ML20062C189
| ML20062C189 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/24/1990 |
| From: | Silver H Office of Nuclear Reactor Regulation |
| To: | Beard P FLORIDA POWER CORP. |
| References | |
| TAC-68200, NUDOCS 9010290457 | |
| Download: ML20062C189 (18) | |
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October 24, 1990 , s% . . . . + / ; Docket No. 50-302 i Mr. Percy M. Beard, Jr. - Senior Vice President, . Nuclear Operations Florida Power Corporation -) ATTN: M nager, Nuclear Operations-Licensing P. O. Box 219-NA-21 ' Crystal River, Florida 32629 '
Dear Mr. Beard:
SUBJECT:
CRYSTAL RIVER UNIT 3 - SAFETY AND PERFORMANCE-IMPROVEMENT PROGRAM . IMPLEMENTATIONAUDIT(TAC'NO.68200)- I Enclosed is our evaluation report on the implementation of the Babcock & Wilcox , Owners Group's safety and Performance Improvement: Program (SPIP) at_ Crystal River, Unit 3(CR-3). This evaluation is' based on a staff audit.at the: Florida 1 Power Corporation (FPC) headquarters in St. Petersburg, Florida-and the CR-3 site in Crystal River, Florida during the week of June through 28 1990. The i audit was conducted with assistance of Idaho National Engineering, Laboratory ' i consultants. . The staff audit of SPIP implementation was conducted in-two' phases: (1)' a programmatic audit to evaluate the commitment'and involvement of corporate management and the site organization in=the SPIP, and the. process for dispost- ' tion of SPIP technical recommendations (TRs), and (2)-an implementation audit. .! i to perform more detailed review of the implementation'and disposition of l individual SPIP TRs. We had completed the programmatic audit in-1989 and-transmitted our report on that audit to you by letter dated November 1,1989. This implementation audit completes. Phase:2 of the SPIP' audit. As a result of our implementation audit, the staff: finds'that-the'TRs: .(1)ha'd. ' been satisfactorily implemented or were in the process of being. satisfactorily implemented; (2) had acceptable analysis that' verified.that:the' existing plant-
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procedures or design met TR-intent;.(3)'had acceptable justification basis,for- 4 rejection; and (4) had acceptable analysis to support non-applicability. The staff also found that good communication channels. existed between FPC head-c quarters and CR-3 personnel.
' In our previous programatic audit, we expressed concern that FPC has not J completed closure of certain TRs that required plant modification in a timely, manner, and that in some cases'. TRs closed prior to implementation of-the: ,
o L current:SPIP program lacked adequate documentetion to support conclusions- ; l regarding TR disposition. During;the implementation: audit the staff foundL. 1 that .FPC and' CR-3 had satisfactorily upgraded the:TR files,in accordance.with the programmatic audit recommendations.: In addition,_ based on our, review ofLthe TR status summary, we found that more'than 80 percent"of-the TRs had:been closed, and by the end of Cycle 8 refueling outage:in late-1992, all~ the remaining TRs; ~! pc . . 9010290457 PDR ADOCK 050 01024 2g W ' IOi
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Mr. P.M. Beard --2'- October 24, 1990 except TR-98-MFW, which is under review as required by GL 89-19, will be completed. Since the current =SPIP program was not implemented until the end of Refueling Cycle 6, and since CR-3 is on a 2. year refueling cycle, we found that the TRs are being implemented in a timely manner and that the previous concern of implementation timeliness had been satisfactorily resolved. Therefore, the staff concludes that FPC had adequately strengthened the areas of concern identified in the programmatic audit report, and that. Florida Power Corporation had established a SPIP program that satisfactorily controls the disposition and the implementation of the BWOG SPIP TRs. This completes our effort on TAC No. 68200. Sincerely, (OriginalSignedBy) Harley Silver, Project Manager Project Directorate.11 2-Division of Reactor Projects - I/II' Office of Nuclear Reactor Regulation'-
Enclosure:
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OFFICIAL RECORD COPY / ' Document Name: LTR BEARD SPIP m
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Mr. Percy M. Beard, Jr. Crystal River Unit No. 3 Nuclear Florida Power Corporation Generating Plant 1 cc: Mr. A. H. Stephens General Counsel State Planning and Development Clearinghouse j < Florida Power Corporation Office of Planning and Budget MAC - A5D P. O. Box 14042 Executive Office of the Governor The Capitol Building St. Petersburg, Florida 33733 Tallahassee, Florida 32301 i j Mr. P. F. McKee, Director Chairman Nuclear Plant Operations Board of County.Comissioners Florida Power Corporat-lon Citrus County P. O. Box 219-NA-2C 110 North Apopka Avenue Crystal River, Florida 32629 Inverness, Florida 32650 Mr. Robert B. Borsum Mr. Rolf C._Widell, Director Babcock & Wilcox Nuclear Operations Site Support Nuclear' Power Generation Division Florida Power Corporation 1700 Rockville Pike, Suite 525 P.O.-Box 219-NA-21. Rockville, Maryland 20B52 Crystal River, Florida 32629 1 Senior Resident Inspector Mr. Gary L' Boldt Crystal River Unit 3 . U.S. Nuclear Regulatory Comission Vice President Nuclear Production
.FloridaPowerdorporation-6745 N.' Tallahassee Road P. 0. Box 219-SA-2C-Crystal River. Florida 32629' Crystal River. . Florida. 32629 Regional Administrator, Region 11 l U.S. Nuclear Regulatory Comission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 l Mr. Jacob Daniel Nash Office of Radiation Control. !
Department of Health and t Rehabilitative Services l 1317 Winewood Blvd. , Tallahassee, Florida _32399-0700 Administrator Department of Environmental Regulation , Power Plant Siting Section State of Florida 2600 Blair Stone Road' Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 i
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1 ENCLOSURE. 1 SAFETY AND. PERFORMANCE IMPROVEMENT PROGRAM IMPLEMENTATION AUDIT FLORIDA POWER CORPORATION i CRYSTAL RIVER, UNIT 3 1.0 SPETY AND PERFORMANCE IMPROVEMENT-PROGRAM AUDIT 1 I 1.1 Introdi:ction From June 25 to 28, 1990, the Nuclear Regulatory Comission (NRC)' staff conducted an implementation audit of the Safet at Florida Power Corporation's (FPC)y and Performance
. corporate office in St. Improvement PetersburgProgram and at (SPIP) the Crystal River Unit 3 (CR-3) site in Crystal River, Florida. The SPIP program' was developed by the Babcock & Wilcox Owners Group:(BWOG) in order to reduce 1, both the number of reactor trips and the complexity of post-trip response. The' p(TR)implementetionatCR-3.urpose of this audit was to evaluate the BWOG SPIP tech' .
1.2 Background
AftertheaccidentatThreeMileIsland,. Unit.2(TMI-2),nuclearpowerplant { owners made a number 7f improvements to their facilities. Despite these ' improvements, the NRC staff was concerned that- the number and complexity of' events at B&W nuclear plants had not decreased as expected. This concern was reinforced by the loss-of-feedwater event at Davis-Besse Nuclear Power Station on June 9, 1985, and the overcooling transient at Rancho Seco Nuclear' Generating Station on December 26. 1985. l By letter dated January 24, 1986, the NRC ExecutiSe Director for Operations.- (EDO) informed designed reactors the Chairman should of the BWOG that a. number of recent events at B&W-be reexamined. In -its February 13,.1986, response'to. the ED0's letter, the BWOG connitted .to lead'an effort to define concerns relative to reducing the frequency of reactor > trips and the complexity of post-trip response in B&W plants. The BWOG' submitted a: description of the B&W program entitied " Safety and Performance Improvement-Program":(BAW-1919) to the NRC staff on May 15, 1986. submitted. Included in BAW-1919 Five revisions to BAW-1919 have also been were specific tasks identified as Technical. Recomendations (TRs) to be completed by each utility. under the SPIP program. The NRC staff reviewed BAW-1919 and its five revisions.and presented its evaluation in NUREG-1231, dated November 1987, and in Supplement No. I to NUREG-1231 dated March 1988. The NRC staff- had previously performed an audit of the BWOG's dis and task groups. position of TRs that were developed by various -BWOG comittees The results of that audit,.which were favorable, were , i reported in NRC Inspection Report 99900400/87/01. However, the staff deter- J"
- mined that an NRC audit program to ensure the quality of each utility's' program used to control the disposition and implementation of TRs is necessary since' l the majority of the recomendations developed by the BWOG did not provide specific design details. '
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s , 2 1 d Initially, a programatic audit was conducted that evaluated the adequacy of . the SPIP programatic process and TR dispositior This was followed by an implenentation audit that evaluated the adeque , of TR implementation. I 1.3 BWOG Reconcendation Categories All BWOG recocuencations are to be' tracked through closure. The following: categories have been selected as " bin " to be used by the utility when. assigning tracking status. These categories, as well as explanatory notes, 4 are addressed in the BWOG Reconsnendation Tracking System (RTS), in BA!!-1919,. .J and in NUREG-1231. I Evaluating for Applicabilit.y (E/A) The recomendation is being evaluated by the utiMty for applicability-to their particular alant. The evaluation may conclude that the recommendation (a) is not applica>1e, (b) was implenented previotsly and is operable, or (c) if applicable, requires further evaluation to determine;if it should be imple-meated. Evaluating for Implenentation (E/I) l l An evaluation of the recommendation for applicability has been completed, and i the reconnendation is now being evaluated to determine if it should be i implemented. l ifi ! implementing (I) -l Utility evaluation is complete and the need for software / hardware changes to 'i meet the intent of'the recomendations has--been identified. l Software changes have been assigned to the appropriate' organization and are
- scheduled and budgeted. Hardware changes have been assigned to-the' appropriate J organization for implementation, funding is approved, and the changes are included in a corporate plan for. implementation. ,
Additional coments on implementation status or' method of implementation are appropr.iate, l 1' Closed /0perable (C/0) Utility neets the intent of the recommendation, and' implementation is- complete. ! Review of exnting plant software or hardware results in a' conclusiori~ that the intent of recomendation is already met. If software changes were required, 1 new/ revised training procedures, training plans,'etc. are approved and issued. Personnel are trained and procedures issued. J
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Cicsed/NotApplicable(C/NA) ' Utility evaluation determines that the recommendation does not apply to plant-specific configuration; no past experience of underlying problems has occurred. i Software / hardware of concern does not exist, and existing software / hardware is such that a similar problem could not develop at their plant. Additional comments on why it is not applicable are required. Closed / Rejected (C/R) Utility evaluation determines sof tware/ hardware changes meeting the intent of the recomendation are unacceptable and will not be implemented. Recommendations may be unacceptable because: (1) Implementation would not result in an overall improvement in plant safety or performance. (1) Implementation of recommendation as described would not effectively resolve problem of concern. (3) Resources required for implementation are excessive for expected pl6nt improvenent or benefit.
/dditional comments on why it is rejected are required.
1.4 Progrannatic Audit.- Scope and Sunnary The NRC staff has performed the SPIP Progrannatic Audits at five utilities having D&W-designed reactors. The Proorar.matic Audit included an~ evaluation of j (1) the process used to control BWOG SPIP TR disposition, (2) the adequacy of l TR file documentation, (3) corporate and site organizational involvement in the l SPIP process, (4) the disposition of approximately 34 selected TRs, and (5) the disposition and implementation status of the approximately 222 BWOG SPIP TRs. As a result of the programmatic aucit at CR-3 in February 1989, the staff- found- ! that: l (1) FPC headquarters and CR-3, using the existing organizational structure, had established a formal, proceduralized SPIP program that adequately-controlled TR disposition; '
.i (2) CR-3 had established and maintained TR files which contained complete . -t and accurate information for those files. developed after the implementa- l tion of the current SPIP program. However, for those TR files developed-i prior to the current SPIP-program, the staff recommended that CR-3 upgrade these files by including statements in the T't files addressing the disposi-tion action taken and any engineering analysis performed and provide a i brief description of the modification implemented and the associatec l modification approval record, including a copy of the TR closure memo; u
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, 4 (3) The decisions made regarding Tk intent and' applicability during the E/A and E/I reviews were satisfactory and led to proper TR disposition . and that the SPIP program included the necessary self assesssent mechanisms l to ensure the continued adequacy of the decisions regarding TR disposition; (4) TRs were being implemented in a timely manner with the exception of those TRs which required a plant hardware change for implementation; (5) -There was evidence of adequate corporate and site management involvement in the SPIP program and that personnel involved in the SPIP program were knowledgeable with respect to their SPIP duties ano. responsibilities and that good communication channels existed between SPIP organizations.
These conclusions were documented in the letter, H. Silver to P.. Beard,
" Programmatic Audit of the Safety and Perf ormance-Improvement Program at >
Crystal River, Unit 3" dated Noven.ber 1 1989. These staff concerns can be summarized as: (1)inadequateTRfiledocumentation, i.e., in some cases TRs which were closed prior to the implerentation of the current SPIP Program lacked adequate' documentation to support conclusions-regarding TR disposition; and (2) tardiness of TR implementation, i.e., closure - t of TRs that required plant modification had not been completed,in a timely manner. o J 1.5 Implenentation Audit - Scope o l The SelP impitrentation audit included an evaluation of selected-TR files. to determ'ne if (1) plant rodification met the intent of the<TR,'(2) the operating, , training and/or maintenance procedures iraplemented met the intent of .the TR, (3) the engineering analysis used to verify that_ the existing plant design ano/or existing procedures met the intent of the TR.was adequate, (4) the basis used to reject a TR was adequate, and (5) communication channels and interfaces 1 between the corporate and site management,. operations, training, and mainten-- ence were adequate. The results of-the implementation audit at CR-3 are documented in the Section 3.2 of this report.- 2.0 FPC AND CR-3 TR IMPLEMENTATI0ll 4 Presently, the Director, Nuclear 0perations Site Support exercises oversight of the CR-3 SPIP program and also serves as the BWOG Steering Ccnmiittee repre-n sentative. The Supervisor, Nuclear Licensing,;is responsible and accountable-for the overall CR-3 SPIP program. Nuclear Operations Department-Procedure-l- N0D-15 formally establishes the methods and responsibilities associated'with-processing the BWOG SPIP TRs. In addition,' Nuclear. Licensing Procedure NL-11 formally defines the methods and responsibilities that assure adequate review, tracking, resolution, and proper disposition of TRs. TR implenentation is1 achieved through normal (existing) plant procedures / processes. The.huclear Operations Department Tracking ~and Expediting System-(NOTES) is used by the FPC. Compliance Group to monitor the progress of.each TR.ano assure that the projected schedules are met. . . co
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q The Supervisor, Nuclear Licensing, is responsible for the implementation and j closure of each 1R. The appropriate documentation is attached to a closure
'l memo and reviewed and signed by the reviewer,-the Supervisor of. Nuclear i Licensing, the director of the department responsible for implementation of H the TR, and the BWOG Steering Committee Representative. Following distribution '
of the closure memo, the FPC Compliance Group updates NOTES to reflect the new closed status. 3.0 REVIEW 0F SELECTED RECONNENDATIONS 3.1 Audited TR Selection Criteria . The staf f reviewed 16 TR files and associated documentation and. evaluated th'e-timeliness and acceptability of TR implementation.- These TRs were selected based on NUREG-1231, " Safety Evaluation Report Related to Babcock and Wilcox: Owners Group Plant Reassessment Program," the most recent' Recommendation . Tracking System (RTS) report, and the " Programmatic Audit Report - Safety and.
- I Perfornance Improvement Program at Crystal River, Unit 3." A broad telection of TRs were selected so that representative samples from the-following categories were reviewed: (1) TRs that required further attention based on the-concerns identified during the programmatic audit, (2) TRs designated " key" hy the BWOG and TRs designated high priority'by the NRC staff, (3) TRs that required a plant software chan hardware change for closure, (ge for closure, (4) TRs that required a plant 5) TRs plant operating experience, and (6) TRs that were rejected by' the individual l '
utilities. A listing of TRs reviewed and'TR status at the conclusion of the - SPIP Implementation Audit is' contained in Appendix A. 3.2 Results of Staff Review
-7 During the course of the SPIP implementation audit, the staff = reviewed the TR '
files, plant drawings, plant modification packages, training docunents, operating precedures, ano maintenance procedures ~ associated.with'the selected ' TRs. In addition, the staff conducted interviews with FPC and CR-3 personnel-to obtein supplemental information-and resolve concerns found during the l audit. The staff also performed in-plant walk-downs to verify the-accuracy of the above reviewed paperwork and information received during the-interviews. As a result of this audit, the staff found evidence that the TRs reviewed had been satisf actorily implemented or were in the process of being satisfactorily l implemented, had acceptable analysis that verified existing plant procedures l or design met TR intent, had acceptable justification basis for rejection, and had acceptable analysis to support non-applicability. The staff also found ' that good communication channels existed between FPC and CR-3 personnel and L' that the TRs were being implemented in a tirely manner. A 'brief discussion of the TR documentation reviewed as well as any exceptions to the above are-dis-cussed below. h
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. 6 TR-138-ItsSi Categorized C/R This TR recorrended that the utility install a check valve downstream of each -
air compressor af tercooler as this would enhance the reliability of!the instrument air system by preventing a ra following failure of an air compressor. pid decrease The rejectioninwas system basedpressure on the , following: 1) each air compressor-is equipped with suction and discharge I Reed (chect)(valveswhichpreventback-flowandloss-of-ai"inallcases , except a cata:,ttsphic air compressor f ailure, (2) installing a check valve immediately downstream of the air compressor would result in severe cycling stresses on the valve as the existing air compressors are of the single cylinder reciprocating type, (3) installing a check valve downstream of the air receiver would result in excessive pressure oscillations in the entire system (wher, a single air compressor is. secured) es the surge volume would be removed, and (4) the installation of check valves as discussed in (2) and (3) would not increase system reliability or performance. 'The' staff reviewed the basis for rejection, found it acceptable, and therefore, concluded that'TR-138-IAS was justifiably rejected. TR-144-IAS Categorized C/0 This TR recorcenced that the utility develop or upgrade its loss of instruirent air procedure as this would enhance the operators ability to respond to~1oss-of-air events. In accordance with the TR intent, the procedure should address the loss and restoration of instrument air, including (1) valve fai. lure air pressures, (2) air using component failure positions, (3) isolation valve i locations, and (4) a requirement for an administrative reactor shutdown / trip l on low air pressure. Also, emphasis should be placed on post-trip control for air-controlled components and on control of the decay heat removal system. TPC and CR-3 developed Procedure AP-470, Loss of Instrument Air. The staff-reviewed this procedure and' associated documentation,-found it acceptable, and therefore, concluded that TR-144-IAS was satisf actorily implemented. TR-178-IC5. Categorized I This TR recenmended that each utility ensure that the plant goes to a.known safe state (KSS) on loss of power to the ICS/NNI systems, as this would reduce the nunber of inadvertent transients caused by unexpected plant responses and would also reduce the demands placed on operators during transient conditions. FPC and CR-3 (1) developed a list of instruments and controls necessary to i achieve KSS, (2) performed an evaluation of the failed positions of the various ICS/NNI controlled components to~ determine what position would have theleasteffectofplantoperations,(3)performedanevaluationtoassure-
' the KSS is: achieved using the present signal lineup, provided indication of the power supply necessary for operation of the components, (4) installed an-alarm for loss of ICS/NNI AC power to' preclude the required manual action to trip MFW, and (5) provided an analysis-that addressed the control system and plant ressonse for the scenarios associated with loss.of ICS/NNI power. The licensee lad developed specific procedures, e.g., AP-581 Loss of NNI-X, to address the above concerns. The staff reviewed the above information, performed an in-plant walk-down using AP-581', found all to be acceptable, and therefore, concluded that TR-178-ICS was satisfactorily implemented to date.
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7 The remaining items ne.essary to complete TR closure is the operations department review of loss of sp9cific component effects on the electrical: loading'of other i instruments and corcrois in associated circuits. The results of these reviews may require changes to specific operating procedures. TR-1BI-OPS, Categorized C/0 This TR recomended that each utility verify the adequacy of the instrumenta-tion and displays used to assure and control the Abno' mal' Transient Operating Guidelines (AT00) :tability parameters, as this would reduce transient severity. ' The steff reviewed the file documentation and perforced an in-plant walk-down-to observe the instrumentatie and displays, hund that the required generic-Category C events, i.e., loss of off-site power, loss of ICS/NNI, small steam' leaks, loss of MFW and EFW, and excessive EFW flow, were adequately addressed, and therefore, concluded that TR-181-0PS was satisfactorily implemented. RT19-0PS Categorized C/0 This TR reccmmended that each utility include plant response for turbine , runback below 45% power in the operator training program,.as this would reduce- > i the number of reactor trips due to turbine trips that occur at power;1evels . less than 45%. The staff reviewed Lesson Plan ROT 5-29 which-addresses'AP-660 " Turbine Trip Without a Reactor Trip", Lesson Plan ROT-4-14, which addresses the Integrated Control System and the Unit Load Demand '(ULD) effects on turbine centrols and runbacks, and Simulator Lesson Plan POT-7-24D, which provides l simulator training for turbine trip at less than 45% power. The' staff found the above to be acceptable, and therefore concluded that TR-219-0PS was satisfactorily . implemented, i I
'TR-013-ICS, Categorized C/0 i
This TR recommended that each utility install the necessary equ.ipment to prevent-the loss of + or - P4VDC power to the ICS/NNI following the loss of.a single power source as this would reduce the probability of loss of NNI and the resultant plant transient that may lead to, a Category C event. Backup power is ' ' l supplied through automatic transfer switches which transfer. power from VBDPk6 1 to VBDP-7 for Ntil-Y, from VBDP-5 to VBDP-1 for NNI-X,- and from VBDP-4 to VBDP-2 i for the ICS, following normal power source failures. The staff. reviewed the. , support documentation for the above, found it acceptable, and.therefore, con-cluded that TR-013-ICS was satisfactorily implemented. ! TR-105-ICS, Categorized as I This TR reconnended that each utility perform a field verification of the. ICS/NNI drawing and update them accordingly as-this would assure that plant! staffs have complete, accurate drawings of.the NNI/ICS systems. -FPC Inter-- officeCorrespondence(IOC)NEA-90-0454, dated March 9, 1990, extends the NOTES due date-for completion of the 1-phase from May 1, 1990 to December'31, 1990. The staff found that utility plant walk-dewns of the ICS/NNI were
' completed, and that some design change notices (DCN) were implemented to' '
resolve deficiencies, and that additional DCN's would be issued to resolve: any additional deficiencies found during the module and termination cabinet i I
8 i i walk-downs. In addition, the photo /nenual enhancement of the drawings'and the resolution of deficiencies found on the ICS/NNI external connection drawings are ciso reouired to close this TR. The staff reviewed the basis for delay,. completed drawing updates, and associated file documentation,.found all to be acceptable, impitmented and therefore concluded that TR-105-ICS was being satisfactorily to dete. TR-190-ICS and TR-191-ICS. both Categorized C/0 TR-190-ICS recomended that the utility develop backup manual or automatic controls for pressurizer level and pressurizer pressure control powered from j 4 an alternate power source, as this would increase the operators ability-to maintain pressurizer level and pressure control during a loss of NNI power or , an input pressure signal failure. TR-191-ICS-recommended that CR-3 separate-condensate flow control from NNI-Y power as this would reduce transient response complexity and eliminate reactor trips due to NN1-Y power failures. During the E/1 phase, the licensee determined that no changes were required for pressurizer pressure control as the analog-controlled pressurizer heater 1 j banks would fail to zero on loss of NNI-X power and that existing: plant design 1 allowed for manual on/off control _of pressurizer heater banks under this condition. Backup controls for pressurizer level control and condensate flow control were developed and im The staff reviewed the above,plemented to meet the found it acceptable, full intent ofconcluded and-therefore: these TRs. i that TR-190-ICS and TR-191-1CS were satisf actorily implemented. TR-119-PES. Categorized C/0 1 This TR recomr. ended that preventive baintenance procedures be implemented. ! for the maintenance of electrical buses, as this would significantly= reduce the i likelihoed of catastrophic bus failure which could create both a plant opera-tional problem as well as a personnel safety hazard. FPC had preventive ' maintenance procedure PM-119, Rev. 7, in place. The staff reviewed the procedure, found that it adequately. addressed the TR, intent,.and therefore r concluded that TR-119-PES had been satisfactorily implemented. 1 TR-066-MFW and TR-179-MFW, both Categorized 1 TR-066 reconsended that each utility check all condensate /feedwater system , protective circuits, interlocks, motors, and other necessary elactrical equip , ment for system operation to ensure that no single electric failure would cause-a loss of both feedwater trains. TR-179 recommended that each Ltility evaluate. and identify areas for enhancing'the reliability of the condensate /feedwater systems and controls with attention given to preventing the failure of:an i active component from causing a-loss of all feedwater (FW),7and to make changes identified in this evaluation:as practical. i i In June 1988 FPC changed the status of TR-066-MFW from E/A to E/1,-contacted D&W and purchased an ongoing study, B&W Report DOC 51-11717.79-01,;"MFW Reliability-Improvement Program," dated December 20, 1988,.that identified all single failures-in the MFW and support systems in. the CR-3 plant. The results of the ) B&W evaluation were contained in B&W report DOC 51-1171279-01', "MFW . Reliability i s
q 9-1 J l Improvement Program," dated December 20, 1968. The B&W study identified a l total of 14 single failures within the feedwater system, condensate system, . electrical distribution MCC/ panels, gland seal steam system, gland seal water system, turbine drain system, and secondary service con,ponent cooling system. These 14 single failure points were prioritized based on their impact on plant operation. In addition.-B&W included 19 recommendations which should be ! considered to enhance system reliability. l l Two of the 14 single failures were identified as having a low probability of ; ! occurrence but having a "high" impact on plant operations, i.e...the single l failure would cause an immediate or very short-term plant trip. These two items are Deaerator Storage Tank high level switch (FW-4-LS) and low-level l l switch (FW-311-LS). A failure of FW-04-LS or- FW-311-LS could result in a-trip l of both condensate pumps or both booster pumps, respectively, and therefore ceuld result in a complete loss of both feedwater trains. These two items-were assigned to TR-066-MFW under FPC modification approval . report. MAR-86 i 09-01. The remaining 12 single failures were determined to have a medium-to- l low impact on plant operations. The 12 single ~. failures, along with-the 19-additional B&W reconnendations and the 10 FPC internally generated' reconsnenda-tions, were considered reliability enhancement; items and were scheduled to be implemented in accordance with MAR-87-02-30-05 under TR-179-MFW. In regard to the single failure concern of TR-066-MFW, FPC determined the need to add two additional level switches, one to' supplement FW-0445 and one- to supplement FW-311-LS, and revise the condensate and booster pumps control switch i schemes such that a single failure would not trip both main feedwater (MFW) trains. Detailed designed work is scheduled to begin in March 1990 with installation scheduled for Refuel VIII (in mid-1992). i l With regard to TR-179-MFW, FPC had reviewed the potential reliability enhance- " ment items to determine whether they were' covered by other SPIP TRs, whether- ; they would enhance CR-3 MFW reliability, and whether they were cost effective. As a result of'this evaluation, a complete list of-items recommended =for implementation was generated. These reliability enhancement items are scheduled ~ { i for implementation during Refuel 8, in accordance with MAR.87-02-30-05. ! In an October 30, 1989 speed letter ;FPC addressed an additional possible i single failure in the MFW system. FWV-28 is an 18" motor-operated gate valve j that cross ties MFW-trains A and B downstream of the'_MFW punip discharges.- The BWOG audit indicated that a failure of the FW. crossover valve to open would. ; reruit in closure of the MFW block valves-(FWV 29/30) via an interlocks thus -{ causing a loss of main feedwater., However, FPC stated that the operators are i trained to respond to this type event and prevent a loss of MFW, therefore a j system modification is not required. ! The staff reviewed and evaluated the above information, found all to be- .I acceptable, and therefore concluded that TR-066-MFW'and TR-179-MFW had been ! satisfactorily implemented to date. l i =
i 10 TR-071-NFW, Categorized C/0 This TR recomended the installation of valve position indication for the j startup ar.d main feeowater regulating valves or low load control valves, as it ' would provide true valve position and eliminate confusion and allow faster operator response during transient conditions. The CR-3 feedwater configuration has three parallel feedwater paths to each steam generator, i.e., main, low load, and startup flow paths. Flow regulating valves are only installed in the startup and low load lines. When flow demand' exceeds 50%, the ICS controls FW flow by varying the speed of MFW pump turbine. ; Therefore, the only CR-3 valves affected by this TR are the-low load valves- l FWV-37 and 38, and _startup valves FWV-39 and 40.
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This TR was closed by implementing MAR 86-05-09-05, "Feedwater Valve Position-Indicatior. " which required: (1) analog position indicators be added to the- q MCB-ICS Section and locatea above their associated valve auto / manual stations, ' (2) Bailey RQ-20 electronic analog position transmitters be' yoke-mounted on-each of the four regulating valves, and (3) that four single-loop 24 YDC power. supplies be installed on the rear of and inside the MCB to supply power to the - 4-20 MADC current loops. The staff concluded that TR-071-MFW was satisfactorily implemented.. TR-098-MFW,. Categorized as E/I l This TR recomended that the MFW system design include an operational automatic L overfill protection system in order to prevent a loss of heat sink or water l inventory in the main steam lines. , The original CR-3 emergency feedwater initiation and control (EFIC) system design submitted to NRC included such an overfill protection sy. stem. However, , the steam generator overfill protection feature was removed from the EFICL ' systerr because of a concern over MFW pump trips due to fluctuations in steam generator level, as level is maintained close to the SG aspiration ports level and the EFIC high level trip setpoint. FPC performed an evaluation and con-cluded that safe operation with the steani generator. overfill trip = feature disconnected could be continued, and therefore, this TR was origina.lly categorized C/R. ~ However, this TR was reopened on June 22,'1990 and categorized E/I, based on. the suggestions of NRC Generic Letter 89-19. " Request for Action Related;to ' Resolution ofs unresolved Safety Issue A-47," which requires that all'PWR (, plants provide automatic steam generator overfill-protection. iThe E/I-' phase l will begin Janaury 2,1991 and is scheduled to be completed by March 31,'.1992, i FPC responded to GL 89-19 in a letter dated March 9. 1990, which stated that an appropriate system will be developed to protect against steam generator overfill. Iriplementation of TR-098-MFW will be during Fuel Cycle 9(1992-1994) commensurate [ with construction work package development and material delivery..This schedule L is based on the following considerations: .(1) there is a low probability of an i overfill event because adequate instrurr.entation and procedures are availeble to , , - aid in manual operator actions, (2) sufficient indications are available for. the operators to recognize overfill problems, and (3)'there is adequate time d
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11 1 for the operators to react to an overfill condition. In addition,.the FPC letter also expressed concerns that the GL 89-19 recommendation maylnot have fully considered the integrated effects of all of.the installed and proposed control systems associated with the MFW system. Since the NRC staff is cur-rently reviewing and evaluating the implementation of GL 89-19, this TR should be implemented based on the resolution of GL 89-19. FPC may be' required to- . take additional actions to ensure satisfactory implementation of TR-098-MFW. TR-155-EFW. Categorized C/0 This TR recomended that each utility: -(1)considerameanstolimit'the l maximum flow rate delivered by' the emergency feedwater (EFW) system.-(2) make - plant-specific modifications to limit-EFW flow when once-through-steam-generator (OTSG) level is increased to the natural circulation level .setpoint forplantswithoutautoflowlimits,.and-(3)determinewhetheranEFWpump runout condition is possible at their plant and evaluate the consequences. The basis for this recomendation is to ensure that EFW flow to the OTSGs is limited in order to reduce the_ potential for overecoling of those' plants having EFW capacity significantly in excess of that Med for decay heat removal, and to prevent EFW puma runout following a ruptur J a 4FW or EFW line, or a steam-line break. CR-3 .1ad previously: installed EFic, which has a-built-in level rate control systen that regulates the OTSG fill rate and maintains-the. natural circulation setpoint. The OTSG fill rate is control;ed using OTSG. outlet , i pressure so that the system will initially supply high EFV flow:(8= inches / minute at 1050 psig) for high decay heat' levels, but _will automatically ' ; throttle back (2 inches / minute at 800 psig) if overcooling becomes apparent:as evidenced by decreasing OTSG pressures. During the Cycle 6 refueling outage, a modification was made via modification
-i approval record MAR-86-05-25-01 to alter part.of the'EFW. control valve = '
circuitry in order to limit the EFW flowrate.and provide pump' net positive suctionhead(NPSH) protection. The EFW flow control circuitry acts to. partiall minute (y close the control valves when-the EFW flow exceeds l600 gallons-per gpm). Despite the EFIC system limitations on EFW flow, fit may be necessary for the operators to further reduce flow in certain-low power / low' decay heat : scenarios. There is adequate guidance-in-procedures to avoid exceeding maximum allowable cooldown rates. Step 3.15 of' AP-450, "EFW Actuation,"- directs the operators to maintain allowable RCS:cooldown' rates-once EFW has, been initiated. With regard to the concern over the EFW. pump runout _during OTSG depressuriza-tion, FPC contracted Gilbert / Commonwealth Corporation to perform:a hydraulic analysis of the EFW system (W.0. 045510140,. dated December:1 t 1987.'The' analysis calculated the available and' required NPSH at various pump /0TSG combinations, control valve positions and OTSG pressures. The results showed" that if the EFW control system limits EFW flow to 650 gpm per pump oriless, sufficient NPSH is available to prevent ~ pumt cavitation. Since the EFIC system limits the EFW ficw to less than 600 gpm, t1ere isino concern-over pump
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4 cavitation. The staffthat concluded reviewed the was TR-155-EFW abovesatisfactor information,ily implemented found it! acceptable, and.therefore
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l 1 ' 12 l I TR-163-EFW, Categorized C/0 This TR recomend that each utility review the EFW surveillance and test pro-cedures to ensure that components used in emergency or abnormal operating i procedures are included in those procedures, and that these components are tested as near as possible to the expected operating conditions to demonstrate EFW system and component operability, thus enhancing the overall plant reliability. FPC had reviewed all applicable emergency and abnormal procedures, compiled a. list of all components and/or functions of the EFW and EFIC systems, and evaluated the specific surveil?ance rocedures that test these components.. As j a result of these reviews FPC conc 1 ded that the EFW system components- used , in the energency and abnormal operating procedures are satisfactorily tested l with four exceptions. These four exceptions, which were addressed by Nuclear 1 -! Plant Systems EnDineering, are summarized below:, 1 (1) The use of the " manual permissive" push button after an EFW initiation, l originally found not to be tested in any surveillance procedure, is tested- 1 every 18 months per SP-416 " Emergency Feedwater Automatic: Initiation." (2) EFW is never initiated and supplied to the steam generators, nor is the j auto level control checked under real flow conditions. However SP-435, l
" Valve Testing During Cold Shutdown," is performed every 18 months and after every mode 5 or 6 outage that lasts for~more than.30 days. It is l CR-3's policy to run a full-flow test whenever a major. change is made!to i i
the control room. Otherwise, the flow test in.SP-435 and electronic l checks in SP-416 are performed to verify adequate flow paths and control. 9' (3) The EFW tank level indication was checked in a functional test during Refuel 6 and it was confirmed that SP-169A has been revised and now includes a calibration of the EFW tank level string. (4) The hot well level transmitter and level switches have scheduled calibra-tion intervals of 24 and 36 months under the approved PM-200' periodic calibration program. The staff reviewed the:above information, found iti " acceptable, and therefore concluded that TR-163-EFW.had been satisfac-- torily implemented. 'However, the safety-related motor-operated valve ' n testing and surveillance program is currently.under NRC staff review per requirement of GL 89-10. This TR implementation should.be consistent with. the resolution of GL 89-10.
4.0 CONCLUSION
S - SPIP PROGRAMMATIC AND IMPLEMENTATION AUDITS ~ During the arogrammatic. audit, the staff reviewed the disposition of 34 TRs-and found taat-evidence of adequeu. FPC corporate and site = management involve-ment in the SPIP process, and determined that a formal,'well-documented , l proceduralized SPIP process had-been used at CR-3 to control the, disposition l ' of TRs. The staff also found that: (1) FPC has not completed closure of d certainTRsthatrequiredplant<modificationinatimely, manner,and'(2)in some cases. TRs closed prior to implementation of the current SPIP program lacked adequate documentation to support conclusions regarding TRLdisposition.. q , t l
. .i 13- I l
During the implementation audit, the staff reviewed the implementation of 16 TRs. Several of these TRs were identified during the programmatic audit as 1 TRs that would require follow-up action. As a result of the review, the staff- I found that the TRs: I process of being satisfactorily implemented;(1) (2 had had been satisfactorily) acceptable analysis.that implemen .i verified the existing plant procedures or design met TR intent; (3) had acceptable justification basis for rejection; L4) had acceptable analysis to support non-applicability. The staff also found that good communication channels existed between FPC and CR-3 personnel. Our previous progrannatic audit found that the timeliness of implementation need strengthening. Based on our review of the TR status summary, we-found that more than 80 percent of the TRs had been closed, and at.the end:of Cycle 8. refueling outage in mid-1992, all the remaining TRs except TR-98-MFW will be- ' completed. Since the! current SPIP program was not implemented until the end of . , Refueling Cycle 6, and sinceLCR-3l1s on a 2-year. refueling cycle, we fcund .that the TRs were being implemented in a timely manner and that the programmatic audit concern of implementation timeliness had been satisfactorily resolved.< In addition, the staff found that FPC and CR-3 had satisfactorily upgraded the TR files in accordance with the programnatic audit recommendations. Therefore, < the staff concluded that FPC:had adequately strengthened the. areas of concern identified in the programmatic audit report. Therefore, the staff' concluded that Florida Power Corporation and Crystal River Unit 3 had established a SPIP ; program that satisfactorily controlled the disposition and the implementation 4 of the BWOG SPIP TRs. b +
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i c :l 14 ; APPENDIX A-IDENTIFICATION OF TRs REVIEWED, TR STATUS, AND CONCLUSION STATEMENT Instrumentation and Control System (ICS) Instrument Air System (IAS) Motor Operated Yalves (MOV) ' Emergency feedwater (EFW) Operations (OPS) + Plant Electrical Systems MainTurbineSystem(MTS)(PES) Main feedwater System (MFW) Status . 3 6/28/90 Comments on Implementation / Recommendations . 013-105 C/0- Satisf actorily . Implemented 105-10S I Satisfactory Implementation to date 119-PES C/0 Satisfactorily Implemented l, 190-105 C/0 Satisfactorily Implemented . 191-105 C/0 Satisfactorily Implemented. 5 066-HFW I Satisfactory Implementation to'date 179-NFW I Satisfactory Implen.entation to date 071-NFW C/0 Satisf actorily. Irplemented , ' 098-MFW E/I St.tisfactory Implementation to date 155-EFW C/0 Satisfactorily Implemented ' 163-EFW C/0 Satisfactorily Implesented . 1 i 138-IAS C/R Justifiably . Rejected - L a
' 144-IAS C/0- Satisfactorily Implemented th8-ICS I Satisfactory Implementationi toLdate i 181-0PS C/0 Satisfactorily Implerented-EI9-0PS . C/0 Satisfactorily Implenented J
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15 APPENDIX B LIST OF AT.iNDEES AT.THE ENTRANCE AND EXIT MEETINGS FPC/CR-3 SPIP IMPLEMENTATION AUDIT JuhE 25-28, 1990- - Attendee Organization /Titie~- Entrance. g Edwin Froats FPC/Supy Nuc Licensing X- X. Rolf Wioell Dir Nuc Ops Site 1 Support X 1 Jares Owen TrainingSupy(acting) X 4 James Kreiker Man Sup Superintendent X- X ( Richard Low Principal Nuc I&C Engr X X-Paul Tanguay- Dir Nuc-Ops Eng & Proj (acting) X Pablo Rubio Nuc I&C Engr-Supy X -X-Chris Doyel Mgr Mech /Struc Eng X Max Yost- INEL/NRC Engr Spec X X John Fehringer INEL/NRC Engr Spec X X: i Y. Gene Hsii NRC/NRR SRXS X- XL J. A. Frijouf Nuc Regulatory Spec X. Richarc Iwachow Senior Nuc I&C Engr X Ronald Zareck SkO/ Tech Consultant ~X Gary boldt VP Nuc Production X Bruce Hinkle Mgr Nuc Plant Ops X W. L. Rossfeld Hgr.Nuc Compliance .X
- Sarah Johnson Ngr Site Nuc Serv XL
, Ken Lind Ngr Lic Oper Training X 4 Ken Wilsun Hgr Nuc Licensing: X 7 i d
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