TABLE 15.1.3-2 - REAC'1VR PROTECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 1. Manual Reactor Trip Not Applicable
- 2. Power Level - High 5 0.40 seconde *4 and 5 14.0 secondeN# l
- 3. Reactor Coolant Flow - Low econ FM Pressuriser Pressure - High' 4
' s 0.9 0 seconds
- 5. Containment Pressure - High 5 1.40 secondo
- 6. Steam Generator Pressure - Low 5 0.90 seconds
- 7. Steam Generator Water Level - Low 5 0.9 0 seconds
- 8. Local Power Density - High '
5 0.4 0 enconde*# and 5 14.0 secondeNN l
- 9. Thennal Margin / Low Pressure s 0.90 seconds *# and 5 14.0 secondoNN l 9a. Steam Generatov Pressure Difference - High 5 0.9 0 seconde '
- 10. Loss of Turbine- - Hydraulic Fluid Pressure - Low Not Applicable '
- 11. Wide Range Logarithmic Neutron Flux Monitor Not Applicable I
Neutron input of detectors are exenpt first electronic from response conponent time testing. Response time shall be measured from detector output or in channel.
# Response time does not include contribution of RTDO. n>%
## i*n3
? !
RTD response time only. This value is equivalent to the time interval required for the RTDs output I2 63.2% of its total change when subjected to a step change in RFD temperature. to achieve h4
* -4 i p - #- ,
A i 15.1.3-4 S Amendment No. 14 (6/95)
Table 1. .-2 ST LUCIE UNIT 1 TRIP FUNCTIONS Function , Allowable Values Values Used in the Analysis - Setpoint Delay
- Setpoint Delay *
(seconds) (seconds). Variable High Power Trip (1 of rated) 107 0.9 112 Low Flow Trip (% of design) 0.9 i 95 MS- 1.517 93 1 High Pressurizer Pressure (psia) 1.15 t 2400 1.4 2422 1.4 Low Steam Generator Pressure (psla) 600 1.4 578 1.4 Low Steam Generator Water Level 20.5(1)
*f (1 of span) 1.4 16.46(I)
- 1.4 '
LPD (described in text) - 0.9 - 0.9 IPf/LP (described in text) 1.4 1.4 Steam Generator Pressure Difference (psi) 135 1.4 185 1.4 i i
- includes a 0.5 second allowance for the holding coils to release (1) Loss of Feedwater transient was reanalyzed by ANF to support th setpoints.
Analysis is presented in Section 15.2.8. e reduction of the low SG 1evel v
]$e?J i
, -fI ;
3
+4
<f T'
, t r* ' 4 b
~
Q i 0149F/8 v 15.1.6-6 -
1 i Facility PSL Unit 01 PC/M No.144-194 i Attachment 2 l Revision 0 1 Page 1 of 4 l l TOTAL EQUIPMENT DATA BASE CHANGE PACKAGE COVER SH i Description of Change: Add dampening cards to RCS Flow transmitters Basis for Change: See PCM 144-194 P
References:
PCM 144-194 Engineering Prepared by (d Date _3 '49 % Verified by ~ L# ' Date 3M~ % Approved by . 4 24 --
' Date 3 M /
(Discipline Supv/ Lead Engineer) - l Configuration Management Reviewed by: Date: Data Entry by: . Date: i Data Entry Verified by: Date: For instruqtions, see Q12.8-1 1
i j Facility: PSL Units 01 LMD E ~ 1 l Component: PDT-ill1A l l l Associate TRANSMITTER l l l l ) lAtts 2. ( l Date Printeds 03/14/96 l lRev: C) l l l }Page 2, o f 2- ] i I l TOTAL EQUIPMENT CATA BASE SHEET EQ Tag N/A EQ Rev: N/A EQ Doc Pact N/A Systemi 01 REACTOR COOLANT SYSTEM Seismici 1 Saf ety Class SR Eng Refs Q Group 18 EQ Sury Notes N/A EQ Speer: N/A RO197: N EQ Relateds N EQ Scews N/A RG197 Cat , Q Basist ,,,, EQ Remarks: N/A RG197 Type: , Corp Types IX Sub Types T Safety Channel MA Fem: 353-188 l
\
Names DITTERENTIAL PRESSURE TRANSMITTER FOR STEAM GENERATOR 1A 1 Locn Code: RCB/47/N-28/E-11 Startup Systems 034A Locn Deses IR 55 1 ! I Instl MFC 8: ROE RQSEMOUNT INC. Engineering Verified: Y Insti Model 1154HH6 RhDO 3'l Reva 000 Orig Pos CE-2851(9200695) Comp Group PD-1111A NPRDS: Y Acct Nos $30 ! l EQ Tab ,,, Insulation Revis , Train , 1 Scaffold Reg: , Critical Comps , Control Room Compt , Work Group IST Reqd N RWP Reqd: Y Maint Pges: , , , , ,,,_, 1 l
\
e l I I
d j Facility: PSL Uniti 01 LMD: E
- l l l Componenti PDT-1111A
- l l l l Associate TRANSMITTER l lAtti 7,.
l 1 i IRevi o t l Date Printed: 03/14/96 l lPage,3 et l l 1 I I TCrfAL EQUIPMENT DATA BASE SHEET Drawing: Sheets Coordinates 8770-G 226 3 8770-1527 8770-9-231 01-10 8770+B 327 385 4770*G-078 (C-7) 110 Tsch Manuales
$770-9834 Procedures 1 0120050 I 1-1400153 I 140C065 l
Notes: FcM : P.O. C89528-80061 (NEW) T E 68 C.T%[e OAmpeJt4 6 60 05 re s t 4d Q Approved Alternate Mfg:
Description:
Model: Rev. Inst 1. Eng Ver. Parameter i Nime: Valuei UOM. SENSING LINE QGROUP _ N/A ' IMPULSE LINE NO N/A TUBE TRACK NO N/A RANGE DESCR 81 N/A l PP.OCESS RANGE 81 N/A I SENSING LINE QGROUP B N/A IMPULSE LINE NO N/A N/A TUBE TRACK NO N/A N/A l PROCESS RANGE 81 0-50 N/A PROCESS UNITS #1 PSIG N/A l SIGNAL INPtTT 81 0-50 N/A I INPUT UNITS s1 PSIG N/A l j S!GNAL CtTTPUT 81 4 20 N/A CITTPITT UNITS 81 mA DC N/A I SCALE RANCE 81 N/A SCALE UNITS 81 N/A t e N '
.m -
't l Facility PSL Unit 01 PC/M No.144-194 Attachment 2 Revision 0 Page 4 of 4 TAG ASSOCIATE Instl Model Notes (add to notes field) PDT-1111B TRANSMITTER 1154HH6RA N0037 Electronic dampening cards installed PDT-1111C TRANSMITTER 1154HH6RA N0037 Electronic dampening cards installed PDT-1111D TRANSMITTER I153HD6PB N0037 Electronic dampening cards installed PDT-ll21 A TRANSMITTER ll54HH6RA N0037 Electronic dampening cards installed PDT-1121B TRANSMITTER I154HH6RA N0037 Electronic dampening cards installed PDT-1121C TRANSMITTER 1154HH6RA N0037 Electronic dampening cards ins,talled ! PDT-!121D TRANSMITTER 1154HH6RA N0037 Electronic dampening cards installed l 1 I
PC/M No. 144-194 Attachment 3 I Revision 0 Page 1 of 1 ENVIRONMENTAL QUALIFICATION (EQ) CHECKLIST
)
- 1. ELgcIBIgAL_3RgIEgggI a)
Does any affect the PC/M EQ Docadd, Pac?modify or replace electrical /I&C equipment or b) If NO proceed to Section 2. X Is the equipment located in a harsh environment per EQ Doc Pac 1000 c) (PSL) or 1001 (PTN)? If NO proceed to Section 2. DoesORthe equipment perform a safety related function? _2 _ _)L Does the equipment perform a quality related function such that its failure could mislead an operator or adversely affect any safety related OR function required to mitigate or monitor an accident? . x Is the equipment classified as Category 1 or 2 by RG 1.97? X If the answer to any question in Section 1.c is YES, then the equipment is subject to 1 10CFR50.49 and the EP must justify that the equipment is Environmentally Qualified. I
- 2. ENVIRQ_NNENTAL_C9NDITIONS **
IEE HQ a) Does the modification add, relocate, raise flow rate of, or increase radiation levels of any piping containing radioactive liquids such that lifetime or accident exposure of electrical equipment may be increased? j b) _1_ Does the modification add or reroute any high energy piping, the I failure of which could impinge upon electrical equipment or increase l the post-break local pressure or temperature? ! c) 2L_ Does the modification alter process fluid characteristics such that i the post accident pressure, temperature, PH, or boron concentration i d) to which electrical equipment will be exposed may be changed? X I Does the modification alter any barriers that shield electrical equipment from high energy lines or radiation exposure? i e) _)L._ Does the modification add equipment to Containment or increase the j post-LOCA makeup water inventory such that the design basis i Containment Flood Level may be increased? 2_._,
** If the answer equipment to any question in Section 2 is YES, then it must be determined if any subject to 10CFR50.49 could be impacted and, if so, that its Environmental Qualification is still valid. Document the determination in the EP. :
Prepared by/Date N/R Verified by/Date N/R
** Leave signatures blank if same as on PC/M coversheet **
Form 3E, Rev 6/94 !
Attachment 4 PC/M No. 144-194 l Revision 0 Page 1 of 3 1 FIRE PROTECTION REVIEW CHECKLIST I I. SAFE SHUTDOWN CAPABILlTX XES*. NQ EEFERENCE ! Does the Engineering Package install, relocate, modify, or affect the operation of: A. Equipment on the Essential Equip. List X__ B. Safe Shutdown Analyses
- 1. Safe Shutdown Circuits X
- 2. Alternate Shutdown Components X
- 3. Associated Circuits X
- 4. Manual Actions X_
II. FJRE PROTECTION SYSTEMS Does the Engineering Package install, modify, or affect the operation of: A. Fire Detection Systems v l B. Fire Water Supply System X C. Water Suppression Systems _X _ D. Halon Suppression Systems X_ E. Standpipes or Hose Stations X F. Portable Fire Extinguishers X III. FIRE RATED ASSEMBLIES l Does this Engineering Package install, modify or affect the function of:
- A. Fire Barriers X_
l B. Fire Doors X Indicate "YES" if an item is impacted or requires justification.
- If "YES", a section in the EP shall dispositi.on or justify the item, with' the EP section
~ number referenced accordingly on this form. Form 3F, Rev 6/94 Sheet 1 of 3 1 III. FIRE RATED ASS _EMB_ LIER (Cont) YES* NQ REFEREN_GLE
.A.----A---_m__.--w_- -= w-.m-. m wa a aweema.m__m.sM.J$*e- hd4_4p.h4.Jhemh 4.'.hd,w-.4,memweham. h E.44.44h W.44 ^4& w _M _.a d u -
- AAM MO4. m # e_m 2
l l
- f .
1 4 4 i k a J i e I + a 4 0 I i 4 n 1 e e
Attachment 4 ! PC/M No. 144-194 ' Revision 0 Page 2 of 3 C. Fire Dampers i X D. Mechanical Penetration Fire Seals X E. HVAC Duct Penetration Fire Seals X + 6 F. Cable Tray Fire Stops 'X ; G. Structural Steel Fireproofing X H. Conduit or Raceway Support Fireproofing X I. Conduit or Raceway Fireproofing X J. Internal Conduit Fireproofing (Stuffing) , X t IV. EMERGENCY LIGHTI}[Q A. Does the Engineering Package install, modify, or affect the I i operation of? B. Does the Engineering Package obstruct the required light i pattern of? X i C. Does the Engineering Package add or relocate essential equipment or components which will require the addition or [ relocation of? i X $ D. Does the Engineering Package install, relocate, modify, : affect, or require the use of hand held emergency lights? X E. Does the Engineering Package relocate, modify, affect, or l obstruct the light pattern of perimeter security lighting to ; equipment requiring manual actions? ' X V. ECP OIL COLLECTION Does the Engineering Package install, modify, or affect the j operation of? y [ I VI . MIRCEL&NEDUE i A. Does the Engineering Package affeet the quantity or protection of insitu combustibles (solids, liquids, or gases) beyond the assumptions in the FHA?
. _X_._
Form 3F, Rev 6/94 Sheet 2 of 3 VI. MIEMELLANEOUS (Cont) i
, -or - = , - - , , . - - - , _ - .l
Attachment 4 PC/M No. 144-194 Revision 0 Page 3 of 3 B. Does the Engineering Package Cause the addition of a large combustible inventory within 50 ft. of essential equipment, components, electrical manholes or structures? X C. Does this Engineering Package modify or affect curbs or dikes used to contain combustible liquid spills? X D. Does this Engineering Package cause the removal of a flame retardant material from non IEEE-383 cables? l E.
-E--
Does this Engineering Package affect fire protection technical specifications or fire fighting strategies? { i _ _ _ X F. Does this Engineering Package install, modify or affect the operation of alternate shutdown communications? ) l L G. Does this Engineering Package install, modify, or affect hydrogen lines (or any combustible gas) in areas of the plant containing safe shutdown equipment or components? j X_ H. Does this Engineering Package install, modify, or affect any HVAC equipment or room heat loads in areas of the plant containing Safe Shutdown equipment or components J X l l l PREPARED BY N/R DATE VERIFIED BY N/R DATE Leave signatures blank if same as on PC/M coversheet " i Form 3F, Rev 6/94 l Sheet 3 of 3 l t i
Attach:nent 5 PC/M _144-194 Revision ._Q Page 1 of 1
- f. M4BA SCRElllIRQ
- l. Is this PC/M administrati'nt only (no physical modifications)?
yes, Further ALARA screer */. is not required. X no, continue screening. (RCA)?
- 2. Does this PC/M involve a location in the Radiation Controlled Area X yes, Location: RcB Elev 67 Continuescreening.
_ no, Further ALARA screening is not required.
- 3. Does the implementation, operation, or maintenance of this PC/M involves
_ No Movement of radioactive material? Modification of systems containing radioactive fluids or resins _ _ NO such that routing or retention characteristics are affected? of existing permanent radiation ___ NO Movement / modification shielding? __ NO Modification or removal of equipment that results in an uncontrolled opening / penetration into a High Radiation Area, Locked High Radiation Area, Very High Radiation Area, or Exclusion Area? NO Diving associated with systems containing radioactive material? NO Entrance into containment during power operation? NO Potential for personnel exposure to a radiation field of 2 1r/ hour (assuming current area dose rates)? _ NO A total lifetime estimated dose due to the modification greater than or equal to one (1) man rem? If any item in Section 3 above is checked "yes", indicate "yes" belows yes, This PC/M has the pot < ntial to significantly impact personnel radiation exposure. Complete Form 72 to ensure total radiation dose is minimized by design, X no, This PC/M has little or no impact on personnel radiation exposure. Form 72 not required, however, normal ALARA precepts should be followed to minimize radiation exposure.
.< arified : N/R Prepared: N/R
** Leave signatures blank if same as on PC/M coversheet Form 71, Rev 6/94 1
O Attachment 6 Document No. 144-194 Revision 0 Page_1 of 3 DESIGN BASIS DOCUMENT (DBD) CHANGE PACKAGE COVERSilEE Plant St Lucie Unit 1 DBD PAGES ATTACHED: RPS-1 Pages 123 and 255 DBD FIGURES ATTACHED: COMMENTS: k 5 k2'1 % % 2- P 0 Prepared by/Date: Verified by/Date: AL BuAsk/u Approved by/Date: Notes: 1. All affected DBD pages and tables shall be legibly marked up and attached. If additional space is required. then additional pages should be provided. New information for inclusion in the DBD shall also be provided. 2. If a figure is provided elsewhere in the design package, then it need not be duplicated in the DBD change package. However, a note should be provided in the comment section referring to its location in the package. This also applies to new drawings which should be l added to the DBD. If a figure revision is to be included in the change package, then a copy of the DBD figure with a bubble around the affected area is sufficient. Form 62, Rev 6/94 l l l
Qw .I'H* IM b i . St. Lucie Unit 1 A 11xher r b ' REACTOR PROTECTION SYSTEM Document No. RPS 1 9.ge t .4.5 i Design Basis Document Revision B
. Page 123 i j
COMPONENT PARAMETER WORKSHEET j l Section 7.14.2b ' 1 Comoonent ID Descriotion - I PDT-1111 A, B, C, D PDT-1121 A, B, C, D Differential Pressure { , Transmitters for Reactor I Coolant Flow Measurement 1 Loops f A. Parameters Response Time i \ B. Value OS51.025 seconds i s C. Source 1. St. Lucie Unit 1 Technical Specification Table 3.3-2. j !
- 2. St. Lucie Unit 1 UFSAR section 7.2, table 7.2-1 and table '
15.1.6-2.
- 3. RPS Technical Manual Volume 1, 8770-7120, Rev. 6 i
- 4. Project Engineering Specification for Instrumentation and j
Control Equipment C-E document 19367-lCE-903, Rev. 2. D. Background / Reason for Value The value shown above represents the total channel response time requirements, as measured from application of a sudden differential pressure change at the transmitter input to the opening of the reactor trip switchgear breakers. PCM 144- 1 194 added electronic dampening to the transmitters to reduce noise. The safety I analysis assumes a low flow trip function response tirne of t451.525 seconds, t
, which is obtained by adding 0.5 seconds (for the CEDM gripper fingers to disengage from the CEA extension shaft) to the channel response time. Although FSAR table 7.2-1 and the original I&C equipment specification sheet identify the RCS flow measurement transmitter response time as 250 800 milliseconds, this I value should not be considered to be part of the transmitter design basis. No -
specific value is required for the transmitter (or any other individual component) as long as the total channel response time is less than &651.025 seconds. As i discussed in source 3, the response time of the RPS actuation logic circuits (from the bistable through the RTSG breakers) is approximately 160 milliseconds, thus leaving ample margin available for the transmitter.
- - - . . - .. - . - . - ~ - - .. _. -.-. -. -..._.-.-.. -. -
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St. Lucie Unit 1 A m w t C. i' Document No. RPS-1 ! REACTOR PROTECTION SYSTEM ry ) .f J ; Design Basis Document Revision B ' i Page 255 * .! The relays settings are not critical as long as it is sufficiently above zero Vac 7 3 reliably de-energizes when the CEDM bus voltage reduces to zero and s . i so that the undervoltage relay is not de-energized by normal 240 V ac volt Based on the above, the Unit 2 setpoints were adopted for Unit-1 and are as foi , The' normally open contact opens at 192.2 i 26V ac line to line, dec . The normally closed contact closes at 173122V ac line to line, decre 144-194 { RCS FLOW TRANSMITTER DAMPENING BOARD ADDITION t ! t i This PC/M adds dampening boards to the RCS Flow transmitters .Ito i reduce noise. The dampening boards willincrease the response i I time of the flow transmitters. The noise was established I to b! related to flow, but due to the geometry of the RCS piping. I : i 1 l l l l I i l
( 1 l i l l . \ l l i l \ MECIBSlIdNACENFORCEMENT P I CONFERENCE l l l ST.LUCIE PLANT 1 I
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l l l NRC INSPECTION REPORT l i i NOS. 50-335/96-03 AND 50-389/96-03 l l l 3 l MARCH 8,1996 i l l l ATLANTA,GA i l
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4 1 OPS , ERDADS ( - ADMIN. OFFICE ICOMPUTER OFFICE l l l l l lllll l i M l tocer.tcwasss.n.nu l 1
UNIT 1 DILUTION EVENT
. JANUARY 22,1996 BRCO RETURNS FROM .
KITCHEN o BRCO DILUTION LEAVES FOR h 4 EVENT KITCHEN TERMINATED 0200 0300 0314 E , .
"W .
E E E 'JL JL JLh JL JL JL o o NPS ASSUMES BRCO 9-- 4 ANPS T-COLD AND SENIOR RO BEGINS SUMMONED TO POWER DILUTION CONTROL ROOM; STABILIZED; RESPONStBILITY ENTER LCO EXIT LCO (0225) (0314) BRCO F RESPONDS TO ALARM 4, E-9 9 DRCO DRCO LEAVES RETURNS l FOR FROM i KITCHEN i KITCHEN tas cwate r re-nt) 2
UNIT 1 DILUTION EVENT - PROMPT ACTION JANUARY 22 JANUARY 23 A A w/ 3
/
0400 NOON MIDNIGIIT E E E E E E h h JL JiJihli JL JL JL JL O O DS-7 OPS TECHNICAL COMPLETED SUPERVISOR DY ANPS RECOMMENDS BRCO HPES 9- 4 OPSTECHNICAL SUSPENSION OPS TECHNICAL b COORDINATOR SUPERVISOR SUFERVISOR HOLDS ON SITE BEGINS EVENT FACT FINDING REVIEW MEETING WITH CREW (0515) AND BARGAINING UNIT NPS REl.ATES 9- -9 MANAGEMENT EVENT TO OPS PHONE CALL SUPERVISOR DISCUSSING EVENT DURING ROUTINE (0740) CALL FROM OFFSITE OPS SUPERVISOR 9-(0545)
-4 DRAFT IHE DISTRIBUTED TO APPROVES CPS MANAGEMENT TECHNICAL SUPERVISOR RECOMMENDATION (0630)
TO SUSPEND BRCO NPS FURTHER 9- -9 ANPS NOTIFIES DISCUSSES OPS TECHNICAL EVENT WITH OPS SUPERVISOR ORALLY SUPERVISOR IN OFFICE; DEllVERS DRAFT IHE (0600) totJCJ96C16132-nt) 3
-_m.m___ -_-__ ____m_____a_. _ - - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - __ -_-__. - _ _ _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _
Root Cause Evaluation . PROBLEM 1: , A reactivity evolution was initiated without adequate controls. Root Cause: Routine boron dilutions to maintain 100 percent power are not treated with the same importance as other reactivity management evolutions. PROBLEM 2: The plant staffs recognition of this event's significance was slow. Root Cause: The root cause of this problem is lack of a well defined threshold for recognizing safety significance. . 6 4 i
PROBLEM ~ 1 . A reactivity evolution was initiated without adequate controls. Corrective Actions .
- Personnel - Procedures / Documents / Policies - - Equipment Performance - Training and Quality Assurance - Supervision and Management 5
PROBLEM 2 The plant staffs recognition of this event's significance was slow. Corrective Actions
- Personnel
- Procedures / Documents / Policies
- Equipment Performance
- Training and Quality Assurance
- Supervision and Management 6
~
Management Lessons Learned - Management's Operational Guidance to Maintain Tc at 549F Adjustment of High Tc Alarm Reinforceinent of Expectations to the Operating Crews Concerning Personal Accountability Operating Crew Communication with Plant Management
~
Personnel Lessons Learned . Operating Staff Has the Highest Levels of Honesty, Integrity, and Accountability Confirmation of Crew Members' Fitness to Perform Licensed Duties 8 i
Procedures and Policies Lessons Learned - Senior Reactor Operator Direct Oversight of Reactivity Manipulations Periodic Dilution of the Reactor Coolant System Is Not an Activity to Be Turned Over to Another Operator Clarification of Short Terni Turnover of Control Station - Responsibility , Implementation of Event Response Teams 9 ,
Equipment Lessons Learned - Continued Focus on " Dark Board" . Use of" Manual" Mode versus " Automatic" Mode of Control for Boration and Dilution 10
Training and Quality Assurance Lessons Learned - 9 Lessons Learned Need to be Included in Continuing Training Effectiveness of Corrective Actions - 11
._____...____..-______..-__._______.:_._.____.-__m-
_ _ . . - . _ __-.___ .._ _ _ _ . _ _ _ .m___._m___2 ___ ._m--__.-__.m -._;___ ____--_____-_.
Safety Significance ., Reactor Power Peak at 101.13% Observed Cold Leg Temperature (Tc ) Maximum of 549.75F Technical Specification Limiting Condition for Operation ACTION Limit with Tc > 549F is 2 Hours; Tc Exceeded 549F for About 50 Minutes UFSAR B6ron Dilution Event Licensing Basis Assumptions Bounds Subject Boron Over-Dilution Event Probabilistic Safety Assessment Evaluati6n Concluded That the Plant's Core Damage Frequency Was Unaffected by the Event 12
Apparent Violations B and C - Apparent Violation B - Inadequate Design Control
- Review & Conclusions - Corrective Actions Apparent Violation C - Inadequate 10 CFR 50.59 Evaluation - Review & Conclusions - Corrective Actions i
e 13 r..-___ -m ._. ___.m. ___-.
Apparent Violation B - i i Proposed Violation:
" Design control was inadequate, ... procedures for adding ... demineralized water and boric acid to the ... [RCS] (in manual and directly to the suction of the charging pumps) did not implement the method in... Chapter 15 (in automatic and to the volume control tank), ...since January I976..."
Assessment: l FPL Concurs with the Apparent Violation ' UFSAR Describes Automatic Mode as Normal in Contrast to Plant Practice: i 15.2.4: 1 "During normal plant operation, concentrated boric acid solution is mixed with demineralized makeup water...and is automatically introduced into the i [VCT]..."
)
i i. 14 i
Apparent Violation B (Cont'd) Conclusions a Design as Described in the UFSAR Is not Consistently Translated into Procedures UFSAR Inconsistencies with Plant Practices Need to Be Eliminated Safety Analysis Has Concluded No Unreviewed Safety Question Corrective Actions Review and Enhance UFSAR Assessment by Multi-Discipline Team (Covered ~ 1/3 UFSAR Content)
. 1 Complete Identification and Elimination ofInconsistencies Unit 1 by mid-December 1996 Unit'2 by September 30,1996 Improve the Procedure Review Process to Include Feedback for UFSAR Update 15
~
Apparent Violation C Proposed Violation:
"A .. 50.59 evaluation was inadequate, ... the licensee made a change to the Unit I boron dilution procedure on January 23,1996 (after the event), to allow adding demineralized water in " Manual" and directly to the suction of the charging pumps, that was different from the method stated in the UFSAR, Chapter 15 (in " Dilute" and to the volume control tank) and without preparing a .. 50.59 safety evaluation."
Assessment: FPL Concurs with the Apparent Violation
- UFSAR Description Is Not Consistent with Plant Practices 16 m.__--.__..m_.__..__m__ ______-_.m_-- . _ . . - _ _ . - _ _ - _ _ -_- k _ _ -__ _ +-__ ma- - t-w __-+ - -_ _ . _ __ - - --__- - _ m ___ ____- -- __.m-'__-___ .-__- .__ _____-
Apparent Violation C (Cont'd)
Conclusions:
- TC 1-96-01710 CFR 50.59 Screening Conclusions Are Not Supported by UFSAR Description - UFSAR and 10 CFR 50.59 Screening Process Need Improving I
l 1 *7
Apparent Violation C (Cont'd) Corrective Xctions
- Improve 10 CFR 50.59 Screening Process
- Improve Process by Documenting the UFSAR and Technical Specification Reviewed During Screening Conduct 10 CFR 50.59 Training for Departments Responsible for Procedure New Emphasis on the Definition of"... procedures as described in the safety an report"
- Clarify Screening Criteria Reference Applicable UFSAR Sections within Procedures
. Improve UFSAR
- Eliminate Procedure / UFSAR Inconsistencies 18
Impact of the Boron Dilution Event - on St. Lucie Plant Management Expectations Programs and Procedures Training i i Personnel Performance t l Management Performance i 19
. I
{ l CHANGE REQUEST NOTICE CRN No. no u - n s - catc-e eg I of ' kl PC/M No. NPWO No. WO No. G SR O QR
,lmplemented by:
EI Ptant ec9-K% E s eO O NSR O Construction Services Affected Document (See page _ for addttional affected documents) . 1 TEBD Affected: O Yes E'INo f 8.ed . (.- 67b Sk E Rev. & Rev. Rev. { Rev. - Rev. Rev. i Rev. I Rev. Rev. j PC/M Title sT wu_ u.- d L c,eute a CRN Affects Work Control Document (s): ! b w%< M w.ag care ten.as ^"D WCD t.i m . ,c e m w c. o a Rev. Step (s) 1 Reason for Change: h Eb o, . M neid m c_m f l i c_-t s J. A clW m_ r. + > m m e, As . ' I% ved - \/ O h WY . 'Pc m cJ t .- w 9tG 1,, te e_ cl-<w s N 0-sn CLtum Coel e %. V o 'll T) 6 < n.~e cl i n.- A a sLJNtd & Change Reques't/ Sketch (If Necessary) ' m, A, d o p . i E hPy1 4 'TC< W g PSL INFO SVC 9. hf tM.n.t46 ; T SM. %ODOS . Dec N39cded % b babJ% 1~01000 (Aw)M htA.pdct 4 - C.,-o ? o Tl DB. Rev.h tk J Opvedib'). ACTIOf(BY ENGINEERING: (Preliminary) This CRN has been discussed with n Person O m b i.3 Nwnh Ofroceed of Engineering. O Telephone (2f Hold SOURCE OF CHANGE: A-O Impt. Dept. Error O Convenience APPROVALS:[ t Prepared by: _ vW' Date 3 / 4 /94 O Engr. Error O yDign Evo!ution Dept. (s#.5 ./A V Phone ~7/#T3 O Existing Cond. ErOther O Mat'l Substitution hpl. Dept. Se C-j Date bbD% SCOP CF WORK: SU/OC Revie to e Date 8 / 6 / f6 nerease O Decrease O Unchanged FPL Const. Supv. Date / / ENGIN,EERING APPROVAL (Final) IMPACT OF REVISION CHECK SHEET Yes No Gr APPROVED O NOT APPROVED N Al 1 nizant Design Org. (CDO) 1. Design Basis or Analysis Affected O3 e Prepared by: A /'s )/2, . . Datel/ P / 9/, 2. Safety Evaluation Affected O2 j Verified by: DN%pd Date 3 /6/4 3. PD/MDs or SRDs Functionally Affected O G
, d Approved by: Y[ lM A[CDate 8 /UJM 4. Start-Up, Operations, or Maintenance h JPN (if not Cdb)[ Date / /
Remarks / Basis for Disposition: '/A e (//A) om r /- A e n u m 4f w d o / / 4 - K r e _,e Or/m 00 Y- 9 'f L M Necw en Cedr 9. th curl IW O <b e M Y m 0 0 Cn llci <r>tw o tm& (Q s, h b f,r)i c is n a , Ar06 1r7 P o to SLA 38 hkr R f, n2 <Iouw em pm 2 ' / 'f4W CA A) . IMPLEMENTATION' IMPL. DEPT. SUPV. i) OUALITY CONTROL COMPLETE MAltlT-WORKED CRN j
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- Help Data Print PLANT CHANGE / MODIFICATION (PC/M)
Roadmap Exit 11/08/96 10:29 I Options J Facility = PSL Unit = 02 PC/M No. = 96004__ Plant Apyd Rev: ?0Status Latest=Sup.: WORKING _ Status Date: 02/27/96 00 PC/M Type = MEP_ Sub= __ Title : UNIT 2 GENERIC DRAWING CHANGE { l I Orig Doc Fac= Type = No ; Othr Doc Fac= Type = No l CDO : JPS Disc.: N Resp Eng : CJWOKJC_ WASIK ! Spnsr Dpt: Hold = _ l
'Expr Date: 06/30/96 Affect SRD/ POD : Y i Cap /O-M : _
Safety Class: SR_ Outage Related : Y Outage / Cycle = 1 Rev Pending?: _ Safety Eval Eqd: _ QC Required : ; ISMOD No.: Security Inf: _ NRC Commitment :
; a =us- _ Implmnt Dept = PLT _ i
----- MILESTONE DATES ========---
=- -- ==- -
-__= l l
' Initial Value: 01/19/96 PNSC/FRG. Approved : 02/09/96 Plant Approved : 02/09/96 ! Work Complete: SATS Complete-PTN : To Finl Updt-PSL: Draw.Up.Cmplt: Audit Complete : Vault Received :
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I Nos Doc ads SDC Nts Att PLs Phs TOs Rev CRNs i Contents: + + + - - - - - -
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-g Help Data Print Roadmap Exit Facility : PSL Doc Type: DRWG Sub Type:
l Doc Ho.: 2998-G-080 Sheet: 2B Status: ACTIVE __ Unit: 02 Note : - Security : N Safety Class : Update Priority : 1 Title : FLOW DIAGRAM FEEDWATER & CONDENSATE SYSTEMS
===__ :-
=====------------- ==========_______================================
Change Change Document Update o Type Number Status Due _ CRN_ U005960 APPROVED PCM_ 96004 WORKING _ More:
*S= SELECT & F9 TO VIEW CHANGE DOCUMENT; *S & F10 TO VIEW AFFECTED DOC. LIST.
Fl= Help F4= Prompt F6= Refresh F7=Bkwd F8= Fwd F10= Perform Msg F12= Cancel i d n 1 e i i t n l l
JAN 09 '96 11:OSAr1 PRES NUCLEAR DIV P.1 Inter Office Correspondence f: P L To: Executive Council DATE: January g,1996 FROM: J. H. Goldb LOCATION: EX-JB
SUBJECT:
NUCLEAR DIVISION MANAGEMENT CHANGES Effective February 5,1996 Mr. D. A. Sager will assume the duties as Vice Presi Nuclear Assurance, replacing Mr. J. E. Geiger who is retiring. Mr. W. H. Bohlke will replace Mr. Sager, on an interim basis, as Site Vice Presi St. Lucie Plant. 1 Dr. K. R. Craig will replace Mr. W. H. Bohlke, on an interim basis, and is designa as an acting Vice President, Nuclear Engineering & Licensing. ! We wish Jim Geiger good health and happiness in his retirement after a career which has spanned more than 40 years. Your cooperation with Messrs. Sager, Bohlke, and Craig in carrying out their new responsibilities will be appreciated. JHG:re ~ Executive Council: W. H. Bohlke J. E. Hertz J. L. Broadhead T. F. Plunkett J. P. Higgins A. Rodriguez Dennis P. Coyle Robert J. Hovey D. A. Sager K. M. Davis L. J. Kelleher Roberto Denis Dilek Samil D. M. Klinger J. E. Scalf P.J.Evanson S. H. Levin Maria V. Fogarty R. E. Stewart J. E. Geiger L. J. Gelber R. M. Marshall J. G. Milne J. C. Norris W. G. Walker, til Michael M. Wilson
\
k C. O. Woody W. W. Hamilton A. J. Olivera M. W. Yackira [/ V
.h .
TABLE OF CONTENTS Page SECTION I VICE PRESIDENT - ST. LUCIE PLANT ORGANI7sATION Vice President Organization . I Business Systems Department - Locn 9166 2 Contracts - Locn 9167 3 Corrective Action - Locn 9168 4 Management Information Systems - Locn 9048 5 Resource Control- Locn 9123 6 Licensing Department - Locn 9041 7 Materials Management Department - Locn 9046 8 Iluman Resources - Locn 9122 9 Site Engineering Department - Locn 9256 10 Quality Assurance- Locn 6270 11 SECTION 11 PLANT GENERAL MANAGER - ST. LUCIE PLANT ORGANIZATION Plant General Manager Organization 12 Maintenance Group 13 Electrical Maintenance Department - Locn 9103 14 Instrument & Control Maintenance Department - Locn 9102 15 Maintenance Services - Locn 9162 16 Rotating Equipment Department (Pumps, Motors & Fans) Department - Locn 9163 17 Vaires and Welding Maintenance Department - Locn 9165 18 Operations Group 19 Chemistry Department - Locn 9106 20 Ilcalth Physics Department - Locn 9104 21 Operations Department - 9107 22 Operations Support Department - Locn 9152 23 Work Control Department - Locn 9115 24 SECTION III SITE SERVICES MANAGER - ST. LUCIE PLANT ORGANIZATION Services Manager Organization 25 Information Services . Department - Locn 9109 26 Land Utilization Department - Locn,9257 27 Protection Services Department - Locn 9129 28 Security Department - Locn 9020 29 Training Department - Locn 9127 30
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- 3.A STAT 1 V. P. SECRETARY K WEST !
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BMtNESS SY$m PLANTCLNERAL MANAGER PROCESS STEAMGEN REPLACP_ IIUMAN RESOURCE I IICENSING MANAGER MATFRIALS MANAGER MANAGFR ' MANAGFR SERVICES MANAGER gyppggtggy PROJ MANAGFR l E. WEINKAM G.I BOISSY 3.SCAROLA l ! R EL DAWSON D. L FADDEN ' l. VOORitEES R SIPOS L C. MORGAN I l
! I NUCIIAR SPhAKOUT INFORMATION SERVICES i CONTRACTS ACTING WORK StTFRVISOR MAINTENANCE MAINTI' NANCE OPERATIONS I K.I.UCIfKA ! {
MANAGER SPFCIAllST MANAqER CONTROL MAN AGFR T. G EREINDERG _ # C GALLAGilrR l
- 1. W. INT ^^ ^
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- 3. MARCIIESE T. A.DILLARD pg it IOf tNSON -
SITE ENGINEF. RING ! LAND UTillZ % TION \I; % \ htANAfWR l [ CORRFCTIVE ACTION SUFFRVi m ELI:CTRICAL M MSMy SUPFRVISOR StTFRVISOR g D 2 DENVER N G L BOL,3KA g W.RORTE D H FAULKNER
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,yAN A^ PAWLEY i II. F. DUCffANAN g D M CALABRESE b YlSITORS CENTTR SECUPITY OPERATIONS i
RESOURCE CONTROL StTERVISOR MAINTFNANCE SUPERVISOR K A. SVODODA ! StMRVISOR PROCFDURES SUPERVISOR W. G. WITITE C.D MARa ' t R ltER(XIX OPEN I [ I TRAfNfNG MANAGER hGAINTFNANCE " " * * {
$rRvtCr3 $UrrRvtSOR M ALLEN M y R. FRECitETTE PREDICTIVE MA!NT a
StTI RVISOR i B. IL SCULTItORPE ' l t ROTATING EQUt* MENT StTFRVISOR > D.O ENGLISit VALVES & WELDING SUPERVISOR ! 1
. L N. MOTLEY 4
)
I SEPTEMBER 10,1996 REV 0 . l' age I y i
St. Lucie Nuclear Plant . Business Systems Group VICE PRESIDENT ST . LUCIE PLANT J. A. STALL 1.oc.9I66 DUSINESS SYSTEMS MANAGER R. E. DAWSON I x.9167 lac.9168 IEc.9048 Inc.9123 inc.9166 M ANAGMENT INIORMATION RESOURCE CONTROL CORRECTIVE ACTION ""' ^ CONTRACTS SYSTEMS SUPERVISOR SUPERVISOR SUPERVISOR T.G KREINDEx0 D. M. CALAURESE R.W. IIEROUX OPEN Page 2 SEPTEMBER 10,1996 REV 0
St. Lucie Nuclear Plant Business Systems Department Contracts Section
- Locn 9167 HUSINESS MANAGER R. E. DAWSON CONTRACTS SUPERVISOR T. G. KREINDERG CONTRACTS AGENT OPEN PLANTTECIINICIAN J. N. GRAYDUSII SEPTEMBER 10,1996 REV 0 p 3
- ._. _ _. . _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ . ~
St. Lucie Nuclear Plant Business Systems Department Corrective Action Section Locn 9168 BUSINESS MANAGER R. E. DAWSON CORRECTIVE ACTION SUI ERVISOR OI 'IN TRENDING /1IPES D.A.LOCKE OPEN CR DATAHASE TEClfNICIAN R. MIXON SEPTEMurR 10,1996 REV O Page 4
St. Lucie Nuclear Plant Business Systems Department ' Management Information Systems Section
. . LOcn 9048 DUSINESS MANAGER R. E. DAWSON M ANAGMENT INFORMATION SYSTEMS SUPERVISOR D. M. CALABRESE I
LAN ADMINSTRATION APPLICATION DEVELOPMENT ANALYSTS R. K. ANDERSON D. L DOTSON D. HROWN J. C. GLASS M. DIXON M. L I!ORTM AN OPEN J. P. MCGLENN OPEN SEPT EMBER 10,1996 REV 0 Page 5
St. Lucie Nuclear Plant
. Resource Control Department Locn 9123 DUSINESS SERVICES MANAGER R. E. DAWSON RESOURCE CONTROL SUPERVISOR R.W.IIEROUX ACCOUNTS PAYABLE DUIXiET TECIINICIAN COST ANALYS13 SUPERVISOR.
J. A. LAIIERA P. W. KENNEDY L. LI'N E. MESLIN C. PITI MAN J. S. SLIMM i SEPTEMBER 10,1996 REV 0
St. Lucie Nuclear Plant Licensing Department Location 9041 VICE PRESIDENI ST. LUCIE PLANT J. A. STALL LICENSING MANAGER E. J. WEINKAM LICENSING ENGINEER LICENSING LICENSING ENGINEER OPERATIONS EXP. NRR ANALYSTS TEClINICAL SUPPORT COMPI.!ANCE ENGINEER R. L DIETZ S.P. TREPANIER E. J. BENKEN C. F. FERRIDAY G. R. MADDEN OPEN R. NOrit.E P. T. QUILLEN . SEPTEMllER 10,1996 REV 0
- l _ _ _ _ _ _ __
h n St. Lucie Nuclear Plant Materials Management Department , Locn 9046 l SITE VICE PRESIDENT J. A. STALL I MATERIALS MANAGER G. J. DOISSY , I t I !
)
CEtrTRAL RECEIVING INVENTORY CON 1ROL ISSUES SUPERVISOR PROCUREMENT f DEDICATION ANALYST SUPERVISOR ' SUPERVISOR SUPERVISOR D. K. QUILTY A. R. NEURERGER M. G. ULLMAN L A. ROGERS W. A.LYNCII l TECIINICIANS PROCUREMENT e TECIINICIANS ANALYSTS AGENTS { J.J.DRACK W.C. ARMES S. D. DARRY ,J.D CIIESSER i C. R. OWENS T. E. EVANS J. C. EDGAR R. J. CIIOt.EWINSKI & S. D. PINKSTON ! D. A.KORNSTADT C. J. IIARRIS E. O. PRICE O. RICIIARDS C D. SUMMERS T. M. SilEIL f W. D. WillTE t
' TECIINICIANS ,
TECitNICIAN ! D, A.DIRNDAUM J. D. ClIAUVIN R. J. CONKLIN [ T. L CIIVILA D.M.GLUUKA . D. K. COX P. S. POULOS N. A.QUESADA
. W. M. QUIGLEY f
Page 8 f SEPTEMBFR 10,1996 REV 0 r
St. Lucie NucIcar Plant
~
Site Human Resources Department Location 9122 IIUMAN RESOURCES MANAGER L C. MORGAN EMPLOYMEt fr LAllOR RELATIONS ADhflNISTRATIVE ADMINISTRATOR ADMINISTRATOR A. A. GEllRING C. D. SCOlT K. M. NELSON SIIPTEMBER 10,1996 REV 0 Page 9
St. Lucie Nuclear Plant Site Engineering Department Location 9256 PSL ENGINEERING MANAGFR D IDENVER ADMfNISTRATiVE PLANT ENGINEER OPS SUPPORT PROJFCT PROJECT ENGR SYSENGR 58 'PF R VTSOR M AN AGF R FNGR SUPTRVTSOR FNGR FSAR MAINT Rtil E MANAGril R 1. OIURClt R D GTL 3 P FULFORD K. Moll!NDROO M $NYDER 3 MTST RE ACTOR FNGR DRAWING UPDAT E ! I I f 1 EAn FSAR Sys l l IFAD CIVIL ENUR DE31GN E A115 PkOCUREMENI P Barnes COMPONENT PflJM ARY 51TF R VISOR StfPrRVt3OR ENGR SUPV C. G OFARRILL R. Custis StPV St ?V DE GATES R FILIPEK ~ "' E A. FATA R WINNARD I f tOGE P rw M Mathaven 3 pagcp j pogypg B Pagneami D'80"T W 58"1 R M Klein civig p g,,,,,,, DD FOR's IAC Enges w p we,4 PSk* ComPanent Sys Primsey Sys R Cone C Buche.g D
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St 'TEMBER 10.1996 REV O Page 10 l
St. Lucie Nuclear Plant Site Quality Assurance Department Location 6270 SITE QUALITY MANAGER L. W. BLADOW QUALITY SUPERVISOR QUALITY SUPERVISOR QUALITY SUPERVISOR AUDITS /PERIORMANCl; INDEPENDENT TECIINICAL QUALITY CONTROL MONITORING REVIEW G. A.SCIINEBL1 D. LOWENS R. DE LA ESPRIELLA L DEARROR K 11UTLER '^ ^ D.liUEY D. P. KOENNICKE . A Rl SS J.1.OWERY C. NORRIS T. DWYER J. MCNEY D. PARKS II. FEINBERG S. RilIA OPEN D. GINGRAS R. WALCIIESKI OPEN R IMWLEY A W.llAYS R WEIS A.IlOSIE G. INGRAM M. MEILEY D. MELODY R. MILI.ER U. MOSS I. PANESSA R. SPARKS R. STORMS A. UREVITil OPEN SEPTEMBER 10,1996 REV 0 Page Il
.i l
i i SECTION II PLANT GENERAL YIANAGER ST. LUCIE PLANT ORGANIZATION
. .~ St. Lucie Nuclear Plant Plant General Manager's Organization . PLANT GI'NERAL MANAGER J.SCAROLA OPERATIONS MANAGER AGING OR NTROL MAINTENANCE MANAGER MAINTENANCE SPECIALIST J. M ARCitESE T. A. D!!. LARD II. II. JO!!NSON , e Pap 12 SEPTEMBER 10.1996 REV 0
I . . St. Lucie Nuclear Plant Maintenance Group PLANT GENERAL MANAGER J.SCAROLA I MAINTENANCE MANAGER J. MARCIIESE ELI:CIRlCAL ACTINGI A C MAINTI: NANCE MAINTENANCE ROTATING eiQUIP VALVES & WELDING SUPERVISOR SUPERVISOR PROCEDURES SUPV SERVICES SUPV SUPERVISOR SUPERVISOR W. KORTE A. PAWLEY OPI:N R. FRECIIETTE D. G. ENGLISil I N. MOTI.EY I I PREDICTIVE MAINT ROTATING ELEC'I RICAL I & C PROC MECil l' ROC WRITER Surf?RVISOR MAINTENANCE PROC WRITER WRITl R sign 7 II. DISIlOP A. DELGAM R. TOSCANO - L W. NEELY** I R. C. OLSON PM SPECIALISTS J R R. W. AKERMAN W. D. JORDAN 'h,
'^'~ '
R. S. KAPLAN OPEN
*
- ON ROTATION TO QA .
SEPTEMBER 10,1996 REV O Page 13
St. Lucie Nuclear Plant Electrical Maintenance Department . LOcn 9103 MAINTENANCE MANAGER J. MARC 1tESE ELECTRICAL SUPERVISOR W. KORTE A/C. I1OP, COMMON ELECTRICAL SUPPORT ELECTRICAL UNIT 1 PRODUCTION UNIT 2 PRODUCI10N SUPERVISOR ANALYST MAINTL' NANCE SUPERVISOR SUPERVISOR SUPPORT R. L IlOSKINS D. GILDERT R. ENSLEY II. GRINER J. TRINGALI SEPTEM11ER 10,1996 REV O Page 14
St. Lucie NucIcar Plant Instrument & Control Department Locn 9102 MAINTENANCE MANAGER
- j. MARCI(ESE I
ACTING INSTRUMENT & CONTROL StTFavdSOR A. PAWLEY I I _ l [ I tHTT I SLTTRVISOR UNIT 2 SUPERVISOR g,,, , SitOP SUPV. M&TE. EQUTP P. EPA!R IAC SUPPORT SUPI RVtSOR ANI) COhfMLHICATIONS A PAW 1.EY R D COLEMAN ,p SUPERVISOR R F. CARROT 1 E SYSTEM SLTERVISORS SYSTEM SUPERVISORS SOFTWARE ENUTNEERS F. E IIUBER i A C SUPPORT TFCil F. P CUSM ANO P1 TONES & DETPERS TFut
~
D S KtRgy J.KAWA D. L ANTitONY E C.ORDWAY F. L KENESSEY D B CARLSON G EDWARDS T. DWS R D PITTS R M LAW 5 L Mit.FS G R SOfMtD T. A NEW110USE N R RADAK R Sitt RMAN OPEN E O TTCTINICTAN MATE TEClfNfCIAN R 5101DREN RADf ATION MONITORING T stOSS!NI TEOt R IBECKER I SEPTEM11ER 10.1946 RE V. O Page 15
St. Lucie Nuclear Plant Maintenance Services Locn 9162 .
~
MAINTENANCE MANAGER J. MARCIIESE I MAINTENANCE SERVICES SUPERVISOR R. J. FRECllETTE I I I l DWA/CCO ANALYST M AIN I ENANCli OU I'AUli MAIN IENANCE SUPPORT PROJLiCIS & SUPPLLMEN I AL lECilNICIANS SERVICES SERVICES SUPERVISOR LABOR SUPERVISOR R. UMSTEAD M.BLOESER IL E. BALL S. MARCII!GIANO C. G. CRIDER P. CORDOVA J. IlULL INTAKIdCCW SUPV ^ SCAL 1 OLDING & SI'l E
. NU UPW W. JENKINS D.YATES M. GOODIEL
^^' "
SG/llEAT EXCIIANUERS/ OPEN STATION \RY SUPV SIIli PAIN IINU/ INSULA'llON Sii E 'IOOLING & MISC PROJEC SUPPORT A. FLOWERS J. C. JONES P. IX)NOVAN
'l URillNE/REAC1OR/ RAMP L.MII.LER SUPERVISOR _
liSIIMAlING VElllCilLES/FACILillES/ D. R. NICllOLS LOCKSMITil B. G. BUTCllER R. STANTON SEPTEM11ER 10,1996 REV 0 Page 16
.1_... - _ _. _ _ _ . _ _ . _ _ ..___.______________...___._____.____.m____._ . _ _ . _ _ _ . _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . . _ _ _ _ _ _ . . . _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ -
St. Lucie NucIcar Plant Rotating Equipment (Pumps, Motors, and Fans) Locn 9163 w MAINTENANCE MANAGER J. M ARCllESE ROTATING EQUIP. SUPERVISOR D. G. ENGLISil CRANE SUPPORT TECII PRODUCTION SUPPORT R. O. STOLL - J. W. MILTON R.A. DECKER SEPTEMBER 10,1996 REV 0
St. Lucie Nuclear Plant Valves and Welding Department Location 9165 PLANT GENERAL M ANAGER J. SCAROLA MAINTENANCE MANAGER J. M ARCIIESE VALVES & WELDING SUPERVISOR L N. MOILEY MMA ALE ONDAR VAN AOV MAINT SUPV MOV MAINT SUPV 3p p, VALVES MAINT SUPV WELDING SUPV M. LITTLE D.DOYETF E. IIASSELt. G. PUSTOVER D. JACOHS R. ERICKSON
- 1. A. COOK - MOTOR OP VALVliS J. KUNKEL- WELDING COORD.
S. OEllRLE - AIR OP VALVES G. ROlXiERS - SITE WELDING ENG Page 18 SEPTEMBER 10,1996 REV O
St. Lucic Nuclear Plant Operating Group PLANT GENERAL MANAGER J.SCAROLA OPERATIONS MANAGER II. II. JO!INSON CIIEMISTRY IIEALTil PIIYSICS OPERATIONS OPERATIONS SUPPORT SUPERVISOR SUPERVISOR SUPERVISOR SUPERVISOR D. II. FAULKNER II F.13UCllANAN C. D. MARPLE R. O. ENFINGER SEPTEMI1ER 10,1996 REV O . Parc 19
St. Lucie Nuclear Plant Chemistry Department - Location 9106 PLANT GENERAL MANAGER J. SCAROLA OPERATIONS MANAGER II. II. JCitNSON CIIEMISTRY SUPERVISOR D. II. FAULKNER ENVIRONMENTAL PRIMARY SUPV PURCIIASING/ DATA RADIOLOGICAI, SECONDARY WTP / STP POLISIIER SUPERVISOR DATA MANAGEMENT MANAGMENT EFFLUENTS SUPV SUPERVISOR SUPERVISOR A.P.13UTLER J. II. I1URGESS S.11ROMSTRUP R. E. COX J. SEAGER D. A. IIART E SENIOR LAD TECIINICIANS E. R. DOSTER D. M. M AJEWSKI D. J. EISERT E. II. MEYER J. R. GEORGE 5. F. PEDDICORD W. M. ITUFFOR V. S. SCllWERER G. J. KINGSLEY M. L SERRUTO L M. LEllLANC W. II. VOLIJt AR ROTATED AMONGST SUPERVISORS SEPTEMBER 10,1996 REV O
St. Lucie Nuclear Plant Health Physics Department Location 9104 PLANT GENEPAL MANAGER J. SCAROLA I OPERATIONS MANAGER II. I1. JOIINSON llEALTil PIIYSICS SUPERVISOR II. F. DUCIIANAN llEALTil PilYSICS IX)SIMETRY IIEALTil PilYSICS OPERATIONS SUPERVISOR SUPERVISOR TECIINICAL SUPERVISOR R. M. McCULLERS 11. N.10!INSON II. M. MERCER SENIOR l PERF.TECilNICIAN TECIINICIANS
- 11. R. JOllNSON Sil!PT SUPERVISORS P. J. DUNNE D DEllOCK-MIDS ^ " AtARA SUPERVISOR IOANE WAm V. E. MUNNE-UI DAYS '" '"
A.G. WIER-U2 DAYS OPEN-PEAKS II. K. MOURING
. . PAM R.13. SOMERS RAD WAS11!
DECON SUPERVISOR ANAI,YST J. R. SMITII P. STONER RADIOACTIVE PACKAGING SUPERVISOR D. C. II AITilCOX SEPTEMilER 10,1996 REV 0 p,.,p ; g
St. Lucie Nuclear Plant Operations Department Location 9107 PLARI GrMERAL MANAGER J. SCAROLA OPERATIONS MANAGER
- 11. II. JOliNSON I OPERATIONS SUPERVISOR C. D. MARPLE l
NUC Pt.AM SUW ASST NUC PLT SUPV D. K. ALDRilTON
' I ^ R D BROWN J. P. IlONEYSETT C.F.DIAZ D. ISAAG R. T. DIEf !L C. D. LADD M. u MB J. S. SANDY J. R. MARTIN
. D. WEI.t.ER A. J. SCALES A. M. TERE 7.AKIS J. D. TOTTON M. E. WRIGilT "
OPEN OPEN OPEN OPEN OPEN SEPTEMBER 10,1996 REV O + On rotation to Work Control Department Page 22
e . 1 St. Lorie Nuclear Plant Operations Support Department Location 9152 OPERATIONS MANAGER
- 11. II. JOIINSON l
OPERATIONS SUPPORT SUPERVISOR R. O. ENITNGER ANALYSTS SRO DEVELOPMENT T1ICIINICIANS R. F. CZACIIOR R. ITUGIIES 13. E. CAMPLIN M. GILMORE K. KORTII R. J. DAILEY M. KOGELSCllATZ W. L PARKS J. VALDES J. S. NAPIER O. SMITil S. M. WIIITEIIEAD S. E. PATTERSON C. SWI ATEK R. M. PENNENGA C. WARD M. D. POTTORrF SEPTEMBER 10,1996 REV 0 Page 23'
St. Lucie Nuclear Plant Work Control Department Location 9115 PLANT GENERAL
. MANAGER 1SCAROLA I
ACTING WORK CONTROL MANAGER C.II. WOOD I I I I l DAILY SCIIED IMPLEMENTATION PLANNING SUPERVISOR PIANNING/ PARIS SYSTEM SUPERVISOR SUPERVISOR SUFFRVISOR ANALYST J. E. IIAUGER OPS ROTATION-2 YRS A. W. M ARVIN A. G. MENOCAL D. I,. If0 WARD I I i [ [ OPS MAINT SCilEDULING SC11EDULING ANALYST I &C /ELLC PLANNING MECll. I't .ANNING PARTS PLANNING SPECIAL PROJECTS ANALYST SUPERVISOR SUPERVISOR SUPERVISOR COORDINATOR- -
& PWO Cl.OSEOUT R. J. M AIS ANO SUPERVISOR R. J. DAVIS J. S. MONALDI R. II. MAY'IEW P. J. SARNO P. N. CONNOLLY A DEROY PARTSFtANNER$
tar ANALYST OTAl nWAU OITI' AGE SCllEDULING P2 TECilNICIANS __
- B W.BLASClfKE TE'C1'NfCIANS Et. ALTFRMATT I D Of*LSON DAILY SCl!EDULERS G. ANDERSON P u cRrrris , mssrNTtNo s s otArt N M. G. IIAVERLAND D. GLEN u- = ,",8, [y77 7 F McKFON C i ItETNULD R. K. ROGERS ** " "* ^ "N A W. D. CilESNEY T.C. GLENN .-
R. CLINE J. M. SCIIORN R. F. GROSS T Mn.AM VALVMVFI D PL ANNFRS L JACODUS K.L.TARDES B. IIAINES D 8 FAGE .
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** On rotation from Operations SEPTEMITER 10,1996 REV 0 Page 24
SECTION III SITE SERVICES MANAGER ST. LUCIE PLANT ORGANIZATION i a I i I
9
. St. Lucie Nuclear Plant Services Group SERVICES MANAGER D. L FADDEN I
INFORMATION SERVICES LAND UTILIZATION PROTECTION SERVICES SECURITY SUPERVISOR TRAINING MANAGER SUPERVISOR SUPERVISOR SUPERVISOR W. G. WII!1E M. ALLEN J. W.IlOLT G. L DOUSKA OPEN SEPTEM11ER 10,1996 REV O Page 25
St. Lucie Nuclear Plant Information Services Department LOcn 9109 INFORMATION SERVICES SECTION SUPERVISOR J. IIOLT I I I l CONFIGURATION AUDIO / VISUALS PROCEDURES SUPERVISOR NUCLEAR RECORDS / MGMT SUPERVISOR DOC. CONTROL M. S. STOCKER S. A. VAI. DES SUPERVISOR D. A. DROWN T. V. OSWALD Si1E GRAPillCS - CONFIGURATION ANALYSTS MANAGEMENT NUCLEAR RI?tT)RDS L M. DONGltlA TECIINICIANS P. A. IIASILE C. P. VANDERSCIIOT J. A. CIMINO J. J. D AILEY T. M. MAITLAND J.G.MANN M. D. WILLIFORD PROCEDURES K. J. MANNING WORDPROCESSINGTRG N. J. WEST TECIINICIANS
- fj. C. lARANO\ ' SKY EUWM tmTRot, TECllNICIANS M. P. DIM ARCO
- 11. S. G1LMOUR l OPEN J. L DUTClIER C. II. GillSON PROCEDURES PRODUCTION K. L IIALVERSON 1TCIINICIANS P. G. IIAMIL1DN M. J. LANE
. L R. DAVIS V. L McDANIEL K. A. WOOD T. L TAYLOR CPEN SEPTEMilER 10,1996 REV 0 Page 26
St. Lucie Nuclear Plant Land Utilization Department
. LOcn 9257 SITE SUPT. I.AND UTILIZATION _
G. L DOUSKA SR ENVIRONMENTAL ENVIRONMENTAL SR ENVIRONMENTAL' LEAD SITE INSP. INSPECTOR INSPECTOR MAINTENANCE RFP. S.M. FOSTER G. W. IIANSEN D. A. SINGEWAI.D J. J. TOEI3E SEPIEMBER 10,1996 REV O Page 27
St. Lucie Nuclear Plant Protection Services Department Locn 9129 FRCTLCTION SERVtCES , SUPERVISOR OPEN FIRE PROTECTION kfEDICAL /3AFETY ^ ' Al3 ENVIRO f%1ANCE
$UFERVtSOR T SUPERVISOR R A ktcDANTFL T. h10SER ,I hl TORDTN N 91 WIffTING D WlitTWFLL FIPI PROTECTION ANALY11 WIfSANDfL FIRE PROTFCTION ANALYST R BritRE SEPTEMDER 10,1996 REV 0 Page 28
St. Lucie Nuclear Plant Security Department Locn 9020 SECURITY SUPFRVISOR W.G WiliTE ACCESS CONTROL FITNESS FOR DUTY SECURITY OPFRATIONS TRAINING SPFCIAlfST COORDINATOR 51,TTRVISOR SPFCI At (ST J H ALLEN A R Cl%%tfNGS R W.CZARNFCKt S. P PLANTZ 51ffFT SPf CIALISTS R T.BITNFR R L ftOSKEY R F DORST 3 A TALLEY SEPTEMBER 10.1996 REV O Page 29
St. Lucie Nucler Plant Training Department Locn 9127 i SITE TRAINING MANAGER M ALLEN I I I I AS$rS5ATENT/ M AINTFNANCE TRAfNTNG TERA ff'HS INITIAL TRAININti STMULATOR ENGINEERING TECyfNTCALTRATNtNG SECTION CONFIGURAT10N SFCT!ON SUPERVTSOR SECdON StiPERVISOR SECT!ON SUPERYtSOR SLTFRY!SOR StTE RVISOR L C CROTEAU l 8 A SPODICK '4ARTIN T. I WARE
'~
K $ METFGf R - RCOSTA TRAfNfNG INST RtCTIONAL ELFCTRICAL TRAINING WSTRtCTORS NRY mmmM INSTRLCTORS SIMULATOR SOFTWART TFCIMM OGIST ENGTNEFRS T DOLANDER D N WOOLDRI1XiB K K Akt]LARIS I ffARRIS J C COUTURE R. I L PACKWOOD A T IfAt.L G T LOREE R. ANDERSON J A. EfAGENNTS S E.MOltN P.IROGERS R P. WATSON ON SHIFT COACH I W WEEK' its: ALTil Pf fYS?CS INST R11CTOR W Bl_OESER INSTRUkfENT A CONTROL R GOI DSTEIN TRAINTNG INSTRUCTORS SRO INSTRUCTOR SIMULATOR LEAD !!. J LFIF]If1 M I ITFfHOL D ltARDWARE St?Y. W W ARMSTRONG D H BORGMANN D. A MALEY D F. HOULDSWORT11 CONF fGURATION TRAINING SUPPORT CONTROL hlHITTAL TF AfMING pp g_ g _ fMSTRUCTORS - MECHANfCAL TRAINING
~
p g, E. L COX FR L A SPAT DfNG INSTRUCTORS 3 A stAnyt M R ANDERSON GrNFRAL EMPt.OYFE ASSES $hfENT TRAR M INSTR U M S COORDINATOR WSTR M W AL _ TFCITHOf.OGTST i A C IMGITAL SPfCIAt IST M L COOPFR PLANNERA4AINT. SUPY. St,KNAPP R E. DODO J A LFWfS T]t AINING tNSTRtCTOR 3 M DAXLEY G L PtitPr$ M ** GROOM TfTNESS FOR DUTY 1 TECffNICIAf4 ITAZAR DOUS MATT RIAL REQtJAL TRAINING $UPV - INSTRUCTt HL K F. McMANUS L M RICH C fl STROUD TEC11NfCIAN LICENSED OPERATOR REQUAl i TRST AfD!tTR 6NSTRUCTOR I R OTX)NP= ELL G C FARKFR D E. BROWN
- f. D CARPENTER K.M CARPENTER W L DAIKilfTRY P F.FARNSWORTil
. u ! IMBRIALE e R A. WALKER STA OPEN CUpT1:h40FD in 8 00f. R FV n Pee 30
a M Inter-Office Correspondence FPL JPN-SPSL-95-0612 T.: S. A. Valdes o: Dec 19,1995 St. Lucie Plant
%: D.J. Denver % JPN/PSL Nuclear Enginee g M se: ST. LUCIE PLANT UNIT 1 PC/M 236-195, revision 0,l',isconnection of ICW/CCW Pump Local Pushbutton Stations.
l i Attached for your use is PC/M 236-195 revision O, which evaluates the disconnection and removal of the local pushbunon control stations for the 1 A, 18 and 1C ICW and CCW pumps. This will prevent the operator workaround associated with control switch reset whan the pumps have been stopped ; locally (reference STAR 94110473). If there are any questions on this package, please coittact Warren Busch at ext. 7484. l l l { fd ilC'.Q
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QI 3-PR/PSL-1 lt Revision 37 .. December,1995 ' Page 20 of 26 1
- ATTACHMENT 2 PC/M REVIEW FORM l
! (Page 1 of 2) 2 l PC/M Number 5 * /W Supplement Number D Expiration Date #2 / / / 75 J ! PC/M TITLE: i b1 s connect;en of ICW /CC W L ec / Pa t h bums sat.b j Safety Classification: -
~
X Safety Related Quality Reisited Not Nuclear Safety Related AdmirusWtive PC/M Division: - M Normal Either/Or As-Requested-Package As-Fail Implementing Documentation: l Department ER/NPWO # W/O # WCD# EM GS //4.1LT 4feaoolJJ
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i - i O! 3-PR/PSL-1 Revision 37 !. December,1995 Page 21 of 26 " ATTACHMENT 2 ; PC/M REVIEW FORM l (Page 2 of 2) i PC/M Review and Approval: . / Date 8/ / V SCE
/h 7
Date S //Y /SS , Maintenance QC I f S br l kM&Date 3 //b ff Plant General Manager i FRG Number: kk/%/ FRG Secretary Date 3 / 8/ M Comments: />j2goOc fibtDfbMs 6^)LtWes /432-6/453-C T lM(fCN)fuws na AbMPM Fod fMs -/6].F 66//tDJ /434 A3f-415, w typ-45~ l f d6 .i w-R , 4 i PC/M Closeout Review: Approved for Closeout: / Date: / i Maintenance OC Approved for Closeout: / Date: / SCE ' Procedure Notification Complete: Date: / / SCE Training Notification Completed: Date: / / SCE Cancellation information: ' Date: / / SCE e 9
FPL NUCLEAR ENGINEERING ENGINEERING PACKAGE PC/M NO: 236-195 REV: O SUPPL: O PLANT: ST. LUCIE Ut.lT: 01 , 1 TITLE: DISCONNECTION OF ICW/CCW PUMP LOCAL PUSH-BUTTON STATIONS REVISION DESCRIPT!ON: ISSUED FOR USE LEAD DISCIPLINE: ELECTRICAL EXPIRATION DATE: 12/1/96 REA NO: STAR O-94110473 DWA NO: N/A , PC/M CLASSIFICATION: SR X QR NNS DESIGN ORGANIZATION: PSL/ SITE DISC CHIEF REVIEW REQD? YES NO X EXTERNAL INTERFACES: NONE DISC CHIEF SIGNATURE: N/A REVIEW / APPROVAL: INTERFACE TYPE GROUP PREPARED VERIFIED APPROVED FPL APPROVED
- ENPUT REVIEW N/A mm --
MECH X . , .
/7 ELECT X
/ A /+ _
I&C X civil X NUC** X d$ 4. TTLEW ESI X NUC FUEL X
- For Contractor Prepared EPs As Determined By Projee *
- Review interf ace As A Minimum On All EPs FP1, PROJECTS APPROVAL:
/ y-
/
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- DATE: ! 21 i I l l
i I
PC/M No 236195 Rev O Page L of 14 LIST OF EFFECTIVE PAGES PAGE NO. g 1 o 2 0 3 o 4 o 5 o 6 o 7 o 8 o 9 o 10 o 11 0 12 0 13 o 14 o l l l l O i l l
.. - . - - -- - -~ . . -- - - . . - - - - - . - - - - - -
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PC/M No 236-195 Rev 0 j Page 2_ of 14
- ABSTRACT i
This modification disconnects the local push-button control station from the I A,18, and 1C Intake ' Cooling Water (ICW) pump and the 1 A,18 and 1C Component Cooling Water (CCW) pump control circuits. This modification is being implemented to delete the operational requirements of manually reseting the RTGB control switch to "stop* and then back to " auto" which may be required to preserve , the automatic SIAS start feature of the subject pumps. The reseting of the RTGB control switches is ' i only required after a local push-button stop of a running pump which was started via the RTGB control . switch (reference, STAR 0-94110473). Disconnection of the local push-button stations from the ) associated control circuit will prevent the stopping of a running pump via the local push-button station 3 and thus preserve the automatic SlAS start fea;ure of the pumps without requiring a manual reset of 4 the RTGB control switches. 4 j The modification has no impact on the safety related functions of the ICW and CCW control circuits 1 and is being implemented to prevent the operator work around associated with the manual reset required after a local push-button stop. The modification has no affects on plant operations since the i local push-button stations are not required during normal, abnormal, emergency or maintenance operations. 4 This modificatiors includes the safety related ICW end CCW pump control circuits, therefore.this EP is classified as Safety Related. t 4 A safety evaluation of these modifications has been performed in accordance with 10CFR50.59. This evaluation concludes that implementation of this EP does not involve an unreviewed saf:ty question nor does it require a change to the Technical Specifications. An FSAR change package is provided to reflect this modification. It has no adverse effects on plant operations or safety. Therefore, prior NRC approvalis not required for implementation of this modification. i l 4 f 4
PC/M No 236195 Rev 0 Page 4 of 14
- 1. DESIGN
- 1. Structure. Svstem. or comoonent (SSC) involved:
ICW Pump 1 A,1B and 1C and CCW Pump 1 A,1B and 1C control circuits, specifically: Local push-button control stations for each of the pumps.
- 2. Purnose/ Function /Desian Basis of SSC Involved:
The ICW and CCW Systems provide a heat sink for safety related and non nuclear safety related components under normal operating and emergency shutdown conditions. The redundant pumps in each system (including the swing pump 1C) are controlled by pull to-lock control switches at the RTGB, control switches at the respective 4160 volt cubicle for each pump and local push-button control stations for each pump. The local push-button control stations are mounted in areas adjacent to the respective pump and allow the applicable pump to be started or stopped locally. The pushbuttons were installed for maintenance convenience and are not used for any design basis operating function or postulated accident. The local push-button stations are not referenced in normal, abnormal, or emergency operating procedures nor are they used during maintenance activities on the ICW and CCW systems.
- 3. Safetv Classification of Desion Chawg SR X QR NNS
- 4. Purcose of Desian CPE22 This EP is being prepared as requested by the St. Lucie Operations Department as an alternative to the existing operational requirements of manually reseting the RTGB control switch to "stop" and .
then back to " auto"in order to preserve the automatic SIAS start feature of the subject pumps. This requirement is only present after a local push-button stop of a running pump started via the RTGB control switch (reference PC/M 182-193, STAR O-94110473). Disconnection of the local push-button stations from the associated control circuit willinsure that stopping of a running pump via the local push-button station is not possible and therefore will inherently preserve the automatic SIAS start feature of the pumps without rcquiring a manual reset of the RTGB control switches. The current configuration of the push-button control circuit is considered an " operator work around" and is thus being rectified via this PC/M.
' The local start and stop features of the pumps are described in the FSAR text, sections 7.4.1, 9.2.1 and 9.2.2. This requires the modifications be implemented under the documentation of an Engineer'ng Package with an accompanying safety evaluation performed in accordance with 10CFR50.59.
C
m . _ _ _ _ _ _ l l i !' PC/M No 236-195 Rev 0 Page L of 14
]
l 1 I
- 5. Descriotion of Desion Chance l
l l The modification consists of disabling the local push button stations for the subject pumps by l disconnecting the local push-button control station wiring at the respective 4160 volt switchgear cubicle for each of the respective pumps and at each local push-button station. Control wiring will
)
be spared and insulated in place at each cubicle and at each push button station. Finally, the ' pushbuttons and the box nameplates will be removed. I I I i l. i 1 i i 1 1 i l r
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l l l 1
PCIM Mo 236-195 Rev 0 Page 6 of 14
- 6. Desian Chance Checkhst Does the Design Change involve / impact / require justification of:
4 y.ES. NQ REFERENCE Internal / External Flooding 2. Heavy Load Handling _ 1 Tornado / Internal Missiles _ l. Single Failure Criteria _ 1 Human Factors _ 1 3 Paging System Audibility _. l j Masonry Block Wall interaction _ 1 Environmental Criteria 1 Plant Security Capability 1 HELBA Criteria / Analyses _ X_ Seismic Qualification _ l
- . Seismic Interaction _ 1 Electrical Separation Criteria l.,
) Accessibility /Laydown/ Clearance Requirements 2L
- Loads Applied to Existing Structures (+ buried) _ l.,
EDG/ Battery Loading / Load Sequencing _ X i Hydrogen Generation in Containment _ 1 Heat Sinks in Containment l_ ) } Emergency Plant Operating Procedures (EOPs/ONOPs) 2 I 8.1 ASME Code _ 1 ! Emergency Lighting Criteria l. 1 Snubber Criteria __ 1 ! ) Meterial Compatibility / Hazardous Materials l. Electrical Equipment Grounding _ l., l } Cable Tray Seismic Loading 2. ,, j Instrument Setpoints 2.
- Hurricane / Tornado Wind Loading 2.
i Thermal / Hydraulic Performance _ l_ Coatings inside Containment __ l.
- Emergency Response Data System (ERDS) 1 Emergency Planning _ l.
j NML Property insurance Requirements
,_ l_
4 Environmental Qualification l_ Att. 4.1 Fire Protection Capability X Att. 4.2 1 4 Safe / Alternate Shutdown Capability 2 _ l-8.2
, ALARA Exposure Criteria Att. 4.3 l_.
4 4 JPN Forms 3E,3F, and 71 shall be attached to justify the conclusion for these items. i
i , . PC/M No 236-195 Rev 0 Page 7 of 14 !
- 7. Desian Evaluation / Justification 1
The design function of the ICW and CCW systems is to provide a heat sink for Safety Related and Not Nuclear Safety components under normal operating, post accident and normal shutdown conditions. The pump control circuit design utilizes RTGB mounted pull-to lock control switches for control of manual and automatic design functions of the applicable pumps. The local push-button stations for each pump allow local starting or stopping of the pump, t The local control features for these pumps are based on maintenance convenience and are not t depended upon for manual starting or stopping design functions of the applicable pumps under j normal or abnormal operating conditions. Tne local push button stations are not a functional part of the " normal", "off normal" or " emergency operating" plant procedures, i j Although the pushbuttons were installed for maintenance convenience, maintenance performed on these pumps is controlled through the plant clearance procedures that control pump power supply ! at the respective 4160 voit switchgear breaker cubicle. Maintenance is carried out on pumps when a clearance is taken out on the respective switchgear cubicle feeding each pump. The ] programmatic clearance procedures ensure that motors are not inadvertently started during i maintenance activities and thus ensure personnel safety. Post maintenance testing on the pumps is coordinated with operations and any starting or stopping is actuated from the control room RTGB control switches. Therefore, the local push-button stations are not used by maintenance departments and are not required for normal maintenance procedures. 1 Finally, there are no OSHA or personnel safety requirements involved with the removal of these j tocal push-button stations because the programmatic clearance procedures used during maintenance of pumps precludes inadvertent motor starts during maintenance activities. Motor i starting and stopping is coordinated with the operations department which utilizes both the 4160 volt switchgear cubicle feeding each pump and the respective control switches at the RTGB. I The modifications provided in this EP do not adversely affect the ICW or CCW Safety Related l systems' abilities to perform their intended Safety Related functions. Separation between Safety I Related functions and Net Nuclear Safety functions has not been affected by this modification. 4 i ] 8. Evaluation of env "YES" resoonses in Desian Chance Checklist ! , 8.1 Emergency Plant Operating Procedures i Off-Normal Operating Procedures 1-0310030 (CCW) and 1-0640030 (ICW) shalibe l revised to reflect the disconnection of the local push-button stations for the ICW and ! CCW pump control circuits. l 8.2 Safe Alternate Shutdown Capability Since separation requirements defined in Appendix R cannot be provided for essential components and circuits in the event of a fire in the control room or cable spreading room, alternative shutdown capability is provided. This ensures that in the unlikely event a fire makes the control room uninhabitable or renders equipment in either room inoperable, the plant can be safely taken to cold shutdown from a remote location. 5 The CCW and ICW pumps are considered Appendix R essential equipment (ref. FSAR table 9.5A-4) and their function is required for safe shutdown. Because the systems are required to achieve safe shutdown, Appendix R requires that the functions.of the
PC/M No 236495 Rev 0 Page J _ of 14 pump b'e controllable from an alternate location. Control for all 3 ICW pumps and all 3 CCW pumps is provided in the control room and at the respective 4160 volt switchgear cubicle feeding each pump. The cubicle control switches which are located in the respective switchgear cubicles feeding each pump (4.16 kv 1 AB switchgear, 4.16 kv 1 A3 switchgear and 4.16 kv 183 switchgear all of which are on the appendix R Essential equipment Lis and subsequently treated as essential (reference FSAR table 9.5A-4)) are the Appendix R alternate means of control and rened upon should an alternate means of control be required. Push-button stations are not relied upon for alternate means of control. The Appendix R Safe Shutdown Analysis for Unit 1 is being revised by deleting cables associated with the pushbuttons.
- 9. Desian Verification Statement The design basis for the modification was reviewed to ensure that the overall design concept meets applicable FSAR, Regulatory Guide, and 10 CFR Part 50 requirements. The design of the modification was verified by a review of the functional requirements of the system and by the performance of a circuit analysis to determine if the proposed modifications perform the functional and safety requirements described in the FSAR. Integration of this modification with other designs in progress and impact on existing safety evaluations was verified by review of the Package Information Tracking System and the Affected Drawings List databases, including the Active Safety Evaluation List. The safety analysis was verified by review of each conclusion statement required to satisfy 10 CFR 50.59 criteria for establishing whether an unreviewed safety question existed.
- 11. SAFETY
- 1. Descriotion and Puroose This Engineering Package covers modifications to the control circuits of the 1 A,18 and 1C Component Cooling Water Pumps and the 1 A,18 and 1C Intake Cooling Water Pumps. Specifically, this PC/M is an alternative to the procedural controls required to preserve the automatic start of respective pumps on SIAS after a running pump stop actuation by the local push-button control station. Under normal operation, two pumps will be running with their control switches in the
" Auto-after-Start" position, and the third pump will be idle with its control switch in the " pull-to-lock" position. The pumps will no longer be able to be controlled from the local control station.
This will not affect the proper response to a LOOP, SIAS, or LOOP /SIAS condition and will prevent the requirement of a manual reset as discussed in I-4.
- 2. Analvsis of Effects on Safety The design bases for the ICW and CCW systems are described in FSAR Sections 9.2.1 and 9.2.2.
, This modification will not affect the bases as described in the FSAR and will not result in loss of any protective function nor adversely affect the function of any Safety Related or Not Nuclear Safety structure, system or component. The removal of the local push-button control stations will not affect normal or off-normal operation of the respective pumps and will have no adverse effect on safety.
The new design prevents the local stopping of an ICW and CCW pump, which in turn reduces the number of procedural controls associated with the manual actuation and subsequent actions
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PC/M No 236 9 95 Rev 0 l- Page 9 of 14 following local push-button control station actuation. , The current configuration relies on the control room operator being alerted to the loss of a pump, . by various flow and temperature alarms. Removal of the local stations prevents the need of a pump j restart by turning its RTGB control switch to "Stop" and then " Start". In addition, if the automati:
- S!AS start feature was desired after a local control station stop command, the control room
' operator must turn the RTGB control switch to "Stop" and then back to " Auto" (the spring return position). A design integration review has been conducted, and the following sources have been reviewed: PSL Package information Tracking System, as of 12/15,1995. Nuclear Engineering Drawing index, as of 12/15,1995. PSL Active Safety Evaluation List, as of 12/15,1995. As a result of this review it has been concluded that no pending modifications or designs in J. progress will be affected by this modification. = 3 Based on the review of the St. Lucie Unit 1 Safety Evaluation List, there are no active safety evaluations issued or in progress which have any effect on, or can be affected by this EP. In addition, there are no Justifications for Continued Operation (JCOs) which affect, or are affected , by this EP. [ 3. Failure Modes and Effects Analysis (FMEA) The Single Failure Analysis and FEMA are provided in FSAR Sections 9.2.1 and 9.2.2 and in Tables ! 9.2-2 and 9.2-6. A review of these sections and tables has determined that this mo6fication, ! which changes the pump control circuits, does not introduce any new f ailure modes to the pumps. l Disconnection of the push-button control stations increases the overall reliability of the pump = control circuits. Therefore, the conclusion of the existing FMEA is not changed. l s
- 4. Effect on Technical Snecifications l The ICW and CCW System requirements are contained in Sections 3/4 7.3 and 3/4 7.4 of the j- Technical Specifications. A review of the Technical Specifications has determined that this l , modification does not have an affect on, nor require a change to them.
- 5. Unreviewed Safetv Question (USO) Determination i 1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR?
4 This modification does not affect any equipment whose malfunction is postulated in the FSAR { . to initiate an accident. The modifications performed by this Engineering Package enhance the .! ability of the pumps to perform as intended during. emergency and off-normal conditions. Therefore, the probability of occurrence of an accident previously described in the SAR is not increased by this modification.
- 2. Does the proposed activity increase the consequences of an accident previously evaluated in the SAR7 t
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PC/M No 236195 Rev 0 Page 10 of 14 This modification does not affect the ability of the ICW and CCW pumps to perform their safety function nor does it adversely affect accident mitigation functions. Disconnection of the local push-button stations minimizes the procedural requirements associated with maintaining the automatic start feature of the pumps after being stopped by a local push button actuation. Therefore, the consequences of an accident previously evaluated in the SAR are not increased by this modification.
- 3. Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR7 The operability of the pumps in this modification is being enhanced by eliminating a possible scenario in which control room operators are required to detect and compensate for a local push-button stop of the subject pumps which requires manual resetting of the RTGB control switches in order to re-start a running pump or preserve the SIAS auto start feature of.an idle pump. Since the reliability of the pump control circuit is being improved and the procedural requirements associated with the function of the pumps are being minimized by this proposed
[ change, it will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR.
- 4. Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR7 The operability of these systems is being enhanced by this modification. Since the pumps will continue to perform and satisfy their design basis requirements, the proposed change will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR.
- 5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the SAR?
This modification does not change the design bases of any structure, system or component important to safety as described in the SAR. Since this modification does not affect any equipment whose malfunction is postulated in the SAR to initiate an accident, it will not create the possibility of an accident of a different type than any previously evaluated in the SAR.
- 6. Does the proposed activity create the possibility of a malfunction of equipment important to i safety of a different type than any previously evaluated in the SAR7 '
Disconnection of the stations ensures that local push-button pump stop cannot occur and therefore eliminates the possibility of equipment malfunction associated with local push-button ; station actuation and the procedural controls resulting from said actuation. No new failure modes are created by this modification. Therefore, the possibility of a malfunction of equipment important to safety which is of a different type than any previously evaluated in the SAR is not created by this modification.
- 7. Does the proposed activity reduce the margin of safety as defined in the basis for any
, Technical Specification?
This modification does not adversely affect the operational requirements for the cooling systems as defined by the Technical Specifications Sections 3/4 7.3 and 3/4 7.4. Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced by this modification.
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PC/M No 236195 I Rev 0 Page 11 of 14 i l i J i 6, Plant Restrictions This EP can be partially implemented without any adverse consequences. The implementation may be performed during any mode of operation on one system pump out of service at a time. The following Operating Procedures and Off Normal Operating Procedures should be revised to , i reflect the removal of the local push-button stations: ONOP 1-0640030, ONOP 1-0310030. I Procedures should not be revised until actual implementation of PC/M on all pumps for the applicable system, l 7. Conclusions The foregoing constitutes, per 10CFR50,59 (b), that the modifications to be performed by this EP l do not involve an unreviewed safety question nor a change to the Plant Technical Specifications. l Therefore, prior NRC approval for the implementation of ,these modifications is not required, I l l l 5 I i
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PC/M No 236105 Rev 0 ;
- - Page 12 of 14
{ i Ill. CONFIGURATION
- 1. Affected Document Checklist 1
YES MQ Reference FSAR 1 _, Attachment 4.5 I Technical Specifications TEDB _ 1 Security Plan 1 Attachment 4.4 , DBDs _ 1 Snubber List _ 1 j _ 1 < ISI/IST Program Of *Yes", ESI review is reauwed) Code Stress Reports _ 1 i _ 1
- 2. Affected Drawines PC/M Dwo # Bgy Descrintion/ Title Dhig !
Affected Dwas Egy h* EP Rev JPN 236195-0001 0 CCWP1Acwd JPN-236195-0002 JPN 236195-0003 JPN-236-195-OOO4 0 0 0 CCWP 1B cwd CCWP 1C cwd I I I 8770 B 327-201 8770 B-327 205 8770-B-327 209 16 18 22 1 1 1 0h 0._ O_ n a ICWP 1 A cwd I 8770-B-327-832 21 1 0 ,._,_ er JPN 236195-0005 0 ICWP1Bcwd I 8770-B-327 833 18 1 0, .. JPN-236-195-0006 0 ICWP 1C cwd i 8770-B-327-834 25 1 0- o JPN 236-195-0007 0 CCWP 1 A sch I 8770-B-326-201 7 1 0 l JPN-236195-0008 0 CCWP 1B sch I 8770-B-326-205 8 1 O JPN 236-195-0009 0 CCWP 1C sch I 8770-B 326-209 10 1 0 D JPN-236195-0010 0 ICWP 1 A sch I 8770-B 326-832 10 1 0 JPN-236195-0011 0 ICWP 1B sch I 8770-B-326-833 8 1 0 JPN 236-195 0012 0 ICWP 1C sch
.....-, JPN-236 i s5-vvu
__ t 8770-B-326-834 11 1' O O Swgr Conn CCW 1 A l' 8 / /U-1638 10 2 ~~0 JPN-236-195-0014 O Swgr Conn CCW 1B I 8770-1548 11 2 0 JPN-236-195-0015 0 Swgr Conn CCW 1C 1 1 8770-1557 11 2 0 l JPN-236-195 0016 0 Swgr Conn ICW 1 A l 1 8770-1637 10 2 0 JPN-236-195-0017 0 Swgr Conn ICW 1B l 8770-1553 13 2 0 l JPN-236-195-0018 0 Swgr Conn ICW 1C I 8770-1555 11 2 0 JPN-236195 0019 0 Box Details E 8770-B-404-5 16 2 0 l JPN-236-195-0020 0 Yard Duct Runs & E B770-G-408-1 15 2 0 ' Lighting JPN-236-195-0021 0 Intake Structure Light. E 8770-G-386 16 2 0 Sect. & Details JPN-236-195-0022 0 Cable and Raceway *E 8770-B-328 37 2 O i JPN-236-1'95-0023 0 Cable and Raceway E 8770-B-328 37 2 0 JPN 236-195-0024 0 Appendix R Safe E 8770-B-048 2 2 0 Shutdown Analysis i Indicate Yes or No if SRD for PSL. Indicate 1 (POD),2 (CRD), 3 (MD),4 (ED), or 5 (other) for PTN. l l
PC/M No 236-195 Rev 0 Page 1 of 14 i-
- 3. Affected Vendor Manuals 5
i Plant Doc N2 Bgy Vendor /Ecuio Qigg Remarks EP Rev 4 3 NONE
- 4. Attachments Egggs
! 4.1 Environmental Qualification Checklist (Form 3E) 1 4.2 Fire Protection Review Checklist (Form 3F) 3 4 ! 4.3 ALARA Screening (Form 71) 1 l 4.4 TEDB Change Pa,:kage 7 1 j 4.5 FSAR Change Package 7
- 5. References 5.1 St. Lucie Unit 1, Final Safety Analysis Report, Amend.14 5.2 St. Lucie Unit 1, Technical Specifications, Amend.140 5.3 Plant Operating Procedure 1-0310020, CCW System, Rev. 43 5.4 Plant Operating Procedure 1-0640020, ICW System, Rev. 40 5.5 Plart Off-Normal Operating Procedure, 1-0310030, CCW System, Rev. 27 5.6 Plant Off-Normal Operating Procedure, 1-0640030, ICW System, Rev.19 5.7 St Lucie Action Report (STAR) 0-94110473 5.8 Plant Off-Normal Operating Procedure, 1-0030135, Control Room inaccessibility, Rev.
23 , IV. MATERIAL None
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i PC/M No 236105 Rev 0 Page 14 of 14 j V. IMPLEMENTATION
- 1. Imolementation instructions /Soecifications a) Wiring should be verified to existing CWDs and cubicle drawings. Discrepancies should be forwarded to JPN for resolution.
The implementor's work scope includes the following as detailed on the JPN drawings listed in the Affected Drawings Section (steps are not limited or confined to a rpecific order): b) Determinate affected cables in the respective 4.16 kv switchgear cubicles and at the local push-button stations per drawings JPN 236-195.0001 through JPN 236195.0006 and JPN. 236 195.0013 through JPN-236-195.OO18. c) At switchgear, insulate effected cable lugs, re tag spare r ables per JPN 194 295.0022 and secure spare cables in place in the respective switchgear. d) At local push-button stations, insulate cable ends, re tag spare cables per JPN 194-295.0022 and coil cables inside push-button station. i e) Remove box nameplates and pushbuttons from each push-button station, f) The 1C CCW pump pushbutton is located in box 93 which includes pushbuttons for other associated system valves. Care should be taken to insure that remaining pushbuttons are lef t intact. 2, Post-Mod Accentence Testino Post-modification tests shall be in accordance with approved plant procedures. a) Testing shall verify proper operation of the pumps for manual start /stop commands from both the control room and respective switchgear cubicle control switches. l i
i Attachment 4.1 i PC/M No. 194 295 i
;* Aev. O j Page 1 of t ;
ENVIRONMENTAL QUALIFICATION (EO) CHECKLIST I
- 1. El FCTRICAL EQUIPMENT 1E[ liq a) Does the PC/M add, modify or replace electrical /l&C equipment
- or affect any EQ Doc Pac? If NO proceed to Section'2. 1 _
1 b) is the equipment located in a harsh environment per EQ Doc Pac ! 1000 (PSL) or 1001 (PTN)? If NO proceed to Section 2. _ 1 c) Does the equipment perform a safety related function? _ _ OR I Does the equipment perform a quality related function such that its failure could mislead an operator or adversely affect any safety 3 related function required to mitigate or monitor an accident? _ _ ! OR 18 the equipment classified as Category 1 or 2 by RG 1.977 ,
'If the answer to any question in Section 1.c is YES, then the equipment is subject to 10CFR50.49 and the EP l must justify that the equipment is Environmentally Qualified.
- 2. ENVIRONMENTAL CONDITIONS 1El liq i
i , a) Does the modification add, relocate, raise flow rate of, or increase radiation levels of any piping containing radioactive liquids such that lifetime or eccident exposure of electrical equipment may be increased? _ .X ! b) Does the modification add or reroute any high energy piping, the ^ failure of which could impinge upon electrical equipment or increase i the post-break local pressure or temperature? , 1 { a e c) Does the modification alter process fluid characteristics such that
- the post accident pressure, temperature, pH, or boron concentration to which electrical equipment will be exposed may be changed? _ .X d) Does the modification alter any barriers that shield electrical equipment from high energy lines or radiation exposure? _ 1 e) Does the modification add equipment to Containment or increase the
, , post LOCA makeup water inventory such that the design basis 4 Containment Flood Level may be increased? _X.,
*
- If the answer to any question in Section 2 is YES, then it must be determined if any equipment subject to l 10CFR50.49 could be impacted and, if so, that its Environmental Qualification is still valid. Document the determination in the EP.
Prepared by/Date N rified by/Date 12//g/gf" , t JPN Form 3E. Rev. 6/94
Atttchmint 4.2 PC/M No. 194 295 Rev. o Page l' of 3 FIRE PROTECTION REVIEW CHECKLIST
- l. SAFE SHUTDOWN CAPABILITY YES* NQ REFERENCE Does the Engineering Package install, relocate, modify, or affect the operation of:
A. Equipment on the Essential Equip. List X _ Section I 8.2 B. Safe Shutdown Analyses'
- 1. Safe Shutdown Circuits .X._ _.__ Section I 8.2
- 2. Alternate Shutdown Components _ X
- 3. Associated brcuits _.__
X
- 4. Manual Actions _ X II. FIRE PROTECTION SYSTEMS Does the Engineering Package install, modify, or affect the operation of:
A. Fire Detection Systems .__ X B. Fire Water Supply System _ X C. Water Suppression Systems X D. Halon Suppression Systems _.,_ X E. Standpipes or Hose Stations . _ _ X F. Portable Fire Extinguishers __._ X lit. FIRE RATED ASSEMBLIES Does this Engineering Package install, modify or affect the function of: A. Fire Barriers X B. Fire Doors X If "YES" for PTN EPs, a Safe Shutdown Analysis Change Package (SCP) is required. if "YES" for PSL EPs, a drawing change is required. JPN Form 3F, Rev. 6/94
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. . - - . = . . .. _
Anschmsnt 4.2 PC/M No. 194 295 Rev. O Page 2 of 3 Ill. FIRE RATED ASSEMBLfES (Cont) YES fiQ REFERENCE C. Fire Dampers __ X D. Mechanical Penetration Fire Seals X E. HVAC Duct Penetration Fire Seals X i F. Cable Tray Fire Stops _ X 4 G. Structural Steel Fireproofing X l H, Conduit or Raceway Support Fireproofing X
- 1. Conduit or Raceway Fireproofing X J. Internal Conduit Fireproofing (Stuffing) X IV. EMERGENCY LIGHTING ,,,_ X i
A. Does the Engineering Package install, modify, or ! ) affect the operation of? _ X B. Does the Engineering Package obstruct the required light pattern of ? _ X r i C. Does the Engineering Package add or relocate essential equipment or components which will require the addition or relocation of ? , _ X i D. Does the Engineering Package install, relocate, i modify, affect, or require the use of hand held , d emergency lights? .,_ X i E. Does the Engineering Package relocate, modify,
- {
af fect, or obstruct the light pattern of perimeter security lighting to equipment requiring manual actions? X l l
- V. RCP OIL COttFCTION 1 l
Does the Engineering Package install, modify, or affect the operation of ? ,,,,,,,_ X VI. MISCELLANEDUS A. Does the Engineering Package affect the quantity ) or protection of insitu combustibles (solids, liquids, or gases) beyond the assumptions in the j FHA? _, X r JPN Form 3F, Rev. 6/94 L.
1 l Attachment 4.2 ~ PC/M No. 194 29b l Rev. 0 l Page 3 of 3 ! VI. MISCELLANEOUS (Cont) B. Does the Engineering Package Cause the addition YES. NO REFERENCE of a large combustible inventory within 50 ft. of essential equipment, components, electrical manholes or structures? _ X C. Does this Engineering Package modify or affect curbs or dikes used to contain combustible liquid spills? ,_ X D. Does this Engineering Package cause the removal of a flame retardant material from non IEEE-383 cables? X E. Does this Engineering Package affect fire protection technical specifications or fire fighting strategies? _ X F. Does this Engineering Package install, modify or affect the operation of alternate shutdown i communications? ,,,,,,,. X G. Does this Engineering Package install, modify, or affect hydrogen lines (or any combustible gas) in areas of the plant containing safe shutdown equipment or components?
, , _ X H. Does this Engineering Package install, rnodify, or affect any HVAC equipment or room heat loads in ;
areas _of the plant containing Safe Shutdown ' equipment or components? i _ X PREPARED BY DATE /2//B/fr VERIFIED BY DATE S M U jr JPN Form 3F, Rev. 6/94
1
. I Atta:nnent 4 3 FC 'M 194 2 %
, Res. O Page 1 0* 1 ALARA SCREENING I
- 1. Is this PC/M Administrative only (no physical modification)?
ycs. Further ALARA screening is not required. X no, Continue screening. 2. Does this PC/M involve a location in the Radiation Controlled Area (RCAl? yes, Location: Continue screening. X no, Further ALARA screening is not required. i l
- 3. Does the implementation , operation, or maintenance of this PC/M involve:
(Circle all "yes" responses) Movement of radioactive material? Modification of systems containing radioactive fluids or resins such that routing or retention characteristics are affected? Movement / modifications of existing permanent radiation shielding? Modification or removal of equipment that results in an uncontrolled opening / penetration into a High Radiation Area, Locked High Radiation Area, Very High Radiation Ares, or Exclusion Area? Diving associated with systems containing radioactive material? Entrance into containment during power operation? Potential for personnel exposure to radiation field of a 1r/ hour (assuming current area dose rates)? A total lifetime estimated dose due to the modification greater than or equal to one (1) man rem? For items in section 3 above is checked "yes", indicate "yes" below: ,
)
yes, This PC/M has the potential to significantly impact personnel radiation exposure. Complete Form 72 to ensure total radiation dose is minimized by design. ! nd, This PC/M has little or no impact on personnel radiation exposure. Form 72 not required, however, normal ALARA precepts should be followed to minimize radiation exposure.
!N ' 'Wehr 4',A .// . nkkr - l Prepared by / Reviewed by i I
Note: If the preparer and reviewer are the same as for the PC/M, duplicate signatures are not required. ! l 1 JPN Form 71, Rev. 6/94 I l
)
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- Facthty PSI Unit 01 PCM. 236-195 Attachment No 44 Revision No. O Page 1 of 7 TOTAL EQUlPMEh.T DATA BASE CHANGE PACKAGE COVER SHEET Description of Change: !
Removal of pushbutton stations on CCW 1A,18 and 1C and ICW 1 A,1B and 1C. Basis for Change: Pushbutton stations are being disconnected and removed from the field therefore the "LCL PB STA" associate tag for each of the above pumps should be deleted from TEDB. j l l
References:
- 1. PCM 236-195 Enoineerina !
l Prepared by _ Date /2[B[r,s-Venfied by IM /M - Date /22#/fr Approved b Date /MUk5
' ' l (Discipline Supv/ Lead Engineer)
Confiauration Manaaement Reviewed by: Date: Data Entry by: Date: Data Entry Verified by: Date: For instructions, see O! 2.8-1. 9
Facility: PSL Unit: 01 LMD: Component: CCW PP 1A _ f'hM 23ts~195 I Associate LCL PB STA Att: $ ll y I Rev: o ) Date Printed: 12/15/95 Page 7 of 7 l TOTAL EQUIPMENT DATA BASE SHEET l EQ Tag: N/A EQ Rev: N/A EQ Dec Pac: N/A l System: 14 COMPONENT COOLING WATER SYSTEM I Seismic: I Safety Class: Eng Ref: O Group: 1E EQ Surv Eote: N/A EQ Speer: N RG197: EQ Related: N EQ Scew: N/ RG197 Cat: _ Q Basis: EQ Remarks: /A RG197 Type: _ Comp Type: CK Sub Type: afety C nnel: Pcm:
\ /
Name: LOCAL PUSHBUTTON STATION F COMPONENT COOLING WATER PUMP 1A Locn Code: CCW/28/N-992/E-1697 Startup System: 048 Locn Desc: Instl MFG #: GEN GENERAL ELE C CO. Engineering Verified: Y Instl Model: CR2940 Rev 000 Orig Po: BY FIELD l Comp Group: NPRDS: Y Acct No: 530
/
EQ Tab: Insular' ion Rmvl: _ Train: . Scaffold Req: _ Critical Comp: _ Control Room Com _ Work Group: IST Reqd: Y RWP Reqd: N Maint Pgms: _ _ _ _ _ _ _ _ _ 40c e4 1 l l l I
Facility: PSL Units 01 LMD: ~~ PCM 23G - 16 Component: CCW PP IB Associate: LCL PB STA Act: 4. Y ; Rev: O ! Date Printed: 12/15/95 Page 3 of 7 i TOTAL EQUIPMENT DATA BASE SHEET EQ Tag: N/A EQ Rev: N/A EQ Doc Pac: N/A System: 14 C PONENT COOLING WATER SYSTEM ,/ Seismic: I -Safe Class: Eng Ref: / Q Group: 1E EQ Surv ote: N/A EQ Speer: N/ RG197t , EQ Related: N EQ Scews ' j RG197 Cat: _ Q Basis:- EQ Remarks: A ! p RG197 Type: _ Comp Type: CK Sub Type: fety C]iannel: Pcm:
/
Names LOCAL PUSHBUTTON STATIO POR COMPONENT COOLING -WATER PUMP 1B Locn Code: CCW/28/N-956/E-1697 Startup System: 048 Locn Descs / Instl MFG #: GEN GENERALELECT)dCCO. Engineering Verified: Y Instl Model:'CR2940 Rev: 0 Orig Po: BY FIELD Comp Group: / NPRDS: Y A et No: 530 EQ Tab Insul ion Rmvl _ Train _ Scaffold Req: _ Crit al Comp; _ Control Room Comp: _ Work Group: IST Reqd Y RWP Reqd: N Maint Pgms: _ _ _ _ _ _ _ _ _ (/
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L .I i Facility: PSL Unit: 01 LMD: I Component: CCW PP 1C ~ 9f>4 2 % -146 i i Associatt. : LCL PB STA Att: %d/ j l Date Printed: 12/15/95 Rev:c > Page y of 7 ; e c l TOTAL EQUIPMENT DATA BASE SHEET EQ Tag: N/A EQ Rev: N/A EQ Doc Pac: N/A j i System 14 COMPO COOLING WATER SYSTEM [ 1 Seismic: I safety.C1 ss Eng Ref: / : _ /
/
3 O Group: 1E EQ Surv Not ' N/A EQ Speer: /A I RG197: _ EQ Related: N EQ Scew: N/A RG197 Cat: j ,, j j Q Basis: EQ Remarks: N/A ,/ RG197 Type: _ I
- Comp Type: CK Sub Type: /
Saf yC el: Pcm: , Name: LOCAL PUSHBUTTON STATION COMPONENT COOLING WATER PUMP 1C i Locn Code CCW/28/N-974/E-1697 ! Startup System: 048
/
- Locn Desc
- ,/
l Instl MFG #: GEN GENERAL ELECTRIC/ !CO. 4 i Engineering Verified: Y I i ? Instl Model: CR2940 Rev: 000 Orig Po: BY FIELD Comp Group: NPRDS: Y Ac t No 530 3 j 2 EQ Tab: Insulatio Rmvl _ Train: l Scaffold Reg: _ Critical omp: _ Control Room Comps _ Work Groups IST Reqd: Y RWP Reqd: N ! Maint Pgms: _ i-t , I ! l i I 4 I 4 e 5 d 1
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Facility: PSL Unit 01 LMD: _ Fowt zu -in Component: ICW PP 1A Associate: LCL PB STA Att: 4. 4/ [ Rev: O ,3 Date Printed: 12/15/95 Page 5 of 7 TOTAL EQUIPMENT DATA BASE SHEET EQ Tag: N/A EQ Rev: N/ EQ Doc Pac: N/A System: 21 CIRCULATING WATER - INTAKE OOLING WATER Seismic I fety Class: Eng Re .
/
Q Group: 1E EQ Note: N/A EQ S eer: N/A RG197: _ EQ Related: N EQ Scew N/A RG197 Cat: - l Q Basis: EQ Remark N/A RG197 Typer _ .
. Comp Type: CK Sub Type: Saf ty Channel: Pem:
Name: LOCAL PUSHBUTTON STA4 ON FOR INTAKE COOLING WATER PUMP 1A Locn Codes INTK/19/S-4/W-C Startup System: 007 Locn Desc: Instl MFG #: GEN GENERAL EL CTRIC CO. Engineering Verified: Y Instl Model: CR2940 Rev: 00 Orig Po: BY FIELD Comp Group: NPRDS: Y Acct No: 531 . EQ Tab: Ins ation Rmvl: _ Train: _ Scaffold Reg: _ Cri cal Comp: _ Control Room Comp _ - Work Groups IST Reqd: Y RWP Reqd: N r Maint Pgms: h CQ j 4 l l
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. 1 i 1 I facility: Ps:. cast, c: po i , i l @M 2 t(e-195 1
l Component ICW PP 13 t a' l l i l Am oesates LCL Ps sTA ( l l At t s Y. I[ ] l Date Pranted: 12/15/95 l laevs O l I l lPage y of 7 l I I i TOTAL EQUIPMEW DATA BASE $NEET EQ Tags N/A EQ Rev N/A EQ Doc Pac N/A i e systems 21 CIR TING WATER . I M AKE COOLING WATER seismic: I safety asa _ Eng Refs Q Croups it EQ Surv Not N/A EQ spears N/A A0197: , j. r I EQ Relateds N EQ Scows N/A
- l RG197 Cat ,
s ) $ Q Basie s . _ EQ Remarks: N/A RG197 Types . Comp Type: CE sub Types _ safety C nnel,i' Pem: Name: LOCAL PUSMatJTTON S'iATION FOR I COOLING WATER PUMP 1B 1. Loen Codes INTE/19/S.3/W-C Startup Systems 005 i Loen Descs # Inst! MFG 8e GEN GENERAL ELECTRIC,CO. \ Enginepring Verified: Y Insti Models CR2940
/ Keve 000 Orig Pos Y FIELD Coop Group NPRD$s Y Acet No 531 .
EQ Tabs _ InsulatP n Revis , Trains , y
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_ ~ . . - . . . _. -,, . , _ - . _ . . ~ . , _~__. _ ~ ~. A 4 j - .. . . ; e i i . f' -' Facility: PSL Unit: 01 LMD: Component: ICW PP 1C _ 9CM 23(e 19 5' s Ass'ociate LCL PB STA Att: Y. M , Rev: p Date Printed: 12/15/95- Page 7 of 7 ; (
! TOTAL EQUIPMENT DATA BASE SHEET i
j EQ Tag: N/A E Rev: N/A EQ Doc. Pac: N/A System: 21 CIRCULATING WATER INTAKE COOLING WATER Seismic: I S ety Class: Eng Reft ~ Q Group: 1E EQ rv Note: NA EQ Speer: N/A RG197: , , j EQ Related: N EQ Se : N/A RG197 Cat: _ , ). -Q Basis: EQ Rema s: /A RG197 Type: _
. Comp Type: CK Sub Type: Safety Channel: Pem:
Name: LOCAL PUSHBUTTO ATION FOR INTAKE COOLING WATER PUMP IC Locn Code: INTK/19/N-3/W Startup System: 005 Locn Desc: Instl' MFG #: GF" GENE ELECTRIC CO. Engineering Verified: Y Instl Model: CR2940 Rev: 000 Orig Po: BY FIELD Comp Group: NPRDS: Acct No: 531
- EQ Tab
- I ulation Rmvl: _ Train:
Scaffold Req: _ Cri
- cal Comps , Control Room Comp: _
Work Group: IST Reqd Y RWP Reqd N Maint Pgms: _ _ _ , _ _ _ _ _ h ,t i I 1 \ f 9 4 a E r , _ __
i i j l PC/M No. M6-1" f Rev. ? ' Attachment 45 l Page 1 of 9 l l FSAR C".ANGE PACKAGE Plant St. Lucie Unit _gl FSAR PACFA ATTAQiED: 7.3-14, 7.3-16, 7.4-8, 7.4-9, 9.2-7, 9.2-18 FSAR FIGURES ATTACMm:
- None although figures 7.4-6, 7.4-7, 7.4-8, 7.4-12, 7.4-13 and 7.4-14 will be affected but updated at PCM drawing update. An "INFORMATION ONLY" note should i be added to the following figures: 7,4-3, 7.4-4, 7.4-5, 7.4-9, 7.4-10, 7.4-11.
COMMENTS: This reflects the deletion of local pushbutton stations for the 3 Icw and CCW pumps. The addition of the notes to the above figures reflects the current status of the drawings referenced by the figures. _h
* ~
Y $ ' 2//9/f[ Prepared by/ Dater [er'ified by/Date: Al- A T Approved by[Date: Notes:
- 1. All affected FSAR pages and tables are to be marked up and attached. If additional space is required, than additional pages should be provided.
New information for inclusion in the FSAR shall also be provided, i' 2. The new or marked up pages are to be legible and of s.ufficient quality to be microfi2med, i i 3. If a figure is provided elsewhere in the design package, then it need not be duplicated in the FCP. However, a note should be provided in the comment section referring to its location in the package. This also applies to new drawings which should be added to the FSAR. If a figure j j revision is to be included in the FCP, then a copy of the FSAR figure i with a bubble around the affected area is sufficient. ' i l l i l 3
.. . __ =._ -- - ._ _...m. . _ - _ .m . - ._.. -_~ ~ . _ _m..._.mm>_m _ _ _ . - . . - .
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2 PC/M No. 236-145 Rey, ? ~ Attactnent_,i 1 Page _;. or - 2 d l a The pumps are started manually by means of coritrol switches located on the main control panel or by means of : 1;;;_ p_.r :'_::;.. the respective switchgear cubicle I control switch. Pump logic and control diagrams are shown on Figures 7.4 3 through 7.4-5. Electrical schematic diagrams of pump control circuits aresshown on Figures 7.4-6 through 7.4-8. Control panel switches are provided to actuate the shutdown heat exchanger outlet l valves (HCV-14-3A and B) . , In the event of a LOCA, the component cooling pumps, heat exchanger and header j isolation valves are actuated automatically upon SIAS. The actuating instrumentation 4 i and controls for SIAS actuation are part of the engineered safety features actuation i system and are discussed in Section 7.3. The component cooling water surge tank is normally vented to the atmosphere through a j 1 three-way valve (RCV 14-1) in the tank vent line. Upon a high radiation signal (4 x 10' ' { *) pci/ce) the valve will change position and venting will be diverted to the waste management system. The high radiation signal is derived from either of the two-radioactivity monitors (see Figure 7.3-45) located at the component cooling water discharge headers. The operation of this interlock is not required for safe shutdown l and is not designed as seismic Class I. I 1 c) Monitoring of System operation Control room process indication alarm and status diverse instrumentation (flow, i pressure) is provided to enable the operator to evaluate system performance and detect ' malfunctions. Component cooling surge tank low level is alarmed in the control room by redundant instrumentation. The outlet temperature, pressure and flow from each component cooling heat exchanger is indicated in the control room. High temperature. low flow and low pressure are alarmed. Shutdown heat exchanger outlet temperature and flow are similarly indicated and alarmed. The shutdown heat exchanger outlet valves and header isolation valves are provided with position indicating lights in the control room. Component cooling pump operating status is also indicated in the control room. Refer to Section 7.5 for further discussion of safety related monitoring instrumentation. d) Interlocks, Bypasses and Sequencing Upon loss of off-site power, the pumps are automatically restarted and loaded on the emergency. diesel generators. Their sequencing is shown in Table 8.3-2. As discussed in Section 8.3.1.2.4, if all three pumps are available for starting, pump 1C which is part of electrical load group AB will not be started if off-site power is lost to avoid
- overloading the diesel generator. If either pump 1A or 1B is out of service, pump 1C will replace that pump and will start automatically as part of the corresponding electrical load group.
l 7.4-8 1 l
o-
+
f r: P N . M D iM Rev. Attachment i.1_ Page _1_ of
- e) Redundancy Separatt switches and actuation circuitry,are provided for redundant components. s Physical and electrical separations are provided as discussed in Section 7.4.2.1. '
f) System Supporting Equipment , control switches are also provided locally and in the control room to operate the cross-connection valves (I-HV-14-1,2,3,4) on the suction and discharge pump headers. This allows the operator to control alignment of pump flow to each of the redundant headers. 7.4.1.5 Intake cooline water system ins t rument a t ien The intake cooling water system is discussed in Section 9.2.1. The system P&ID is shown on Figure 9.2-1. Location of system components is shown on the plant general arrangemerst drawings. The system instrumentation and controls necessary to achieve plant shutdown are discussed as follows: a) Actuation of System Components To achieve safe shutdown the only system component actuation step requirad is starting the intake cooling water pumps. b) Control of System Operation The pumps are started manually either by means of keee4 switchgear cubicle control switches or control room switches. pump logic and control diagrams are shown on Figures 7.4-9 through 7.4-11. Electrical schematic diagrams of pump operation are shown on Figure 7.4-12 through 7.4-14. In the event of a LOCA, the intake cooling water pumps and essential header- ! isolation valves are actuated automatically upon SIAS. The actuating instrumentation and controls for SIAS actuation are part of the engineered saf*' features actuation system and are discussed in Section 7.3. l Following actuation of the pumps, the intake cooling system is designed to operate with autocaatir. temperature controlled modulation of the intake cooling water flow through the camponent cooling heat exchangers. The heat exchanger outlet flow control valves (TCV-14-4A and TCV-14-4B) are controlled by pneumatic temperature I controllers TIC-14-4A and TIC-14-4B which sense outlet temperature on the component cooling water side of the heat exchangers. The temperature controllers are provided for efficient system operation during normal plant operation. The control valve pneumatic controls have been designed and qualified as seismic Class I to assure ' proper operation of the control valves during safe shutdown. As temperature increases, intake cooling water flow is automatically increased. The control valves are pneumatically operated and f ail wide.open on loss of instrument air. In the event of loss of air the intake cooling system will operate in the full unmodulated l flow mode. I 7.4-9 Amendment No. 12 (12/93)
o 6 Pe m No ,' Mg.ies Rev, Attaenment 45 i Page 1 of - ! The a ntake cooling water pumps are rotated in service periodically so that their { cont.uced availability for emergency conditions is ensured. ] 9.2,1.5 Inne rument at ion Acelicatien Table 9.2-3 lists the parameters measured by the intake cooling water system instrumentation. The heat exchanger parameters and pump status are monitored either I locally or in the control room. In every case where saf ety-related equipment is involved, more than one parameter is measured to assure design performance of the j equipment. I The intake cooling water pumps can be started or stopped either 1;;;_.y at the j switchgear or from the control room. The pumps-receive a start signal upon SIAS. j The logic and instrumentation for this syatem is discussed in Section 7.3.1,3.2. All valves in the system are manually operated with four exceptions. The turbine cooling water and steam generator open blowdown systems are automatically valved l l off of the intake cooling water on receipt of an SIAS. This operation can also be i initiated either locally or in the control room. The second automatic valve in the ! system is the butterfly valve (one in each headers at,the outlet of the component l cooling water heat exchanger. This valve automatically controls outlet water flow l from the exchanger. It is modulated opened and closed according to the outlet water temperature of the shell side of the component cooling water heat exchanger. In addition, valve closure is limited to 7.2 degrees from full closed position to , prevent turbulent flow and system damage. The third valve is located at the outlet l of the turbine cooling water heat exchangers. This valve is temperature controlled f rom the shell outlet side of the heat exchanger and controls intake cooling water flow. The fourth valve is located at the outlet of the steam generator cpen l blowdown heat exchangers. This valve is temperature controlled from the shell outlet side of the heat exchanger and controls intake cooling water flow. 4 9,2-7 Amendment No. 14, (6/95)
1
- I e
Ft/M N. 234.iec Rev. " Attachment 4_? rr? L c: - \ The valves in the component cooling water system allow: a) isolation of the nonessential (N) header from the A and B headers on SIAS b) routing of component cooling water through the shutdown heat exchangers o.. SIAS c) isolation of the reactor coolant pumps quench tank coolers, and CEDM air cooler on SIAS l d) routing of component cooling water surge tank ventilation to the waste management system upon a high component cooling water radiation signal i e) controlling the level in the component cooling water surge tank f) directing of discharge of the 1C component cooling water pump to either the A or B header and g) controlling of flow to the containment cooling units The component cooling water pumps cas. be started and stopped both 1;;;1:y at the switchgear and from the control room. The pumps receive a start signal on SIAS. The Ir 71c and instrumentation are discussed in Section 7.3.1.3.1. I I I l 6 9.2-18
- . , - - - . - - - . . . . - _ . .~.-- -__ ~- -. - ~ . . . . - - - . - .- . . . . -
s
- s 1
4 Pc:M N: H te Rev. Attachmen: 4 s Page _t_ cf - Control room indication alarm and status instrumentation is provided to enable the operator to evaluate system performance and deceet malfunctaens. The manual control - switches and equipment status lights are provided in the control room for system manual control and indication. Redundant and diverse instrumentation is provided
'f or annulus pressure indication. Annulus high pressure (+5 in. w.g.) and low pressure (-4 in. w.g.) are annunciated in the control room.
1 The differential pressure across the HEPA filters is indicated in the control room and high differential pressure is annunciated. High heater element temperature as monitored by a loca3 controller and high temperatures are alarmed in the control room. High humidity a.. J - system is locally indicated and alarmed in the control room. The charcoal aosoever filter and air flow temperatures are. recorded and high temperatures sie alarhed. The shield building exhaust fan failure, low air flow, or f ailure to start on CIS is also alarmed. Refer also to Section 6.2-3.5. Upon loss of cItsite power, the fans are automatically restarted and. loaded on the emergency diesel generator's. Their sequencing is shown in Table 8.3-2. The electrical bypass circuit 16 provided to permit stopping of one redundant fan when systems were started automatically upon CIS. This bypass automatically resets when CIS is reset. The standby system is restarted automatically upon f ailure of the running system. This interlock is designed to meet IEEE-279 separation criterion as discussed in Section 7.3.2.3.2. Separate actuation switches and circuitry is provided for redundant components. Physical and electrical separations are provided as discussed in Section 7.3.2.2. 7.3.1.3 Enoineered safetv Features sueeertino systems inst rumentatien and centrol 7.3.1.3.1 Component Cooling System Instrumentation The Component Cooling System is discussed in Section 9.2.2. The system P&ID is shown on Figure 9.2-2. Location of system components is shown on the plant general , arrangement drawings. - The sysrma instrumentation and control necessary, to support i
. ESF are as follows:
a) Actuation of System Components l To support ESF, the system component actuation steps required are:
- 1) automatic starting of the component cooling water pumps,
- 2) automatic opening of the outlet valves from the shutdown heat exchangers.
In the event of a LOCA, the component cooling pumps, heat exchanger and header isolation valves are actuated automatically upon SIAS. The actuating instrumentation and controls for SIAS ' actuation are part of the Engineered Safety Features Actuation System and are discussed in Section 7.3.1.1. i The pumps can also be started manually by means of control switches located on the i 1 main control panel or by means of : '.;;_1 push bu;;;r control switches at the respective switchgear. Pump J0;ic and aontrol diagrams are shown on Figures 7.4-3 through 7.4-5. Electrical schematic diagrams of pump control circuits are shown on Figures 7.4-6 thro",a 7.4 8. 7.3-14 l
6 FC'M NO. 281
- Rev.
Attachment 4.5 Fage _*_ of
- e) Redundancy Separate switches and actuation circuitry are provided for redundant comp,one nt s .
Physical and electrical separations are provided as discussed in Section 7.4.2.1. f) System Supporting Equipment Control switches are also provided locally and in the control room to operate the cross-connection valves (I-HV-14-1,2,3,4) on the suction and discharge pump , headers. This allows the operator to control alignment of pump flow to each of the redundant headers. 7.3.1.3.2 Intake Cooling Water System Instrumentation The intake cooling water system is discussed in Section 9.2.1. The system P&ID is shown on Figure 9.2-1. Location of system components is shown on the plant general arrangement drawings. The system instrumentation and controls necessary to support ESF are discussed as follows: a) Actuation of System Components To achieve safe shutdown the only system component actuation step required is starting the intake cooling water pumps. In the event of a LOCA, the intake cooling water pumps and essential header isolation valves are actuated automatically upon SIAS. The actuating instrumentation and controls for SIAS actuation are part of the engineered safety features actuation system and are discussed in Section 7.3.1.1.8. The pumps may also be started manually either by means of 4eee+ switchgear or , control room switches. Pump logic and control diagrams are shown on Figures 7.4-9 through 7.4-11. Electrical schematic diagrams of pump operation are shown on Figures 7.4-12 through 7.4-14. b) Control of System Operation t Following actuation of the pumps, the intake cooling system is designed to operate with automatic temperature controlled modulation of the intake cooling water . flow through the component cooling heat exchangers. The heat exchanger outlet flow control valves (TCV-14-4A and TCV-14-4B) are controlled by temperature controllers TIC-14-4A and TIC-14-4B which sense outlet temperature on the component cooling water side of the heat exchangers. As temperature increases, intake cooling water flow is automatically increased. The control valves are pneumatically operated and f ail wide open on loss of instrument air. In the event of loss of air, the intake cooling system will operate in the full unmodulated flow mode. The temperature controllers are provided only for efficient system operation during normal plant operation. No other automatic or manual control of system operation is required to support ESF. I 7.3-16
3 s b Inter-off:.ce cerrespendence FPL r:R & M ' ' * * * *
- U 5 9 (, go4 4] zgjq g l I i l To: S. A. Valdes Date: i St. Lucie Plant DEC 0 71995 l From: D. Denver ,
Department: JPN/PSL Nuclear Engineering l
Subject:
ST. LUCIE PLANT UNIT 1 1 PC/M #: 177-195M TITLE: DG 1A & 1B PROTECTIVE TRIPS DESIGN CHANGE l REA: STAR 951344 FILE: PCM 177-195M l l Attached for review, approval and use .is the MEP for the subject PCM. PCM 177-195M was generated to make existing DG mechanical trips active during the 10 minute idle cooldown period. I This PCM completes activities associated with STAR 951334. I 1 If you have any questions, please contact Gerry Novarro at l 467-7483. DJD /JES Copies: H. L. Fagley - DCC-CS/PSL (w/ original) D. M. Stewart - TS/PSL A. S. Suggs - JPN/JB (w/dwg list) L. Bossinger - EM/PSL Originator: Roger Kulavich - SCE/PSL Other: R. Gonzalez - ICM/PSL ,
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. PC/M 177-195M e Revision 0 Page 1 of 5 } MINOR ENGINEERING PACKAGE (MEP) PLANT PSL UNIT _.0L_ PC/51 NUMBER 17719FM SUPPL 0 ORIGINATLNG DOCUMENT STAR 951344 EXPIRATION DATE 7/31/96 PC/M CLASSIFICATION _X_SR QR __._ NNS __ ADMIN TITLE DG 1 A and IB Protective Trins Desi<m Chance 1 ADDITIONAL REQUIREMENTS / INSTRUCTIONS %S NO AS-BUILDING TO COMMENCE UPON ISSUANCE OF PACKAGE? ___ 1 THIS PACKAGE HAS THE POTENTIAL TO SIGNIFICANTLY _ 1, IMPACT PERSONNEL RADIATION EXPOSURE (See QI 3.13). IF YES, JPN FORM 72 IS REQUIRED. I 10CFR50.59 SCREENING YES NO
- 1) DOES THE CHANGE REPRESENT A CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR?
2) _ _X_ j DOES THE CHANGE REPRESENT A CHANGE TO ' PROCEDURES AS DESCRIBED IN THE SAR? 3) _ _X_ IS THE CHANGE ASSOCIATED WITH A TEST OR EXPERIMENT NOT DESCRIBEDIN THE 4) _ _X_ COULD THE CHANGE AFFECT NUCLEAR S AFETY IN A WAY NOT PREVIOUSLY EVALUATED IN THE SAR? __ _X_
- 5) DOES THE CHANGE REQUIRE A CHANGE TO THE i TECHNICAL SPECIFICATIONS?
_ _X_ NOTE: IF THE ANSWER TO ANY OF THE AB0VE 10CTR50.59 SCREENING QUESTIONS IS YES. THE W REVIET/APPROVA!; INTERFACE TYPE GROUP PREPARED VERIFIED rwtv arvirw mus APPROVED FTL APPROVED
- MECll n N/A ELECT N/A N/A N/A
, _ff _ / h'/A l&C x l.
74 . N/A CIVIL x N/A NU C*
- z N/A N/A by [ - _ _ N/A FS! s N/A N/A A N/A NUC FUEL N/A N/A N/A N/A
- For Contractor Esals As Determined By Projects " Review Interface As A Minimum On AllNon-Admin MEPs FPL PROJECTS APPROVAL: 4 DATE:_S //9F
Ol 3-PR/PSL-1 R2 vision 39 April,1996 i Page 22 of 26 ATTACHMENT 2 PC/M REVIEW FORM (Page 1 of 2) PC/M Number _177-t95 Supplement Number O Expiration Date '7 /5//_T_5 PC/M
Title:
Dr_,lAt16 W ecnVE t61F5 A7 (SM CFA C PC/M Classification: Major Modification or Minor Modification Safety Classification: Safety Related Quality Related Not Nuclear Safety Related Administrative PC/M Category: [ Normal Either/Or As-Requested Package As Fall Generic Does the PC/M contain a Safety Evaluation? O Yes or I No is this a proposed change or modification to a ujnit ystem or equipment that affects nuclear safety? Ld Yes or O No if either of the above questions is YES, FRG approvalis required. FRG Review Require, ?
~
Yes or O No M-
~ Configuration kgement Date: _D l9 /%
' Plant General Manager Date: N / 6/ IL FRG Nurnber: % -/64 FRG Secretary: h i
/ Date: / M /f6 FRG Comments:
Q13-PR/PSL-1 RIvmon 39 I April,1996 ; Page 23 of 26 ATTACHMENT 2 t l PC/M REVIEW FORM ' 4 (Page 2 of 2) implementing Documentation: { Department ER/NPy0 # W/O e , Gld G'3/I944 Ern 9 ACO4"RO6-Ol 6(/tMS_ 9AcO450G-ol
)
0 PC/M Review and Approval: di 1 m u Date 4 / I9 /N Configuration Mana
' 'M Meent k Date b /b/
, Comments: n u ' % rol 1 Areas Affected: Yes or No Description , Operator Training Yes ! As per Training Department Requirements Operating Procedures yg5 Surveillance Procedures yp:s k i Maintenance Procedures Spare Parts
@ i i
Drawings / Technical Manuals yE$ SRDs? (b N i FUSAR Change N(') TEDB Change NO Human Factors (CREDIT) O Environmental Concems MO In-Service inspection Maintenance Rule g Plant Restrictions OQf Others
, MOM ,
Il t ! Date N / O b l/ Configuration Ma ment
, PC/M 177-195M Revision 0 Pace 2 of 5 ENGINEERING .TUSTIFICATION The purpose of this Minor Engineering Package (HEP)is to alter the design of the Unit 1 Diesel Generator protective trips cirruit to allow cenain mechanical trips, the low oil preuure tnp and the high jacket water temperature trip, to be active and not bypassed during the 10 minute idle speed cooldown period. This HEP provides thejustification and the guidelines for rewiring the appropriate ponion of the circuits. Ahhough the idle speed control circuits do not represent a nuclear safety function. they do share winng with circuits that are performing nuclear safety functions (the Emergency Auto Start). Therefore, this MEP is considered as Safety Related. The current design has the low oil pressure and high jacket water temperature trips bypassed during the 10 minute idle speed cooldown period. These same trips are not bypassed during the stan at idle speed or the run at normal speed. These trips are always bypassed during an emergency stan and will not be affected by this MEP. Because the idle. period is not a nuclear safety function and because these trips are currently allowed during idle starting, they should also be allowed during the cooldown idle period. It is sensible to state that the cooldown function is not working properly if the oil pressure goes low or the jacket water temperature goes high and that the immediate need is to trip the DG to allow natural cooling to take place. Any further running could elevate the DG temperature and cause damage. Therefore, the concept of allowing these trips to function during the idle cooldown period is justifiable. A careful review of the existing circuits has shown that the wiring changes proposed by this MEP and as shown on the drawings referenced in this HEP will not affect any other circuit function. Currently, as soon as the DG stop is initiated, relay contacts K16 and ESR open and de-energize reh y K12 even though the DG is still running. The relay K12 contact providing power to the mechanical trips in turn opens and causes the trips to be bypassed. Contacts K16 and ESR serve no function that is not accommodated by the 200 rpm speed switch also wired in series with the K12 relay. The K16 and ESR contacts are closed by any stan command whereas the 200 rpm speed switch indicates that the DG was started and is also running at greater than 200 RPM. By reconnecting the feed to relay K12 around the K16 and ESR contact, relay K12 will remain energized after the stop in initiated. K12 will then de-energize only after the speed drops below 200 rpm signifying the end of the cooldown idle period. On starting the DG, the K12 relay was never energized prior to reaching 200 rpm so the contacts K16 and ESR showing that a start command was given is not necessary. Finally, the wiring change around the K16 and ESR contacts will still allow these contacts to continue their function in parJlel portions of the same circuit. Therefore, all wiring changes proposed by this MEP are acceptable and meet the intent of the functional changejustified above. CONCLUSION The Diesel Generator system serves a critical Nuclear Safety function as discussed in section 8.3.1.1.7 of the FS AR. The FSAR lists all the trip signals that should shut down the DG in the absence of any emergency start. The mechanical trips being added by this MEP are already listed in the FSAR as being required for any non-emergency running state of the DG. Also, the addition of the mechanical trips in the idle cooldown mode does not conflict with the FSAR since the idle cooldown does not occur curing an emergency condition and the changes will have no affect on the emergency operation of the DG. Therefore the changes contained in this MEP results in no degradation, either directly or indirectly, to any safety functions required for analyzed accidents, and do not increase any radiological hazards.
i s PC/M 177-195M Revision 0 I I Page 3 of 5 t OTHER AFFECTED DOCUNfENTS l Reference ParagrapWPage or Do ument Attach Revised Papes None 1 SPECIAL INSTRUCTIONS \INIPI FNfENTATION REOUIRENfENTS
- 1. Coordinate a'l work with Operations to obtain appropriate clearances.
- 2. Use #14 AWG SIS wiring, PC-2 or better, when rewiring per this MEP. Relabel existing wires as required per this MEP.
- 3. Surveillance and Operations procedures should be reviewed and updated as appropriate to reflect the change in tripping function during the DG cooldown.
- 4. After completion of the as-building of drawings represented by the JPN sketches JPN-177-195.001 through JPN-177-195.020, replace the existing revision of the appropriate drawings in the Volume 2 ofInstruction Manual 8770-6703 with the newer revision.
POST MODIFICATION TESTING 1. Verify that wiring around ESR contact 6-7 was properly reconnected per the CWD. Verify the engines trip when in the idle cool-down period upon given a simulated mechanical trip signal. REFERENCES l
- 1. St. Lucie Unit No.1 Final Safety Analysis Report, Section 8.3.1.1.7, Amendment 14.
1
- 2. St. Lucie Active Safety Evaluation List (ASEL) access date 11/01/95. I l
- 3. PSL Package Information Tracking System (PITS) reviewed 11/01/95.
- 4. PSL Affected drawings (DCTS) reviewed 11/01/95.
- 5. St. Lucie Unit No.1 Technical Specifications, Amendment 140.
ATTACHAfENTS Attachment Number Number Descrintionfritle of Paces None L .-. _ _ .
3 PC.'M 177-195M Revisier. 2 Page 4 of 5 k DRAWINGS (l) ALL AFTECTED (2) (31 (4) PrN DPA%TNG NO. E DFScortrnnverrT1 p == PLaS T DpARTNc.5 PH P FH t JPN 177-195.001 0 SLI Schm. DG 1 A Stan Sol. I 8770-B 326 sh. 959 12 1 0 JPN 177195.002 0 SLI Schm. DG 1 A Stan Sol. I 8770-B 326 sh. 959 12 1 0 JPN 177-195.003 0 SLI Schm. DG IB Start Sol. I 8770-B-326 sh. 969 12 1 0 JPN 177195.004 0 l SLi Schm. DG 1B Stan Sol. I 8770-B 326 sh. %9 12 1 0 l JPN 177-195.005 0 SLI CWD DG 1 A Stan Sol. I 8770-B 327 sh. 959 14 1 0 JPN 177-195.006 0 SLI CWD DG 1 A Stan Sol. I 8770-B 327 sh. 959 14 1 0 l JPN 177-195.007 0 SLI CWD DG IB Stan Sol. I 8770-B 327 sh. 969 13 1 0 JPN 177195.008 0 SLI CWD DG IB Stan Sol. I 8770-B 327 sh. %9 13 1 0 JPN 177195.009 0 SLI DG 1 A Idle Stan/Stop I 8770102N I 2 0 Pnl.Schem. Diag.sh I of 3 - JPN 177195.010 0 SLI DG I A Idle Stan/Stop I 8770-102N 1, 2 0 PnJ. Schem. Diag.sh 1 of 3 JPN 177195.011 0 SLI DG 1 A Idle Stan/Stop I 8770-10295 2 2 0 Pnt.Schem. Diag.sh 2 of 3 JPN 177195.012 0 SLI DG 1 A Idle Stan/Stop I 8770-10295 2 2 0 Pnl.Schem. Diag.sh 2 of 3 JPN 177-195.013 0 SLI DG IB Idle Stan/Stop I 8770-11437 0 2 0 Pnl.Schem. Diag.sh I of 3 JPN 177-195.014 0 SLI DG IB Idle Stan/Stop I 8770-11437 0 2 0 Pnt. Schem. Diag.sh I of 3 JPN 177-195.015 0 SLI DG IB Idle Stan/Stop I 8770-11438 0 2 0 Pnt. Schem. Diag.sh 2 of 3 JPN 177 195.016 0 SLI DG IB Idle Stan/Stop 8770-11438 1 0 2 0 Pnt. Schem. Diag.sh 2 of 3 JPN 177-195.017 0 SLI DG I A Idle Stan/Stop 1 8770-10297 2 2 0 Pnt. Door & Rear Subpnl.
, Wiring Diag. Sh 1 of 3 JPN 177-195.018 0 SLI DG 1 A Idle Stan/Stop I 8770-10297 2 2 0 Pnt. Door & Rear Subpnl.
Wiring Diag. Sh I of 3 JPN 177-195.019 0 SLI DG 1 A Idle Stan/Stop I 8770-10298 1 2 0 Pnt. I. eft Subpnl. Wiring Diag JPN 177195.020 0 SLI DG 1 A Idle Stan/Stop I 8770-10298 1 2 0 Pnl. left Subpnl. Wiring Diag JPN 177 1 c i.021 0 SLI DG IB Idle Stan/Stop I 8770-11579 0 2 0 ,.' i Pnt. Door & Rear Subpnl. Wiring Diag. Sh I of 3 JPN 177-195.022 , O SL1 DG 1Bldle Start /Stop I 8770-11579 0 2 0 Pnt. Door & Rear Subpnl. Wiring Diag. Sh I of 3 JPN 177195.023 0 SLI DG 1B Idle Start /Stop I 8770-11440 0 2 0 . Pnl. Left Subpnl. Wiring Ding JPN 177-195.024 0 SL1 DG 1B Idle Stan/Stop I 8770-11440 0 2 0 Pnl. Left Subpnl. Wiring Diag
. . _ . _ . . . . . . _ . _ . _ . . . _ _ _ - . . . _ _ _ _ - . . _ . ~ . .I PC/M 177-195M Revisior. O Page 5 of 5 i b i
j VENDOR MANUALS j (2) " Mim a-PLAST DOCUMENT NUMBER M VENDOR /EOUIPMENT 1* RF3tARKS E 8770-6703 10 Diesel Generator Instruction Manual E Rep!xe dwgs '0 rm wahne ) see speciai l l Instruction 84 )
- - on rure 3 l s
~ (1) DISCIPLINE: C = CIVIL: E = ELECT; I = I&C; M = MECH; N = NUCLEAR i (2) REVISION of affected drawing / vendor manual. Indicate "new"if the drawing / vendor manual is being created. Th
- engineer is accountable for rese,rving the new drawing / vendor manual number.
j .. (3) UPDATE PRIORITY Indicate 1 (SRD),2 (non-SRD), or 5 (other) for PSL. Indicate 1 (POD) 2 (CRD 1 4 (ED), or 5 (other) for ITN. (4) MEP REVISION under which last drawing change was made, t 1 i 1 i s i 4 { i
i inter-Office Correspondence FPL JPN-SPSL-95-0449 v.: S. A. Valdes o: Nov 22,1995 i
- t. Lucie lant '
N D.J. Denver # 8 t = _ .: JPN/PSL Nuclear Engineering s e a: ST. LUCIE PLANT UNIT 2 PC/M 194 295, revision 0, Disconnection of ICW/CCW Pump Local Pushbutton Stations l l Attached for your use is PC/M 194-295 revision 0, which evaluates the disconnection of the local pushbutton stations for the 2A,28 and 2C CCW and ICW pumps. - The disconnection of the local pushbutton stations deletes the operator workaround associated with procedural requirements of resetting the RTGB control switches for the pumps after being stopped locally in order to retain the pump SIAS auto-start function. if there are any questions on this package, please contact Warren Busch at ext. 7484. I l POST DJD EJS l\ MA' ~lON F. R. G. cc: file N!.,y est into svc VERIFY pH t; Co.*i loi.t~vr
, C O r ;- ;; ,t;; t
\
e - Rr. vision 36 November,1995 Page 20 of 26 $. ATTACHMENT 2 PC/M REVIEW FORM (Page 1 of 2) a PC/M Number #N' Supplement Number O Expiration Date (; // /f6 4 PC/M TITLE: D 'os c cim E c Nc,, ab I O l~' l CC W I++;' L i c - / I . ! s i. ti- m,; l ~ /. .. , s . i Safety Classification: l i Y Safety Related Quality Related Not Nuclear Safety Related Administratrve PC/M Division: 1 l X Normal Either/Or As Requested-Package As-Fail 1 Implementing Documentation: 1 Department ERNPWO # W/O# WCD # C CM fl Ern 6 6 f f 7un cy503377s Nfjo
" R A C Cu)
' C6/7W 95c3 s 778 n '2 6 M' \ h _ G b / $ 7 ti t, q c % s 7Er.:
- 2-G C d t c / I 7 st 9 'tw u 7s/ s 2A EC U 66 /1747 %,o n 792 8. 2MW u // 7 50 95c T3 yg3 n 2CXcsd Areas Affected: Yes or No Description Operator Trainina Yes ,
Operatino Proceduree As per Trainino Department Reauirernents 4es Surveillance Procedures 4p( Maintenance Procedures /V : Spero Parts Al e Drawinos/ Technical Manuals /es fta p, ., /1 + ( A FUSAR Chance W; ' SRDs? &J N i TEDB Chanoe re.s W./ ' Human Factors (CREDIT) )es Environmental concoms M, in-Service inSDeCtion Ale Maintenance Rule N-Plant Restricbons TCM WrformeJ usfA ci/v cv .r u , h n1 rue c0> a t a e tw . Others ' '
/
-_ . . = . . . . - - . - - - - ' - -
w a re, . ra a {' Rsvision 36 Nov;mber 1995 Pege 21 of 26 ATTACHMENT 2 PC/M REVIEW FORM (Page 2 of 2) PC/M Review and Approval: db U Date / /2/ fc
/ SCE l bt Date / / b / 9b Wntenance OC I f Date I/ 4 /Jh__
Plant GenersfManager FRG Number: % -21 FRG Secretary 2 Date ,1/ /d 94 V V Comments: I PC/M Closeout Review: Approved for Closecut: 1 Date: / / i Maintenance QC Approved for Closecut: Date:
/_ /
SCE l l Procedure Notification Complete: Date: /- / SCE l i Training Notification Completed: Date: / / SCE Cancellation information: ~ Date: /. / SCE e
i FPL NUCLEAR ENGINEERING . ENGINEERING PACKAGE PC/M NO: 199-295 REV: O SUPPL: O PLANT: ST. LUCfE UNIT: 02 TITLE: DISCONNECTION OF ICW/CCW PUMP LOCAL PUSH BUTTON STATIONS _ REVISION DESCRIPTION: ISSUED FOR USE LEAD DISCIPLINE: ELECTRICAL EXPlRATION DATE: 6/1/96 REA NO: STAR O-94110473 DWA NO: N/A PC/M CLASSIFICATION: SR X QR NNS_,__ DESIGN ORGANIZATION: PSL/ SITE DISC CHIEF REVIEW Gi.QD? YES _ NO _,,X_ EXTERNAL INTERFACES: NONE DISC CHIEF SIGNATURE: N/A REVIEW / APPROVAL: INTERFACE TYPE GROUP PREPARED VERIFIED APPROVED FPL APPROVED
- INPUT REVIEW N/A MM MMMM MECH X
[) ELECT X , -
!V ,A f/
l&C X CML X NUC' X
/-f hM ES) X NUC FUEL X
- For Contractor Prepared EPs As Determined By Projects
" Review interfa:e As A Minimum On All EPs FPL PROJECTS APPROVAL: , Nb% DATE: [
v
PC/M No 194 295 ' Rav 0 Pcg2 L of 1 LIST OF EFFECTIVE PAGES PAGE NO. BLY 1 0 2 0 , 3 0 i 4 0 l I 5 - 0 6 0 1 7 0 l 8 0 9 O 10 0 11 0 : 12 O ) 13 0 l 14 0 l 15 0 1 { i l i e
PC/M N3194 295 ) hv 0 . Pagt 1 of_15,,_ ABSTRACT l l This mndification disconnects the local push button control station from the 2A, 28. and 2C Intake Cooling Water ilCW) pump and the 2A, 28 ano 2C Component Cooiing Water (CCW) pump contic: l circuits This modification is being implemented to delete the operational requirements of manually ~ reseting the RTGB control switch to *stop* and then back to
- auto
- which may be required to preserve ,
the automatic SlAS start feature of the subject pumps. The reseting of the RTGB control switches is
- only required after a local push-button stop of a running pump which was started via the RTGB control
- switch (reference PC/M 183 293, STAR O 94110473). Disconnection of the local push button stations 1
from tne associated control circuit will prevent the stopping of a running pump via the local push. ~j. button station and thus preserve the automatic SlAS start feature of the pumps without requiring a manual reset of the RTGB control switches. The modi *ication hes no impact on the safety related functions of the ICW and CCW control circuits and is being implemented to prevent the operator work around associated with the manual reset required after a local push button stop. The modification has no affects on plant operations since the local push-button stations are not required during normal, abnormal, emergency or maintenance operations. This modification includes the safety related ICW and CCW pump control circuits, therefore this EP is classified as Safety Related. A safety evaluation of these modifications has been performed in accordance with 10CFR50.59. This evaluation concludes that implementation of this EP does not involve an unreviewed safety question nor does it require a change to the Technical Specifications. It has also been determined to have no adverse effects on plant operations or safety. Therefore, prior NRC approval is not required for implementation of this modification. I 4
\
1
PC/M No 194 295 Rsv 0 Pcga 4 of 1
- l. DESIGN
- 1. St,uct"<r Svetam or comoonent (SSD Involved:
ICW Pump 2A,28 and 2C and CCW Pump 2A,2B and 2C control circuits. specifically: Local push button control stations for each of the pumps.
- 2. Puroose/ Function /Desion Basis of SSC Involved:
The ICW and CCW Systems provide a heat sink for safety related and non nuclear safety related components under normal operating and emergency shutdown conditions. The redundant pumps in each system (including the swing pump 2C) are controlled by pull to lock control switches at the RTGB, control switches at the respective 4160 volt cubicle for each pump and local push button control stations for each pump. The local push-button control stations are mounted in areas adjacent to the respective pump and allow the applicable pump to be started or stopped locally. The pushbuttons were installed for maintenance convenience and are not used for any design basis operating function or postulated accident. The local push button stations are not referenced in normal, abnormal, or emergency operating procedures nor are they used during maintenance activities on the ICW and CCW systems.
- 3. Safety Classification of Desian Chance: SR X OR NNS
- 4. Purnose of Desian Chance This EP is being prepared as requested by the St. Lucie Operations Department as an alternative to the existing operational requirements of manually reseting the RTGB control switch to "stop* and then back to
- auto"in order to preserve the automatic SIAS start feature of the subject pumps.
This requirement is only present after a local push-button stop of a running pump started via the RTGB control switch (reference PC/M 183 293, STAR O-94110473). Disconnection of the local push-button stations from the associated control circuit willinsure that stopping of a running pump via the local push-button station is not possible and therefore will inherently preserve the automatic SIAS start feature of the pumps without requiring a manual reset of the RTGB control switches. The current configuration of the push button control circuit it considered an
- operator work around* and is thus being rectified via this PC/M.
Although this is considered a " minor" modificatinn, the local start and stop features of the pumps l
. are described in the FSAR text, sections 9.2.1.5 and 9.2.2.5. This requires the modifications be implemented under the documentation of an Engineering Package with an accompanying safety evaluation performed in accordance with 10CFR50.59
PC/M N3194 295 . R:v O Pega L of 11_
- 5. Desertation of Desion Chance The modificat i on consists of disabling the local push button stations for the subject pumps by disconnecting the local push-button control station wiring at the respective 4160 volt switchgear cubicle for each of tiee respective pumps and at eacn local pusn button station. Contro;
- n..g ...!:
be spared and insulated in place at each cubicle and at each push button station. Finally, the pushbuttons and the box nameplates will be removed. l
l PC/M Na 194 295 l R:v 0 Pigt L of L
- 6. Desion Chance Checklist Does the Design Change involve / impact / require justification of; y.ES HQ REFERENCE Internal / External Flooding _ 1 Heavy Load Handling _ 1 Tornado / internal Missiles _ 1 Single Failure Criteria _ 1 Human Factors _ 1 Paging System Audibility _ 1 Masonry Block Wall interaction _ 1 Environmental Criteria _ 1 Plant Security Capability _ 1 i HELBA Criteria / Analyses 1 l
Seismic Qualification _ 1 i Seismic Interaction _ 1 l Electrical Separation Criteria _ 1 Accessibility /Laydown/ Clearance Requirements _ 1 Loads Applied to Existing Structures (+ buried) _ 1 EDG/ Battery Loading / Load Sequencing _ X l Hydrogen Generation in Containment _ 1, j Heat Sinks in Containment _ 1 Emergency Plant Operating Procedures (EOPs/ONOPs) 2 ,_ l 8.1 ASME Code _ 1 j Emergency Lighting Criteria _ 1 ! Snubber Criteria _ 1 Material Compatibility / Hazardous Materials _ 1 Electrical Equipment Grounding _ 1 l Cable Tray Seismic Loading _ 1 l Instrument Setpoints _ 1 Hurricane / Tornado Wind Loading _ 1 Thermal / Hydraulic Performance _ 1 Coatings inside Containment _ 1 Emergency Response Data System (ERDS) _ 1 Emergency Planning _ 1 NML Property Insurance Requirements _ 1 Environmental Qualification _ X Att. 4.1 Fire Protection Capability _ X Att. 4.2 Safe / Alternate Shutdown Capability J ,,_, 1-8.2 ALARA Exposure Criteria _ .2L, Att. 4.3 JPN Forms 3E,3F, and 71 shall be attached to justify the conclusion for these items.
- PC/M N3194 295 j R:s 0 :
Pcga 7 of 1.5_
- 7. Desien Evaluation / Justification The design function of the ICW and CCW systems is to provide a heat sink f or Safety Related and l d
Not Nuclear Safety components under normal operating, post accident and normal shutdown ] conditions. The pump control circuit design utilizes RTGB mounted puil to-lock control switches for control of manual end automatic design functions of the applicable pumps. The local push- ) I button stations for each pump allow local starting or stopping of the pump. I l The local control features for these pumps are based on maintenance convenience and are not j depended upon for manual starting or stopping design functions of the apphcable pumps under normal or abnormal operating conditions. The local push button stations are not part of the
" normal *, "off-normal" or ' emergency operating
- plant procedures.
Although the pushbuttons were installed for maintenance convenience, maintenance performed on these pumps is controlled through the plant clearance procedures that control pump power supply at the respective 4160 volt switchgear breaker cubicle. Maintenance is carried out on pumps when a clearance is taken out on the respective switchgear cubicle feeding each pump. The ! programmatic clearance procedures ensure that motors are not inauverter'tly started dunng maintenance activities and thus ensure personnel safety. Post maintenance testing on the pumps is coordinated with operations and any starting or stopping is actuated from the control room RTGB control switches. Therefore, the local push button stations are not used by mamtenance departments and are not required for normal maintenance procedures, i Finally, there are no OSHA or personnel safety requirements involved with the removal of these l
- local push button stations because the programmatic clearance procedures used during maintenance of pumps precludes inadvertent motor starts during maintenance activiti es. Motor starting and stopt ing is coordinated with the operations department which utilizes both the 4160 volt switchgear ubicle feeding each pump and the respective control switches at the RTGB.
1 The modifications provided in this EP do not adversely affect the ICW or CCW Safety Related systems' abilities to perform their intended Safety Related functions, Separation between Safety Related functions and Not Nuclear Safety functions has not been affected by this modification.
- 8. Evaluation of any "YES" resoonses in Desian Chance Checklist i
, 8.1 Emergency Plant Operating Procedures , Off-Normal Operating Procedures 2-0310030 (CCW) and 2 0640030 (ICW) shall be revised to reflect the disconnection of the local push button stations for the ICW and i CCW pump control circuits. 8.2 Safe Attemate Shutdown Capability Since separation requirements defined in Appendix R cannot be provided for essential components and circuits in the event of a fire in the control room or cable spreading room, alternative shutdown capability is provided. This ensures that in the unlikely event a fire makes the control room uninhabitable or renders equipment in either room inoperable, the plant can be safely taken to cold shutdown from a remote location. The CCW and ICW pumps are considered Appendix R essential equipment (ref. FSAR table 9.5A-2) and their function is required for safe shutdown. Because the systems are required to achieve safe shutdown, appendix R requires that the functions of the i i
PC/M No 194-295 Rev O Pega_f._ of,,1 L pump be controllable from an alternate location. Control for all 3 ICW pumps and all 3 CCW pumps is provided in the control room and at the respective 4160 voit switchgear cubicle feeding each pump. The switchgear cubicle control switch is therefore the Appendix R alternate means of control and relied upon tref. FSAR table 7.4-3) should an afternate means of control be required. Push button stations are not relied upon for alternate means of control. The Appendix R Safe Shutdown Analysis for Unit 2 is being revised by deleting cables associated with the pushbuttons.
- 9. Desion Verification Statement The design basis for the modification was reviewed to ensure that the overall design concept meets applicable FSAR, Regulatory Guide, and 10 CFR Part 50 requirements. The design of the modification was verified by a review of the functional requirements of the system and by the performance of a circuit analysis to determine if the proposed modifications perform the functional and safety requirements described in the FSAR. Integration of this modification with other designs in progress and impact on existing safety evaluations was verified by review of the Package information Tracking System and the Affected Drawings List databases, including the Active Safety Evaluation List. The safety analysis was verified by review of each conclusion statement required to satisfy 10 CFR 50.59 criteria for establishing whether isn unreviewed safety question existed, it. SAFETY
- 1. Descriotion and Purcose This Engineering Package covers modifications to the control circuits of the 2A, 2B and 2C Component Cooling Water Pumps and the 2A,2B and 2C intake Cooling Water Pumps. Specifically, this PC/M is an alternative to the procedural controls required to preserve the automatic start of respective pumps on SIAS after a running pump stop actuation by the local push-button control station. Under normal operation, two pumps will be running with their control switches in the
" Auto-after-Start" position, and the third pump will be idle with its control switch in the " pull-to-lock" position. The pumps will no longer be able to l'e controlled from the local control station.
This will not affect the proper response to a LOOP, SIAS, or LOOP /SIAS condition and will prevent the r'quirement of a manual reset as discussed in 1-4.
- 2. Analvsis of Effects on Safety The design bases for the ICW and CCW systems are described in FSAR Sections 9.2.1 and 9.2.2.
This modification will not affect the bases as described in the FSAR and will not result in loss of any protective function nor adversely affect the function of any Safety Related or Not Nuclear Safety structure, system or component. The removal of the local push-button control stations will not affect normal or off-normal operation of the respective pumps and will have no adverse effect on safety. The new design prevents the local stopping of an ICW and CCW pump, which in turn reduces the number of procedural controls associated with the manual actuation and subsequent actions following local push-button control station actuation. The current configuration relies on the control room operator being alerted to the loss of a pump, 1
)
1 PC/M No 194-295 R2v 0 Pega 9 of 11_ by vanous flow and temperature aiarms. Removal of the local stations prevents the need of a pump restart by turning its RTGB control switch to "Stop* and then
- Start *. In addition. if the automatic SIAS start feature was desired after a local con:rol station stop command, the control room l cperator must tu'n the RTGB control switch to *Stoo* and then back to " Auto" (the spring return )
position). A design integration review has been conducted, and the following sources have been reveewed: PSL Package Information Tracking System, as of 11/18, 1995. Nuclear Engineering Drawing index, as of 11/18,1995.
- PSL Active Safety Evaluation List, as of 11/18,1995.
As a result of this review it has been enncluded that no pending modifications or designs in progress will affected by this modification. J Based on the review of the St. Lucie Unit 2 Safety Evaluation List, there are no active safety evaluations issued or in progress which have any effect on, or can be affected by this EP. In addition, there are no Justifications for Continued Operation (JCDs) which affect, or are affected l' by this EP. l l
- 3. Failure Modes and Effects Analvsis (FMEA)
The FEMA is provided in the FSAR Sections 9.2.1 and 9.2.2 and in Tables 9.2 2 and 9.2-6. A review of these sections and tables has determined that this modification, which changes the pump control circuits, does not introduce any new potential f ailure modes to the pumps. Disconnection , of the push button control stations increases the overall reliability of the pump control circuits. Therefore, the conclusion of the existing FMEA is not changed. I 1 l
- 4 Effect on Technical Soecifications 1
, The ICW and CCW System requirements are contained in Sections 3/4 7.3 and 3/4 7.4 of the l Technical Specifications. A review of the Technical Specifications has determined that this j modification does not have an affect on, nor require a change to them, a
- 5. Unreviewed Safetv Question fUSQ) Determination
- 1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR?
This modification does not affect any equipment whose malfunction is postulated in the FSAR to initiate an accident. The modifications performed by this Engineering Package enhance the ability of the pumps to perform as intended during emergency and off. normal conditions. Therefore, the probability of occurrence of an accident previously described in the SAR is not increased by this modification.
- 2. Does the proposed activity increase the consequences of an accident previously evaluated in the SAR?
This modification does not affect the ability of the ICW and CCW pumps to perform their safety j function. Disconnection of the local push-button stations minimizes the procedural requirements associated with maintaining thG automatic start feature of the pumps after being a
- m _ _ m _ . . . ____m . . _ .. _ . - _ __ _ __ _ _ _ _ _ . _ _ _ _ _ . _ . . . _ . . _ _
, PC/M No 194 295 ,
Rav 0 l Pcge L of 1 l l 1 t 4 stopped by a local push-button actuation. Therefore, the consequences of an accident previously evaluated in the SAR are not increased by this modification. l 1
- 3. Does the propes*:d activity increase the probability of occurrence of a malfunction of equioment
' important to safety previously evaluated in the SAR?
The operability of the pumps in this modification is being enhanced by eliminating a possible scenario in which control room operators are required to detect and compensate for a local push button stop of the subject pumps which requires manual resetting of the RTGB control
- a; witches in order to re start a running pump or preserve the SlAS auto start feature of an idle pump. Since the reliabilsty of the pump control circuit is being improved and the procedural requirements associated with the fun'ction of the pumps are being minimized by this proposed
. change, it will not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR. 4 Does the proposed activity increase the consequences of a malfunction of equipment erhportant I. to safety previously evaluated in the SAR? The operaoility of these systems is being enhanced by this modification. Since the pumps will continue to perform and satisfy their design basis requirements, the proposed change will not
- increase the consequences of a malfunction of equipment important to safety previously f
evaluated in the S AR. i S. Does the proposed activity create the possibility of an accident of a different type than any
- p'eviously r evaluated in the SAR7 This modification does not change the function or design bases of any structure, system or
~
component important to safety as described in the SAR. Since this modification does not affect i ?. any equipment whose malfunction is postulated in the SAR to initiate an accident, it will not l
- create the possibility of an accident of a different type than any previously evaluated in the SAR.
? l 6. Does the proposed activity create the possibility of a malfunction ci equipment important to safety of a different type than any previously evaluated in the S/".i i j Disconnection of the stations ensures that local push-button pump stop cannot occur and I therefore eliminates the possibility of equipment malfunction associated with local push-button
- i station actuation and the procedural controls resulting from said actuation. Therefore, the l possibility of a malfunction of equipment important to safety which is of a different type than
! any previously evaluated in the SAR is not created by this modification. } 7. Does the proposed activity reduce the margin of safety as defined in the basis for any ? Technical Specification? I
.This modification does not affect the operational requirements for the cooling systems as
- defined by the Technical Specifications Sections 3/4 7.3 and 3/4 7.4. Therefore, the margin {
j of safety as defined in the bases for any Technical Specification is not reduced by this modification. 4 I i-r . . - . , ~ _ _
_ ~. - - - . _ - . . - -_ . -. . . - = . - . . - . . _ - PC/M No 194-295 R3v 0 Page 11 of 1 l
- 6. Plant Restrictrons l
7 is i El' can be partially implemented without any adverse consecuences. The implementation may , be performed dur#ng any mode of ooeration on one system pump out of service at a time. i The following Operating Procedures and Off Normal Operating Procedures should be revised to reflect the removal of the local push button stations: ONOP 2-0640030. ONOP 2 0310030. Procedures should not be revised until actual implementation of PC/M on all pumps for the applicable system. . I 1
- 7. Cor.clusions The foregoing constitutes, per 10CFR50.59 (b), that the modifications to be performed by th;s EP do not involve an unreviewed safety Question nor a change to the Plant Technical Specifications.
Therefore, prior NRC approval for the implementation of these modifications is not required. I I l l l 1 i 1
PC/M No 194 295 R;v 0 Pcge 1 of 1 - i 1 111. CONFIGURATION
- 1. Aficcted Document Checkhst
.Y11 HQ Reference FSAR 1 _ Attachment 4.5 l Technical Specifications ,,,,_ 1 TEDB 1 _ -Attachment 4_4 Security Plan ,,,_. 1 1 DBDs _ 1 Snubber List _ 1 I ISI/IST Program of Yes . Est review is teou' red) - - . 1 Code Stress Reports _ 1 l
- 2. Affected Drawinas i PC/M Dwo # Bgy Descriotion / Title Qi:ig Affected Dwas Egy h* EP Rev l JPN 194-295-OO1 0 CCWP 2A cwd 1 2998 B 327 201 15 y 0 JPN-194 295 002 O CCWP 2B cwd 1 2998 B 327 205 15 y 0 l JPN 194 295-003 0 CCWP 2C cwd i 2998 B 327 209 16 y 0 i JPN 194 295-OO4 0 ICWP 2A cwd i 2998-B-327-832 17 y 0 I JPN 194-295-005 0 ICWP 2B cwd 1 2998 B-327 833 16 y 0 JPN 194 295-006 0 ICWP 2C cwd i 2998 B-327 834 17 y 0 JPN 194-295-007 0 CCWP 2A sch 1 2998 B-326 201 6- y 0 JPN 194-295-008 0 CCWP 2B sch I 2998-B-326 205 6 y 0 JPN-194 295-009 0 CCWP 2C sch 1 2998-B-326-209 7 y 0 l JPN 194-295 010 0 ICWP 2A sch 1 2998-B-326-832 9 y O l JPN 194-295-011 0 ICWP 2B sch 1 2998-B 326 833 8 y 0_ !
JPN-194 295-012 0 ICWP 2C sch 1 2998 B 326-834 10 y O' l JPN 194 295-013 0 4.16 kv swgr 2A3-6 1 2998-1680 !. 10 yv 0 JPN 194 29E014 0 4.16 kv swgr 283-4 1 2998-1689 e 9 yg 0 JPN 194-295 015 0 4.16 kv swgr 2AB-2 1 2998-1697 i 9 y# 0 JPN-194-295-016 0 4.16 kv swgr 2A3-7 1 2998-1681 11 Yu O l JPN 194-295-017 0 4.16 kv swgr 2B3-9 1 2998 1694 7 yg O l JPN-194 295-018 0 4.16 kv swgr 2AB-3 1 2998-1698 9 yg O JPN 194 295-019 0 Intake Structure Cond. E 2998-G-385 10 ya 0 and Lighting (Before) JPN-194-295-020 0 Intake Structure Cond. E 2998-G-385 10 y# 0 and Lighting (After) JPN-194-295-021 0 Intake Structure Light. E 2998-G-386-1 e 14 .y# 0 Sections and Det. (Before) JPN-194.295-022 0 Intake Structure Light / E 2998-G-386-1 i- - 14 v4 O Sections and Det. (After) JPN 194-295-023 0 Yard Duct Runs & E 2998-G-408-1 f 10 yd O Lighting (Before) JPN 194-295-024 0 Yard Duct Runs & E 2998-G-408-1 '- 10 yg 0 Lighting (After) JPN-194-295-025 0 Yard Duct Runs & E 2998 G-408-1 f- 10 yd 0
I PC/M No 194 295 l R:v O Page 11_ of L,_ l Lighting (Before) l JPN 194 295 026 0 Yard Duct Runs & E 2998-G-4081 10 vg 0 Lighting (Atter) ; JPN 194 295 027 4A 'O Cable end Recewey E 2998 B-328 99 vM 0 ! JPN 194-295-028 0 Appendix R Safe E 2998-B-048 1 arm 0 Shutdown Analysis
- Indicate Yes or No if SRD for PSL. Indicate 1 (POD). 2 (CRD). 3 (MD). 4 (ED). or 5 (otherl for PTN.
- 3. Affected Vendor Manuals Plant Doc No Sg.y Vender /Ecuin Djag Remarks EP Rev l
NONE ! 4 Attachments Paaes i 4.1 Environmental Qualification Checklist (Form 3E) 1 4.2 Fire Protection Review Checklist (Form 3F) 3 i 4.3 ALARA Screening (Form 71) 1 l 4.4 TEDB Change Package 7 4.5 FSAR Change Package 3 l
- 5. References !
5.1 St. Lucie Unit 2, Final Safety Analysis Report, Amend. 9 ' 5.2 St. Lucie Unit 2, Technical Specifications, Amend. 79 5.3 Plant Operating Procedure 2-0310020, CCW System, Rev. 33 5.4 Plant Operating Procedure 2-0640020, ICW System, Rev. 28 5.5 Plant Off Normal Operating Procedure, 2-0310030, CCW System, Rev.18 I l 5.6 Plant Off-Normal Operating Procedure, 2-0640030, ICW System, Rev.18 ! 5.7 St Lucie Action Report (STAR) 0-94110473
PC/M No 194 295 R;v 0 P ge 14 of 1 5.8 Plant Off Normal Operating Procedure, 2-0030135. Control Rocm inaccessibility. Rev. 23 IV. MATERIAL l I None l l l l
)
\
- PC/M No j 94 295 [
Rav 0 ? Page 15 of_L,_ ? i V. IMPLEMENTATION r l 1, lmolementation Instructions /Soecifications t The implementor's work scope includes the following as detailed on the JPN drawings listed in the Affected Drawings Section (steps are not limited or confined to a specific order): al Determinate affected cables in the respective 4.16 kv switchgear cubicles and at the local push button stations per drawings JPN-194 295.001 through JPN 194 295.OO6 and JPN 194-i 295.013 through JPN 194 295.018, t b) At switchgear, insulate affected cable lugs, re-tag spare cables per JPN 194 295.027 and secure spare cables in place in the respective switchgear. , c) At local push button stations, insulate cable ends, re tag spare cables per JPN 194 295.027 and coil cables inside push-button station. l d) Flemove box nameplates and pushbuttons from each push button station. I
- 2. Post-Mod Accentance Testino Post-modification tests shall be in accordance with approved plant procedures, a) Testing shall verify proper operation of the pumps for auto and manual start /stop commands from both the control room and respective switchgear cubicle control switches. Testing should af. a verify proper operation of the Norm-isol switches.
I I O e
, Attictiment 41 PC M % 19F 29! '
Rev 4 e j Page 1 c81 i I ENVIRONMENTAL QUALIFICATION (EQ1 CHECKLIST l
- 1. ELECTRICAL EQUIPMENT N@
a) Does the PC/M add, modify or replace electrical /l&C equipment or affect any EQ Doc Pac? If NO proceed to Section 2. 1 _ b) is the equipment located in a harsh environment per EO Doc Pac i 1000 (PSL) or 1001 (PTN)? If NO proceed to Section 2. _ 1 j c) Does the equipment perform a safety related function? OR ' Does the equipment perform a quality related function such that its failure could mislead an operator or adversely affect any safety related function _ requited to mitigate or monitor an accident? ' OR is the equipment classified as Category 1 or 2 by RG 1.977
'If the answer to any question in Section 1.c is YES, then the equipment is subject to 10CFR50.49 and th must justify that the equipment is Environmentally Qualified.
- 2. ENVIRONMENTAL CONDITIONS N"E a) Does the modification add, relocate, raise flow rate of, or increase l radiation levels of any piping containing radioactive liquids such that lifetime or accident exposure of electrical equipment may be increased?
_ 1 b) Does the modification add or reroute any high energy piping, the f ailure of which could impinge upon electrical equipment or increase the post break local pressure or temperature? _ 1 c) Does the modification atter process fluid characteristics such that the post accident pressure, temperature, pH, or boron concentration to which electrical equipment will be exposed may be changed? _ 1 d) Does the modification alter any barriers that shield electrical equipment from high energy lines or radiation exposure? l _ 1 e) Does the modification add equipment to Containment or increase the post-LOCA makeup water inventory such that the design basis Containment Flood Level may be increased? 1 i
*
- If the answer to any question in Section 2 is YES, then it must be determined if any equip 10CFR50.49 could be impacted and, if so, that its Environmental Qualification is still valid. Document the determination in the EP, Prspared by/Date V Verified by/Date * #
JPN Form 3E, Rev. 6/94
l
- Ott achment O Pc w No 194 :s 4.
Pepe 1 o' ! FIRE PROTECTION REVIEW CHECKLIST
- l. SAFE SHUTDOWN CAPABILfTY YES* NQ REFERENCE Does the Engineenng Package install, relocate, modify, or affect the operation of:
A. Equipment on the Essential Equip. List X _ Section I 8.2 B. Safe Shutdown Analyses'
- 1. Safe Shutdown Circuits .)L. Section I 8.2
- 2. Alternate Shutdown Components ,_ .2L,_
- 3. Associated Circuits X
- 4. Manual Actions X ll. FIRE PROTECTION SYSTEMS Does the Engineering Package install, modify, or affect the operation of.
A. Fire Detection Systems __ _2L. - B. Fire Water Supply System 2L_ C. Water Suppression Systems _., X D. Halon Suppression Systems X E. Standpipes or Hose Stations l X ' F. Portable Fire Extinguishers X 111. FIRE RATED ASSEMBLIER Does this Engineering Package install, modify or affect the function of: A. Fire Barriers
)L.
B. F, ire Doors
- _)L.
If "YES" for PTN EPs, c Safe Shutdown Analysis Change Package (SCP) is required, if "YES" for PSL EPs, a drawing change is required. JPN Form 3F, Rev. 6/94
i
, 'A rt schment 4 PC M No 194-29 Rev 6 Page 2 o' .
Ill. FIRE RATED ASSEMBLIES (Cont) XES MQ REFERENCE 4 C. Fire Dampers X D. Mechanical Penetration Fire Seals X E. HVAC Duct Penetration Fire Seals X F. Cable Tray Fire Stops _. X G. Structural Steel Fireproofing _ X l H. Conduit or Raceway Support Fireproofing ,X
- l. Conduit or Raceway Fireproofing X J. Internal Conduit Fireproofing (Stuffing) X IV.
EMERGENCY LIGHTING X A. Does the Engineering Package install, modify, or affect the operation of ? , , , ,X, B. Does the Engineering Package obstruct the required light pattern of? X C. Does the Engineering Package add or relocate essential equipment or components which will require the addition or relocation of ?
- 1 D. Does the Engineering Package install, relocate, modify, affect, or require the use of hand held emergency lights? ,_ X E. Does the Engineering Package relocate, modify, affect, or obstruct the light pattern of perimeter security lighting to equipment requiring manual actions?
1 V. RCP OIL COtt FCTION Does the Engineering Package install, modify, or affect the operation of? _ X VI. MISCELLANEOUS A. Does the Engineering Package affect the quantity or protection of insitu combustibles (solids, liquids, or gases) beyond the assumptions in the FHA? _ X JPN Form 3F, Rev 6/94
' Antc+vewet 4 PC W No 19l 29 i Aev 9 Page 3 of VI. MfSCELL ANEOUS (Cont)
B. Does the Engineenng Package Cause the addition YES NO REFERENCE i of a large combustible inventory within 50 ft. of essential equipment, components, electrical manholes or structures? i _ X i J C. Does this Engineering Package modify or affect curbs or dikes used to contain combustible liquid i spills? X D. Does this Engineering Package cause the removal of a flame retardant material from non IEEE 383 cables? _ )L. E. Does this Er.gineering Package affect fire 1 ) protection technical specifications or fire fighting strategies? X t F. ! Does this Engineering Package install, modify or 1 affect the operation of alternate shutdown communications? _ X j G. Does this Engineering Package install, modify, or l affect hydrogen lines (or any combustible gas) in areas of the plant containing safe shutdown { equipment or componentr,7 ;
- .1-H. I
- Does this Engineering Package install, modify, or affect any HVAC equipment or room heat loads in areas of the plant containing Safe Shutdown l
equipment or components?
._ X PREPARED BY ZM' _ DATE /' 4I VERIFIED BY DATE 8
/
U
/
I JPN Form 3F. Rev. 6/94
!.tta:wt 4
- ?.' " 1M .'i Re, 4 Ca;t '. ;*
4 i ALARA SCREENING l
- 1. Is this PC/M Administrative only (no physical mndification)?
yes, Further ALARA screening is not required. _X_ no, Continue screening,
- 2. Does this PC/M involve a location in the Radiation Controlled Area (RCA)?
yes, Location: Continue screening. _X_ no, Further ALARA screening is not required.
- 3. Does the implementation , operation, or maintenance of this PC/M involve:
(Circle all *yes* responses) Movement of radioactive material? Modification of systems containing radioactive fluids or resins such that souting or retention characteristics are affected? Movement / modifications of existing permanent radiation shielding? ! Modification or removal of equipment that results in an uncontrolled opening / penetration into a i High Radiation Area, Locked High Radiation Area, Very High Radiation Area, or Exclusion Area? Diving associated with systems containing radioactive material? l Entrance into containment during power operation? Potential for personnel exposure to radiation field of a tr/ hour (assuming current area dose rates)? A total lifetime estimated dose due to the modification greater than or equal to one (1) man rem? For items in section 3 above is checked "yes", indicate "yes" below; yes, This PC/M has the potential to significantly impact personnel radiation exposure. Complete Form 72 to l total radiation dose is minimized by design. no,
- This PC/M.has little or no impact on personnel radiation exposure. Form 72 not required, however, nor ALARA precepts should be followed to minimize radiation exposure.
l
- N "l* Nff // Q$
Prepared by Reviewed by Note: If the preparer and reviewer are the same as for the PC/M, duplicate signatures are not required. JPN Form 71, Rev. 6/94
Faoltry PSL Un d PCM 194295 Attachment No 44 Revision No O Page 1 of 7 TOTAL EQUIPMENT DATA BASE CHANGE PACKAGE COVER SHEET Description of Change: Removal of pushbutton stations on CCW 2A,2B and 2C and ICW 2A,28 and 2C. Basis for Change: Pushbutton stations are being disconnected nd removed S , the 2' therefore the 'LCL PB STA* associate tag for each of the above purnps should be deleted from TEDB.
References:
- 1. PCM 194-295 Enaineerina Prepared by l Date l'lt tlff Verified by h Date //[22[#5 Approved by + Date #/7 ff (6ise ne Supv/ Lead Engineer)
Confiouration Manaaement Reviewed by: Date: l Data Entry by: l Date: i l l
1 l Pscility PSL Desti C2 (Jc l Co*ponent 3 C18 PP 20 l l lkM lh eMQ g [ l l l l A j !
< l Assocaste Lc. to STA l lAtt, 4 g I i i i.ev, o i ;
l Late Printed: 11/10/95 l lp.9 og l TOT C EQU1PMENT LATA Sass snart 30 Tag: gg seve W/A EQ Doc PaCr N/A e 4 System 21 RCULAT!WG WATER = IFTAK3 C00LIle; EATER seisnier ! he! y Clases Eng Ref: O Groups 18 20 su ote N/A 30 speer, y/A RG197: , 30 Related: N gg scoes W RG197 Cat , Q Saats: _ BQ pesarks 20197 Typer , ( Coop Typer CK sub Typea _ $4 ty Channe _ Pcei 5 Wase LOCAL PUSHBUTTON STATIDW AK3 Cool 3pg gAyg2 PUMP 2C Loca Codes 3FfR/19/$-3/W D Startup systee 805 Imen Deeca Instl MFC $3 N/L NOT LOCATED R.Rs EAAC gagineering verified: Y l Inst) Models N/L Sevi 000 g Pot NT-422589 I Coup Group FPRDS: Y Acct W 531 4 30 Taba _ 1 sulation A*v1s . Train , l Sestfold Rig , ritical Camps , Control Room Comps , i i Eork Group IST Sogde T prP Reqdi N Mint Pgees , , , , , , , , , 1 1 O
4 I i 1 1 I Pari! sty: PSL Uniti C2 IJC 2 , f ampone ett i 3CW PP 2A 1 J kb *Mh ! 8 1 Assocante LCL PB $*A l l g l l lAtt T em l t f ltevi O l
- ate Pr t r.t e d ; 11/12/95 l lPage i
af g i __I I / I 70'TAL SQU PME3r! LATA RA38 sMBIT i 80 g: N/A BQ Rev: N/A EQ Doc Pace N/A System 21 CIRCULATING MATER
- INTAKE COOLING NtTER
@C Seassici I Safety Claset _
&ng Sofs i
O Groups at - Surv Notes N/A 30 speers N/A RG19ti, 20 Relateds N 30 ews N/A RG197 Cats , Q Eseios _ 30 De kes N/A RG197 Typei , Coop Type: CK sub Types safety 1: _ Pes: Name: LOCAL PUSMBITTTON T3ON FC INTAR3 COOLING MATER PUMP 2A Locn Codes trTK/17/N*3/N.D Startup Systems 005 Locn Deeca Instl MFG 0 N/L NCrf LOCATED a RSS C Engineering verified T 2nstl Model N/L Rev 000 Orig Pos NY-422509
- Coop Groups FPED81 Y A*ct No 131 EQ Taba _ 3 ulation hay 1: , Train; ,
Scaffold Reg , ritical camps , Control Room Comps , l Nork Groups ist Reqd Y RNP Regd: N 3
*
- at ei _ _ _ _ _ _ _ _ _
I l 1 l I j i i i
I Pernlaty: PSL lina t i 02 IJc: i Compcnent> lCv PP 28
, 1J k lk4*2k
, ? Assocastes LOL P4 STA l l g j l lAtti T l l leen. ;
, 0.te Pt..ted. ,,,,,,,,
a 1 I
,,.,e g i i
T0"AL 1;*.' PMDr" LATA EME SWir* EQ g N/A 30 Reve N/A BQ Doc Paca N/A , System 21 CIRC 7 ATING MATER W INTAKI CDCLING MATER Q Groups 38 L...t, ,,.... _ ..,.... Eury Note: N/A 30 Speers N/A aclets , p SQ 4elateds W SQ ews N/A B0197 Cat , O Deeles _ SQ he hei N/A 33337 yype, , Coop Type: CK Sub Type Safety Cha el _ pen: Womee LOCAL PUSNBt71 TON ION INTAKE COCLING MATER PUt(P 23 Locn Code: JWTR/19/8*3/M D Startup system 005 ken Desco Itastl KrG 3: N /1, Wat g,ocA Ayygg gaggAR Ragineer109 Verifieds T Inst! Models N/L Reva 0 Orig Pos NT=422589 Comp Group NPRD5s Y et Not $31 30 Tabi Insulation navis . Train , Scaffold Req , Critical coups , Control Room Comps , j Work ups IST Reqd6 Y RMP Reqd N Mint 5 ss , , , , , , , , , e
,._m . . _ . . _ _ ._ _ _ , _ _ _
e PJtility: PSL ttns t e C2 LJC ; l Cowpenent, CCw P, 3C
, l j
M l4 '/
- 2 h l
}
e i Assetsates LCL ,B FTA
, lAtt: e {
i leev: I
- m te ,,,,.ted, m :,,,,
l
, ,,.,e o, 7 .
l l l TOT E EQ"3PMENT LATA SASE $KICT Tag N/A 30 mov: N/A RQ Doc Pats N/ Syst i 14 COMPONDr7 CDOLING NATER SYSTEM O# setentes safety Clases _ ang Ref O Group: 2B SQ Surv Note N/A 30 Speers N/A BG197: , 20 Related: N 3 Scews N/A SG191 Cat , O Basie s _ SQ R rkes N/A RG197 Type , Coop Type: CE sub Typen safety N - : Pen Namei LOCAL PU5RBUTTON ATION POR OMPONENT COOLING NATER PUMP 2C Lorn Code: CCW/21/N 9/N 3 Startup Systems cet I l Imen Deses i Insti MPG 8 N/L NOT LOCATED R RAS C Engineering Verifiedi Y j Instl Modeli N/L Re 000 Orig Pos NY.422589 1 I Coop Groups wrRDE T Acet Nos $10 50 Tabi _ 2nsu tion Ray 1: , Train , 1 Scaffold Reg , Cr & cal Comps , control Room Comps , 1 tork Group: 287 Reqdi Y kWP Reqd: N Maint Pguar l i 9
+
_..m... _ _. - ___ s i Pac 13 8 t y P8L tina t i C2 LJC J
*oeper.e nt : CT1r PP 29
, l l l l k .# hl 6
l Assocaste. LOL PD $~A l lAtts ' .*
- l leev, f., 7
- .t. ,r..t.d. . m .,,,
l ,,e,e TCTAL SQU2 PMDrT C.ATA &ASE SMEE" IQ T s N/A SQ kev N/A SQ Doc act N/A systems COMPoprsNT COOLING MATER SYSTEM senemac 3 safety Class, _ ang aef , Q Groups 18 8 surv Note: N/A 30 speera N/ SG197: , SQ Related W SQ $ we W/A kG197 Cata , Q hesis BQ 8e ks: W/A 20197 Type: , Coop Type: CE Sub Types Safety noel Pcu s Names LOCAL pCSEBtMTOW T20W A Cos:POWENT CCCLING WATER PUNP at Loca Crr,en CCN/28/W D/Wat startup systeen oss i Loen Deac: Instl MPG 8: N/L NOT N TED R Ess C Sagineering Verified: Y Inst! Model W/L Rev 000 Orig Poi pry.42250 9 Coop Group FPRD$1 Y Acet No: 530 SQ Taba _ 2neu tion Rev1s , Train , Scaffold teg . Crit cel Comps , Control Roos Camps , a Work Groupe IST Regde Y RNP tegd: W
-i=t m e< .
L
t l l I f Pac 111tyt P5L Daiti 82 1JC
- emponent C:v PP 2 A
, I i l M *2 Y[
e Assocseter LCL Pa STA l '
! (Att Ag I
. s Late Pranted; 12/12/95 i.ee.
lPage o"7cf ,
"CTAL SQVI, HINT DAT A bA&E EMIE" .
SQ Tag N/A EQ povs N/A BQ Doc pact s/A ' Systems 14 Cool COOLING WATER 575T338 sete te 2 sarety es, ,_, Rn,Rer: J Q Group: 2B SQ Surv Not N/A BQ Speers N/A RC191 , SQ Related: N SQ Scow: N/A R3191 Cat , Q teste s _ 30 samarks: N/A B0191 Typea , Coop % CR Sub Type Safet la _ Pes: Names LOCAL PUSHBLTTTott STATION FOR POWENT COOLING WATER PUMP 2A Loca Codei CCw/20/8/W 1 Startup Systemi 040 Loco Deses Inst! MFG es N/L WCFT 14CATED TR RESEARC gineering Verified: Y Inst! Models N/L Reva 000 Ori Pos NY 422589 Camp Groupt NPRDS: Y Acct Noi 0 . l sg Tabs _ eulation kevli . Train .
]
Scaffold Regs , ritical Comps , control Room Cospa , tork Groups IST peqds Y pur Ragde N Maint Pgast . . . . . . . . . i l j 1 l 4
_- - _ - . ~ i s i l PC/M 194 295 0 Attachment 4.5 ! Rev.O i Page 1 of 3 FSAR CHANGE PACKAGE (FCP) COVERSHEET Plant: PS' , Unit: 2 fSAR PAGES ATTACHED: 9.2 5 9.2 10 l FSAR FIGURES ATTACHED: None
]
l l COMMENTS: None 1
/RL tirs- Ww y/ubc PREPARED / DATE VERIFIED / DATE v/n/S(
APP VED / ATE O e
.. _ -. .- . - - - - -. .- ~ - - -
l h no: seismic Cate; cry : and are isclared au:cma: PC,M 19't295 ; o Category I portions upon S!AS. The isola:icn val Attachment jk/h Rev.O , The pumps and hea; exchangers are located insi Page 2 of 3 building. Seismic qualification of system demonstrated by manuf acturer calculations anc :s 3.9.3. The Component Cooling Water System is protectec missiles by the Component Cooling Water Structu r Building. Component Cooling Water System Equipment suscep i protected by locating all safety related components above the maximum ex-pected water level and wave runup during probable maximum hurricane as ; di-scussed in Section 3.4. 9.2.2.4 Testing and Inspection Hydrostatic tests at 150 percent of design pressure are performed in the , shop on the heat exchangers. Performance tests to demonstrate design re- : quirements are conducted and heat transfer characteristics are verified l af ter installation. Eddy current tests of all tubes are performed in ac- , cordance with ASTM B-111, Paragraph 10, for entire tube cross sections. - All pressure-containing welds are checked by radiographic examination. Hydrostatic tests to 150 percent of maximum operating head is performed in the shop on each pump casing. Performance tests are performed on each c.nmp characteristics. Nondestructive testing is performed on welds, forg- !
. gs and castings in accordance with the requirements of ASME Code, !
'setion III, Code Class 3 equipment.
Preopstational testing and inspection of the CCWS is discussed in Section r 14.2; Periodic testing is a part of the Technical SpecificLtion. 9.2.2.5 Instrumentation Application i Tacle 9.2-7 gives a functional listing of component cooling water instrumentation. ' i The pumps and heat exchangers have diverse parameters measured to confirm , the correct operation of the equipment involved. The monitoring of flow, ! temperature and pressure at the points indicated in Table 9.2-7 and Figure 9.2-2 provides the control room with information for operating the essential and normal header systems. ! 5 The pumps receive a start signal on SIAS. The pumps can be also started ! and stopped bctF Iccc11y cnd from the control room. l t 1
. 9.2-10 ,
t i
_._ _ . . _ . . . _ - _ _ _ . . _ _ . . _ . _ . . .. ._ -__..m_ _ _ _ _ _ - . . _ _ . . _ . _ _ _ _ _ _ _ . . _ . /' t di High water velocities tn the ICW System seen '
' Some a umulation in the water boxes :s expe. PCIM 194 29,5 tunes is unl:kely cue to :ne s: curing effe_. Anachnwnt - I ,
l Rev.0 ; Ine CW pumps are :apatie Of pumping :ntake water . Page 3 of 3 i suspenced ma:erial or sil: w -heut a=versely effe: , l 3.2.*.4 . Testi..: a..d :nste::icns Each intake cooling water pump is shop-tested at n. ' capacity points including the design point to measi input and efficiency. Shop hydrostatic tests on the at 150 percent of the maximum operating pressure. and forgings are nondestructive tested in accordani i Section III, Code Class 3. l Preoperational testing and inspection of the ICW S,..... ~ l Section 14.2; periodic testi,ng is a part of the Technical Specifications. 9.2.1.5 Instrumentation Application Table 9.2-3 lists the parameters measured by the Intake Cooling Water System instrumentation. The heat exchanger parameters and pump status are monitored
- 3. either locally or in the control room as identified in Table 9.2-The intake cooling water pumps can be started or stonped eith:: ::::::y ::
from the control room. The pumps receive a. start signal upon SIAS. 9.2.2 COMPONENT COOLING WATER SYSTEM l The Component Cooling Water (CCW) System is a closed loop cooling water l system that utilizes demineralized water to cool various components as shown schematically on Figure 9.2-2. Design data for the CCW System - components are tabulated in Table ~9.2-4. l l 9.2.2.1 Design Bases The Component Cooling Water System is designed to: a) provide operatinga and heatshutdown sink for the reactor auxiliary systems under normal . conditions. i b)- provide an intermediate radiological barrier between radioactive , 7 systems and the Intake Cooling Water (ICW) System. h
=
i 9.2-5 Amendment No. 8, (9/93)
.m ,.
_ _ _ _ . . m . _. . - - - .__ - . - . . - . - .. - .- . o. . . .. . .- y .e Page No. 27 - 10/24/95 ST. LUCIE U' NITS 1 & 2 ,, PLAFC CHANGE / MODIFICATION TRACKING CONTROL LOV SYS DATE OUT IM % 0 PEN ACT DATE TO DATE TO PCM SP RECD STAR NO. FRG'D AGE DP COM D 0 T C P S L T/F VAULT A/B NO. No NO. NO. TITLE / DESCRIPTION 08/23/94 NO ARP .F. / / / / 94069-1 0 5409 CW/ICW INSTR IMPULSE LINE PRODUCT & MAT CHG 34 / / 0 .F. / / / ./ , 94070-1 0 5506 STEAM GENERATOR 1A SUPPORTS 0 .F. / / / / 0 5507 STEAM GENERATOR 1B SUPDORTS 34 / / .F. 07/01/94 06/14/94 94071-1 94072-1M 0 5508 BORIC ACID BATCHING TANK HTR NOZZLE LINE INCR67 4407/26/94 06/06/06 NO NOMM AFE D IC 100 .F. 09/22/94 08/23/94 94073-1M 0 5509 ISOPHASE BUS AIR COOLERS SET PT CHANGE 12/08/94 11/28/94 COPGAINMEPG RADIO REPEATER 68 09/13/94 94 EM 100 .F. 09/22/94 06/06/06 94074-1M 0 5510 48 08/30/94 NO MM AFE .F. 94075-1M 0 5511 CCW PUMP MISSILE SHIELD INSULATION 0 .F. / / / / PLAtG LIGHTING UPGRADE 125 / / 09/09/09 09/09/09 94076-9M 0 5512 09/09/09 CAN .F. 94077-2M 0 5513 125 07/05/94 94 NW CRN .F. / / 03/28/95 94078-1M 0 5514 GENERIC PKG FOR DRAWING CORRECTIONS .F. 03/28/95 03/28/95 G"NERIC PACKAGE FOR DRAWING CORRECTIONS 125 07/05/94 NO NW CRN D 94079-2M 0 5410 20 10/11/94 NO MM 100 .F. 02/07/95 01/18/95 94080-1M 0 5411 CHEMICAL FEED TANK LEVEL GLASS DRAIN VLV RE 0 .F. / / / / 0 5412
/ / 12/13/94 11/29/94 94081-VALVE 3480 REPLACEMENT 125 09/26/94 94 BA 100 .F.
12/13/94 11/29/94 ' 94082-1M 0 5413 125 10/27/94 94 BA 100 .F. 94082-1M 1 5521 VALVE V3480 REPLACEMENT 0 .F. / / / /
/ / / /
94083- 0 5414 / / 0 .F. / / 94083- 0 5415 / / 0 .F. / / / / 94084- 0 5416 66 09/26/94 NO EM ARP .F. / / / / 94085-2M 0 5417 STARTUP & AUX TRANS M ANSFER SWITCH REPLACMT .F. 12/08/94 12/01/94 i 94086-1M 0 5418 WESTNGHS DEH CLOSE INTER VLVS CONTROL CKT MOD 22 08/09/94 94 IC 100 01/10/94 12/16/94 ! MOD SPRING HANGERS LCVH-30A1,30B1,30C1,30D1 11 08/18/94 NO MM 100 .F. 94087-1M 0 5419 34 09/01/94 94 MM 100 .F. 12/08/94 11/16/94 94088-1M 0 5420 PRESS SPRAY LINE SUPPORT SPS-237 MOD 09/22/94 09/01/94 PRIMARY WATER STORAGE TANK LOW LEVEL ALARM 45 07/28/94 NO IC 100 .F. 09/22/94 09/08/94 94089-2M 0 5421 PRIMARY WATER STORAGE TANK LOW LEVEL ALARM 45 08/25/94 NO 100 .F. 94090-1M 0 5422 .F. 01/05/95 01/04/95 94091-9M 0 5424 ADD OF EGRESS BIOMETRICS (HAND CON 7ROL UNITS) 108 09/27/94 NO BA SAC 01/05/95 12/16/94 RM-26-11 SAMPLE DISCHARGE LINE MOD 77 10/20/94 No BA 100 .F.
/ /
94092-2M 0 5425 5 02/23/95 MM ARP .F. / / 94093-2 0 5426 CW/ICW INSTR IMPULSE LINE PRODUCT & MAT C C 11 08/06/95 595 IC 100 .F. 08/18/95 08/17/95 94094-1 0 5427 LAMBDA POWER SUPPLY .F. / / 12/06/94 94095-2M 0 5428 AS-REQ REPLCMT EDG LUBE OIL SYS TIME DELAY RE 53 08/05/94 NO 37 09/26/94 94 MM 100 EM ARP .F. 12/13/94 11/29/94 ' SIT AIR OPERATED VALVES LA!GERN RING REMOVAL 01/10/95 12/16/94 94096-1M 0 5429 TRAVELING SCREEN MOTOR REPLQ4T 5 08/30/94 NO EM 100 D .F. 94097-1M 0 5430 5 08/30/94 NO EM 100 .F. 06/12/95 06/12/95 94098-2M o 5431 TRAVELING SCREEN MOTOR REPLCMT 12/12/94 11/14/94 CW FLOW MEASUREMENT TAPS MOD 7 08/11/94 NO MM 100 .F.
/ /
94099-1M 0 5432 REPLCMT OF ITE JL-2 125VDC BREAKERS 62 09/26/94 NO EM ARP .F. / / 94100-1M 0 5433 REPLCMT OF CW PUMP LUBE MATER GLOBE VALVES 7 09/09/09 NO MM CAN .F. 04/18/95 09/09/09 94101-1M 0 5219 GLAND STEAM PIPING MODIFICATION 13 10/25/94 94 BA 100 .F. 12/08/94 12/02/94 94102-1M 0 4968 27 09/13/94 NO MM 100 .F. 02/07/95 02/02/95 94103-2M 0 3786 TURBINE LUBE OIL PIPE HANGER MOD 01/05/95 12/16/94 0 4745 DELETION OF HPSI/LPSI PUMP LOW FLOW ALARMS 37 11/10/49 N3 IC 100 .F. 02/07/95 01/06/95 94104-2 PZR PRESSURE (PIC-1100XLY) LOOP ISOLATION 72 10/25/94 NO IC 100 .F. 11/10/94 06/06/06 94105-2M 0 5444 ADD BONNET RING FEELER GAUGE HOLES COND/FW VL 18 09/26/94 NO MM AFE0
.F.
94106-1M 0 5445 / / .F. / / / / 94107- 0 5446 FIRE DETECTION AND SECURITY INVERTER FUSE RE 62 11/15/94 NO EM 100 .F. 08/07/95 08/07/95 94108-2M 0 5447 SETPOItG CHANGE TO HYRAZINE LOW LEVEL ALARM 39 10/25/94 NO IC 100 .F. 02/07/95 01/09/95 94109-2M 0 5448 VEHICLE BARRIER SYSTEM 108 05/04/95 NO BA 0 .T. / / / / 94110-9 0 5449 SAFEGUARDS BYPASS INDICATION PANEL MOD 131 06/01/95 NO IC 100 D T .F. 08/03/95 08/03/95 q 94111-2 0 5450 RX VESSEL CLOSURE HEAD PASS S" IUD TENSION EVAL 33 10/27/94 94 NW 100
.F. 12/?2/94 06/06/06 D 94112-1M 0 5451 2A & 2B EDG SYNCH CHK RELAY REPLACEMENT 53 10/11/94 NO EM ARP .F. / / / /
% 4113-2M 00 5453 5452 37 11/04/94 NO EM AFE .F. 12/12/M 06/06/06
- 94114-1M HPSI/LPSI MOTOR BEARING TRD REMOVAL 78 05/30/95 IC 0 .T. / / / / ,
M'5"~94115-2M 0 5454 RAD MONITORING SYS NRCM NIN BIN ASSEM/ TIMER
. \
Florida Power & Ught - ST. LUCIE PLANT ! PLANT CHANGE / MODIFICATION (PC/M) ROUTING TITLE: SETPOINT CHANGE TO NSR REA # PC/M # PC/M l' THE HYDRAZlNE LOW LEVEL OR RFD # SUPP. SR XX S R# ALARM LIS-07 9 9,o80 .1 109-294 M 0 i l 4 ORIGINATOR: PEG OtfrAGE T/O MODE IMPLEMENTING DEPARTMENT:
- CHRIS P. CHARTIER RELATED l&C i Yes i I
No 4 ROUTING INIT DATE INIT DATE REMARKS l 4
- 1. PC/M Coordinator AEK 9/19 S4 DfS./
- 2. Document Control // s f/ #d,.
- 3. Project Field Engineer
- 4. Construction OC
- 5. ISI Coordinator
- 6. Start-Up Engineer (Test Required) YES NO DEPT SIGNATURE If yes, Testing Requirements:
INIT DATE INIT DATE REMARKS
- 7. Tech Staff F f/ f"! /# '/
- 8. Plant OC Dilh IM3M fdlr# fY-@
- 9. FRG Approval % /0259/
- 10. PC/M Coordinator /4y
- 11. Document Control d
- 12. Turnover -
k
- 13. Plant OC COMMENTS:
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Inter Office Correspondence b FPL JPN-PSLP-94-0561 To: S. A. Valdes St. Lucie Pla Date: 9 (6,; C) k
\
From: D. Denver Department: JPN'/JB rt. :_ fr;FC TN C Nuclear Engineeri ,, p i n
Subject:
ST. LUCIE PLANT UNIT (S) 2 I '$ ! REA/ PROJECT # STAR 2-94080081,l ! '.5 i TITLE: Setpoint change to the Hydrazine j"! - Low Level Alarm (LIS-07-9) ] DCC PC/M/ FILE # 109-294 ~ - - Attached for your review, approval, and use is the (EP/MEP) for the subject Plant Change Modification. This package provides the details necessary to implement the following modification: Chances the setooint for the Hydrazine Low Level Alarm (LIS-07-9). By copy of this letter, Document Control is requested to distribute this package to the normal distribution plus the following: REA Originator: J.P. Honeysett - PSL/0PS 1 This completes our effort on STAR 2-94080081.l If you have any questions, contact Chris P. Chartier at (407) 691-2103. l DD/cpc y g , I Attachment l q . - - __. j
~
- v. hh/-
j*{.J c_ : Otp t ' POST H.L. Fagley/DCC - CS/PSL (w/ Repro) tLO S. Xozlin-JDC/JB(w/dwglist&origdwgsh. CC' 1HE y~. TRou.aF. R. G
- D.M. Wolf - JPN/JB PSL INFO SVC J.A. Porter - JPN/JB (w/dwg list) ,
C.L. Schaeffer - JMC/JB L
.C.M. Spalter - JPN/JB (w/ orig word doc & copy of dwgs)
Others: L. Roaers - ICM/PSL F:\NIS\FORMSiPSL5 FORM.009 form 1008 IStocked) Rev 2'89
Cl 3-PR/PSL-1 Revision 30
. February,1994
!* Page 26 of 39 4 ATTACHMENT 2 PC/M REVIEW FORM
- (Page 1 of 3)
/M -J WM Supplement Number o PC/M Numoer i
1 Plant / Unit Ps c / z, Expiration Date /t/z//of
)
Safrty Classificanon: ] Safety Related Ouality Related Not Safety Related Originating Document Number. Originator / Organization: (REA, NCR, RFD, etc.) J. P. wm* sTA a z- w o a o o s i . , Title of PC/M: 5 + + pii.+ C,ww 4 A %dmh Lm leu c \ ) Alu m ( u s - on - q) PC/M
Description:
Cheavp s-c+po'.4 f% 35 5 ' += 3 b ' " l
'I ^ 5 '" "'*\"""
\'"'\ fM J h +* #"\'d S P c 'l Ebb Need for PC/M-(&ns g) i s n, & e,cmacJ.c & c.6 . s 3s.s" ( r,, g) i implementing Documentation: CWO #:
Rackfit worked
/ Plant-worked WPS #:
No testing required / Plant-tested NPWO #: tygg7 , y , g 5 g 3. , Yes or No Descnotion Other Areas Affected: As Der Traintne Department Recurrements Operator Trainino Yes Operatino Procedures %s Surveillance Procedures % Maintenance Procedures do _Soare Parts Ah e%_ 5*+oekt Lis & SRDs? Y //NI_ Drawines/ Technical Manuals # FUSAR Chance rJo U.
'_T EDB Chance tJ .
Human Factors ERDADS Affected 4 In-Service inspection LJo Others h
Ol 3 PR/PSL-1 Revision 30
' February,1994 Page 27 of 39 ATTACHMENT 2 PCIM REVIEW FORM (Page 2 of 3)
/ 0 4- 194 M Supplement Numcer o PC/M Numcer Plant /Und P5t / L Expiration Date 17 / ~5 I / &
PC/M Type: Physical Work Required
/ Normal .ARP As Fail Either or Technical Statt Review and Comments: No Yes Is this PC/M Safety Related? /
- 1. /
- 2. Is the answer to the 10 CFR 50.59 -
screening question Yes?
- 3. Does the PC/M require any physical work to be performed?
/
if the answere to these questions are NO, no FRG review is required. Safety Evaluation complies with 10 CFR 50.59: N/A M de# Date / / Technical Staff PC/M Review Complete: Technical Statt //t< Y--- ^Date 10 / / t- l 94'
/
m' Maintenance CC Review and Comments: MD g, C PC/M Review Complete: Data f 0 / O / CC Signature I m - ,/ - Plant General Manager Approval: Signature [# _ Date # / #[ / FRG Approval: [ Date /d/6 M_,, /, FRG Secretary FRG Number: - -
l
~ ,
1 QI 3-PR/PSL-1 Revision 30 February,1994 Page 28 of 39 l j l ATTACHMENT 2 l PC/M REVIEW FORM (Page 3 of 3) 4
/ 04 -2 W m Supplement Numcer o l PC/M Number i
Plant / Unit PW / 2-J ! Maintenance QC Closecut Review: Required Date: /2 / et / Sy Not Required 4
/ Technical Staff I
/ Date: / / 6 / 9(
Approved for Closecut: ~ l Maintenance 4C 1 ' Comments: 1 l 4 9 Technical Staff Closecut Review: A l' / Date: / /6 /_1 Approved for Closecut: Technical Statf Date: / /__ Procedure Notification Completed: _
- Technical Staff Date: / /_.
Training Notification Completed: Technical Staff i Cance!!ation Information: Date: / /_ Technical Staff
PC/M IO9-294 Revision 0 Page I of 4 MINOR ENGINEERING PACKAGE (MEP) PLANT St. Lucie UNIT 2 PC/M NUMBER I09-294 SUPPL 0 ORIGINATING DOCUMENT STAR 2-9408008I.I EXPIRATION DATE I2/31/95 PC/M CLASSIFICATION X SR QR NNS ADMIN TITLE _SETPOINT CHANGE TO THE HYDRAZINE LOW LEVEL ALARM (LIS-07-9) ADDITIONAL REQUIREMENTS / INSTRUCTION 5 YES NO AS-BUILDING TO COMMENCE UPON ISSUANCE OF PACKAGE? X THIS PACKAGE HAS THE POTENTIAL TO SIGNIFICANTLY X IMPACT PERSONNEL RADIATION EXPOSURE (See QI 3.I3). IF YES, JPN FORM 72 IS REQUIRED. IOCFR50.59 SCREENING YES NO I) DOES THE CHANGE REPRESENT A CHANGE TO THE FACILITY AS DESCRIBED IN THE SAR? X
- 2) DOES THE CHANGE REPRESENT A CHANGE TO PROCEDURES AS DESCRIBED IN THE SAR? X
- 3) IS THE CHANGE ASSOCIATED WITH A TEST OR EXPERIMENT NOT DESCRIBED IN THE SAR? X
- 4) COULD THE CHANGE AFFECT NUCLEAR SAFETY IN A WAY NOT PREVIOUSLY EVALUATED IN THE SARI X
- 5) DOES THE CHANGE REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS? X NOTE: IF THE ANSWER TO ANY OF THE ABOVE 10CFR50.59 SCREENING QUESTIONS IS YES. THE MEP CANNOT BE U REVIEW / APPROVAL:
INTERFACE TYPE GROUP PREPARED VERIFIED APPROVED FPL APPROVED
- INPUT REVIEW N/A MECH X ELECT X
. , 4 _ ,,
!&C X O /
CIV!L X uUe - x AM/WM ES! X NUC FUEL X
- For Contractor Evals As Deterrined By Pro ec ** Review Interface As A Minian on All Non-Adnin MEPs FPL PROJECTS APPROVAL: DATE: I 1
i
PC/M 109-294 Revision 0 , Page 2 of 4 ENGINEERING JUSTIFICATION This Minor Engineering Package provides the justification to change the setpoint of level indicatro switch (LIS-07-9) for the Hydrazine storage tank. The following will be modified:
- 1. The mpoint for Level Indicating Switch (LIS-07-9) will be changed from 35.5" to 36.7".
- 2. The setpoint list and drawings will be revised accordingly.
JUSTIFICATION The design basis for the hydrazine storage tank is to provide controlled amounts of hydrazine to the contaimnent spray water following a loss-of-coolant accident to remove l radio-iodines from the containment atmosphere. The hydrazine storage tank has several alarming setpoints which are used to alert the operator when the hydrazine level in the i storage tank is reaching a predetermined level. . 1 Currently, annunciator window S-10 for the hydrazine low level alarm (LIS-07-9) is set at 35.5" (674 gallons)(Ref. 3). However, a more conservative setpoint for the low level l alarm is required, since the existing setpoint is very close to the technical specification ! required minimum level of hydrazine (675 gallons)(Ref. 2). A more conservative setpoint ! can be accomplished by taking the existing setpoint of 35.5" (674 gallons) + .1" (1.6 l gallons) = 35.6" (675.6 gallons) + 1.10" (instrument inaccuracies) = 36.7". Calculation ! PSL-2FJI-92-024 has been reviewed and it has been determined that the new setpoint is acceptable and will envelope the technical specifications minimum requirement for hydrazine levels for the Containment Spray / lodine Removal Systems. This setpoint change does not change the design basis of the Containment Spray System / Iodine Removal System (CSS / IRS) or alter any plant Technical Specification (Ref. 2). This setpoint change does not change the facility as described in the FSAR (Ref.1). While the annunciator window (S-10) is non-safety, this Minor Engineering Package is classified as safety related since the instrumentation associated with the setpoint (LIS-07-9) is classified safety related.
l l PC/M 109-294 , Revision 0 l Page 3 of 4 ' CONCLUSION: A design integration review was performed and it has been concluded that this modification does not affect nor is it affected by any other designs in progress. This review included the following: l (1) The computer database PC/M Log as of 9/7/94. (2) The computer database Affected Drawing List as of 9/7/94. (3) The PSL Active Safety Evaluation List as of 9/7/94. As a result of this design review, no outstanding designs were identified which would interact with this modification. In addition, there are no Justification for Continued Operation (JCO's) which affect, or are affected by this MEP. l The CSS / IRS Systems hre described in the FSAR,'Section 6.5.2. However, the level l indicating switch (LIS) setpoint for the hydrazine storage tank is not described in the FSAR and it is concluded that the setpoint change from 35.5" to 36.7" is acceptable and has no direct or indirect impact on any safety functions required for analyzed accidents, and does not increase any radiological hazards. The portion of the CSS / IRS System being modified is safety related, since it is required to meet certain technical specification levels and is used during a DBA. The CSS / IRS serves only as an Engineered Safety Function (ESF) and performs no normal operating functions. The annunciator window (S-10) is non-safety and provides additional assurance that the hydrazine level is maintained within the Technical Specification levels. This setpoint change does not alter the CSS / IRS system ability to mitigate or monitor the consequences of an accident. In addition, the modification to the level switch setpoint does not alter any specific design, operational or performance requirements for the Ca/ IRS System. Furthermore, these changes are not considered a change to the facility ' as described in the FSAR, Section 6.5.2, nor does this modification change or alter the Technical Specification. While the annunciator window (S-10) is non-safety, this modification affects a safety related piece of equipment (LIS-07-9), therefore this package is classified as Safety-Related. This conclusion supports the 10CFR50.59 screening results. Therefore, there is no unreviewed safety question. SPECIAL INSTRUCTION / IMPLEMENTATION REOUIREMENTS
- 1. Notify operations of work to be performed in the control room.
- 2. Change setpoint for LIS-07-9 from 35.5" to 36.7". See attached setpoint list.
- 3. Verify operation ofloop (new setpoint & annunciator S-10).
PC/M 109-294 Revision 0 . Page 4 of 4 OTHER AFFECTED DOCUMENTS Reference Paragraph /Page or
- Attach Revised Pages Plant to identify any affected procedure ~which may require revision due to this modification, including the annunciator procedures.
P_OST MODIFICATION TESTING ICM shall verify proper operation of loop I 07-9. REFERENCES
- 1. PSL Unit 2 FSAR, Amendment No. 8
- 2. PSL Unit 2 Technical Specification, Amendment No. 67
- 3. I&C Calculation PSL-2FJI-92-024, Rev. 0 " Unit 2 Hydrazine Storage Tank Level"
- 4. Computer Database Affected Drawing List as of September 09,1994.
- 5. Computer Database PC/M I.og as of September 09,1994.
- 6. PSL Active Safety Evaluation List dated September 09,1994.
ATTACHMENTS Attach Number Number Descriotion/ Title of Paces 1 Setpoint List 1 6
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Facility PSL Unit 2 PC/M or DCR No. PCM 109-294
- Attachment No. N/A Revision No . N/A Page 1 of 2 TOTAL EOUIPMENT DATA BASE CHANGE PACKAGE COVER SHEET
- Description of Change: INCORPORATION OF SET-POINT LIST DATA TO TEDB FOR PCM 109-294 Basis for Change: PCM 109-294 ATTACHMENT 1 4
References:
Encineerine Prepared by SEE PCM 109-294 Date Verified by SEE PCM 109-294 Date Approved by SEE PCM 109-294 Date (Discipline Supv/ Lead Engineer) DRAWING UPDATE Reviewed by: Date: Data Entry by: /9( . 8 ~# < - Date: / // S /9 5 Q .,
/
- Data Entry Verified by: .. Date:
[ /./ 9f For instructions, see 013.6 JPN Form 53, Rev 12/94 O e
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~ Component: Sys: 07 Train: Fac: PSL Unit: 02 Associate: Assign Priority: C3 MASTER Name: PCM-109-294 SETPOINT CHANGE TO Work Type: 3 WORK ORDER TASK HYDRAZINE LOW LEVEL ALARM Location: RAB/RTGB-206 LMD- e 94025805 01 Defect / Request: SETPOINT CHANGE / ( ER/PWO: 64 / 3600 y[y Chg Loc: 910 gg , h [g b-25-M [ "
/ PAGE 1 of 5
(( Detailed Explanation: . 'T IMPLEMENT PCM-109-294 SETPOINT CHANGE NRRD ) s EKM E Mu Work Request: 94015181 Def Tag: PCM-109-294 Loc: . Trbl/Brkdown: N LCO: N Unit Cond Reg: 8 i NPRDS: N Fail Date: Time: Det: Stat: Symp: Originator: WAP000F PAULSEN WA Date: 10/12/94 Dept: IC Approve By: WAP0QQF PAULSEN WA Date: 10/12/94 Task Determination Data: IST Required -N . NCR/CR : NA Safety Class: i PMT Required -Q Group
- Y PCM -
109-294 : B RWP Required : Y RWP No : Assign To : r4/ 10 CFR 50.49 : N EQ Doc Pkg : N Est M/H : 16.00 Reg Guide 1.97 : N Seismic Cat : I Crew Qty : 2 ASME XI(ISI) Reqd : N Scaffold Req . N Insul Rem : N Security Clearance: N Fire Prot Reg: N
- learance Required: Clearance No .
~
QC Requirements:
,. _ ,~, _ , , , -
QC Requ' red - Y l 6 u 4.4 su _,./& vv& 4 44im Mv4 *VA asu y V A 4tuu
.................................. ......... <!?.!'................" * ...
Work Order Task
Description:
Standar : 1. l l I SEE PAGE 2 FOR TASK DESCRIPTION. More: Y Planned By : WAP000F PAULSEN WA Date: 10/12/94 Pkg Appr By . WAP000F PAULSEN WA Date: 10/12/94 Time: 09:44 QC Approval : DSM00FK MELODY DS Date: 10/12/94
******************* ** OPERATIONS APPROVAL TO START ***********************
NPS Start Permission: Start Date/ Time : ill i / qq kS h / ov34 LCO(Y/N): Y
- t
- NPS Completion Notif: [ /_ J V k' _ /_ Major Failure:
Compl. Date/ Time: /F//3/ W ,ll /g /i .W r/r Major Action . Deficiency Tag Removed ( '/NS : ' /cV
_ _ . . _ _ . . . , _ _ _ _._ ._.~ _. -- . . . . _ _ . . _ - _ _ . _ . _ _ . - . _ I
~
Component: Sys: 07 Train: Fac: PSL Unit: 02 Associate: Assign Priority: C3 MASTER Name: PCM-109-294 SETPOINT CHANGE TO Work Type: 3 WORK ORDER TASK
- HYDRAZINE LOW LEVEL ALARM Location: RAB/RTGB-206 LMD: 2 94025805 01 Defect / Request: SETPOINT CHANGE ER/PWO: 64 / 3600
, Chg loc: 910 j PAGE 2 of 5 Continuation of Task
Description:
/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/4/ \/
- 1. VERIFY THAT YOU ARE WORKING ON kHE CORRECT UNIT -
--------- - -r - -
~ AND COMPONENT (S) : SIGNATURE /DATE
/\/\/\/\/\/\/\/\/\/5/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\/\
NPWO WORK DESCRIPTION FOR PCM-109-294 IS AS FOLLOWS: b.. NOTIFY OPERATIONS THAT A LOOP CALIBRATION OF THE HYDRAZINE STORAGE TANK LEVEL INDICATION WILLL PROCEED. 2.OBTAIN A TEMPORARY CHANGE AND / OR PROCEDURE CHANGE REQUEST FOR THE NEW LOW LEVEL ALARM SETPOINT OF 36.7 I H VERIFIED BY : o., DATE: g /43 g
- 3. ADJUST THE ALARM BISTABLE FOR NEW S .0F 3'.~ JNCHES.
VERIIFIED BY : N//:. DATE:
, '-As/n[
- 4. PERFORM LOOP CALIBRATION PER I&C P OCEDURE 2-1400064L," INSTALLED PLANT INSTRUMENTATION CALIBRATION (LEVEL) PER THE QI-3-1 TESTING DOCUM VERIFIED B DATE: / !
SYSTEM SUPVISOR T. NEWHOUSE l i j i e
_ . . - . _ _ _ _ _ - _ . _ . _ . _ _ _ _ . . . . _ . . . . - - ~ . _ . . __ . ...__ _ ._ _ - __ . _ _
~ Component: Sys: 07 Train: Fac: PSL Unit: 02 Associate: Assign Priority: C3 MASTER i Name: PCM-109-294 SETPOINT CHANGE TO Work Type: 3 WORK ORDER TASK HYDRAZINE LOW LEVEL ALARM Location: RAB/RTGB-206 LMD: 2 94025805 01 Defect / Request: SETPOINT CHANGE ER/PWO: 64 / 3600 '
Chg loc: 910 ; PAGE 3 of 5 , I JOURNEYMANS WORK REPORT i Actual Start Date: Time: Actual Completion Date: Time: Y
/ / O YOO Note: Journeyman shall sign and date text after their entries.
/Rf/3hf ff30 l
^
Trouble Found: This Section is NOT Applicable for PMs or other planned jobs IEST EQUIP. FSL d? W PSL stW F51. /72- . nni FDL DCl l Wor)c Perf ormed: f '" ' SW y ffao& h.f //7(ff~ d
^3 7Y f/
l l 1 l l l Continued on Additional Sheets: Y N t-Suggestions For Future Planning / Variance Reason:
- Supv/ Foreman / Chief Date Supervi or Date 7CQspector Date
.h b .
Aue 12 lR V ff9f m - . l I
e , ATTACHMENT FAC. W/O# UNIT ER PWO# LOC. , PSL 94025805 02 64 3600 910 EQUIPMENT #: LIS-07 9 1 JOURNEYMAN'S WORK REPORT ACTUAL START DATEmME: COMPLETION DATEmME: 12/01/94 0400 12/k794 1530 ) TROUBLE FOUND: NONE, PCM - i WORK / REPAIRS PERFORMED: OBTAINED PERMISSION FROM OPS. SUBMI'ITED TC # 2-94 337 FOR APPROVAL BUT COULD NOT OBTAIN PERMISSION DUE FRG,BEIN REQUIRED. INFORMED SUPERVISOR OF RESULTS. j G.SHACKETT 12/01/94 y OBTAINED APPROVED TC AND NOTIFIED OPS OF START. PERFORMED SETPOINT CHANGE PER PCM-109-294 SAT IAW NPWO TASK DESCRIPTION. PMT SAT PE f.^)tF. LEASED TO OPS. INFORMED SUPERVISOR OF RESULTS. G.SHACKETT Q[l-3-PR/P
/r
//
~/.
12/13/94
'7 SUGGESTIONS FOR FUTURE PLANNING / VARIANCE REASON:
TEST EQUIPMENT TYPE PSL# PSL # PSL # KEITHLEY 881 HEISE 861 TORQUE WRENCH 793 PART # / NAME M&S# SERIAL REMOVED LNSTALLED SUPERVISOR DATE D. w s2 14-94
~
QI 3-PR/PSL-1 Revision 30 February, 1994 ATTACHMENT 4 PC/M TESTING DOCUMENT l l PC/M Number 109-294 Supplement Number 0 , Plant / Unit PSL/2 NPWO Number 23GDOC) l l 1.
Title:
SETPOINT CHANGE TO THE HYDRAZINE LOW LEVEL ALARM (LIS-07-9) 2.
Purpose:
j THE PURPOSE OF THIS PCM IS CHANGE THE SETPOINT FOR THE l EYDRAZINE LOW LEVEL ALARM. THE PURPOSE OF THIS QI-3 IS TO l PROVIDE INSTRUCTIONS FOR TESTING MODIFICATIONS PERFORMED BY THIS PCM. l
- 3.
References:
4 1 3.1 PCM 109-294 I 3.2 I&C PROCEDURE 2-1400064L, INSTALLED PLANT INSTRUMENTATION CALIBRATION (LEVEL),
- 4. Prerequisites:
NONE
- 5. Instrumentation:
5.1 MILLIAMMETER - TRANSKATION 1040 OR EQUIVALENT. 5.2 VOLTMETER - KEITHLEY 197, FLUKE 8600 OR EQUIVALENT.
- 6. Related System Status:
l l NONE 1
l OI 3-PR/PSL-1 I Revision 30 ) February, 1994 l ATTACHMENT 4 PC/M TESTING DOCUMENT PC/M Number 109-294 Supplement Number 0 Plant / Unit PSL/2 NPWO Number MOD l I
- 7. Special Materials or Equipment:
NONE i l , 8. Temporary Connections: l l NONE
- 9. Precautions and Limitations: '
USE EXTREME CAUTION WHEN CONNECTING AND DISCONNECTING , ENERGIZED CIRCUITS.
- 10. Records Required to be included in the PC/M Package:
4 4 THE COMPLETED PLANT WORK ORDER AND THIS TEST DOCUMENT SHOULD BE INCLUDED WITH THE PCM.
- 11. Acceptance Criteria:
l Attach additional sheets if necessary. ' THE HYDRAZINE STORAGE TANK LEVEL ALARM (S-10) SHALL ALARM AT 36.7" +/- 0.4". l i r a b . 4 f 2 1
QI 3-PR/PSL-1 Revision 30 February, 1994 ATTACHMENT 4 PC/M TESTING DOCUMENT PC/M Number 109-294 Supplement Number 0 Plant / Unit PSL/2 NPWO Number ~3 CUO
- 11. Acceptance Criteria: (continued)
- 12. Detailed Procedure:
Steps meeting the acceptance criterir must be identified as such and contain verification signatures and dates. Attach additional sheets if necessary. PERFORM A L-07-9 LOOP CALIBRATION PER ERE 3.2. Verified by _f '
" V' j l
3
i 1 . QI 3-PR/PSL-1 i Revision 30 February, 1994 ATTACHMENT 4 , PC/M TESTING DOCUMENT ' PC/M Number 109-294 Supplement Number 0 Plant / Unit PSL/2 NPWO Number j CO d
- 12. Detailed Procedure: (continued) i I
5 4 a 4 7 s i a 4 Test Form Prepared By: i Mike Altermatt Date 10_/12_/94_ Engineer / Supervisor I 1 4 ; i h
. ._ .m. . . . . _ _ . .-_..__._-.._ _ ._ _ . . . _ . - _ _ _ . _ _ _ - - . . _ _ _ _ _ . . . _ . _ _ _ . _ . . _ .
E QI 3-PR/PSL-1 Revision 30 February, 1994 ATTACHMENT 4 PC/M TESTING DOCUMENT PC/M Number 109-294 Supplement Number 0 Plant / Unit PSL/2 NPWO Number _36 90
- 13. Testing Documentation:
(Record all pertinent testing information either here or on I the Journeyman's Work Report). l l i 1 i j Date /d/J /M Test Conducted by:' p ourn yman/ Supervisor Test Reviewed by: U. p Date /t./I4 / W En ineer/ Supervisor
=
( Test Approved by: ,
~
Date l2 / /f-/N l , Depar% ment Head i i o 5 i
Revision 58
. Octobar,1994 Page 70 of 76 FIGURE 4 TEMPORARY CHANGE REQUEST (Page 1 of 2)
A Reference Information: (Originator to complete) St. Lucie Unit # 0 1. TC # 2- 9k33/ Procedure
Title:
h.aco A,, menom.av:.a 04rin .v:. , h ead Procedure Number: 2 %eotNL Rev. 27 Steps affected: .:rav p,ic Con L.o?.4 (#yn n e:, sr....c m a teuh Reason for change: Scm tao -2 ev (araan ser-o:ur1.toI.2 l
\
Originator:(leev A. e4 >4-5 < Department Hsad or DesigneeN MA- - Phone: >< 2e r o Date: n// /W B Procedural Controis: (Originator to complete) Yes No O E is this Temporary Change for a one-time use? (if no, the responsible Dept. Head or designee shall ensure a procedure change request incorporating this TC has been
, developed, appr.oved and submitted for FRG approval.) This TC may be used up to 90 days.) ,
~-
O DI is this T.C. for a Q.l.? (if yes, written concurrence shall be obtained from the Quality Manager ant written approval shall be obtained from the Dept. Head who is jurisdictionally responsible for the O.l.) C Temocrary Chanoe Contenn: (Originator to complete) Does this Change: Yes No O E Alter the intent of the procedure? O E . Incorporate complex or extensive changes?
@' O Modify instrument setpoints?
O E Delete an independent verification? O E Alter a OC holdpoint? O E Modify a procedural step which alters a regulatory requirement as identified in the procedure? O E Alter the first execution of a procedure? (Preop, LOI) O E Addition of any chemicals? NOTE If any of the above criteria are marked yes, prior FRG review is required. s t k e
.].s.
- \
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.=. -- __ - .-. -
Hovision DB Octobar,1994 Paga 71 of 76 4 FIGURE 4 TEMPORARY CHANGE REQUEST (Page 2 of 2) D Does this change: (NPS to complete) - Yes No O ( Compromise the separation of redundant trains of equipment? O [B' Potentially isolate pressure reliefs? O Y Defeat automatic signals? O [ Defeat mechanical or electrical interlocks? NOTE If any cf the above enteria are marked yes, discuss possible attematives with the originator. . 10 CFR 50.59 Screening Yes No
- 1. Does the change represent a change to the facility as described in the SAR? Y
- 2. Does the change represent a change to procedures as described in the SAR7 X 4
- 3. Is the change associated with a test or experiment not described in the SAR? X
' 4. Could the change affect nuclear safety in a way not previously evaldated in -
- l the SAR?
X
- 5. Does the change require a change to the Technical Specifications? X NOTE If the answer to ALL the above 10 CFR 50.59 screening questions are no, (Questions 1 5),
then a safety evaluation is not required. STA review bbb b1 M IL / I /91/ ' STA Sig'n}fyre Date E Review: (This change shall have prior appthfal by a NPS and one member of the plant management staff.) NPS Signature [ Authorization Date _ /2 j AjW Plant Management Staff Signature C.kuw Date 11/ ? /d Plant General Manager Approva //_)S_/b, ~R 6/.-$N#1bN. Date TA / 6 ]'74
/R58 FRG Number M- s4 This change shall be reviewed (if prior FRG review is not
[MosU required) by the Facility Review Group within 14 days of the authorization date. F Canc'ellation Authorization (NPS/ANPS) Date / / Reason: e
s -- . ST. LUCIE UNIT 2 1 & C PROCEDURE NO. 2-1400064L, REVISION 27 INSTALLED PLANT INSTRUMENTATION CAllBRATION (LEVEL) L-07-9 System: Containment Spray Test Equipment Tag No-Loop No. Service: Hydrarine Storeae Tank Level Data Started: PSL-2 FJt-92-024 Date Completed: Work Performed by: Independent Verificaton: Date Supervisor-Irvut Test Point 2998-0-227 (13) MA Test Point HIGB 208 Output of: SH 1 Total Output of: LIS47 9 Totat MADC LT47-9 Tot. "H,0 Tot Var. Accuracy W .25 i .04 MA hecy W t.0 g ,4 gn. Pt. % *H,0 Destred As Found Asleft Error Destred As Fourd As Left Enor 1 0 .8 4.00 0 2 25 11.3 8.00 10.5 3 50 21.8 12.00 21.0 4 F5 323 18.00 31.5 5 100 42.8 20.00 42 4 75 32.3 18.00 31.5 3 50 21.8 12.00 21.0 2 25 11.3 8 00 10.5 1 0 .8 4.00 0 e Distable f unction Armunciator RTGB-208 Comments: Action liequked. CWD-306 Total Verfy Annunciator Windows S-10 Lo-Alarm
,ff 7
- 35 1.7 '
Tot. i .4" 351/2" = 17.53 me-f.,4.76 man
~{! [ 'k - J 3 Desired As Found As Left CWD-288 Ir5 y34((y, lieset /
Action itequired: E 20 lo'ai Lo-Lo Alarm . 2* Tot. i .4 % Desired As Found As Left ,_ _ ,,, , 2- {,;s.)eg t,t ' E%;;,yg 3 3 l'b . , liesst s * ,8 % '.
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Inter-Office Correspondence a To: All Plant Personnel Date: October 10,1995 From: D. A. Sager Depanment: St. Lucie Plant
Subject:
/ < Vice President St Lucie Plant Ore _ani tion and Staffine Recently the plant has made a number of organizational changes. These changes have been made to produce a strengthening of our technical, maintenance, and operational capabilities, as well as to achieve greater productivity and efficiency.
As a result of this effon, several plant staff positions can and will be eliminated. During the past several months, we have experienced a number of difficulties associated with the forced outage of Unit 1. The analyses that have been conducted of these events have identified the underlying root causes to be an acceptance of long-standing equipment deficiencies, the experience level of some staff positions, and the usage of plant procedures. We are implementing corrective actions to address each of these areas. However, in no case was the number of people at this plant an issue. In order to meet our long term strategic goal of becoming a more cost effective producer of electricity, a reduction in the number of staff positions at St. Lucie Plant will occur between now and February,1996. This will involve both bargaining and non-bargaining positions and will amount to a reduction of about 5% of our authorized staffing level. l This is a difficult and stressful time for everyone, but your continued dedication to your assigned responsibilities during this period is both expected and appreciated. Our constant l emphasis is and will be the safe operation of our plants. This will not be compromised. l l DASkw DAS.PSL =155 05
/
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Inter-Office Correspondence I ME JQQ-95-143 To: J. H. Goldberg . Date: August 22,1995 From: J. T. Voorhe s, Department: JNA/PSL
Subject:
Review of P? ops:: Decisions Leading toJhe Copies to: D. Sager Inadvertent Spraydown of the PSL-1 Containment C. Burton J. Geiger W. Bladow Executive Summars The subject review examined two processes and the accompariylag decisions that contributed to the inadvertent spraydown of the PSL-1 containment on 8/I?/95. The processes examined were;
- 1. The decision to operate PSL-1 with FCV 07-1 A in the open position.
- 2. Development and execution of new procedure OP l-0420060
" Venting of the Emergency Core Cooling and Containment Spray Systems."
The decision to operate Unit-1 with FCV 07-1 A in the open position was made by Operations management in consideration of a potential four day extension of the PSL-1 SNO outage. The FRG approval of the engineering safety evaluation to allow operations with FCV 07-1 A in the open position, the fact that the same condition has been in existence on Unit-2 since May 1994 and the strained resources within the mechanical maintenance organization eliminated any serious consideration of repairing the FCV. The new ECCS venting procedure did not adequately consider the configuration of the contauunent spray system. The review of this procedure did not sufficiently address specific attributes called out in site Quality Instructions governing review of new procedures. Although allowed by site policy and typical of many operations evolutions, the performance of procedural steps out of sequence allowed the return of the containment spray systera to its normal operating configuration (MV 07-3A open) print to venting of the ECCS system. This r-moved the final barrier to spraying down the containment. A contributing factor to this event was the perceived pressure to expedite the return of Unit-1 to service. In addition. a review was conducted of the historical maintenance and surveillance test results of FCV 0.7-1 A. and its equivalent valves, on both units. The review indicates that while these valves provide periodic operational problems. long periods of maintenance free performance have been achieved. The availability and cost of spare parts for these valves is an area of concern. I It should be noted that this investigation was focused only on events surrounding the PSL-1 containment spray event. Within this context. specific recommendations are provided for plant k ; l consideration. Should information beyond that contained in the report be required, please contact , \ me. r
/
)
i D : Review of Process Decisions l Leading to inadvertent Spraydown of PSL-1 Containment
SUMMARY
OF EVENT: At 1756 Hrs. on 8/17/95, during progression towards a unit start-up, the Unit-1 containment was inadvertently sprayed with approximately 10,300 gallons of refueling tank water. This occurred when the l A LPSI pump was started to facilitate venting of the LPSI header in accordance with e a new operating procedure for venting of ECCS systems. The new procedure required that valve V3452 a cross connect between the LPSI and the containment spray systems, be repositioned open. Operation of V3452 occurred under circumstances where FCV 07-1 A a normally closed contaimnent spray header isolation valve had been placed in the open position in accordance with an approved safety evaluation. The configuration created by the two open valves provided a flow path from the discharge of the runnmg l A LPSI pump to the containment spray ring header (Figure 1). i DETAILS: Two separate processes, both occurring in parallel, led to the 'spraydown of the PSL-1 containment. The first of the two, was the process that led to the decision to operate PSL-1 with FCV 07-1 A in the open position. The second was the process that led to the development and i execution of a new operations procedure, OP l-0420060, " Venting of the Emergency Core l Cooling and Containment Spray Systems". Figure 2 illustrates the two processes, including l relevant decision points; The Decision to Ooerate PSL-1 With FCV 07-1 A in the Ooen Position i Four identical air operated isolation valves exist in the containment spray headers of PSL Units 1 and 2. These valves are required to stroke open upon a containment spray actuation signal to provide a containment spray flow path during a loss of coolant accident. The valves fail open upon loss of air or power. On 8/l1/95 FCV 07-1 A failed a normally scheduled stroke time test. The valve stroked open in 20 seconds in contrast with the required value of 8 seconds. This time was substantially longer than that of previous surveillance tests. Operations declared the valve out of service. The mechanical maintenance engineering support group (MMEG) was tasked with the determination of the required corrective measures. The four valves in this category have a recent history of performance problems. Unit 2 FCV 07-IB is entrently out of service in the open p6sition in accordance with a FRG approved engineering safety evaluation. MMEG personnel stated that little information can be gained from external troubleshooting of these valves. It was felt at the time that valve disassembly would be required to restore the valve to an operable condition. FCV 07-1 A had been disassembled and rebuilt in November 1994 dming the last Unit I refueling outage. Vendor assistance had been utilized at that time. MMEG personnel felt that this assistance would again be required and that
i I . l
. l - - - - the 90 hour-duration of the previous rebuild tasiewouldegain bemecessary-His information was ---
forwarded to outage and plant management for an evaluation of schedule impact. Concurrently, MMEG personnel were aware that a safety evaluation was in existence authorizing operation of Unit-2 with the equivalent valve in the open position. As a contingency alternative to repair of the valve, a PSL STAR was initiated by Maintenance which recommended that
)
Engineering provide justification for operation of Unit-1 with FCV 07-1A in the failed open l position. Utilizing the Unit-2 safety evaluation as a basis, the Unit-1 safety evaluation was I quicidy generated and subsequently FRG approved on 8/12/95. The Unit-1 safety evaluation l provides precautionary statements regarding operation, of the containment spray and LPSI systems with FCV 07-1 A in the open position. These precautions were translated into an Operations Night Order on 8/16/95. Caution tags were hung on the containment spray pump and FCV 07-1A control switches in the control room. Develooment and Execution of Procedure OP l-0420060 "Ventine of the Emercency Core Cooline and Containment Sorav Systems" On 2/21/95 the 2A LPSI pump was determined to be inoperable during an ASME Code pump run. Subsequent root cause analysis determined the pump had been air bound. As a result, specific venting procedures were developed for each unit to preclude the possibility of future ECCS air binding. The subject Unit-1 procedure was authored by the technical staff system engineer and submitted to operations for technical review on 7/20/95. No specific priority was assigned to this procedure. During the recent Unit-1 SNO outage difliculties with LPSI system relief valves suggested the possibility of air in the systems. Review of this procedure was then expedited to enable implementation of the procedure during Unit-1 start-up. The Technical subcommittee review and FRG approval were completed on 8/13/95. Temporary changes to operating procedures were issued to sequence ponions of the venting procedure into the operating procedure for the plant stan-up evolution. The inserted procedure ponions were sequenced so that the LPSI headers would be vented immediately following termination of shutdown cooling. The portion of the procedure for LPSI venting causes the 1A LPSI pump to be started and opens valve V3452 providing a flow path to the containment spray system (see Figure 1). This procedure ponion does not specifically address the configuration of the containment spray system during the LPSI venting process. Under normal plant start up conditions, both MV 07-3 A and FCV 07-1 A would be closed at this point. The containment spray system was sequenced to be vented at a later point in the stan-up procedure. During the recent Unit 1 stan-up, problems with LPSI system relief valve replacement delayed venting of the LPSI headers. The containment spray portion of the venting procedure was executed prior to venting of the PS! system. Upon completion of containment spray venting, procedural requirements restore MV 07-3A to the open position required for power operation. I With FCV 07-1 A in the failed open position no isolation existed to prevent flow from reaching the containment spray ring header upon staning the 1 A LPSI pump and opening valve V3452 during subsequent LPSI venting. 2 , r-
-Discussion of Decision Points- - --- -- - - - - - - -
----)
Three decision points within the above processes warrant discussion (see Figure 2). These points l are significant because had altemative decisions been made at these junctures a spray event I would not have occurred.
- 1. The Decision to Ooerate with FCV 07-1 A in the Onen Position Discussion with plant management indica 2cs that at the time of FRG approval of the safety
__ evaluation allowing operation of FCV 07-1A in th,e open position, this course of action was viewed as a contingency plan only, h was assumed that repair options would continue to be pursued. Discussion with maintenance indicates that resources did not exist to seriously pursue this option. Maintenance resources at that point in time were devoted to resolution of PORV and relief valve issues. Maintenance personnel had experienced difficulty in determining the causes of past FCV failures on both units and were reluctant to engage in this effort. A four day schedule impact was estimated to be necessary to rework the valve. No documentation or information was provided during this review to indicate that any serious consideration was given to repairing FCV 07-1 A after the FRG approval of the safety evaluation was completed. Outage managemem has the responsibility for soliciting viable options for emergent work and determining their schedule impact. Operations has the final decision for determining the acceptability of plant operation with equipment inoperable or in a marginal condition. Management of both these organizations was involved in the FCV 07-1 A decision. Discussions i with plant management indicated that, although not the case in this event, this decision making l process is informal, and creates the potential for decisions to be made at lower levels in the l' organization than may be desirable. NP-910 " Plant Readiness for Operations" requires that the Site Vice President personally review and approve the decision to restart a unit containing any one of a number of listed deficiencies. This policy is specifically targeted at systems important to reliability. The equivalent process for management review and approval of the desirability of restarting a unit with safety related equipment out of service is conducted prior to entry into Mode 4. Review at this stage may not provide adequate perspective for an objective evaluation of attematives to operating with safety related equipment in a degraded condition. The maintenance and testing history of Unit-1 FCV 07-1 A is summarized in Attachment 1. The other three valves of this type at PSL share similar histories. Discussion with maintenance ' i personnel revealed the perception that a design issue or misapplication exists with these valves. In this connection, STAR 94110432 was submitted in November 1994 to request a feasibility study and cost alternatives for engineering options including replacement of FCV 07-1 A. It is not clear from the maintenance and surveillance test data that replacement of this valve is warranted. This valve has exhibited long periods of maintenance free operation in addition to periodic problems. No systematic, structured root cause analysis has been performed on the valve. 3 i
1-I. 2c-Technical Subcommittee and FRG -Anprovalrf New-OneratingProcedure OP 1-0420060 - - The referenced procedure was developed to provide systematic venting of the ECCS and containment spray systems. Discussion with personnel involved with development of this procedure indicated that the procedure was developed under the assumption that FCV 07-1 A would be in the closed position. QI 5-PR/PSL-1 Appendix C provides guidelines for the review of new (initial issue) procedures by the author and the technical subcommittee. Attributes to be considered include the following ;... __ 2. Determination of INFREQUENT TES
- 4. Cautions properly entered in ninstructio,T section OR EVOLUTION applicability
- 8. Initial conditions specified and compatible with other procedures The review of this procedure was inadequate in that it failed to adequately specify or provide cautions regarding the assumed configuration of the containment spray header even though the procedure opened an isolation valve to this system with a LPSI pump running. l PSL procedure AP 0010020 " Conduct ofInfrequently Performed Evolutions at St. Lucie Plant" specifies that evolutions which are seldom performed even though covered by existing plant procedures be conducted under specific management controls. These. controls were not imposed by the Technical Subcommittee or FRG on the initial conduct of the ECCS vent procedure.
- 3. Performance of OP 1 0030121 Plant Heat-un crocedure out of secuence The above listed procedure provides instructions for the heat-up of the plant from cold to hot standby through specific instructions contained within this procedure and through reference to other procedures. Discussion with operations and management personnel indicates that plant policy is to allow procedural steps to be conducted out of sequence unless performance in order is specifically called for in the procedure. Complex evolutions such as a plant heat-up are felt to require this latitude. The plant procedure entitled " Conduct of Operations" does not discuss or provide guidance on the performance of procedural steps or evolutions in other than the numbered sequence.
Temporary changes to OP l-0030121 were issued ind inserted in appropriate locations to direct the conduct of ECCS and containment spray venting in accordance with OP l-0420060. Discussion with Operations personnel who originated the temporary changes indicated that the intent was to perform LPSI venting prior to return of the containment spray system to normal operating configuration. Discussion with Operations personnel indicated that there was a significant amount of perceived pressure to expedite the Unit-1 start-up. Operations super ision was in the field overseeing the venting evolution to expedite plant start-up. An additional TC was being processed to eliminate the venting evolution on the B LPSI header due to the small amount of air being found on the A side for the same reason. Operations personnel stated that the frequent requests for updates on 4 a
I
\
4 Discussion with~ Operations personnel ~ indicated ~that~therewara s~ignificant amount of"perc~eived- ~' ~'~ pressure to expedite the Unit-1 start-up. Operations supervision was in the field overseeing the i venting evolution to expedite plant start-up. An additional TC was being processed to eliminate the venting evolution on the B LPS! header due to the small amount of air being found on the A side for the same reason. Operations personnel stated that the frequent requests for updates on plant status , from outside the control room were a distraction from plant operation. The sequence of operational i and maintenance problems encountered since the shutdown for hurricane Erin and the well publicized situation regarding FPUs capability to meet system demand exacerbated these i perceptions. ' Recommendations: I The following recommendations are offered to the plant organization as a result of this review; I
- 1. Initiate a systematic root cause investigation into the performance problems of FCV 07-1 A &
IB on both units.
- 2. Reevaluate the decision making process and level of management approval necessary to allow restart with safety related components in a degraded condition. l
- 3. Revise AP 0010020 " Conduct ofInfrequently Perfonned Evolutions at St. Lucie Plant" to l require that consideration be given to executing new, first time use procedures under the l management controls provided within AP 0010020.
]
- 4. Review the adequacy of the Technical Subcommittee process for new and revised procedures.
- 5. Provide guidance within AP 0010120 " Conduct of Operations" regarding the execution of procedure steps or evolutions out of the numbered sequence.
- 6. Strengthen protocols for disseminating plant and cperations status information from the control room to non-operations personnel.
1 5
L-l Att chme:t 1 pk- ' d[ TESTING AND MAINTENANCE HISTORY N i j Containment Spray flow control valve FCV 07-1 A is normally closed and is designed to open I j under accident conditions in less than 10 seconds. The valve is included under the provisions l of the ASME Section XI Pump and Valve Testing program, and is tested quarterly to verify that stroke time meets an administrative limit of 8 seconds. If the administrative limit is exceeded, the valve is placed out of service until corrective action can be accomplished (ACTION ' ! condition). If the stroke time during a test varies from the stroke time rec'orded in the previous test by +/- 50%, the valve is place in alert end the, testing frequency is increased to monthly l 4 (ALERT condition). 1
- Records of valve testing and maintenance were examined for the period of May 1986 through I the present.
- Of 54 tests recorded for FCV 07-1 A during this period, two tests met the definition of an ALERT 4
condition (5/87,l/92) and three tests met the definition of an ACTION condition (8/88,l/93,8/95). i In four out of five of these instances (excluding the most recent test) the unsatisfactory stroke j times were resolved through additional lubrication and stroking. The mort recent test recorded j a stroke time of 20.3 seconds, more than double the longest stroke time that had previously been recorded. Following this test, provisions were made for unit operation with FCV 07-1 A in the open position. [ Lubrication periodicity for the valve has varied over the time period examined, from
~
approximately quarterly to slightly longer than annually. Based upon the data examined, there does not appear to be a correlation between frequency of lubrication and unsatisfactory test performance. The present surveillance test procedure calls for the valve to h- lubricated immediately prior to each performance of the surveillance. l Although required to be tested quarterly, discussion with Operations personnel indicated that there was a period during the mid-1980's when the valve was exercised daily. Significant corrective maintenance action was undertaken on four occasions during the interval examined. In two cases problems with the actuator were repaired (12/91,5/93) and in two cases ! valve intemals were repaired or replaced (11/91,11/94). The most recent maintenance activity detected scoring and galling on both the wedge and seats of the valve. The maintenance activity was hampered by the need to re-use certain 0-rings due to unavailability of the necessary spare parts in Stores. A review of the earlier history of the valve, based upon data provided by Mechanical Maintenance, showed fairly concentrated maintenance activity during the period of February 1976 , through July 1983, including observed galling of the valve seat and disk. Following repair of the valve in July 1983, a period with little required maintenance resulted for the next 8 years. Based upon the records examined. a systematic investigation of the root cause for the valve failure appears necessary An area of concern is the availability and cost of required spare parts.
l.- i l FIGURE 1 PSL1 LPSI - _
. 3 ,
CONTAINMENT SPRAY a , 1 i SYSTEM l a l i r
- V07160 FCV 07-1 A MV 07-3A LT2 L 2 L 2 r, r, r,
! 1 A SDC HX 1 1 V3452 a RWT 1 A LPSI PMP
FIGURE 2 i CONTAINMENT SPRAY DECISION SEQUENCE l 2/21/95 - -- --- - ---- - - - - - -
-- (LP8 N -
i i aces wwT Paoc. 7/20/95 i oP14 o00eo Dewtoreo
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8/11/95 sinom'Tma Test 1 P a l us y ,, 8/11/95 "" '
" ouTAoE EXTEN9oN 1 -
V ' l FRG APPROVES 8/12/95 coun m Na
- s E To LEAVE FCV 071A OPEN l 3I gg UNSAT 8/13/95 accavuur Phoc.
1 SAT 3 rnG arvevs WSAT
- 8/13/95 6
- cs vuNT
>=oc.
V SAT i oes N6GHToRDEM 8/15/95 'O "E .' Fevo71AiNoPtN PosmoN t
- FCV 071A 4
+ 8/16/95 cAunoN TAGGED oPEN 4 l 1 4 l YES W 074A 'N P"0C-8/17/95 a osgo q seousw:s a mar. up >=oc. i i . V START LPSI Ptr. Bli7/95 ceEN vus2 w7u w074A oPEN i 1 SPRAh
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' 1 y ., (* v 5 y d E' inter-Office Correspondence A>: n p -. , \ WL Ga , L.L a 8 To: Paul Fulford - oste: October 5,1995 3 j% 44e. From: Joh Hallem oepanment: Ops Test
Subject:
STAR 951063 Response I have reviewed the test and su:TeiEance p xedures pertaining to the inservice testing of ) pumps and valves for both Units 1 and 2. I found no further instances where preventive I maintenance is specided by the procedu e w N performed prior to the test or smveillance.
)
1 The plant does routinely perform PMs prior e scheduled quanerly surveillances. His is done i as a matter of conveniences since a su:Td".2nce run is required following the PM. By l scheduling the PM wurk to be performed p:ix to the scheduled surveillance the number of l surveillances performed is reduced. This is desirable because this limits the number of demands placed on the pumps, tk amoun: of operator manpower needed to support the tests, and reduces the unautilabihy for pumps which are taken out of senice to perform the l surveillance. Attached is a hsung of the PMs performed prior to scheduled surveillances. These PMs are in two major categories: pcmps and fans, and valves and AFW Teny Turbine. l
) I have reviewed these PMs with the applicaNe SCE and Predictive Maintenance Engineers 9 ,fQ and have collectively come to the following two conclusions.
N : For the first category. the PMs are for pump oil change out, coupling lubrication and fan lubrication. These are considered to have a minor, if any, impact on the performance of the pumps and fans during the surveillances. Tnese PMs are performed less frequently than the quarterly surveillances with no indication that the performance or lack of the activities has influenced any surveillance. Also. some cf the components are run during normal plant operations, and problems with Charpng pu p accumulator pressure and other pumps and fans are detected by Operatiocs dunng no .i observation of the components. Therefore this first category of PMs should continue to be .erformed prior to scheduled sun'elllances. The PMs performed on FCV-07-1 A. FCV-07-1B, and the AFW Terry Turbine, and Governor valves may have an impact on the performance of these components. De tests performed on the Unit 2 MFIVs hase no efect of the safety related function of the MFIV surveillance. Ris category ofPMs wi ie exception of the MFIVs should be rescheduled from just prior to the performance cf the s: .eillances associated with these components. C
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D' UNIT ONE M/M PMS REQUIRING SURVEILLANCE RUN AFTER PM IS COMPLETE g .........................******...**............**............... 1 I PM # EQUIPMENT PM TASM PROCEDURE # . ~~~~~~~~~ ~~~~~~~ ~~~~~~~~~~~ 190 AFW PP 1A CHANGE OIL 1-M-0018P . 191 AFW PP 1A COUPLING LUBE 1-M-0018P 192 AFW PP 1B CHANGE OIL 1-M-0018P : 193 AFW PP 1B COUPLING LUBE 1-M-0018P i 194 AFW PP 1C CHANGE OIL 1-M-0018P i 195 AFW PP 1C COUPLING LUBE 1-M-0018P j 242 B.A. MKUP PP 1A CHANGE OIL 1-M-0018P i i 243 B.A. MKUP PP 1B CHANGE OIL 1-M-0018P 044 CCW PP 1A CHANGE OIL 1-M-0018P i 045 CCW PP 1A COUPLING LUBE 1-M-UUlbP 046 CCW PP 1B CHANGE OIL 1-M-0018P 047 CCW PP 1B COUPLING LUBE l-M-0018P ] 048 CCW PP 1C CHANGE OIL 1-M-0018P *
- 049 CCW PP 1C COUPLING LUBE 1-M-0018P l 034 CHG PP 1A CK ACCUM. PRESS. 1-M-0018
- 041- CHG PP 1B CK ACCUM. PRESS. 1-M-0018 043 CMG PP 1C CK ACCUM. PRESS. 1-M-0018 035 CHG PP 1A CHANGE OIL 1-M-0018P 036 CHG PP 1A CPLG. LUBE 1-M-0018P 037 CHG PP 1B CHANGE OIL 1-M-0018P 1 038 CHG PP 1B CPLG. LUBE 1-M-0018P i
! 039 CHG PP 1C CHANGE OIL 1-M-0018P j
- '040 CHG PP 1C CPLG LUBE .-M-0018P 1
)
! 377 FCV-07-1A LUBE 1-M-0018C -- PCR s' I , 378 FCV-07-1B LUBE 1-M-0 018C -- Pc( g/ 272 FIRE PP 1A 1-M-0018F LUBE ' l 273 FIRE PP 1B LUBE 1-M-0018F i 266 FIRE PP 1B CPLG. LUBE 1-M-0018F i 269 FIRE PP 1A CPLG. LUBE 1-M-0018F
- 136 HPSI PP 1A CHANGE OIL 1-M-0018P ;
i 137 HPSI PP 1A CPLG. LUBE 1-H-0018P 138 HPSI PP 1B CHG. OIL 1-M-0016P i 139 HPSI PP 1B CPLG. LUBE 1-M-0018P i 302 HVE-13A LUBE 1-M-0018 303 HVE-13B LUBE 1-M-0018 , 314 HVE-16A LUBE 1-M-0018 L 315 HVE-16B LUBE 1-M-0018 4 330 HVE-6A LUBE 1-M-0018 331 HVE-6B LUBE 1-M-0018 i 294 HVE-9A LUBE 1-M-0018 . 295 HVE-9B LUBE 1-M-0018 142 LPSI PP 1A CPLG. LUBE 1-M-0018P i 143 LPSI PP 1A CHG. OIL 1-M-0018P , l 144 LPSI PP IB CHG. OIL 1-M-0018P
-145 LPSI PP 1B CPLG. LUBE 1-M"0018P l~ ,
/
206 TERRY TURBINE GOV VLV & LINKAGE 1-M-0018 - PC A i 202 TERRY TURBINE CHG. OIL 1-M-0018 - e ce y-2 203 TERRY TURBINE THROTTLE VLV LINKAGE 1-M-0018 - ect ge 4 t
LOGU295.XLS s llll k ki hD NI'b y ' .3'vb Vt* I '/C0Ct 2 1
.PM# COMPONENT TAG ' AESOCIATE . DESCRIPTION.. SYS 3 , Fire Doors inspection 15 4 Missile Shields for Drs. 259 & 273 lubrication / inspection 99 5 Steam Leak inspection 8 6 Washing Machines / Dryers lubrication 99 300 Svc Air Compr Ini Fitr Slner change filter media i 18 301 Svc Air Compr Compressor lubrication / inspection ! 18 302 Svc Air Compr Compressor replace valves i 18 303 Svc Air Compr Compressor rebuild valves i 18 401 lA Compr 2A Compressor lubrication / inspection 18 402 IA Compr 28 Compressor lubrication 18 4U3 lA Compr 2A Compressor change out valves 18 404 lA Compt 28 Compressor change out valves 18 405 IA Compt 2A Compressor rebuild valves 18 406 lA Compt 2C Compressor oil sample 18 407 IA Compr 2D Compressor oil sample 18 408 IA Compr 2A Inl Fitr Siner change filters 18 411 lA Compt 2B ini Fitr Slnct change filters 18 412 A lA Dryer 2A change desiccant 18 412 B IA Dryer 28 change desiccant 18 413 IA Compt 2C Compressor lubrication / inspection 18 414 lA Compr 2D Compressor lubrication / inspection i 18 415 IA Compr 2C Compressor inspect / clean 18 416 lA Compr 2D Air Fitr inspect / clean 18 417 lA Dryer 2A inspection 18 418 IA Dryer 2B inspection -
18 419 IA Compt 2C Compressor change out valves 18 420 lA Compr 2D Compressor change out valves 1. 18 421 1A Dryer change dessicant 18 422 IA Compr 2C/2D Compressor rebuild valves 18 501 SS 21-3A1 Strainer ir' ;,cction/ clean 21 502 SS-21 -3 A2 Strainer inspection / clean 21 503 SS-21-3B1 Strainer inspection / clean 21 504 SS 21-3B2 Strainer inspection / clean 21 505 Strnrs for Circ PPs Seal Wtr Sply lubrication 21 605 Recirc PP 2A Pump lubrication 13 1001 Trvlo Scrn 2A1 Screen lubrication / inspection 21 1002 Trvig Scrn 2A2 Screen lubrication / inspection 21 1003 Trvig Scrn 2B1 Screen lubrication / inspection 21 1004 Trvig Scrn 282 Screen lubrication / inspection 21 j 1005 Trvig Scrn 2A1 Screen lubrication 21 ! 1007 Trvig Scrn 2A1 Drv Rdcr lubrication 21 1008 Trvig Scrn 2A2 j Drv Rder lubrication 21 1 1009 Trvig Scrn 2B1 Drv Rder lubrication 21 1010 Trvig Scrn 2B2 Drv Rder lubrication l 21 1011 Intk Trh Rks Tyh A Hoist lubrication / inspection 74 1012 Intk Trh Rks Tyh B Hoist lubrication / inspection 1015 Trvig Scrn 2A1 74_ Screen inspection 21 1016 Trvig Scrn 2A2 Screen inspection 21 Page 1 1 I l
4 . LOGU295.XLS V 1017 Trvig Scrn 2B1 Screen inspection 21 1018 Trvig Scrn 2B2 Screen inspection 21 1201 A Cond PP 2A ' Pump rotate 12 1 1201 Cond PP 2C Pump rotate 12 1601 HCV-09-1 A Act leak test 9 1602 HCV-09-1 B Act leak test 9 1603 HCV-09-2A i Act leak test 9 2 1604. HCV 09 2B l Act leak test 9 1608 HCV-09 2A,B & 1 A,B Air Hydric PP change out pp/mtr assy 9 1609 MFIV Air Mtr/Hyd PP Assy , rebuild 9 1610 Sec Side WL PP 2A Pump lubrication 39
- 1611 Sec Side WL PP 2B Pump lubrication 39 1701 AFW PP 2A Pump lubrication / inspection 9 1702 AFW PP 2A Cpig lubrication / inspection 9 1703 AFW PP 2B Pump lubrication 9 d
1704 AFW PP 2B Pump lubrication cplo 9 1705 AFW PP 2C Pump lubrication 9 1706 AFW PP 2C Cpig Pump lubrication 9 1707 AFW PP 2C Turb change tube oil 9 ree / 1709 AFW PP 2C Gov linkage lubrication 9 PCC < 17JO AFW PP 2C ! Gov gov / trip viv settings 9 tc t- / 1801 Cond Xfr PP Pump
- lubrication 12 1802 Cond Xfr PP Pump lubrication epig 12 1901 CRS PP 2A Pump lubrication 28 1902 CRS PP 2A Pump lubrication epig 28 1903 i CRS PP 28 Pump lubrication 28 1904 CRS PP 2B Cpig Pump lubrication 28 2101 Cyclohex PP 2A i Pump lubrication 20 2102 Cyclohex PP 2B Pump lubrication 20 2103 CFS Hydrzn PP 2A i Pump lubrication 20 2104 CFS Hydrzn PP 2B Pump lubrication 20 2105 Chem Inj PP 2A Pump lubrication 20 2106 Chem Inj PP 2B Pump lubrication 20 2107 Chem inj PP 2C Pump lubrication 20 2201 Isol Phs Bus Air Clr A lubrication / inspection 53 2202 Isol Phs Bus Air Cir B lubrication / inspection 53 2203 Gen SI Oil Unit Unit Sample 22 2501 Cndsr Pit Sp PP 2A Pump lubrication 25 2502 Cndsr Pit Sp PP 2B Pump lubrication 25 2510 Oil Separating Boxes inspection 99 2511 Temp Oil / var Separating Box inspect / clean 99 2701 TLO Rsvr Tank sample 22 2704 Turb Gen Generator lubrication 22 2705 Brg Oil PP lubrication 22 2706
- Brg Oil Lft PP Pump lubrication epig 22 2707 TLO Rsvr Tank clean spare strnr 22 2708 Kaydon Conditioner change filters 22 2801 Gind Stm Cond Xfr PP Pump lubrication 12 2802 Gind Stm Cond Xfr PP Pump lubrication colo 12 Page 2
LOGU295.XLS l y ) 3401 Cis Bldn Clg PP 2A I Pump lubrication 34 l 3402 i Cls Bidn Cig PP 2B I Pump lubrication 34 3601 I Bam PP 2A Pump lubrication 2 ! T3602-l Bam PP 2A Pump lubrication epig - 2 3603 Bam PP 2B Pump lubrication 2 ) 3604: Bam PP 2B Pump lubricatit rolp ' 2 _3606. I Chg PP 2A Pump check accuri. .- + or i 2 f3607P{ Chg PP 2A Pump changelube.at i 2 3609 Chg PP 2B l Pump check accumulator 2 3610 l Chg PP 2B i Pump lubncation 2 3612 Chg PP 2C i Pump check accumulator 2 3613 l Chg PP 2C Pump lubrication 2 , l 3614 ' Mtrng PP , Pump change oil 2 3701 Hpsi PP 2A : Pump lubrication 3 3702 Hpsi PP 2A ' Pump lubrication cplo 3 3703 Hpsi PP 2B Pump lubrication 3 3704 Hpsi PP 2B Pump' lubrication epig 3 3705 Hpsi PP (spare) ! Pump lubrication / inspection 3 3901 IRS Hydrzn PP 2A - Pump lubrication 7 3902 IRS Hydrzn PP 2B -
, Pump lubrication 7
'3903
- FCV-071 A i Valve lubrication 7 PC4., ./
3904 i FCV-071 B Valve lubrication 7 rca /
"4001
- Laun Drn PP 2A . Pump lubrication 6 4002 - Laun Drn PP 2B ,
Pump lubrication 6 ) 4003 - Wst Cond PP 2A Pump lubrication . 6 4004 i Wst Cond PP 2B ! Pump lubrication 6 I' 4005 i Bac 2A Dist PP A i Pump lubrication 6 4006 ! Bac 2A Dist PP B ~i Pump lubrication 6 4007 Bac 2B Dist PP A 1 Pump lubrication 6 4008 - Bac 2B Dist PP B - Pump lubrication 6 4009 SS-6-9A ; clean strainer 6 4010 SS-6-9B clean strainer 6 4101 Wst Gas Compr 2A Compressor lubrication 6 4102 Wst Gas Compr 2B Compressor lubrication 6 4103 T6911 clean 6 4301 A&B Safeguards Rm Sp PP Pump lubrication 6 4304 SS-6-7 A clean strainer 6 4305 Equip Drn PP 2A ! Pump lubrication 6 4306 Equip Drn PP 2B Pump lubrication 6 4307 Equip Drn PP 2C . Pump lubrication 6 4311 Chem Drn PP 2A Pump lubrication 6 4312 SS-6-8 i clean strainer 6 4313 Laun Drn Sp PP ' Pump lubrication 6 4314 Chem Drn Sp PP ! Pump lubrication 6 4401 Rx Drn PP 2A Pump lubrication 6 4402 Rx Drn PP 2A Pump clean / lubricate epig 6 4403 Rx Drn PP 2B Pump change oil 6 4404 Rx Drn PP 2B Pump clean / lubricate cpIg 6 4409 H/U Drn PP 2A ' Pump lubrication 6 Page 3
1 l LOGU295.XLS e 4411 H/U Drn PP 2B l Pump lubrication ! 6 4413 H/U Recire PP 2A i Pump lubrication 6 4 4416 BA Cond PP 2A ' Pump lubrication 6 4418 BA Cond PP 2B Pump lubrication 6 f' 4420 BA Hldo PP 2A Pump lubrication I 6 4422 BA Hldo PP 2B Pump lubrication 6 4510 Pri Wtr PP 2A Pump lubrication . 15 4511 Pri Wtr PP 2A Pump inspect / lubricate epig I 15 4512 Pri Wtr PP 2B Pump lubrication 15 4513 Pri Wtr PP 2B Pump inspect / lubricate epig . 15 4601 OBHX 2A clean / inspect 34 4602 OBHX 2B clean / inspect 34 4801 CCW PP 2A ' Pump change lube oil 14 j 4802 CCW PP 2A i Pump coupling lubrication 14 l 4803 CCW PP 2B i Pump change lube oil 14 4804 CCW PP 2B Pump coupling lubrication 14 i I 4805 CCW PP 2C Pump change lube oil 14 4806 CCW PP 2C Pump coupling lubrication 14 4807 CCW HX 2A clean / inspect 14 4808 CCW HX 2B ' clean / inspect ' 14 5001 Fuel Pool PP 2A i Pump lubrication 4 5002 F}}