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| | number = ML18010B084 | | | number = ML18010B084 |
| | issue date = 05/05/1993 | | | issue date = 05/05/1993 |
| | title = NRC Licensing Submittal Review of Licensing Conditions Imposed by NUREG-1216. | | | title = NRC Licensing Submittal Review of Licensing Conditions Imposed by NUREG-1216 |
| | author name = | | | author name = |
| | author affiliation = TDI (TRANSAMERICA DELAVAL, INC.) OWNERS GROUP | | | author affiliation = TDI (TRANSAMERICA DELAVAL, INC.) OWNERS GROUP |
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| {{#Wiki_filter:TEVhÃSMNERICA DELTAL, INC.OWNERS GROUP NUCLEAR REGULA TORY COMMISSION LICENSING SUBMITTAL REVIEMf OF LICENSING CONDITIONS IMPOSED BV NUREG 12 I6 9305050i65,930503 | | {{#Wiki_filter:}} |
| =POR ADOCK 05000400 P PDR Revision 1 May, 05, 1993 h t NUCLEAR REGULATORY COMMISSION LICENSING SUBMITI AL ON BEHALF OF THE TRANSAMERICA DELAVAL, INC., OWNERS GROUP FOR REVIEW OF LICENSING CONDITIONS IMPOSED BY NUREG 1216 5-3-93 Rev.1 lej THE TRANSAMERICA DELAVAL, INC.OWNERS GROUP LICENSING CONDITIONS TABLE OF CONTENTS 1.0 2.0 3.0 EXECUTIVE
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| ==SUMMARY==
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| INTRODUCTION AND BACKGROUND
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| .COMPONENT PERFORMANCE REVIEW 3.1 Engine Overhaul Frequency 3.2 Air Start Valve Capscrews..3.3 Engine Mounted Electrical Cable 3.4 Engine Base and Bearing Caps 3.5 Connecting Rods.3;51 lssssss 3.6 Connecting Rod Bearing Shells 3.7 High Pressure Fuel Injection Tub 3.8 Crankshafts 3.81 3.9 Jacket Water Pump 3.10 Cylinder Blocks/Liners 3.11 Piston Skirts 3.12 Cylinder Heads.3.13 Push Rods.....~........3.14 Cylinder Head Studs 3.15 Rocker Arm Capscrews.3.16 Turboch erg ers 4.0 SYSTEM UNRELIABILITY
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| ===5.0 SYSTEM===
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| UNAVAILABILITY........
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| HOW TO USE THIS APPENDIX.~...APPENDIX A Part A Part B APPENDIX B APPENDIX C APPENDIX D APPENDIX E ing.~~~~~~~~~~~~~~~~I~~~~~~~~3~~~6 7~~~~~~~~~~~~~~~~9~~~~~~~~~~~~~~~~1 2~~~~~~~~~~~~~~~~~~~~14 14~~~~~~~~~~~~~~~~~~~~~1 6 18~~~~~~~~~~~~~~~~~~~~~~~~1 9~~~~~~~~~21~~~~~~~~~~~~~~~~~~~21 23~~~~~~~~~~~~~~~~~~~~~~~~25~~~~~~~~~~~~~~~~~~~27 29~~~~~31~~~~~~~~~~~~~~~~~~33 35 36 37~~~~~~~~~~~~~~~~~~~~~~~40 42 ,44 5/3/93 Rev.1
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| The Transamerica Delaval, Inc.(TDI)Owners Group recommends the removal of the licensing conditions imposed by NUREG 1216.Based on substantial operating experience and the Design Review/Quality Revalidation (DR/QR)effort for the critical components, the TDI emergency diesel generator (EDG)has demonstrated that special concerns of NUREG 1216 are no longer warranted.
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| Therefore, the TDI EDGs shall be regarded the same as other EDGs within the nuclear industry, and subjected to the standard regulations without the special requirements of NUREG 1216.These conclusions are supported by the information in this document.In addition, this action will improve unavailability of the engines for service, especially during outages, while maintaining current low unreliability levels.Removal of these conditions from the license will not prevent these activities from being performed in the future.These types of activities should be performed when the components are disassembled for other reasons.The Technical Specifications for each plant currently require that an inspection of the diesel generator be conducted every refueling outage and these inspections should include items needed to-maintain the engines in a reliable and available condition.
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| The Owner's Group is currently working with the manufacturer to develop a new maintenance program that incorporates the experience of the owner's of the equipment combined with the experience of the manufacturer.
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| This joint effort will assure that high reliability is maintained in the equipment.
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| For each EDG license requirement that is being removed as a license condition, the Owner's Group will review the future maintenance needed and adopt a program, consistent with manufacturer recommendations, to fulfill these needs.The basis of the TDI surveillance matrix deals with preventative maintenance, monitoring, and inspections.
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| The latter of this list is by far the largest contributor to the significant out of service times experienced in outages.In addition the requirement to perform an overhaul every 10 years (a complete overhaul has not yet been performed after 10 years of operation) will add even more to the unavailability of the engines 5-3-93 Rev.1
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| during outages.The overhaul frequency is discussed in detail in Section 3.1.This submittal addresses a solution to reduce unavailability by reducing engine teardowns and inspections.
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| This will be accomplished by more closely monitoring and trending the data that is already being collected.
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| Teardown and inspection will be performed when indicated by the maintenance/monitoring and trending programs for the engines.Acceptance of this submittal will reduce unavailability and will comply with Station Blackout levels of unreliability which will reduce the risk of core melt as noted in work that has been performed on Station Blackout Issues.Acceptance will also help these utilities prepare for the issues to be addressed by the Maintenance Rule.The TDI Owners Group therefore requests the NRC to review the revised recommendations contained within this report and issue a generic Safety Evaluation Report (SER)endorsing removal of the component based License Conditions that are currently required by certain power plant Operating Licenses.This generic SER would then be referenced by individual licensees to process Operating License amendments on each docket for plant with TDI diesels to remove these License Conditions.
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| All aspects of the maintenance and surveillance programs would then be controlled by the licensee and reviewed by the NRC under current regulations which is the condition that all other plants operate under.5-3-93 Rev.1
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| The Design Review/Quality Revalidation (DR/QR)effort of 1984 has been performed on Emergency Diesel Generators (EDG)supplying emergency AC power for the following utilities that are in support of this licensing submittal:
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| Texas Utilities, Inc Entergy Operations, Inc.Duke Power, Inc.Carolina Power and Light, Inc.Georgia Power/Southern Nuclear Operating, Inc.Comanche Peak Grand Gulf Catawba Shearon Har'ris Vogtle Cleveland Electric illuminating Co,/Centerior Energy II Gulf States Utilities, Inc.Tennessee Valley Authority Peiry River Bend 8ellefonte (Note that not all engines at all plants have completed DR/QR as indicated in the particular docket;but each utility has a representative sample of engines that have completed this inspection and have operational hours since the inspections).
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| This effort was in response to NRC concerns regarding the reliability of large-bore, medium speed diesel generators manufactured by TDI for application at nuclear power plants.The scope of this submittal and review is limited to the utilities and concerns of their specific'engines.
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| Concerns and items of other engines at other utilities are not addressed and are considered valid and applicable to those utilities by the Owners Group.An explanation of the other utilities originally involved in the DR/QR effort but zg a part of this action follows: Southern California Edison remains a current member of the Owners Group, however due to a decision to decommission, Unit 1 of the San Onofre plant is not a Rev.1
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| participant in this action.Long Island Lighting and Sacramento Municipal Utility District have ceased membership in the Group due to decommissioning actions and are not participating in this action.Washington Public Power Supply and Consumers Power have deferred or canceled plants and are not a participant in this action.This accounts for the thirteen utilities that originally began development of the DR/QR effort.This effort was originally outlined and documented with the NRC as the TDI Owner Group Program Plan.This plan was accepted by the NRC in an Safety Evaluation Report (SER)dated August 13, 1984.Following issuance of the SER, the Owners Group member utilities developed and implemented the DR/QR in response to the Program Plan.The specific details of the DR/QR were submitted to the NRC for review and this information was reviewed and referenced as part of the NRC position which was documented in NUREG 1216.The recommendations of the NRC consultants hired to assist in this effort is also referenced in NUREG 1216 and is documented in PNL-5600.These details resulted in specific license conditions for each utility as the individual DR/QR reports were submitted under the utilities respective dockets.These utilities have operated for a substantial time period and logged many operation hours on these EDGs and this operational data is being submitted for review to remove the license conditions imposed by NUREG 1216.It should be noted that the scope of the original NRC review was to took in detail at the Phase I components as defined by the DR/QR program.NUREG-1216 documents the NRC reviews of Phase I and II components.
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| Phase I components are addressed later in this submittal.
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| Phase II components constitute approximately 150-170 components on the engine.The NRC review of Phase II components documented in NUREG-1216 concluded that a detailed review of these items was not necessary and would be redundant."The Phase I components were chosen as those that had potential for generic concerns.Through an extensive review of TDI and other engine performance data in both nuclear and non-nuclear applications, 5-3-93 Rev.1 8
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| the Owners Group identified 16 components with such concerns.These are: air start valve capscrews connecting rods/connecting rod bearing shells crankshafts cylinder block cylinder heads cylinder head studs cylinder liners engine base and bearing caps engine mounted electrical cable high pressure fuel injection tubing jacket water pump piston skirts push rods rocker arm capscrews turbochargers These engines have operated under the requirements of the program reviewed and approved by NUREG 1216.This document presents the results of the operation of a large sample of engines under that program and demonstrates that the reliability of these engines is comparable to the reliability of other EDGs and that the time required to continue to perform teardowns and inspections as outlined in specific licensing It conditions substantially adds to the unavailability of the engines.Subject to the findings of this r'eport, the Owners Group concludes that these engines can be operated in a safe manner without degrading reliability and still achieve improvements in unavailability by removing license conditions to perform inspections requiring engine teardown.The Owners Group will develop a performance based maintenance program outside of the licensing environment to assure that the goals outlined'above will continue to be met.5-3-93 Rev.1 This section discusses the original component concerns, the proposed modifications/inspections that were subsequently required, the results of the modifications/inspections, and a proposed disposition of each item.The proposed resolution of these items has been discussed with the manufacturer and they are in agreement with them.The modifications/inspections that wiii be discussed are listed in the DR/QR report, , Appendix II, Part B.A copy of the current version of Parts A and B of this Appendix is included as a part of this submittal as Appendix A.Appendix A and NUREG 1216 are the basis for the license conditions that are imposed on some utility dockets.The original review contained in the above documents along with the results of the inspections performed since that initial review was completed will be the review basis for the amended recommendations to be approved by the NRC.5-3-93 Rev.1 The overhaul frequency for the TDI engines was originally recommended to occur at an approximate 5 year interval.This interval was later revised to 10 years because (1)of the comprehensive DR/QR effort conducted for each of the engine components, (2)of the limited number of operating hours for the engines in nuclear standby service, and (3)a sample inspection of major engine components will be performed on a one-time basis following 5 years of service.Details of the results of inspections performed during this teardown are outlined in the discussion of the individual components.
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| Overall, the teardowns did not indicate any major problems or suggest that any component had experienced any significant wear.The 1 average number of operating hours logged on an engine in a year is approximately 100 hours.This number is much less than the number of hours typically experienced by non nuclear engines.This mode of operation lends itself to using monitoring/surveillance programs in lieu of hours of operation to determine overhaul frequencies.
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| All utilities have and will maintain a monitoring, trending, and surveillance program to determine the health of the engine and determine when corrective actions, including overhaul, arerequired.Collectively, these engines have accumulated over 9000 hours of operation.
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| This provides a significant data base on which to base removal of the license conditions imposed by NUREG 1216.Recent studies performed for the NRC (
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| ==Reference:==
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| NUREG/CR-5078, PNL-6287, NUREG/CR-4590, NUREG/CR-5057) indicate that for approximately 2 years following a major engine overhaul, EDGs, regardless of their manufacturer, exhibit increased unreliability.
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| This increase is attributed to several reasons.One reason offered is that during disassembly there is a high potential to introduce dirt and other substances that may harm the engine.Another is that disturbing a precision fit system that'wears in'o seat mating surfaces (eg rings and liners, crankshafts and bearings, connecting rods and bearings)can result in alteration of wear patterns that may increase wear or actually cause wear to start and decrease 5-3-93 Rev.1 the life of the component.
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| As noted in the above reference, the period following overhaul is a"shakedown" period that is required to produce a smooth running reliable engine.Utilities have and will continue to minimize this impact by performing"break in'uns per the manufacturer recommendations; however, the period for'shake down'xtends well beyond the break in run time.The Owners Group agrees with the findings of the above study.In addition, the results of the 5 year'mini'verhauls have shown no component failures that resulted in a loss of component function and have also shown that operational component wear since installation has been very minimal.All plants listed have completed the 5 year"mini'verhaul for their engines with the exception of Comanche Peak and Bellefonte.
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| To perform a complete engine overhaul for a typical engine could take approximately six weeks duririg an outage and could make the diesel more unavailable during the outage.Extending the period betw'een overhauls reduces the overall cost that would be incurred for additional parts and labor to install and refurbish components that are no worse from wear than the new parts to be installed.
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| In order to prevent increased unreliability and to reduce unavailability,~hlhh surveillance program and will continue to use maintenance/monitoring and trending data similar to the information gathered in Table 1 of Appendix II of the DR/QR report, to determine when a particular component would need refurbishment or replacement.
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| This would also give the utility the flexibility to plan for this work to be performed over an extended period in lieu of one outage period and would serve to lower unavailability and lower unreliability.
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| The concept of performing overhauls based on trending and monitoring has been discussed and endorsed by the manufacturer.
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| 5-3-93 I I h Rev.1
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| There are no PM recommendations associated with this component in Part B, Appendix A.Revision 2 of Part B, Appendix A recommended that upon installation of a new capscrew, retorquing should be performed at specified intervals to compensate for gasket creep.When no change in torque is detected, the gasket is fully compressed and the torque will be maintained.
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| This item was removed by revision 3 to Part B as the manufacturer has agreed that this is a proper recommendation and has put this item in their PM recommendations.
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| The air start valve capscrew have not had a history of failure.The original concern with the component dealt with the component being too long and'bottoming out'n the cylinder head.In SIM 360, TDI recommended a change to use a shorter capscrew and recommended a suitable torque value.This was in response to reports at Shoreham and Grand Gulf where these capscrews had been found to loosen.Loosening of this component or other related problems have not been detected since the utility has either made the change noted above or has verified that the existing capscrew does not bottom out.All capscrews have been properly torqued.This is the Justification for removal of this item from Part B and placing this information with the vendor recommendations.
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| 9 This item was closed under NUREG 1216 and no further problems have been reported.Utilities should continue to follow vendor torquing procedures upon replacement.
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| 5-3-93 10 Rev.1
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| There are no PM recommendations associated with this component in Part B, Appendix A.TDI SIM 361, revision 1 notified the engine owners of potentially defective engine-mounted cables associated with the Woodward governor/actuator and the AIR-Pax magnetic pickup.This memo led the Owner's Group to review in detail the suitability of all class IE auxiliary module wiring and terminations currently installed on the diesel engines.Of special interest was the suitability of this wiring with respect to flame-retardancy of the insulation, qualification to industry standards, routing of conduit, compatibility with circuit requirements, and the need for special requirements such as shielding.
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| Modifications were, in some cases, recommended and all of these modifications were completed.
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| No further problems or issues have been found dealing with this component.
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| The modifications specified address the concerns with this component and this issue was closed during the initial NRC review.This item was closed under NUREG 1216 with no additional concerns found since that time and this item remains closed.5-3-93 Rev.1
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| The base and bearing caps preventative inspections are listed in Part B of Appendix A.Specifically, PM recommendation 1 can be made without a disassembly; PM recommendation 2 does require disassembly but is only required to be performed when the caps are removed for other reasons.The original Owner's Group design review for this component found adequate factors of safety for all components.
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| Problems encountered with this component are not generic in the engines supplied for nuclear service.Problems that were encountered were with non nuclear service engines resulting from inadequate bolt preload and in one case, marginal strength due to inferior quality of a casting.The NRC review noted specifically that once the caps are installed according to the Owner's Group recommendations and torqued to TDI specifications, they should not require further attention until they are removed for some other reason.It should be noted that inspections proposed in Part B of the maintenance matrix were to validate the findings of the analysis discussed above and were a conservative step to aid the licensing process.5-3-93 12 Rev.1
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| For all engines in current service, a metallurgical exam for Widmanstaetten graphite has been made or the recommended three cycle inspection for cracks have been completed and none of the bases have indications of inferior material~Twenty-five separate base inspections have been made with no signs of cracks noted.In addition, hundreds of inspections have been made of the bearing cap and saddle interface for PM item 2 and no problems have been detected.Based on the positive results of the monitoring and the conservative nature of the PMs, the base inspections should be no longer necessary.
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| The inspection of the cap mating surfaces should continue as good maintenance practice only when the caps are removed for other reasons.5-3-93 Rev.1 I
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| The connecting rod preventative inspections are listed in Part B of Appendix A.Specifically, PMs 1,2,4, and 5 require teardowns to perform.PM item 3 is excluded from this discussion as it is the scope of a previous license submittal and is already under review by the NRC.These inspections have been performed on the River Bend engines as outlined in Appendix B.During the DR/QR review, only one rod failure was reported and that was on a non nuclear application and the failure was due to the possibility of pre-existing defects on the surface of the rod eye and to the higher peak firing pressures used in the engine that had the rod to fail~The design review performed found no design problems with the rod.However, the NRC recommended that a rod eye and bushing be inspected using an acceptable NDE technique and that all bolts and washers be inspected at the same time.The rods at River Bend have been inspected on a sampling basis at the 5 year interval with no problems found.This was performed on two connecting rods per engine and the associated bolts and washers and bearings.5-3-93 14 Rev.1
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| .0 Sufficient operating hours have been accumulated on most engines such that the connecting rods have been in operation and subjected to a number of cyclic loadings to demonstrate unlimited fatigue life.Subsequent inspections have also shown bearing wear to not be a problem.Based on this information and the initial design review and the positive inspection results, it is concluded that these inspections should not be performed unless the rod is removed from the engine for other reasons.These inspections should be viewed as good maintenance practices and not as requirements.
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| '-3-93 15 Rev.1 The connecting rod preventative inspections are listed in Part B of Appendix A.Specifically, all PMs with the exception of PM 9 require teardown to perform.PM item 3 is excluded from this discussion as it is the scope of a previous license submittal to the NRC and has been approved.During DR/QR review, a total of six rod failures were documented.
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| TDI had identified two failure mechanisms in SIM 349.The first was due to fatigue of the link rod bolts resulting from loss of bolt preload.The second mechanism was fatigue cracking of the connecting rod bolts and/or the link rod box in the mating threads.The Owner's Group Design review performed a detailed stress analysis of the rod and looked at fatigue as suggested by TDI.The results of that analysis showed the peak stresses induced by the loading mechanisms are slightly below the fatigue initiation curve for rods with 1-1/2'olts and slightly above the fatigue initiation curve for rods with 1-7/8" bolts (Reference FaAA Report FaAA-84-3-14).
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| Grand Gulf (Entergy)is the only utility that has engines with the 1-7/8" bolts still in use.The summary of this work d is that as long as the bolts are properly torqued the rods will perform with no problems.Oil Analysis should continue to be performed as this will provide indication of premature bearing wear or bearing problems as babbit will be recognizable in the oil.In addition, any significant fretting of the mating surfaces of the connecting rod will be evident as well.This will be detectable as ferrographic analyses is performed for the oil samples indicating the types of metals in the oil.Also, vibration measurements should continue as well as operation monitoring which will also provide an indication of potential problems with this 5-3-93 16 Rev.1
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| component.
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| A total of 42 connecting rods have been completely disassembled and subjected to the PMs described above.A total of 1776 bolts have been checked for proper tension during the time since DR/QR.These inspections have revealed no problems and these rods continue to provide good service.'ased on the above, the Owner's Group recommends that further connecting rod disassembly to perform the inspections above on a particular time frequency is not warranted.
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| However, it is the recommendation of the Group that as rods are removed from service for any reason, they should be subjected to the PMs in Appendix A as a good practice but this should not be a requirement.
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| Connecting rods in service at most 0 of the utilities have recorded sufficient hours producing a sufficient number of cyclic loadings to demonstrate unlimited fatigue life for connecting rod assembly.In addition, no problems have been found with connecting rod bearings and inspections have revealed normal wear.-The engines at Grand Gulf are currently limited to 185 BMEP.This derating reduces the stresses associated with fatigue cracking of connecting rod bolts and/or the link rod box and bolts.Based on past positive inspection results and engine derating, the recommendations for 1-1/2" bolting then applies to Grand Gulf as well.5-3-93 17 Rev.1 This item has been covered in Section 3.5, Connecting Rods and in a previous license submittal currently under review with the NRC.The previous submittals are documented in letters to Mr.Om Chopra dated October 31,1991 and supplemented February 27, 1992 from Messrs JB George and RD Broome.Therefore this item is addressed by reference to previous submittals.(Copies of these submittals are included as Appendix C and D.)5-3-93 18 Rev.1
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| The high pressure fuel injection tubing preventative inspections are listed in Part B of Appendix A.The PMs do not require teardown to perform;however, the requirement to eddy current the non shrouded tubing prior to bending does result in considerable cost and delay of replacement tubing.Use of shrouded tubing has been approved by the Owners Group and the vendor to provide protection of leakage that would potentially result in a fire hazard.Fire hazard and personnel safety are the primary concerns with failure of this component.
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| The review of this component during the DR/QR process revealed that failures had occurred at Shoreham and Grand Gulf Nuclear Stations.A 10CFR21 notification was issued on 7/20/83 by TDI alerting Owners and the NRC of the condition and identified that the cause of the failure stemmed from a draw seam that acts as a stress riser on the inner surface of the tube.One of the points stated is that a draw seam is induced during the drawing phase of the manufacturing and generally will extend over most of the length of the tube and be readily detectable.
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| The design review noted that the tubing is acceptable as long as no preexisting flaws greater than a depth of.0054'xisted.
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| This prompted the recommendation to eddy current the tubing prior to bending.The reason for the concern was to prevent leakage that, could potentially result in a fire and for personnel safety.5-3-93 19 Rev.1
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| The tubing is visually inspected for leaks during each engine run.Since the DR/QR effort, four tubing failures have occurred.This inspection has resulted in hundreds of inspections of this component.
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| Most engines are now equipped with the shrouded tubing which permits the leak check to be performed by removal of a plug.Shrouded tubing is a double wall tube that contains the high pressure fuel spray in the event of a leak and prevents fire and hazards to personnel~The Owners Group recommends that visual inspections for leaks continue during the engine runs.Any problems should be readily identified by this process.In addition, replacement tubing must be shrouded.Further, because of its double wall design, use of shrouded tubing would eliminate the need to eddy current this tubing and this requirement should be deleted for shrouded tubing.5-3-93 20 Rev.1
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| The site specific preventative inspections are listed in Part B of Appendix A.All of these inspections require disassembly to perform.These inspections have been performed on a per PM basis as detailed in Appendix B.In August 1983, the crankshaft in the EDG 102 engine at the Shoreham Nuclear Power Station fractured during plant preoperational tests.The fracture occurred at the crankpin journal of cylinder No.7 and involved the web connecting the crankpin to an adjacent main bearing journa.Following this failure, several cracks were discovered in the crankshafts of the other two TDI diesels at Shoreham.These crankshafts were found to be deficient and were replaced with a different design that increased the diameter of the crankpin from the original 11" to 12".The replacement crankshafts were analyzed by the Owner's Group and by NRC and found acceptable for use.The EDG engines at the River Bend Nuclear Station have crankshafts of the same dimensions as the replacement shafts at Shoreham.However, the generators and flywheels differ between the two installations, resulting in differences in crankshaft torsional stresses.Also the fillet radii at Shoreham are shotpeened while those at River Bend are not.The review and inspection made by the Owner's Group found that there were no relevant indications in the oil holes of the crankpins.
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| However, the analysis revealed that crankshaft torsional stresses in the Shoreham engines at an operational load of 3300kw was 5-3-93 21 Rev.1 equivalent to the torsional stresses in the River Bend engines at an operational load of 3130kw which accounts for the differences in the torsional systems.Therefore, the River Bend engines have been derated for nuclear operation to 3130kw with the crankshafts that are currently installed.
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| No indications or other problems have been found by the inspections and the shaft has accumulated sufficient loadings to demonstrate unlimited fatigue life.The inspections that have been performed are in accordance with Appendix A and has been performed in number as indicated in Appendix B.No indications or problems have been found with this component.
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| Based on the positive inspection results and on the previous design review, the Owner's Group recommends that future inspections of the crankshaft are not warranted as required by the DR/QR as long as the engine is operated at loads below 3130kw.Should this load be exceeded for an extended period, the engine should be removed from service and the crankshaft inspected in accordance with current procedures.
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| Should no indications be found, the unit may return to service and no further inspections made unless the load limit is again exceeded.5-3-93 22 Rev.1 The crankshaft preventative inspections are listed in Part B of Appendix A.All of these recommendations require teardown to perform.The crankshafts for the DSRV-16 engines have a crankpin diameter of 13'nd the overall crankshaft length is approximately 20 feet 7 inches.These engines have eight crank throws with 16 pistons driven by 8 articulated connecting rod sets.Differences in the generators and flywheels.at the various installations result in differences in the torsional stresses.Therefore, each of the crankshafts at each installation were individually evaluated.
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| The results of these investigations produced similar results.The results are that the component is adequate for its intended service at full rated load and the 110%rated overload.Extended operation at speeds at or near the fourth order torsional vibration frequency modes should be avoided.(These speeds have been documented in Owner's Group site specific reports.)In addition, the engine should not be operated for extended periods in an unbalanced condition.
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| 5-3-93 23 Rev.1
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| Appendix B indicates how many times each of the inspections detailed in Appendix A have been performed.
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| None of these inspections have produced any indication of cracking and most of the engines have operated above the period that would subject the crankshafts to a number of cyclic loadings to demonstrate unlimited fatigue life.Based on the positive inspection results and the original design review, the Owner's Group recommends that future inspections as required by the DR/QR are not warranted and should be eliminated.
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| The manufacturer has reviewed this conclusion and is in agreement with it.5-3-93 24 Rev.1
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| The jacket water pump preventative inspections are listed in Part B of Appendix A.All PM recommendations require teardown to perform.The pumps for the DSR-48 and DSRV-16 engines are somewhat different.
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| The original design of the pump for the DSR-48 engines had two failures on the engines at Shoreham that resulted from a fatigue failure originating at the gear/shaft keyway.This pump was subsequently redesigned.
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| The new design removed the keyway on the impeller end and changed the impeller material to ductile iron.The impeller is now driven through its interference fit on the shaft..This later pump design is installed on the engines at River Bend.Pumps for the DSRV-16 engines were reviewed as a result of the problems with the model DSR-48 engines.At the time of the review, there were no reported failures and the design review concluded that the pumps were capable of serving their intended function with no problems.Since the DR/QR, there are.reports of drive gear failures on non-nuclear engines and these have been addressed by the manufacturer through 10CFR21.There have been no problems with the original concern related to the shaft, keyway and impeller.5-3-93 25 Rev.1 There have been no failures of jacket water pumps in nuclear service since the design changes made as a result of the DR/QR review.Inspections performed as outlined in Appendix B reveal that some pitting of the gear teeth on DSRV-16 engines has occurred during the pump operation.
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| The resolution of this issue will be dealt with through the 10CFR21 process.Additional problems related to the shaft, impeller and keyway have not been identified.
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| Based on the positive inspection history, future inspections of this component on a time dependent basis as a requirement is not warranted.
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| However, should the pump be removed or an engine overhaul be necessary, the pump should be inspected per the existing guidance.5-3-93 26 Rev.1
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| The block preventative inspections are listed in Part B of Appendix A.Specifically, PM recommendations P 1, 2, and 3 require teardowns.
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| The PM for the cylinder liners does not require a teardown but removal of the injector for access to the liner is required for visual inspection.
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| The cylinder block provides support for the upper-engine components and contains passageways for the engine cooling water.The block is subjected to both mechanical and thermal stresses and is a grey-iron casting.Although the cylinders in the DSRV-16 engines are arranged in two banks while those in the DSR-48 engines are in a single bank, the two configurations do not differ in block top thickness, cylinder head spacing, upper support of the cylinder liner, and the stud boss region that anchors the cylinder head studs.1 Minor design changes have been incorporated as a result of DR/QR to reduce the protrusion of the cylinder liner collar above the block top and to increase the cold radial clearance between the cylinder liner and the block, thereby reducing stresses in the block top.Cracks have been reported in cylinder blocks of both DSR-48 and DSRV-16 engines in nuclear and non-nuclear applications.
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| A thorough design review of this component was completed during the initial DR/QR review.The results of that review were that some of the castings made during the period may contain Widmanstaetten graphite which is an inclusion that weakens the grey iron casting.It was shown that blocks containing this material have a greater potential for crack development.
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| However, it was also shown that should these cracks develop, regardless of the cause, that the block would continue to perform its intended design function and that the cracking would potentially produce a flow path for water to the block exterior.A cumulative fatigue 5-3-93 27 Rev.1 usage index formula was created and an inspection frequency was established based on that usage.Further, it was noted by the Owner's Group and by the NRC that this analysis was conservative and that"If cumulative results of these inspections over several power plant fuel cycles show that one or more of the inspections reveal nothing of significance, the scope and frequency of the inspections could be reconsidered.'Source:
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| PNL-5600)Block top inspections have been performed in accordance with the numbers outlined in Appendix B.Note that some of these inspections are being performed on a partial basis;however,,none of the inspections (including those of blocks with widmanstaetten graphite)have revealed any cracks.In addition, no significant liner wear or indications have been found.Based on the positive inspection results, the Owner's Group recommends that future block top inspections be performed when a head is removed for other reasons for plants that have blocks with no widmanstaetten graphite.For those sites having blocks with widmanstaetten graphite, the recommendation is to perform a visual inspection of the block top under strong lighting during a test run once a refueling cycle.Should cracks be found, the engine should be evaluated for continued service and a more detailed inspection performed at the next available refueling outage.,The manufacturer has reviewed these"conclusions and" agrees with them.5-3-93 28 Rev.1, l
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| The scope of this review will be limited to Type AE piston skirts.These are the only type skirts currently used in nuclear applications.
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| Recommendations for other type piston skirts are not addressed by this submittal and previous findings by the Owners Group and NRC remain in effect.The piston skirt preventative inspections are listed in Part B of Appendix A.Specifically, the PM listed requires disassembly of the engine~The design review of this component revealed that design stresses are within the allowables and that based on experimentally measured data, neither crack initiation nor propagation is expected to occur.The AE skirts were tested and validated during DR/QR.The purpose of this validation was to determine the calculated fatigue life of this component.
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| Following the validation, a detailed inspection was made of these skirts with no problems found.These skirts have previously been approved by NRC for use at the rated engine loads and all engines in current service have been equipped with these skirts.Thirty nine piston skirts have been removed and inspected in detail.No problems have been found with this component and these skirts continue to provide good service.See Appendix B for the numbers of inspections.
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| 5-3-93 29 Rev.1
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| Based on the positive inspection, results of this component and documented design quality, further inspections under the DR/QR program for this component are not required unless a piston is removed from the engine for some other reason.Research identified by this report regarding aging of this component has identified unnecessary teardowns as a real source that contributes to unreliability.
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| 30 Rev.1 The cylinder head preventative inspections are listed in Part B of Appendix A.Specifically, PM 1 requires teardown.LuJS round The basic cylinder head configuration is common to all TDI DSR-48 and DSRV-16 engines.However, during periods of manufacturing, TDI made changes to manufacturing practices, quality control, and design.The heads manufactured have been categorized into three groups: those cast prior to October 1978 are referred to as Group I, those cast between October, 1978 and September, 1980 are Group II, and those cast after September 1980 are Group III.Cylinder heads in Group I and II are subject to core shift, inadequate control of solidification, and inadequate control of the Stellite valve seat weld deposition process.In addition, Group I heads are not stress relieved and are subject fatigue crack growth in thin areas.Heads in Group III are much less prone to all of these problems.It should be noted that heads from all three groups remain in service.Casting defects were found at Shoreham, Grand Gulf, Catawba, and Comanche Peak during the DR/QR process.The net result from the design reviews and flaws, would have been to allow leakage of jacket water to the exterior of the head or to the cylinder.Exterior leakage is of no real concern from a reliability standpoint, but leakage into a cylinder can result in major engine damage.As a result, the Owner's Group recommended that the engine be barred or air rolled prior to starting with the air start cocks open to detect any potential leakage.Also, the manufacturer has changed its weld repair procedure to correct previous problems with weld repairs in the fire deck region of the head.5-3-93 31 Rev.1 II Inspections have been performed as detailed in Appendix B.Indications were found on the exhaust valve stem during RFO 4 at River Bend.The indications were caused by a sharp chamfered edge on the rocker arm swivel pad and are direct result of excessive valve lash.The root cause of the excessive valve lash has been attributed to back pressure in the exhaust system during the start sequence of the engine..The chamfered edge on the swivel pad was removed by machining.
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| An improved swivel pad has been developed by the vendor.The water leak found a River Bend has been investigated by the owner of the engine and the manufacturer.
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| The leak was caused by a thin wall section in the cylinder head casting near a tapped bolt hole.This defect was reported to the NRC under 10CFR21 by the manufacturer.
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| The manufacturer's recommended corrective actions include inservice repair techniques and a permanent repair that will be made during an overhaul of a cylinder head.Based on the above positive inspection results, PM recommendation 1 is not warranted and should be discontinued.
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| It is the recommendation, of the Owner's Group that pre run air rolls and inspections for leaks, prior to any planned start or as dictated by plant configuration, continue to preclude a leak from resulting in major engine damage.Any other type of degradation that could occur will become evident during compression checks, with exhaust temperature monitoring, and monitoring jacket water standpipe level for losses.The previously referenced NRC NUREG reports again point out that major disassembly, such as head removal, may result in increased unreliability and unavailability.
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| 5-3-93 32 Rev.1 8
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| The scope of this review will be limited to push rods of the friction welded design as this is the only design currently in use.Other designs are not addressed by this submittal and the previous recommendations made remain valid.The push rod preventative maintenance inspections are listed in Part B of Appendix A.The recommendation requires an engine teardown.Design analysis of this design showed that potential buckling under the loads to be imposed was not a 0 concern.Metallurgical evaluations showed no major discrepancies in the chemical composition, hardness, or microstructures of any components.
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| A fatigue crack growth analysis showed that, under cyclic loading, no potential fabrication cracks are expected to propagate in either the main or intermediate push rods using this design.A fatigue test that included 10 to the seventh cycles compressive load from zero load to a vaiue approximately 25%above the maximum theoretical service load, was also conducted.
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| No cracks or indications were found.Over 900 push rods have been inspected following extended service and have shown no problems.5-3-93 33 Rev.1 45 Based on the positive inspection results and the conservatism of the design, future inspections as required in the DR/QR are not warranted and the Owner's Group proposes to delete this item.Should these components be removed for other reasons, Owner's may elect to conduct these inspections depending on the service life'and reasons resulting in engine teardown.5-3-93 34 Rev.1
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| Studs in nuclear service engines have been replaced with the latest design and installed in accordance with the procedures recommended by the manufacturer.
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| This issue was closed in the original NRC review resulting in no preventative inspections for this component.
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| There has been nothing found in subsequent operation of these engines to change this finding.5-3-93 35 Rev.1
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| The rocker arm preventative maintenance inspections are listed in Part B of Appendix A.The inspection is a'one time'nspection and has been completed for all engines.The inspection does require teardown.The review during the initial DR/QR revealed that capscrews failures had occurred on an isolated basis.The cause of the failures was due to insufficient preload on the capscrews.
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| This failure history resulted in the requirements outlined under the PM Recommendations.
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| The Owners'roup performed a detail design review of the component which calculated appropriate resultant stresses, endurance limits, and looked at the material requirements to determine that the material is suitable.Subsequent to incorporating the torque requirements there have been over 500 inspections of this component with no major problems found.River Bend has reported two pop rivets missing;this was disposition as not being a problem as lubrication could still get to the needed areas.This inspection is currently performed only on reassembly of the rocker arms.This should continue when the rocker arm is removed from service for any reason.5-3-93 36 Rev.1
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| The turbocharger preventative inspections are listed in Part B of Appendix A.Specifically, PM Recommendations 2,4,5, and 6 require teardowns.
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| These inspections have been performed on a per PM basis as detailed in Appendix B.These turbochargers typically see operation hours of approximately 500 hours per 5 year interval.Turbocharger performance directly affects the design rating of the engine.During the DR/QR review,, several bearing and lubrication problems were identified.
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| In addition, there was a concern dealing with the potential for damage of the rotating vane group due to ingesting fragments of material, specifically bolts and blades from the stationary vanes assembly that had failed due to fatigue loadings.The response to these concerns were answered as follows: 1)Lubrication and Bearing Wear The Owners Group recommended modifications to install the drip and full flow prelubrication system to provide an oil film to the turbo bearings that would drain away during standby and that this system should be activated to prelube any planned start.This recommendation has been implemented by the Owners.In addition, oil sampling was recommended as a means to detect significant bearing wear.PM items 1,3 and 4 relate specifically to this concern.5-3-93 37 Rev.1
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| 2)Potential For Damage to Rotating Vanes During DR/QR review, it was learned that at least one engine in nuclear service had experienced loss of a stationary vane, and from the rotating vane group', bolting material.The net effect of this event was that no significant damage occurred, and the turbocharger performance was not effected.This is documented in NUREG 1216 as referenced.
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| This issue resulted in PMs 1,2,5,6, and 7.PM items 2,5, and 6 require teardown.Appendix B shows the number of times that each PM has been performed.
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| The results of the inspections have shown that in most cases the oil system modifications have resulted in eliminating significant bearing wear.In a case where some moderate amount of wear was found, this was detected via the oil monitoring trends.There is no case where failure occurred due to excessive bearing wear.Since the original discovery of stationary vane failure and passing of this material through the rotating vane group, three other occurrences have occurred with the same result that the vane fragments passed through the rotating vane group with no significant damage and no significant degradation of turbocharger performance.
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| Based on the positive inspection results described and detailed in Appendix B, PM items 2,4,5,and 6 are not required.PMs 1,3 and 7 will be continued as a part of the future maintenance program.PMs along with results from the oil sampling program and exhaust temperature trending will show degradation in turbocharger performance ancVor indicate increased bearing wear or vane damage.This will permit the 38 Rev.1
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| utility to evaluate and take actions necessary to correct the problems.Should the turbochargers be removed from service for any reason, the PM recommendations 2,4,5, and 6 should be considered as good maintenance practice.5-3-93 39 Rev.1
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| ~~
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| System unreliability for the TDI EDGs has been consistent with the industry median for the period since DR/QR was completed.
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| A review of the INPO data for the period 1/90-12/92 gives a median unreliability I for TDI EDGs as 0.0094.This is well within the expectations of NRC guidance for either a plant needing a 0.0250 unreliability or 0.050 unreliability as directed by Station Blackout and equal to the current industry median.Some unreliability has been attributed to the engine teardowns and inspections.
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| Industry experience indicates that elimination of frequent teardown and inspections has resulted in an additional i decrease in unreliability.
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| The following table lists the INPO data furnished for unreliability:
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| INPO UNRELIABILITY VALUE FOR TDI DIESELS 1/90-12/92 0.0000 0.0000 0.0000 0.0000 0.0000 0.0000 0.0103 0.0109 0.0085 10 0.0250 0.0313 Entire Page 5-3-93 40 Rev.1 Revised 12 0.0336 0.0333 14 0.0364 15 0.0115 16 0.0450 17 18 0.0000 0.0000 MEDIAN 0.0094 A review has been made by the utilities having engines 12, 13,14, and 16 as to the cause of the higher unreliability and what is being done to improve the status.The findings are as follows: 1)Some of the INPO numbers have reporting errors and some of these numbers are really better than reported.These utilities are working with INPO to resolve these problems;2)some utilities have reviewed the failures that were reported as being valid and feel some of these'failures" were conservatively reported and are reviewing the data to determine if the number of valid failures reported is accurate, and 3)in the cases where the numbers are accurate, recent improvements have been noted and the individual utilities are working to address improvements in the program.It should be noted that some failures are hard to detect;for example, a field breaker failure did not show up until the monthly test run.For this item investigation showed that it had failed prior to the run and significant additional time had to be added per the INPO guidelines for the diesel being out of service.It is concluded from the data provided that the unreliability of the TDI EDGs is within the bounds and expectations of the regulatory guidance and other diesels within the nuclear industry.Entire.Page Revised 41 Rev.1 System unavailability has been reasonable for the TDI Enterprise engines since DR/QR as measured by the INPO indicators.(The INPO Indicators are based on unavailability during power operations.)
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| The Industry median (for all engines)is 0.0182.The median for the TDI engines is 0.0177.The following table gives the unavailability three year values for the TDI engines in service for the period 1/90-12/92:
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| INPO UNAVAILABILITY VALUES FOR TDI DIESELS 1/90-12/92 0.0196 0.0105 0.0106 0.0134 0.0141 0.0190 0.0318 0.0348 10 12 13 0.0165 0.0413 0.0343 0.0405 0.0432 Entire Page Revised 5-3-93 42 Rev.1
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| <4 16 17 0.0650 0.0125 0.0160 0.0101 18 0.0110 MEDIAN 0.0177 Recent industry events have focused more attention on unavailability of safety related systems especially the diesels during modes of operation other than full power operation.
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| The above numbers reflect standard industry practice of determining unavailability during periods of power and non power operation.
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| Review of data from utilities involved with this submittal, accounting for unavailability during outages would substantially increase the median.As an example, assume an outage of 6 weeks for an overhaul on a 0 diesel.This would result in 1008 hours out of service and if this were translated, would result in an unavailability of 11.5/o for the year without any other unavailability factored in.In review of data from utilities supporting this licensing request, unavailability numbers in the range of 10-15/o (on a per engine basis)would not be uncommon with outage out of service time figured in.By not performing major teardowns, out of service durations during outages could be shortened to two weeks and significantly reduce this unavailability.
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| The numbers presented above also include outage time related to raw water and other systems that contribute unavailable time to the engine;not Just the engine itself.In the case of any engine having an unavailability of greater than 0.4, a review has been made and the unavailability for these engines is improving.
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| Entire Page Revised 5-3-93 43 Rev.1 i
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| HOW TO USE THIS APPENDIX Appendix A is a reproduction of Appendix II, Revision 3 of the TDI DR/QR report and is placed here for the convenience of the user.Appendix A provides, for information, the specific Preventative Maintenance (PM)Recommendation that is currently performed on the Enterprise engines.These recommendations describe the inspections performed as well.Appendix B is a tabular listing of the collective results of the inspections performed that are listed in Appendix II from the utilities listed in Section 2/0.Each table in each Appendix is listed by Component number.Thus, one may look for an item such as Connecting Rods in Appendix B to see the results of an inspection.
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| The component number for Connecting Rods is 02-340A/8 which is found in the text by the performed.
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| Som recommendations.
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| section number.If one were to need to know what inspection was performed to obtain these results, then one would refer to Appendix A using this component number to find a description of the inspection e components have multiple inspections listed in numerical order under PM New Page 5-3-93 44 Rev.1
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| THE TRANSAMERICA DELAVAL, INC.OWNERS GROUP LICENSING CONDITIONS APPENDICES TABLE OF CONTENTS APPENDIX A PART A-Overview and Definitions.
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| Operating and Standby Surveillance Parameters.
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| PART B-DR/QR Appendix II, Part B and Part D, Selected Pages From Site specific Matrix APPENDIX B Results of Inspection For TDI Diesel Generator Phase I Components.
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| APPENDIX C Position Paper on Radiograph Requirements For Connecting Rod Bearing Shells APPENDIX 0 Position Paper on Radiograph Requirements For Connecting Rod Bearing Shells APPENDIX E Safety Evaluation Report on Radiograph Requirements For Connecting Rod Bearing Shells 5/3/93 Rev.1 k
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| APPENDIX A PART A 5-3-93 Rev.1 11 TDI OWNERS GROUP APPENDIX-II GENERIC MAINTENANCE MATRIX PART A OVERVIEW AND DEFINITIONS OPERATING AND STANDBY SURVEILLANCE PARAMETERS I
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| RI TDI OWNERS GROUP A BAND V I N PR RA PPE IX-I D The purpose of this appendix is to provide the TDI Owners with a set of maintenance and surveillance recommendations for diesel generator components which have been developed by TDI and/or the Owners Group as a result of the overall Owners Group Program and including subsequent testing and insp'ections performed following the review conducted by the original program.This appendix is intended to enhance the existing TDI Instruction Manual, Volume I and Volume IH, which will maintain the qualification of the diesel generators for the life of the plant.During the implementation of the Owners Group Program Plan, the Owners Group Technical Staff reviewed many sources of information regarding the maintenance and surveillance for the diesel generator components identified in this appendix.These sources included TDI Instruction Manuals, Service Information Memos (SIMs), and TDI correspondence on specific components.
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| The basis of this matrix is formed by the following:
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| ~Owners Group Technical Staff review of TDI Instruction Manuals, SIMs, and TDI correspondence on specific components.
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| ~Technical Staff input regarding the adequacy of recommendations found in sources mentioned above.~Additional maintenance recommendations identified during the DR/QR review and from 10CFR21 reports and operating experience at nuclear plants.~Results of subsequent testing and surveillance (i.e., Shoreham EDG103 750-hour endurance run and subsequent engine teardown)performed following the review conducted during the original program.~Additional review by the Owners Group representatives.
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| It should be noted that this revision in some cases modifies the original program results based on this additional information and review.II-A-1 Rev'ision 4
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| LT AND L I Proper maintenance is important in ensuring long, reliable and satisfactory service of the emergency diesel generators.
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| Maintenance work, in order to be effective, must be carried out thoroughly and regularly.
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| It is for these reasons that a detailed schedule of maintenance service has been laid out by the Owners Group for the TDI Diesel Generators.
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| This schedule should be followed as closely as the operating conditions will permit.This maintenance service as specified supersedes previous general maintenance requirements, but is separate and does not supersede Quality Revalidation and/or modifications previously recommended.
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| The schedule details specific components requiring maintenance on a regular basis.This schedule separates the maintenance activities into frequencies as set forth in the subsequent list of definitions.
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| Inspections, as outlined in this maintenance schedule, are to be performed and parts refurbished or replaced as required by the program or deemed necessary by the inspection.
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| Any adverse findings shall be investigated and corrective action, including amended inspection frequencies, shall be implemented unless sufficient justification is present to do otherwise.
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| This generic matrix, Parts A, B, C, together with Part D entitled"Site-Specific Maintenance Matrix" and the sources defined in Section II form the TDI Maintenance Program.Note that component numbers used in the generic matrix are for Texas Utilities'omanche Peak Steam Electric Station-Unit 1.Part E provides a cross reference to identify corresponding components for other engines.Also note that a blank in the cross reference signifies that a component is not on a particular engine and, thus, that the Owner would not perform that maintenance item.Tables 1 and 2 of part A provide engine operating and standby surveillance parameters and standby surveillance parameters and frequencies.
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| It is recommended that the utility address these tables in its operating and monitoring program.Table 1 addresses operating parameters and is not duplicated in the maintenance schedules; these parameters are to be recorded and/or checked during the monthly testing and any other period of operation.
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| Table 2 addresses the standby parameters that occur on a daily frequency and are not duplicated in the maintenance schedules.
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| IV.DE I Overhaul Frequency a)A complete engine teardown inspection will be performed every 10 years.The utility has the flexibility to inspect one engine/reactor unit at the End of Cycle (EOC)prior to 10 years and the other engine at the EOC following 10 years.Alternately for PWR units, the inspection may be performed coincident with the 10-year reactor vessel inservice inspection, This will permit both engines for each unit to be disassembled in parallel since one engine will not have to remain in service with the reactor vessel off loaded.(For reactor units having three engines, the inspections are to be carried out as above with the third engine to be inspected at the second EOC following 10 years.)The 10-year interval will typically be taken from issuance of the Low Power H-A-2 Revision 4
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| Operating license or from subsequent teardown and inspection for plants already in operation.
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| b)A one time inspection will be performed at the EOC closest to five years.For a unit, one engine may be inspected at the EOC prior to five years and the other at the EOC after five years to minimize plant outage length.(For reactor units having three engines, the inspections are to be carried out as above with the third engine to be inspected at the second EOC following five years.)This inspection will generally involve the same components as the 10-year teardown;however, only a sample of items for some components will be inspected as set forth in the maintenance schedule.During this five-year inspection, any significant adverse findings of a particular component will result in an inspection of all such components of that engine to determine any adverse trends.Favorable fiindings will result in reassembly of the engine for service.3.Daily Frequency-To be performed once per day.Monthly Frequency-To be performed once in a month;normally during, before, or after test run per plant Technical Specifications.
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| V 4.EOC (End of Cycle)-To be performed once during outage for refueling.
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| 5.Alternate EOC-To be performed once every other outage for refueling.
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| ~~6.Five Years-To be performed once at the EOC occurring nearest to the end of a recurring five-year period or at the EOC midway between the one time EOC 2 inspections and the first overhaul inspection and subsequently midway between each overhaul.7.As Required-To be performed as often as good maintenance, site procedures, manufacturer's recommendations, or experience dictate as determined by site personnel.
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| 8.Maintenance
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| -Monitoring and/or surveillance on a periodic frequency to assure the component will perform its intended function in a safe reliable manner.9.Accessible
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| -Any item on which the required function can be performed without disassembly of an engine component.
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| Removal of defined access cover is~n considered disassembly.
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| 10.Appropriate NDE-Nondestructive.
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| examination selected by site personnel that is most suitable to obtain the information sought by an individual inspection item;choice of NDE shall be made to assure that the technique will detect indications consistent with the acceptance criteria.II-A-3
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| ~TBL~1 Diesel Engine Operating Surveillance Parameters and Frequency 1)Lube Oil Inlet Pressure to Engine 2)Lube Oil Filter Differential Pressure 3)Lube Oil Temperature (engine inlet and outlet)4)Lube Oil Sump Level 5)Turbocharger Oil Pressure 6)Fuel Oil Filter Differential Pressure 7)Fuel Oil to Engine Pressure 8)Fuel Oil Day Tank Level)9)Jacket Water Pressure (engine inlet 10)Jacket Water Temperature (in, out)Engine Cylinder Temperature Exhaust-All (if temperature in any one cylinder exceeds 1050', refer to MP422/023 Item 7).FRE E Y Log hourly Log hourly Log hourly Log hourly Log hourly Log hourly Log hourly Check hourly Log hourly Log hourly Log hourly 12)Manifold Air Temperature (RB, LB for DSRV Engines)13)Manifold Air Pressure (RB, LB for DSRV Engines)14)Starting Air Pressure (RB, LB for DSRV Engines)15)Crankcase Vacuum 16)Engine Speed 17)Hour Meter 18)Kilowatt Load 19)Visual Inspection for Leaks, etc.Log hourly Log hourly Check hourly Log hourly Log hourly Log hourly Log hourly Check hourlyII-A4 Revision 4
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| ~TAB 2 Diesel Engine Standby Surveillance Parameters and Frequency MP 1)Lube Oil Temperature (in, out)2)Lube Oil Sump Level 3)Check Operation of Lube Oil Keep-Warm Pump Motor 4)Monitor Lube Oil Keep-Warm Strainer and/or Filter Differential Pressure Log daily Log daily Daily Daily 5)Perform a visual inspection for leakage of the Lube Oil Heat Exchanger.
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| Verify that no leakage through the leak-off ports of the lantern~ng is present.Daily 6)'uel Oil Day Tank Level 7)Jacket Water Temperature (in, out)Log daily Log daily 8)Perform a visual inspection for leakage at packing for Jacket Water Heat Exchanger whenever the engine is in the emergency STANDBY mode.Verify that no leakage through the leakwff ports of the lantern ring is present.Daily 9)Verify proper governor oil level 10)Verify proper oil level of generator pedestal bearing Daily Daily 11)Starting Air Pressure Log daily 12)Drain air receiver float traps and/or drain Starting Air Storage Tank and monitor the quantity of moisture produced.If quantity of moisture is excessive, correct immediately.
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| Daily 13)Check Operation of Compressor Air Traps 14)Test Annunciators 15)Check Alarm Clear Daily Before Engine Operation Before Engine Operation II-A-5 Revision 4
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| ~TB,g 2 (cont'd)Diesel Engine Standby Surveillance Parameters and Frequency 16)Inspect for Leaks Daily 17)Visually inspect intercooler for external leaks including intake manifold drain connection.
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| DailyII-A4 Revision 4 APPENDIX A PART B 5-3-93 Rev.1
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| TDI OMNERS GROUP APPENOIX'II GENERIC MAINTENANCE MATRIX PART 8 PHASE I COMPONENTS
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| ~'p+-g A~')"''.~s." II~l, I' GH)ERIC NQMKNCE MNII-PESE I Co@anent Coupon ent Identification HP-022/23 Turbocharger PH Reconnendation I.Measure vibration and check with baseline data.2.Inspect>npeiler/diffuser and clean if necessary.
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| 3.Measure rotor end play (axial clearance) to identify trends of increasing clearance (i.e..thrust bearing degradation).
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| i.Perforn visual and blue check inspections of the thrust hearing.5.Disassenble, inspect, and refurbish.
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| 6.The nozzle ring conponents and inlet guide vanes should be visually inspected for nissing arts or parts showinq dis-ress.If such conditions are!Qt Monthly BOC EX 5 Year 0/erhaul X Connents To be acconplished after obtaining stable exhaust tenperature conditions.
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| Rev)ew thrust beanng axial clearances after inspection to deternine if a trend exists.hny trend toward increasing axial clearance could signify thrust bearing degradation.
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| Note: Thrust bearing inspection should also be perforned after experiencing each 40 nonprelubed (autonatic) fast starts.In addition.a one-tine inspection should be conpleted after the first 100 engine starts.Note: During reassenbly, ensure that capscrews are properly installed with the reconnended torgue.If gR inspection was perforned prior to accunulating significant hours (i.e., the nunber of hours accunulated during plant preoperational testing, approxinately l00 hours), the turbocharqers should be reinspected at the next EOC.Or perforn a visual inspection on one turbo-charger per nuclear unit at each EOC.Revision 3 i
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| ConIon eat Coaponent Identification N Reconnendation GENERIC lhiH?ENATE MAIBII-PHLSH I Alt Rethly EOC EOC 5 Year OIerhaul 02-305h Base Assenhly noted.the entire ring assenbly should be replaced.7.Monitor inlet tenperature to ensure gas tenperature does not exceed nanufacturer s recon-nendation of 1200'F if exhaust tenperature for any cylinder exceeds IOSO'F (Refr: Table I).1, Perforn a visual inspection of the base.The inspection should include the areas adjacent to the nut pockets of each bearing saddle and be conducted after a thorough Mipe dovn of the surfaces, using.good lighting.Any turbocharger in which nozzle ring anonalies are found is to be reinspected at the next EOC.Note: Discontinue inspec-tion with appropriate re-design.Monitoring nay be per-forned using pernanent in-line therrecouple.
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| strap-on thernocouple.
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| heat..gun.or other suitable neans that has been appropriately tested and calibrated per plant pro-cedures.Note: Also perforn noni-toring any tine the engine operates in an unbalanced condition.
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| Note: Any cracks detected nust be investigated further before the engine is allotted to return to service.The nating sur-faces of the base and cap shall be thoroughly cleaned uith solvent before any reassenbly.
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| Perforn on EOC basis for 3 cycles.then overhaul provided there are satisfactory results.Note: 3 EOC inspections nay be elininated by petforning a netal analysis te confirn consistent to class 40 grey iron requirenents; perforn-ing anaEysis does not elini-nate need for overhaul in-spections.
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| II+2 Revision 3
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| GENERIC HAIETEEAKE KATRII-PHASE I Coupon eat Coaponent Identification 02-305C Main Bearing Caps-Studs and Nuts PH Becoaaendation 1.The nating surfaces at the hearing cap/saddle interface should be inspected when disassenbled to ensure the absence of surface inperfec-tions that night prevent tight boltup.Note: Upon renoval of hearing caps, clean sating surfaces wiPh a solvent pnor to reassen-bly of the caps to the base.Alt Monthly EX EOC 5 Tear Overhaul Comauts 02-310A 02-315A 02-315C Crankshaft Cylinder Block Cylinder Liners See site specific reconnenda ions See site specific.reconnendations 1.Perforn a visual inspection of liners for progressive wear.To be perforned for one EOC following piston renoval;then discontinue until next piston renoyal.Boroscopzc)nspection is acceptable if heads are not renoved.Con-lete TOI Inspegtion and aintenance Record Porn No.315-1-1 as applicable, TDI.Instruction Hanual, Voiune l.Section 6.02-340A/B Connecting Rods.Bushings and Bearing Shells (Generic)l.Inspect and neasure all con-necting rod bearing shells to verify lube oil naxntenance, which affects wear rate.X Conplete TOI inspection and Haintenance Record Porn No.3i0-I-I as applicable, TDI Instruction Hanual.Yolune I, Section 6, appendix Ill for clearance values.Per-forn inspection at 5 years, on item accessible, consistent with iten 2 of this conponent.
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| II-B-3 Rovision 3
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| neat Coupon ent r, Identification PH Recoaueudation GMRIC HHHYERQCE Hl?RIX-PHBSE I Ht Honthly Em'.EOC 5 Year Srerbaul Coenents 02-340 h/B Connecting Rods, Bushings DSRV's and Bearing Shells only 3.5.6.Inspect and neasure the connect>ng rods.Hote: Perforn inspection and neasure four connecting rods for DSRVs and two for DSRs at randon at one tine 5-year inspection.
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| Perforn an x-ray exanination on all replacenent bearing shells to acceptance criteria developed by Owners Group Technical Staff.All connecting rod bolts, nuts, and washers should be visually inspected, and danaged parts should be replaced.The bolts should be HT inspected to verify the continued absence of cracking.Ho detectable cracks should be allowed at the root of the threads.During any disassenbly that espokes the inside dianeter of a rod-eye (piston pin)hushing, the surface of the hushing should be LP inspected to verify the continued absence of linear indications in the heavily loaded zone width+I-I5 degrees of the botton dead-center position.Heasure the clearance between tbe link pin and link rod.This clearance should be zero;i.e., X Conplete TDI Inspection Haintenance Record Eorn Ho.340-2-1.-2 as applicable, TDI Instruction Hanual, Volune I.Section 6.This is to be perforned prior to installation of any replacenent bearing shells as required.X Perforn inspection at 5 years, on itens acces-sible.consistent with Iten 2 of this component.
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| Perforn inspection, as required anl on itens accessible, consistent with Iten 2 of this conponent.
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| To be perforaed at each reassenbly of link pin to link rod.Revision 3
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| Coaponent Coaponent Identification PM Beconueudation GEl6EIC QILIIImhKB QIRIZ-PE!SE I hit Monthly BX EOC 5 Tear Overhaul Crcments 7.no neasurable clearance when the specified bolt torque of 1,050 ft-lbs is applied.ht the overhaul.visually inspect the rack teeth surfaces for signs of fretting and at one tine 5-year inspection for rods disassenbled.
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| Inspect nating surfaces to verrfy that the nininun nanufacturers reconnended percent contact surface is available.
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| To be perforned once for new and/or replacenent parts.9.If connecting rod bolt stretch vas neasured ultrasonically during reassenbly folloving the reservice inspection.
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| the enqths of the two pair of bolts above the crankpin should be reneasured ultrasoni-cally before the link rod box is drsassenbled.
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| If ultrasonic neasurenent was not previously used.begin use at next inspection that accesses the connecting rods.Measure bolt stretch before disassenbly.
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| IO.hll connecting rod bolts should be visually inspected for thread danaqe (galling)and the tvo pairs of connecting rod bolts above the crankpxn should be MT inspected to verify the absence of cracking.hll vashers used with the bolts should be exanined visually for signs of galling or cracking and replaced if danaged.If prestressor package>s installed, this iten does not X hlso to be perforned at any tine the connectinq rod is disassenbled.
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| Perforn inspection at 5 years, on itens accessible, consrs-tent vith Iten 2 of this conponent.
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| X hlso to be perforaed at any tine the connectinq rod is disassenbled.
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| Perforn inspection at 5 years.on itens accessible.
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| consistent with Iten 2 of this conponent.
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| ll-B-5 Revision 3
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| nent Coaponent r Identification PH Beacnendation GE)rBBIC HhimuXE Hmr-PBASE I Alt Honthly EOC EOC 5 Tear Rerhaul 02-34IA 02-3MA Pistons Cylinder Bead 11.A visual inspection should be perforned of all external surfaces of the link rod box to verify the absence of any signs of service-induced distress 12.All of the holt holes in the link rod box should be inspected for thread danage (qallinq)or other siqns of abnornalities.
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| Bolt holes subject to the hiqhest stresses (the pair ir~ediately above the crankpin)should be exanined uith an appropriate non-destructive nethod to verify the absence of cracking.Any indications should be recorded for evaluation and corrective action.If prestressor package is installed, this iten does not apply, Inspect and neasure skirt and isbn pin.This iten assunes hat AE skirts are installed.
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| For other types, see site-specific reconnendations.
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| Visually inspect cylinder heads (all cylinders).
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| Also to be perforned at any tine the connecting rod is disassenbled.
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| Perforn inspection at 5 years, on itens accessible, cons]s-tent uith Iten 2 of this conponent.
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| Also to be perforned at any tine the connectinq rod is disassenbled.
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| Perforn in-spection at 5 years, on itens accessible, consis-tent uith Iten 2 of this conponent.
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| Conplete TDI Inspection and Haintenance Report Fora No.34I-I-I as applicable.
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| TDI Instruction Hanual, Volune 1, Section 6.Use Volune I, Section 8, Appendix 111 for clearances values.To be perforned at 5-year interval on sanpling basis consistent uith Conponent 02-340h/B-Connecting Rods.Conplete TDI Inspection and Haintenance Record Porn Ro.360-1-1 as applicable, TDI Instruction Hanual, Volune I, Section 6-one sheet for each head.To be per-forned at 5-year interval on sanplinq basis consis-tent nth Conponent 02-340 A/B-Connecting Rods.II-B"6 Revision 3 h 19 I nent Couponent r Mentification PIl Recommendation GEHERIC MhIHtBIhlCL IlhTRIX-PEhSE I hit Ilonthly EOC EOC 5 Tear Overhaul 02-365C Puel injection Tubing 2.3, 4.Record cold compression pres-sures and maximum firing pressures.
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| Blow-over the engine at least 4 hours but not more than 8 hours after engine shutdown." The cylinder cocks should be oren for detection of water lea'ka e into the cylinders.
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| h second air roll should be performed in the same manner approximately 24 hours after engine shutdown.In addition, the en ine should be air rolled shortly before any planned start.Visually inspect the area around the fuel injection port on each cylinder head during the norma)monthly run for signs of leakage.Check tubing for leaks at conpression fittings.Visually inspect tubing lengths for fuel oil leaks or cracks if tubing is unshrouded.
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| If shrouded, fuel oil leakage can be detected at the leak-off ports in the base nuts, which are provided for this purpose, or by annun-ciator if so equipped.If so indicated-remove cylinder heads, grind valves, and reseat.Refr: TDI Instruction Hanual, Volume I, Section&.In the event water is detected the cylinder head should be replaced or re-turned to the vendor for repair.Delete post-run air roll requirements for eng]nes with Group III heads after one cycle with positive inspection results.If water leaka e is detected, the ead(s)should be replaced.hll fuel oil leak in-s ctions to be performed w ile the en ine is running or whenever he compression fittin s have been distur d.Eittin inspection for leaks be rformed at engine opera ion following shutdown.Suhsequent inspections to be erformed I'zodicall as in icated.nshrouded ubing, used as re lacement, should be fu ly inspected consistent with Fahh NDE Procedure 11.10 prior to bending.Rerision 3
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| GMRIC HHMIEILLEE STRlX-PHLSE I nent Coapouent r Identification 390C Push Rods N Recommendation 2.Each push rod of the forged-head design should be inspected by liquid penetrant prior to installation or.if anstalled.
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| at each overhaul.This should be repeated,'t deternined b 75 urs of operation at oad level used for surveillance testing that the push rod vill not develop service-induced cracks.Push rods confirned in this vay need be exanined only visually at subsequent overhauls.
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| Push rods of he forqed-head design exhibiting cracRs larger than 0.25 inch should be replaced, referably vith push rods of he frict>on-welhed design.Each forged-head rod should also be visually inspected one tine to confirn that the head vas fully inserted in the tube prior to velding.Each push rod of the friction-veldel design should be inspected initially by liquid penetrant.
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| 1f this initial inspection was not perforned prior to placing the push rods xn service, it should be g erforned at the first over-aul.If the friction-welded push rod has been previously anspected by liquid penetrant, then visual ezaninatxon will suffice for future inspections.
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| hll friction-welded push rods vith cracks should be replaced, referably with push rods of he sane design.hit monthly BOG BOG 5 Year Overhaul Cocuents X Refr: PHL-5600 X Ref r: PN1-5600.If initial inspection vas not perforned, perforn on sanpling basis at 5-year inspectzon consistent.
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| vith Conponent 340A/B-Connecting Rods.II+8 Revision 3 1 Id nent Coupon ent r Identification P)I Recoaaendation GHHHRIC QIETHEhKH MhTRII" PHLSH I hit Rethly HOC EOC 5 Year Overhaul Coanents 02-425h Jacket Hater Punp-Gear 02-390G Rocker hrn Capscrevs, Drive Studs (Pop Rivets)1.2.l.2.3.Verify capscrev torque values during QR inspections.
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| If not Lx.erforaed at gR, verify at next , then as required at reassenbly.
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| Verify that rocker am drive studs.are intact and tight during QR inspection or EOCI.then as required at reassenbly.
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| Visually inspect jacket water uap gear for chipped or broken eeth, excessive vear, pitting or other abnoraal conditions.
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| Check the key to keyvay interface for a tiqht Kit on loth the punp shaft to iapeller and the spline to punp shaft during puap reassenbly.
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| ht next disassenbly.
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| verify iapeller is one piece (i.e..without a bore insert).If it is not a one-piece inpeller.replace.It is reconnended that the castle nut that drives the external spline on its taper have aininun and naxiaun torque values of 120 ft-lbs and 660 ft-lbs.respectively for DSRYs and a naxiaun torque value of 77 ft-lbs for DSRs.Use TDI Instruction Hanual, Voluae I, Section 8, hppendix IY for proper torque values.hny abnoraal situations or indications of progressive itting should be reported or an engineering evaluation.
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| For engines with less than 750 hours, also inspect by EOC2.X This along vith the drive fit of the iapeller onto the shaft vill preclude past probleas where relat)ve aotion betveen-shaft and iapeller caused fretting and upset of the keyvay sides.Torque valves vill be checked each tiae castle nut is reassenbled.
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| II-B-9 Revision 3
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| SITE-SPECIFIC HAIHTENAHCE HATRIX Ggncnt 02-3)OA Conponent Identification Crankshaft PH Becoamerdation 1.Heasure and record crankshaft ueb deflections (hot and cold).2.Exanine the fillets and oil holes of three nain bearing l ournals (4, 6,&8)usinq LP.f indications are evident, a sore thorough exanination should be nade usinq appropriate NDE nethods.3.Exanine the fillets and oil holes in three of the crankpin ournals.(choose 3 fron Nos.3 hrough 8 inclusive) using LP.If indications are evident, a rare thorough exanination should be nade usinq appropriate NDE netliads.4.Heasure dianeter of crankpin journals.5.Ana)yze the trends of cylinder pressure and tenperature neas-urenents to detect inbalances.
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| hlt Honth)y EOC EOC 5 Year X Overhaul Conplete TDI Inspection and Haintenance Record Forn No.310-1-1 as applicable, TDI Instruction Nanual.Volune 1, Section 6, Refr: TDI Instruction Hanual, Volune 1, Haintenance Schedule.Also to be perforned once at 5 years.Refr: PNL-5600.Also to be perforned once at 5 years.Refr: PNL-5600.Conplete TDI Inspection and Haintenance Record Eorn No.310-3-1 as applicable.
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| TDI Instruction Hanual.Volune 1, Section 6.hso perforn inspection at 5 years, on itens accessible, consistent uith this conponent and Conponent 02-340A/B.
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| If an engine operates in a severel unbalanced con>>on, reinspect the oil holes for fatigue cracks uihin a tine-franc deternined by the utility considering the particular circunstances of the abnornal operation.
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| Refr: PNL-5600.Revision 3 SITE-SPECIFIC HhlHYmtQCE MhTRIX nent Coapon eat r Identification PII Reanaendation Note: To avoid the effect of the 4th order resonance, steady nornal-loaded operation at speeds rore than a few rpn below the rated speed of 450 rpn should be avoIded.hppro-riate precautions should be aken to prevent sustained engine operation with significant cylinder inbalance.
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| Lower speeds Xor testing and break-in are pernissible.
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| hvoid resonance frequencies.
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| hit Honthly BX EOC 5 Year Overhaul Coments Refr: PNL-5600.Revision 3
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| 'ITE-SPECIPIC QIHTIDUKE MhTRIY nent Conponent r Identification 02-315h Cylinder Block PM Recocaendation 1.Perforn inspections per DR/gR Report 02-315h.2.Perforn visual inspection for cracks.Note: Visual inspection not reyired if an appropriate NDE is perforned, hit Honthly IO!EOC 5 Year Inspections.
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| based on cunulative engine hours in conjunction nth Eahh X reports Eahh-84-9-11 and SP-84-6-12(j).
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| Revision 3 APPENDIX 8 5-3-93 Rev.1
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.RECOMMENDATION NO.OF INSPECTIONS RESULTS AND COMMENTS TURBOCHARGER MP 022/023 Note 1 No problems found.50 No problems found.87 47 47 No problems found.No problems found.Some normal bearing wear has been reported.This wear has been dispositioned by the vendor as being within acceptable limits.No problems found.60 No major problems found.Vogtle and Grand Gulf have reported broken or missing bolts passing through the rotating element without identifiable degradation.
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| Vogtle, Grand Gulf and Catawba have reported missing stationary vanes without identifiable degradation.
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| Missing or damaged items were replaced.Note 2 Performed on each test run.Note 1: Inspections performed monthly.The number of inspections are greater than 200.Note 2: Performed on multiple occassions during test runs.A large data base exists.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 1 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.OF RECOMMENDATION INSPECTIONS NO.RESULTS AND COMMENTS BASE ASSEMBLY 02-305A 43 No problems found.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 2 of 14
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| &I APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.RECOMMENDATION NO.OF INSPECTIONS RESULTS AND COMMENTS MAIN BEARING CAPS-02-305C STUDS AND NUTS 28 No findings on bearing caps.Note that inspections are based upon the number of bearing caps examined.Perry has reported one shell with rolled edges due to contact with counter weight.Bearing performance was determined to be satisfactory and the reported item corrected.
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| Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 3 of 14 Cl I APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS CRANKSHAFT 02-310A 188 No problems found.Inspection is number of.hot and cold deflection measurements taken.67 Inspection is number of oil holes inspected.
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| No problems found.Upon bearing rollout to perform inspections, River Bend has experienced minor cavitation, including pitting on bearing surfaces.This was evaluated and dispositioned as not a problem.The bearings in question had performed their function and could continue to operate withouy adverse effects.Bearings were replaced as good engineering practice.42 No problems found.Inspection is number of fillet and oil holes inspected.
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| 35 Note 1 No problems found.Inspection is number of crankpin journals measured.No problems found.Note 1: Inspections performed monthly.The number of inspections are greater than 200.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 4 of 14 APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT COMPONENT PM NO.OF RESULTS AND COMMENTS NAME NO.RECOMMENDATION INSPECTIONS NO.CYLINDER BLOCK 02-315A 105 159 No problems found.Inspection is related to number of areas inspected under individual heads when removed.No problems found.Number of inspections include inspections made by several utilities during operation.
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| Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 5 of 14 S
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| 0 APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS CYLINDER LINERS 02-315C 512 No problems found.Number of inspections represent number of liners inspected.
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| Vogtle has reported light and moderate scratches with bright spots and carbon build-up.This has been evaluated and dispositioned as acceptable.
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| Grand Gulf has found indications of porosity.The liners performed as designed and were dispositioned as acceptable, but-were replaced as good engineering practice.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 6 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS CONNECTING RODS, 02-340A/B BUSHINGS AND BEARING SHELLS (GENERIC)No problems found.Inspections indicate the number of connecting rod bearings.River Bend has reported some cavatiation induced pitting.The bearings remained capable of performing as designed, but were replaced as good engineering practice.The oil analysis did not identify bearing material in the lube oil prior to replacement.
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| Vogtle has found three shells with evident wear and/or indications.
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| These shells were evaluated and dispositioned as acceptable.
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| They were replaced as good engineering judgement.
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| 36 No problems found.Inspection is the number of connecting rods examined.NA See Referenced submittal to NRC, Appendix C 5 D 89 No problems found.Inspection is the number of connecting rods examined.34 71 No problems found.Inspection is the number of rod-eye bushings examined.No problems found.Inspection is the number of connecting rods examined.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 7 of 14 APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT COMPONENT NO.PM RECOMMENDATION NO.OF INSPECTIONS RESULTS AND COMMENTS CONNECTING RODS, BUSHINGS AND BEARING SHELLS (GENERIC)02-340A/B 20 73 296 No problems found.Inspection is the number of rack teeth'examined.
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| No problems found.Inspection is the number of sets of rod teeth examined (required for new or replacement rods).No problems found.Inspection is the number of connecting rods examined.10 20 No problems found.Inspection is the number of connecting rods examined.20 No additional problems found.12 20 No problems found.Inspection is the number of connecting rods examined.Vogtle has found 1 indication in a hole.It was evaluated and dispositioned as acceptable.
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| The rod was replaced as good engineering practice.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 8 of 14
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| APPENDIX 8 RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS PISTONS 02-341 A 39 No problems found.Inspection is the number of pistons examined.Grand Gulf has found 3 piston pins and plugs to be slightly loose.This was evaluated and dispositioned as acceptable.
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| The plugs were replaced as good engineering practice.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 9 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS CYLINDER HEAD 02-360A 1.51 Note 1 No problems found.Inspection is the number of heads examined.Vogtle has found minor pitting and nicks in 4 valves.This was evaluated and dispositioned as acceptable.
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| Perry has found 2 exhaust valve seat cuts.Performance was not effected.This was evaluated and dispositioned as acceptable.
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| The heads were replaced as good engineering practice.River Bend has found problems with swivel pads.This is discussed in Section 3.12 No problems found.Note 2 No problems found.Some mist has been detected on several ocassions, leading to an in-depth investigation as to the cause.The results are incorporated in Section 3.12 and PM Recommendation No.1 Note 3 Inspection performed each run.No problems found.Note 1: Inspection performed each EOC and more frequently by several utilities.
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| This inspection collectively amounts to greater than 200 inspections.
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| Note 2: Inspection performed prior to each start and collectively amounts to greater than 200 inspections.
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| Note 3: Inspections performed monthly.The number of inspections are reater than 200.Reference Attachment I for Phase I Components Rev 1 4/19/93 Page 10 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS FUEL INJECTION TUBING 02-365 C Note 1 No problems found.Minor fitting leaks have been found and repairs are made as leaks are discovered.
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| Catawba has examined 1 tubing failure of unshrouded tubing due to vibrations.
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| River Bend has experienced 1 failure of the shrouded tubing due to the fuel injection pump base cap screws failing.The tubing was replaced and the engine restored to service.Root cause was evaluated and dispositioned as not being a problem.Note 1 Same as for PM Recommendation No.1 Note 1: Inspections performed monthly.The number of inspections are greater than 200.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 11 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS PUSH RODS 02-390C NA 940 Push rods of this design are not in service.Inspection is the number of push rods examined.No problems found.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 12 of 14
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS COMPONENT NAME COMPONENT NO.PM RECOMMENDATION NO.NO.OF INSPECTIONS RESULTS AND COMMENTS ROCKER ARM CAPSCREWS, DRIVE STUDS (POP RIVETS)02-390 G 551 551 Inspection is of rocker arm assemblies.
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| No problems found.No problems found.Inspection is for rocker arm assemblies.
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| Two pop rivets have been found missing.One each on the River Bend EDGs.Result was no degradation in EDG operability since oil flow continued to the required locations.
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| Grand Gulf has found bearing wear.An evaluation has dispositioned this as normal.However, they were replaced based on good engineering judgement.
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| Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 13 of 14 I
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| APPENDIX B RESULTS OF INSPECTION FOR TDI DIESEL GENERATOR PHASE I COMPONENTS
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| 'COMPONENT COMPONENT PM NO.OF RESULTS AND COMMENTS NAME NO.RECOMMENDATION INSPECTIONS JACKET WATER PUMP-GEAR 02-425 A 22 No significant problems found.Inspections are for the number of verifications.
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| No problems found.Inspections are for the number of verifications.
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| No problems found.Reference Attachment 1 for Phase I Components Rev 1 4/19/93 Page 14 of 14 APPENDIX C 5-3-93 Rev.1
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| 230 Soum Tiyon Q.PO.Bcc>004 ChRCL NC 2&20l-1004 Bus P04)373-c.Fac l704)373-2E October 31, 1991 Mr.P.Om Chopra Office of Nuclear Reactor Regulation El'ectrical Systems Branch{MS 7 E4)U.S.Nuclear Regulatory Commission Washington, DC 20555 Re: Cooper-Enterprise Clearinghouse Gzoup Diesel Generators Position Paper on Radiograph Requirements for Connecting Rod Bearing Shells File: MTS-4086
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| ==Dear Mr.Chopza:==
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| Enclosed is Cooper-.Enterprise Clearinghouse Group's position concerning the current radiographic examination requirement foz the diesel generator's connecting rod bearing shells as detailed in Appendix II of the Design Review/Qualification Revalidation (DR/QR)Report.The position paper provides the necessary technical justification to permit elimination of requirements to inspect replacement bearings shells by radiographic techniques.
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| The Clearinghouse Group is request'ng relief from the radiographic examination zequirements because the bearings supplied by Cooper Industries are presently being manufactured by Federal-Mogul, rather than the f orm'er manuf acturer/supplier, ALCOA.Federal-Mogul manufactures their bearing using a centrifuge process,a more advanced method than the static mold process used by ALCOA.The centrifuge'rocess eliminates the potential for void formation and therefoze radiographic examination is not zequired.The Clearinghouse Group requests you review the enclosed document and based upon the technical justification provided, determine on a generic basis, that the curzent radiographic requirements are not necessary.
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| Response to this issue by January 31, 1992 will be greatly appreciated by the Clearinghouse and the individual utilities membezs.Should you have questions, please direct them to Rick Deese at (704)875-4065.
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| Mr.P.Om Chopra October 31, 1991 Page 2 of 2 Very truly yours, R.D.Broome Project Manager Cooper-Enterprise Clearinghouse Duke Engineering
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| &Services, Inc.''.Ge e Chairperson Cooper-Enterprise Clearinghouse TU Electric RDB/VMA/1 00 9 91 Enclosure cc: E.B.Tomlison (NRC)Clearinghouse Representatives
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| *R.J.Deese
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| POSITION PAPER FOR RADIOGRAPHIC EXAMINATION OF CONNECTING ROD.BEARING SHELLS (02-340B),FOR ENTERPRISE DSR-8, DSRV-16 AND DSRV-20 ENGINES/@roose The purpose of this position paper is to provide sufficient technical justification to permit the elimination of the DR/QR Appendix II requirement to inspect replacement bearing shells by radiographic techniques.
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| c ou During the period of 1983-1985, thirteen utilities formed the TDI Owners Group and contracted Duke Management.
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| and Technical Services, Inc.(now Duke Engineering
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| &Services,.Znc.)to perform a Design Review and Quality Revalidation (DR/QR)of the TDI engines following the crankshaft failure at Shoreham.A portion of this review focused on the connecting rod bearing shells.The experience based review of this component revealed a very small amount of bearing failures.These failures were attributed to two causes: (1)inadequate clamping force in the connecting rod assembly due to inadequate pre-load of the connecting rod bolts, and (2)potential voids and/or impurities induced into the bearing during the casting process.These two items were corrected by: (1)increasing connecting rod bolt pre-load, and (2)performing (NDE)(radiography) of the bearing shells to detect voids or impurities.
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| Technica Discussio The original bearings reviewed and supplied by TDI were cast by ALCOA in static molds.These castings were taken by TDI, machined, electroplated with babbit, and then re-machined to final tolerances.
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| Cooper Enterprise (formerly TDI)has informed the nuclear customers that they will begin supplying bearings purchased from a sub-supplier, Federal Mogul Corporation.
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| These bearings are cast via a centrifuge process that is superior to using a static mold in that the centrifuge assures a more uniform placement of equal density material.Attachment 1 from Federal Mogul offers more details on this issue.Material esti Federal Mogul performed radiographic inspections of bearing shells cast by the centrifuge techniques.
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| These radiocfraphs exhibited dark spots or"ghosts".Several bearings containing these indications were sectioned and metallurgically examined.These images were the result of either (1)material with columnar grains
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| as opposed to equi,-axed or (2)slightly lower tin content in the columnar grain areas.The results of the metallurgical examinations concludee that the metal in these areas is equal to the remaining material in mechanical properties; and therefore the shells will perform as required.Cooper Enterprise has purchased and installed these bearings in several non-nuclear engines.Theses engines have accumulated thousands of operating hours without failure.co e to Due to the manufacturing change that produces quality casting and favorable operating history, it is recommended that the requirement to radiograph connecting rod bearing shells be deleted.Note that Cooper Enterprise concurs with this recommendation (see Attachment 2)~
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| i FEDERAL-MOGUL
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| 'I'ECH;lICAL CZ tTER Engine and Transmission Products April 16, 1991 491-Q4 Page I ATTACHMENT 1 H Coooe;E.".v P/Y 02-3<0-04-AG:
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| Bearin s Reiected bv Radio ra hv~s:.ract Hca;ings rejected by Cooper Energy (25 pcs.)were examined using n>etallography, microhardness, and SEM/EDS analysis.Conclusion is that dark spots in radiograph (normally indicative oi lower density material, porosity, or oxide inclusion) are in this case due to one or both of two possible causes: either (1}small patches of material with columnar grains as opposed to equiaxed, or (2)slightly lower tin content in these columnar grain areas.Consultation wi!h a.adiographic expert cont!rrn that the colunU!ar grains can cause such an e.'feet in the ra'iograph.
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| All metallurgical
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| ',ests indica!e that this metal is equal in mechanical oroperties to the cquiaxed grains, and'.hcrefore predic.that parts will perform acceptably in service.Coov ro: 8.Bridgham, D.Pazuk, A.Sparks, R.Moore, D.Jackson, R.Poehler, G.Pratt,$.Jon s, H.Gibson, Vf.Cook, Ann Arbor File File Under: 8-850, Mooresville, Cooper Energy In!roduc!ion Cooper Energy purchases heavy walt B-850 bearings from Mooresville for general use.When required for special applications, the bearings are inspected by radiography, prior to use, by an outside lab, on behalf of Cooper.As of April 11, 1994 Cooper reported to Moorcsvillc that they have approximate1y 25 bearings which they.are rejecting duc to indications found in radiography.
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| The defect in radiography appears as a fuzzy dark area on the radiographic film.The dark spots appear sporadically, but are more prevalent on one half of the bearing than the other{in other words, the prevalence differs between the top and bottom half of the part as cast.)Unfortunately, there is no way to determine once the part is machined, which half was the top and which w'as the bottom.Normally a dark patch in the radiograph would indicate a low-density area such as porosity, oxide, inclusion, or lack of high density phase (in this case.tin).Discussion Qn AprB 11, a team consisting of B.Bridgham,%'.Cook, H.Gibson and the wntcr attempted to determine the cause of thc darkspots.
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| Whatwc found was that the darkspots corresponded to small areas of columnar grains in the materiaL Figures 4 2 and 3 show cross sections of the bearing wali, heavily etched with KeHer's etch, to reveal the difference in grain structure.
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| In al1 cases, the columnar grains appear near the ID of the bearing.
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| t~~4 o~~'~~~~~~~~~~~I~~0~I~0~~~'~~'~~1~~~'~~~~~~~e~~~~~~~'~~~~~~~'~~~~~~~~~I~~~~~~~~~
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| i FEDE~-lvIOGUI.
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| TECHNICAL CENTER Engine and Transr.".ission Products April 16, 1991 491-Q4 Page 3'urthermore, one...icrohardness in each area was taken with the 1 kg load.This load would be less subject to extrer.".eiy localized aberrations such as grain boundaries an rnicroporosity.
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| Results are as follows: Equiaxed: Elongated:
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| Hv 60.3'v 58.4 0 The olifer nce oetween these, two num'oers is deemed to be tnsigntficant.
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| In this study, no deQnite reason for the areas of different grain structures could be ascertained.
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| The most plausible explanation is that the stnail manifestations of columnar grains represent small parcels oi material which froze either on the bottom of the mold or on the sidewalls prior to the beginning oi tnoid rotation.7/hen the mold began rotating, the smail pieces of frozen material (v ith coiumr:ar structure, since it froz in contact wit the.cold suriace)was wasned away and e..ded uo in its Qnal resting point approximately 15 mm f;om the casing OD.In order to test this theory, a section was made through a rough casting (unmachined) at the bottom.It is shown in Figure 7.The grain structure revealed can be seen to be the same, columnar structure which was seen in the auestionable areas.This lends credibility to the proposed;heory.
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| The dark patches aopearing in the radiograph consist of metal with columnar grains as opposed to equiaxed grains.The columnar grains may be slightly lower in tin content.l Metaiiurgical tests indicate:hat this nM:ai has n>echanical oroperties favorao.y comparable to that of the surrounding metal.Therefore th appearance if these dark patches on the radiographs is not cause to scrap the bearings.4 A W.J.Whitney It tt~'~"I g iL, Tl:-CHiNICAL CENTER t=r""'"i'"ss'c" Products.nr:.".',."i.'.'-:>'..Q
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| 'age 4 Figure 1, Nacre Etc.'".ed San".:e A.6X.Heavy Keller's E!ch.ID.'s'.o:."e r!g:lt.Figure 2.Macro Etched S~~p)e B.6X.Heavy Keller's Etch.em~~
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| Cl h
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| ).'=,-i,"-".,'.'..'..
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| ~~:i': TECHi fICAL CENTER~.-'-,-.::.: ':.::...-.::.>.:oa Products~~T1 m~L.Q J Qyg a il~~74 Figure 3.Macro Etched Sampie C.6X.E avy Keller's Etch.ID is to.he top.
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| L
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| ~DEP~.~Pggg
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| ~gHNICAL~Eng.'ne and Transr~sion Products April 16, 1991 491-Q4 Page 6 1+I~~I g!)O', I.: '~4 II 8'I~z~~~I~t I~I~~J, y~I'a g~lA~~'~~P~I'I~~~~I~~~4'~Iv'I~I,+1~I~'I'~iI:-Y.'ll'~j-:g Figure 4, Equiaied grain area.200X.Heavy Keller's Etch., t/t+I 1;'l"')(,'1 t%o~~~~~~''o r~'~i~~'5.Columnar grain structure.
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| 200K Heavy KeQer's Etch.
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| 'FEDERAL-MOGUL tECH;/ICAL CENTER Engine and Transmission Prcducts April 16, 1991 091-Q4 Page 7~~~~0 1~0~VJ 0~~4~~~0 i0\0~J~~J~.~0 0 0 I 0~1 Figure 6.Panoramic view through the columnar area Showing the hardness test indentations.
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| 38X.Heavy Keller's Etch.00 0 ,t~~~~~~~~0~~, I~
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| FEDEP.AL-.'vfOGUL TECHMCAL CENTER Er gine and Traasr.-.ission Products Aorii'6.1991 k9-Q<Page 8~'4f Figure 7.Cross section through surface of unmachined casing.i4ote simi]arity to ce.".'.er area of Figure 6.50X.Heavy Keller's Etch.~~
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| I REALTIHZ X-RAY R:-VHALS BONUS ZNZORMATX01'T Jeezs I Ml eel's.Magnnflu".
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| Corporation I Presented August 16.l989 at the Air Transportation Association Nondestructive Testing E'orurn S.os~4 s 1 V~0 V&V acknowledge...ant for tachni:al support from: James Donaldson G=rwl.l Mi.;on l!ichadl llooz'8 Zs'0 t)utv'4 0 Ru dir!O i Abstract: A radiographic phenomenon, termed"Ghost.indications".
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| vhich appear to be but: are not necessarily rejectable defects.is deecribed~The ambiguous nature of these indications can result in a sound structure beinq rejected, or unsound structure being placed in critical service.The mechanism of the occurrence and a means to differentiate between."ghost." and true indications is explaine6.
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| K
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| Hi" tory: Page 1 Tile"ghos t" or x-ray di f'rac t'n p he no.".enon has plagued the rad'cgraphic inspocticn business sine crystalline structures were firs" radiographed.
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| G6.".or al Lnow'5"a of the a::is t ance o f this phono@enon coupled with ezt~rsive destrqc"ive vari"ication, has allowed some very experienced radiographers to make judg.",.ent calLs in noncritical areas.An a:ccel'anc paper uas pressured in gaeeri(l)ye>uacicns, Dec., i966, Runnel 6 Grapery"'Ghcsc Lack of Pusinr'n Alu.-.inurn Alloy But t Fusion Molds", d'f aren tie ting"ghost(a indications from t ue defects in a specific inspection application.
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| The increased use of e::otic (especiaLly copper bearing aluminum and h'gh nickle)alloys increases'the number and save-.i ty of d=f racted e ndications.
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| Diractionall y solidi fi ad and single c ps t'1 s t uc".ur'as ar~ne'arly''a)f's=%bi)to radi)-ra~hi a: ly inspect wa tnout vary costly and'ime consuming Cechn'ques.
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| ToCa I, d:se tc these limitat'nns and the extremely critical nature, oL'he air transportation industry, ref'iogrxph~rs are justif'ably reluctant to maLe judgment calls.A method which Mould assist the radiographer
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| 'n confidently differentiating"ghost" from re3ectab'indications, cou'd lowe" scrap rat"s awhile a"suring that t"u'y r a:.actabL:s part" do nct r ach cr-'..cal s:r'i".es.
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| Obs a veri P'xcnomcnon:
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| TNre<rl pv~p'+I+(4 t lp Q (a o i)I['s oI)fc)rtt~((~~~~~a ((~0 l (J L.:t:.-r..y d"...'.y app.sr throe.qh)ut t.'le iaaf,'., ranf[in solotch's".stern'y)e'ne'.ns5i'z-.kp)n.":.
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| B e:"."..'.r f: an~ia(S44(e~((Q sI~o~~i~(~~)~'')m ha"(NC'2';Phil indicat'one a" e quit.>evidint on P~.[time mcni-..o:.
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| fidelity negates reproduction This unusual phenomenon was nearly aLrcays accompanied by: l.A mottled background to the image.2~3.A, du).1 thud in the traditional tap or"ring" t;est.audible Acus tie Emmi a@ion indicator.Poor ability to hold a sound weld repair(a I.~ee p 54~
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| Page 2 w'ndow was cut from a part displaying this ghosting and was replaced with a piece of'ew material so tha" a direct co.parison could be mnde.Upon ra-inspection of the windowed part, it was cbsŽ:a" the new materia'isplayad neither'the mottled backgrcu.-.d nor"ghcst" indications.:
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| Further investigation tevealad tha-the ghost images d'd not move in coordinat.ion wit..par=motion.Mhen v'ewed dynamically, the indications moved oppos='"a to the part motion: i.a., if the.part was moved from ths left to right, the indications would move from the right to lef": i.the part wa" moved up, the indicat'on mooed down.This"antimotion" made it obvious that the indications were d'ffractel x"ray pat"erns rather than a.fec" indications.
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| To fully understand these observations, a s tudy o.the ma.."-'al and the mechanism of x-ray diffract on was undertaken.
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| Hateriai Study: h section of Cha part containing both orig-nal and new ma"ar'al was removed foz analysis.Thc chemical analysis shcwed little deviat'on from tha Has telloy X~analysis supplied by t.".a a'oy vendor, Cabot.Xt was noted'that th-su'ur cent'.".C of the st r face analysis was a factor of 10 tim~:s higher o..t.'za old mat~rial than either the vendot analys is or the ne'v material an!lysis.~~.'i a.".u's'.:n with llo{'1'.f.8'.p.r::crn.~a.".no.;n".>:.
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| ~~v a r.)1 P..''hat t.: '.c a LJ)'1 sul u.'nt>nt'Qus th.s tr<<'pi 14 process used to r'ove.tlat h~it r.'-is.'-="-'tin~t du nc re'wcr>.The na'n cnm~onen t 0 f t o'l s tr.p-n'-.P a t" is su'=u"":: ac=.excss"ive r~tent1in cf st"xppxic ao ut:0 or""cr neztra'i=ntion may account for increase)sulfur cont On closer observation the surface of the old material shows aa extremely rough appearance.(Photo 1)The open and saw-tooth appearance of the cracking also indicated a large grain presence.Theso observations were supported.with e 500x view of thk same surface (Photo 2).This view shows very Large grains and severe etching at the grain boundaries.
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| Some grains appear as Rf they could be lif ted from the surface.When compared to,the.new material at 1000m (Photo 3), the evidence supporting the high sulfur content theory is conclusive.
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| The extromelf large grains'lso indicate that this part was not propert.y annealed.
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| S IP The open boundaries
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| ~ould account for the mottled image, the dull'hud'n the tap o ring test, as sell as the inahliity to" hold a good veld repair.The ghosts in the imago ar a r suit of the x-ray beam being diffracted from the indic s of the large grain structur.Zn th's c e the appearxrc of any"ghost ng" is an indication of poor or no annea'ng and is causa for rejection on its oon.Th's information in itself is an unaxpected bonus for the real t';..e inspection.
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| Y8t, the study of the x-ray diffraction ph ncaencn also revealed more universally useabla information.
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| X-Ray Dif rection T?ie Ru~mel'Gregory paper~as us ed as base pnin t to s tudy the diffraction mechanism.
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| E:;.c rpt: v~rav f g t gwyp w~a~Hhun a baa." of Xrays strik>>a cry's tal.pa"=o the bean Js transmitted.
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| pa t of the baal is scattered.
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| One of the mechanisms f'r X-ray':ca t ter ing is by diffraction ram the same manner as gratinc Bit'fr ccs ordina'!light.How.if a sari~s of cry's t<is (crystal't pl anes)~ra pYoper'ot'ien ted~Hi th res pe t t 0 a:1 v r ay cvs')r'f oo r%4)+r h o f"'.'.""'ee~!Jan~~~raw ro~'eo e op~so o Q j g classicgl lv%+r)ct" Q~n j,Qna i~)Q and grip 1L 1 s'~.a~/and applyin:;it to the observed phenom~non 1 sf t one of t'~o'onclusions.
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| Either the original observations Mere not dif fraction related or a much more complex mechanism is.occuriing..
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| Close comparison of Fig 1 and the recent observed conditions revoaled several differences.
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| 1"Ghost Lack of Fusion: in ALuminum AlLoy Butt Fusion Voids Mard Rummel and B.R.Gregory Material Evaluations Dec 1965
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| Pith Zi1m radiogr phy, the source-to-ob::e t distanc8@as sufficiently long to assume n~nr parallel incicent rays, Mith ijic"o focus Realtime X-ray, the source-to-object distance~as under 5 inches and the divergence othe:t-ray beam must be considered 2.Pith ff.la radiography the ob ject-to-il=i di tanc is always kept to a minimum, prof.trebly hiero~Hith Realtime NicroEocus, the image plane-to"ob oct djstxnc.was 15 inches or a 3: 4 projection ratio.The travel length of the'diffract,ad ray must nov be cons.'d=r d.3.Meld inspect'on has a linear ar"a of interest~Xn this case the d.'.'reaction phenomenon could.cons'e onl i those indications appearing parallel to th>>weld.Burner"wn inspe<.-ti)n is conga".-..
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| <.-a=-.h any'd'at'n in any'..is and the ciiC1raction planes a." corplecely random with no pr..Eer"ntial aliqnme..t.
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| Grappling~ith thes~differences, at length with scratch oad and pencil.lead to the underscarding thac the mechanism had not chanc ed f em thv classic p".a<en..ation (Fig 1), but hcd eu 1"i pli.d'."." variable" su:h that it+as v=r>dif cul.to con'ive a graphic a cz'r'jan tat cn to ci<pre~such vR 1Abl s An i<~t":-.!"';i".h':" C~sr.<.".-": xr.'~Qhcz'(E f o+<+,<p c~~~~~~a~~~i'<~~+fjs++Qp+r+Q~~<v<lf f+lp)a>>>g'<y<-L'<e<)~1 y q'q~, y4g v 1~1 aL~~s%~re l 4~'i out ill 9)l l.el?r'1 4 e~'"~>l'r un'+s'4<<411'7 c th'-'v-''~~'~~a<'<r~~a~'onceived.
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| An unj rstandirg o8 th s phr'n]m<;nun 1".ada Cc I::1<x abil'ty to test conculsiv~ly wh~rher any indication in any material is caused by x-ray diffraction.
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| e Page 5 Figure 2 is a graphic repzasentation at observed raaltime geome>>'r'F
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| .Here~.beam dive.pence ance source-to-oh j z t-to'-page plane re'ti onsh'."s are tak vn in to cons ilare tian, Ta clearly understand:
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| the antimotion ph<>namenon, ae must consid:sr a;.indioidual ray erase.Prem Thecrv Ot>(-say Dit<".=el On in Cr isenls (M.E.Kschsriasen, Dover Publishers, nublislied 1967), ve ac.ept the given that the diffracted beaw vill exit the indices at an equal and opposite angle to the entrance af the in<.id<:nt or pr'mary ray.Using this giv~n.ae can now lnok a".cne event (<:igure 3), in th:.A, posit'an, then moving only the d'"'"act>>ng indic"s to the B posicion.The result.".a opposite shift of the di r'.c'.".0 benra now supports ch<: "ant.'-mo ion'n th..observed realti-e:c-cay i:;age.Ev.;n~aors inta sting, is the effect of va": i..g t.".a cbject-t--
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| i.-.age plane distance.ZE the ratio or.source to abject vs object to imaao plane is 1: 0.equal mn t ion occurs.ZE the ra tic is 1:1', no motion is apparent in the did" a<tad Lndicztion when the object is moved.At 1:" ratio.egual but opposite motion occurs.Displaying this in thr e e di;.ens i ons (F igur e 4)thus ac<".ounting for the c ne af divergent radiation tend t.".e v.3rtical as.9 s)lagoonal etfacts can be camprbhended.
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| <: nc;:::ian:
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| i'i'>>see'>>Fee~e si J ice>>a~>>~m<ee e,<<nys>>g<q>>sn e e>>6~use, ee~6 i."agile egu p<".<nt can'on<:Iu i<r'dLy i"tnt'>>j-f-.ss:t'an ah.:no,qncn.
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| 3v<;,~r Fi p y th~p}8:.'<: n~.ln"..6<.
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| kl:ov!: J ail}.vary in a pr d='ctable man.".~r.When usinz an x-ray source of suf fief.ently small focal spot ta allow some variation in ob j~c t-to-f i'lm dis tance.a f ilm't'.i ograph could be ra shot to confirm the orig'n at sus ptcions ind ica tiona.
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| S k~'
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| ~0~~..+,",~~e;1'.I.~~)t<a~:ia~aw: (Photo')~A I'I p~a~~'~lh~~Ho I~4 I~~~Q~(Photo Z)~~ll!I(1ih~~8~~o i~.'C~~)o all 0 I II~I 1.
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| PEG.1 r ys&is Diffracted Beam 1'I I' X-RAY SOURCE'TE'Rf AL RANOOM CRYSTAL IHTTf M TR AN SMI TTEO..BEAM IFFRAC BEAM GHOST IMAGE RECIEVING M OIA FJGURE 2, MAGNAFLUX 7/89
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| X-RAY SOURCE f KHT RAM DlFFRACTE'0 TRAN MtTTED SFAM$'fQURK'MACMAFLUX l/89
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| ~~~~%%-KV-'91 i'F:25 tD:FEDERR HOG@.~TH N3:31V~L783S SOURCE INCIDENT BEAM TRAN SMI TTED BEAM 08 JKCT
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| ATTACHMENT 2 CPSES 9117826 SU 910310 July 15, 19e1 TO: J.B.George
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| ==SUBJECT:==
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| Radiography Requirement for Enterprise Bearings REPENTANCE:
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| DR/QR RReport 02-340 B Referenced report, prepared by a consultant to the owner's group, suggests that TDI bearings will be acceptable provided they pass a radiographic examination performed by that consultant.
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| This study was initiated as part of the owner's group effort to qualify TDZ diesels and included such events as discovery of cracked connecting rod bearings at Shoreham in 1983, and reports from TDI Vee Engine'owners of cracked bearings.Portions of this report have not been endorsed by Enterprise as discussed below.Bearing shell cracking has never been a problem in the in-line engines such as used at Shoreham.It has always been our contention that the cracking noted there was caused by use of connecting rods with an.extremely large bore end chamfer, which allowed the bearing ends to be unsupported, combined with significant engine overloading.
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| The con-rod condition was corrected immediately.
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| No more cracking occurred.Vee engines in those days utilized connecting rods assembled with what we now know was insuf f icient f astener preload, causing excessive flexure, or micro-distortion of the big end of the rod.This condition caused the highly publicized con-rod rack tooth fretting phenomena.
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| Of greater importance however, was the effect of this flexure on the rod bearing, especially if that particular bearing was brittle, i.e.of extremely low ductility.
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| Most of the f ailure analysis studies done at Enterprise on bearings which cracked for no immediatley apparent reason reported bearing shell elongation numbers either nil or less than 14.Some had regions of casting porosity on or near the crack surface, but most did not.
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| page 2 TDI supplied bearings made and plated in their factory from Aluminum/Tin castings made at Alcoa in Cleveland.
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| These castings were statically cast in a permanent mold and, from time-to-time exhibited less than adequate mechanical properties.
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| Porosity was also sometimes a problem, and resulted in inability to satisfactorily electroplate the lining on the piece, easily detectable in the plate shop.Note also that pores as small as~010"/.020" were easily visible.In no case would pores of.050" allow plating to be acceptable.
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| I In the early 1980's the fastener preload on Vee Engine con-rods was significantly increased.
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| Rack tooth fretting, while still not zero has been reduced from very significant to almost nil.In the mid 1980's, destructive testing of each heat of bearing castings was begun to verify adequate mechanical properties.
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| Operating experience after these changes was most satisfactory, bearing shells routinely lasting 20,000/25,000 hours (BY NO MMS 38,000 HOURS).Shells are replaced based on wear limits rather than base metal condition, in conjunction with general overhaul activities near this hour level.None of these bearings were radiographed.
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| In 1988, Enterprise ceased manufacture of bearings, opting to purchase these parts in finished form from Federal Mogul, a worldwide supplier of all kinds of engine and compressor bearings, including bearings for engines which could have been installed in nuclear generating stations.F-M is not aware of any radiograph requirement for these parts.F-M uses the centrifugal casting method to obtain consistantly high quality castings.This method af f ords the f oundryman various options such as mold spinspeed, pour rate and cooling rate to further enhance casting quality.F-M asserts this fine Omiag is normal and on~oing, and may be the cause of radiograph ghost imaging, as the report I gave you suggests.F-M furthermore applies a flash of plating to the back of the bearing, the lead/tin content aggravating X-Ray problems, but improving its grip in the housing.F-H bearings have been in use in Enterprise Vee Engines for thousands of hours.No reports of bearing quality problems have been received.None of these bearings were radiographed.
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| i page 3 In summary, I submit that the onerous radiographic suggestion of referenced report was of questionable value in the beginning, and certainly is of no value now.Not only have the con-rod problems finally been solved with the use of adequate fastener preload applied by hydraulic tensioning tools, but also the bearings are manufactured by avendor specializing in this work, utilizing a completely different methodology than the TDI/Alcoa method employed.~l7 M.H.Lowrey Cooper Industries Distribution:
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| M.L.Bagale Ken Dixon Bo Neir
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| APPENDIX D 5-3-93 Rev.1
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| APPENDIX D 0
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| DMS Eih6AEERING 8 SERMCER INC PO.8ox 1004 Cencaa W am~Oat 8'04)373-24'U Fac f704)373-2695 February 27, 1992 Hr.P.Om Chopra Office of Nuclear Reactor Regulation Electrical Systems Branch (MS 7 E4)U.S.Nuclear Regulatory Commission Washington, DC 20555 Re: Cooper-Enterprise Clearinghouse Group Diesel Generators Position Paper on Radiograph Requirements for Connecting Rod Bearing Shells File: MTS-4086'ear Mr.Chopra: Enclosed is additional information to clarify questions in regards to certain proposed process changes related to radiography of the connecting rod bearings.This information supplements our previous letter dated October 31, 1991.The Cooper-Enterprise Clearinghouse Group requests you review the enclosed document and based upon the complete technical justification provided, evaluate and concur with the Clearinghouse that current radiographic requirements are not necessary for Cooper Enterprise EDGs.Response to this issue by March 20, 1992 will be greatly appreciated by the Clearinghouse and the individual utilities members.Should you have questions, please direct them to Rick Deese at (704)875-4065.Very truly yours, fm.A~+m R.D.Broome Project Manager ,Cooper-Enterpxise Clearinghouse Duke Engineering
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| &Services, Inc.B.George Chaixperson Cooper-Enterprise Clearinghouse TU Electric RDB/VMA/021492 Cl February 27, 1992 Mr.P.Om Chopra Enclosure cc: E.B.Tomlinson (NRC)Clearinghouse Representatives R.J.Deese
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| '
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| Wi 45WCounty Line R~4 M~eaviil+, inNan4 4$158 Tef.N74314830'ax 317431.7035 vanaR~L.MDRUL January 24, 1992 Jules Hudson Cooper Energy Services 14490 Catalina St.San Leandro, CA.94577 Mr.Hudson: In response to your fax dated January 10, 1992;there are many processing techniques to reduce or eliminate the existence of gas entrapment within the bearing.Here at the Hooresville facility, we use the.centrifugal casting process.This process inherently, lends itself to the elimination of gas bubbles, drosses, and oxides due to the outward radial force (approximately 30-60G)acting on these par ti cles.'ince the densiti es of the af orementi oned par'ticles are considerably less than any element in the AA 852.0 alloy, they are forced to the inside diameter of the~casting vere they are removed by subsequent machining.
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| To further-insure the.removal of gasses, hexachloroethane
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| 'ablets.are dispersed into the meit.The tablets.decompose to evolve chlorine gas which, in turn, ties up the hydrogen (the primary cause of entrained gas in aluminum)and removes it.from:the melt.Past foundry testing using reduced pressure:tests confirm'the expu1sion of hydrogen gas via.this method.In.addition to production techniques, the process is closely monitored to verify the continued success of these techniques..
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| These include: Individual Process Set-Up Sheets for every job, First Piece Inspection of casting;<uorescent Penetrant Testing of each heat, and Verification of Mechanical Properties of each heat.
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| For every job cast, a Process Set-Up Sheet (see attached)is generated and released to the foundry prior to production.
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| The Process Set-Up Sheet contains all of the vital process parameters needed to produce a particular casting.En addition, it provides documentation of any changes to an existing parameter.
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| Standard practice dictates that first piece inspection be performed on'the first casting poured on a job.After cast, the casting is allowed to cool to approximately 300-400 F.The casting is then fractured to reveal four (4)distinct cross sections.These cross sections are evaluated under i0x magnification and i'nspected for dross inclusions, layering, gas voids, and excessive shrink cavities.This evaluation is documented on the Process Set-Up Sheet.Pen The Requirement for fluorescent penetrant inspection (2yglo)is indicated of'he Process Sat-Up Sheet.The majority of large castings (>10-11 in.dia.)are tested in this manner.A sample casting is poured prior to production and bored to the blue print dimension.
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| The bore surface is evaluated for surface discontinuities which may or may.not have been apparent during analysis of the fractured casting.Me n{1 r'{o At.present, a representative casting (termed"lab sample")i'.poured for each individual heat.This casting provides for both chemical and mechanical testing.Test bars are cut from the lab sample and tested for tensile and elongation pf operties.This'esting provides conf irmation that no detrimental.
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| defects exist within the test casting.Under current evaluation is the potential for using separate l y cast test specimens (.505" standard ASTH tensile bars)to predict the acceptability of production castings.Since the separately cast bars are not under the influence of'ead pressures greater than 0 x gravity, they will be affected by discontinuities to a greater degree.Therefore, acceptable results obtained via separately cast specimens would insure a degree of confidence in the centrifugally cast producti
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| ~a JAN-24-'92 13'.B9 ID:FEDORA NZRI (TEL H3t 317~~(I hope that this information assists you in your communication with the HRC.Xf you need, any additiona) information, please feel free to contact me.Sincerely,.Brett L.Bridgham Plant Metallurgist Copy: D.Jackson R.Hoore 0.Pazuk Mooresyi lie Lab File
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| .
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| N-'.ustqmers
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| {:EB rt No g R 3313 Ofe (Conv)s PEL-8 Me{CNC)Ca~t Wt{Z): 1eo~C at Nt<Z)i loy: a850:erti fi cation', Chemistry i Y mechanical
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| -Y Bpociticationa CKS D4998 Zyg lo (y/n 3: Y Other ittttttttttttttt INITIAL SETUP tttttttttttttttt Spr ay Tower"A" Noxxle Type (a)!50/10 Locations (a)e 1,3,5 fetal Temp)ia Temp (pe=lux=ash Start 1350!300 400 2¹11 2 SEC Spray Tower"B" Noxxle Typo{b)x 50/10 Locations fb)i 2,4,6 SW C S~S=as%.I D~-.10" ast Weight: 1SO¹D Stock: 1.328 Mater Delay: 15 SEC S D.Stock:.5a9 Coul Time: 4 5 IN 3AL Stock: 4~201 Water Temp tttttttttttttttttttt PRODUCTION HISTORY ttttttttttttttttttt JOB NO o 3iEi NO~3ATK ta}Temp 31 a Temp RPN*=lux=ast Start as+I.D ast, Weight Change!91-3035!Reason CAST QTY.PCS/HR Spr ay Tower"A" Noxx1 a Type Locations Spray Tower"B" Noxxle Type Locations Mater Delay Cool Time Mator Temp Result 7/14 4 ftttttttttttttttttt WEIGHT REPORTING tttttttttttttttttttt
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| ~~~~ag<<~NS~W~~SSN~WW
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| ~~%82~~Data Shit t.'crap,'ood Lba', Lba Fract CaILt Other Scrap I l Good~Bag Req'c
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| APPENDIX E 5-3-93 Rev.1
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| J~~C I Received O~~g C I J" i QJGLEAR SGG+**<<+UNlTED STATES REGULATORY COMMISSION WASHINGTON.
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| D.C, 29555 OEC-o QP/2~~au~w~~60~P+Ml I.J wee.~y ply Mr.J.B.George, Chairperson Cooper Enterprise Clearinghouse C/0 Duke Engineering 230 South Trion Street Charlotte, NC 28201-1004
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| ==Dear Mr.George:==
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| '7~~$oc~i~a
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| ==Reference:==
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| Letter from R.O.Broome to O.Chopra, dated October 31, 1991 Ln tne poove r eI.ei~iIc u Ielte you requc>tcU I el roTli radiographic examination requirement used to detect voids and impurities previously found in cast aluminum connecting rod bearing shells used in your TDI diesel engine.This requirement was originally proposed by the TOI Owners Group and accepted by the NRC staff for purposes of inspection of TOI diesel engine connecting rod bearing shells manufactured by ALCOA.This requirement was detailed in Appendix II of the Design Review/quality Revalidation Report for TOI diesel engines.We have reviewed the technical justification to delete the radiography requirement in your position paper attached to the above referenced letter.We note that the replacement bearings originally supplied by ALCOA are presently manufactured by Federal-Mogul Corporation.
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| Federal Mogul fabricates its bearings by centrifugal casting, an alternative to Alcoa's static casting process.IJnlike static casting, centrifugal casting significantly reduces the potential for void formation.
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| Furthermore, the manufacturer has demonstrated that, through choice of manufacturing processes and quality assurance measures, the cast aluminum engine bearings will have an acceptable level of quality and safety.'These alternative approaches and inspections should be as effective as the previous requirement of radiographic inspection of static cast bearings to detect voids and the presence of impurities.
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| On this basis, the requested relief from requirements to inspect d'.esel generator connecting rod bearino shells by radiographic techniques for*OSR-8, 3SRY-16, and DSRV-20 engines is granted.A copy of our safety evaluation is enclosed.Sincerely,
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| ==Enclosure:==
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| As stated hs R.Ric ardson, Director vision of Engineering J D Office of Nuclear Reactor Regulation J>>
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| ~4 0-~i n 0 I c wo~O+**++UNiTED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, p.C.20555~NC OSUR~EAF Y YA IIAY All POS ON PAP R FOR NOT P RFORH N RADIOGRAPH C EXAHINAT ON OF R P AC H N S G N RATOR CONN CT NG ROD AR NG SH S FOR T R R S DSR-8 DSRV-16 AND DSRV-0 D SE NGINES
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| ==1.0 INTRODUCTION==
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| By letter dated October 31, 1991, Cooper Enterprise Owners Group requested relief from the current radiographic examination of connecting rod bearing shells for Enterprise DSR-8, DSRV-16, and DSRV-20 engines (originally TDI diesel engines).This requirement was originally proposed by the TDI Owners Group and imposed by the staff for replacement connecting rod bearingAshells manufactured by ALCOA as detailed in Appendix II of the Design Review/quality Reval.idation Report for TDI diesel engines.Our evaluation of the technical justification provided by the Cooper<nterprise Clearing House Group is as follows.2.EVALUATION The subject position paper proposes to eliminate the need for radiographic examination of the currently supplied centrifugally-cast aluminum bearing shells from federal Mogul.When the radiographic requirements were established, the aluminum bearing shells were made from aluminum casting supplied by ALCQA that were manufactured by a permanent mold static casting process.Past failures of the bearing shells were attributed to I)inadequate clamping force in the connecting rod assembly due to inadequate pre-load of the connecting rod bolts, 2)inadequate support at a bearing end because of a 1/4 inch chamfer, and 3)potential voids and/or impurities induced into the bearing shell during the casting process.The problems were corrected by 1)increasing connecting rod bolt pre-load, 2)reducing the size of the chamfer to I/16", and 3)inspection by radiography of the bearing shells to detect voids or impurities.
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| i The position paper and its supporting documentation addresses the problem of the unnecessary rejection of bearings for radiographic indications.
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| The indications are fuzzy dark areas on the film;these can indicate porosity or inclusions, causes for rejection.
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| Tests, however, showed that the dark spots may correspond to areas of columnar grains and minor differences in chemical composition.
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| Evidence shows the spots are likely caused by diffraction of the X-ray by this grain structure.
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| Although the paper showed that these indications can lead to rejecting sound castings, it did not describe how to differentiate columnar grain structures from rejectable defects or other ways to assure the quality of the bearings.Federal Mogul provided this information in a letter dated January 24, 1992.This information demonstrated that there were production procedures and quality control tests which provide adequate assurance that these castings will be produced without defects of significance.
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| The combination of centrifugal casting versus static casting, the removal of hydrogen by chemical means, destructive first piece inspection, fluorescent penetrant testing of a machined part prior to production, and the static casting of mechanical test specimens for centrifugal cast products provide an acceptable level of quality which should assure that voids and impurities will be detected in test specimens prior to their being generated into a finished production part.Based on a review of the information provided by the Cooper-Enterprise Owners Group as discussed above, the staff concludes that the manufacturer has demon-stated that through choice of manufacturing processes and quality assurance measures, the centrifugally-cast aluminumdiesel engine bearings will provide an acceptable level of quality and safety.These alternative approaches and inspections should be as effective as the previously required radiographic inspection of static cast bearings to detect voids and impurities.
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| Therefore, f the requested relief from the current radiographic examination of centrifugally-cast aluminum bearing shell Enterprise DSR-8, DSRV-16, and DSRV-20 di manufactured by Federal-Mogul for el engines is granted.
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| Change to Facility as Described in the FSAR Title: PCR-01181, Westinghouse Evaporators Sample Point Modifications Functional Summar This modification to the Waste Processing Sampling System installs sample coolers on the Waste Evaporators 1 A&B and 2 A&B distillate sample lines to prevent personnel from becoming burned due to high sample temperatures.
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| The cooling water supplied to the coolers is demineralized water which is routed to waste via the sample sink after exiting the coolers rather than returning to the demineralized water header and possibly contaminating the system.A check valve is installed in the common demineralized water inlet line to the coolers to prevent possible demineralized water header cross contamination.
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| The demineralized water flow enters the coolers at the bottom and exits at the top to assure the coolers remain full of cooling water at all times.Ball valves are installed in the demineralized water inlet line to the sample coolers to control the flow.The sample flow enters the top of the cooler and flows counter current to the cooling water to provide maximum heat transfer capacity.After the sample line exits the cooler, but prior to the sample returning to the water evaporator, a 1/4" sample line with a valve tees off into the sample sink to allow plant personnel to sample the water evaporator distillate.
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| Safet Summar The newly installed equipment (coolers, valves, tubing, fittings)exceeds the system design temperature (200'F)and pressure (150 psig)ratings.The Waste Processing Sampling System is non-safety related.If a sample cooler failure occurred it would result in a spill of possible contaminated water into the sample sink which would drain to the Radioactive Drain System.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 11.2.2-02 MEM/HO-930100/3/OS1
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| .9306040245 Change to Facility as Described in the FSAR Title: PCR-01652, Reduction of Sealwater Pressure to Various Radwaste Pumps Functional Summar This plant modification to the Liquid Waste Processing System adds seal water pressure regulators to the seal water lines going to the Solidation System Pretreatment Tank Pumps 1-4A and 1-4B, Waste Evaporator Concentrates Tank Pumps l&2A and 1&2B, and Spent Resin Transfer Pumps 1-4A and 1-4B.This modification was required to limit the excessive pressure of the demineralized water being supplied to the seals which was forcing the seal faces open allowing demineralized water to enter the process stream.Instrumentation and electrical involvement in this modification is limited to replacement of the existing pressure switch with one of the narrower range and the establishment of setpoints associated with these switches for monitoring low seal water pressure.Also minor rework was required on pressure switch process tubing and existing solenoid valve conduit and wiring due to relocation of components'af et Summa The three (3)pumps associated with this modification are all part of the Liquid Waste Processing System and are classified as non-nuclear safety (NNS).Addition~~~~~~~~~~of seal water pressure regulators to the seal water lines serve to reduce pump seal failure from excessive line pressure.Loss of pump seal water from failure of the pressure control regulator is not a valid concern since the regulator fails to the open position providing the equivalent of the original design.Materials of construction used in the fabrication of the regulator are compatible with existing system piping and consistent with industry standards.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 11.2.2-3, 11.4.2-1, and 11.4.2-4 MEM/HO-930100/4/OS1
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| Change to Facility as Described in the FSAR Title: PCR-01963, Reactor Auxiliary Building (RAB)Switchgear Room'A'low Balancing Functional Summar This plant modification provides revised airflow quantities to RAB Switchgear Room'A'upplied by Air Handling Unit AH-12.These changes incorporate the results of the revised heat load calculation completed for the affected area.In addition, a new return register is provided in the heating and ventilation Room¹4.Safet Summer The AH-12 System is safety class 2, Seismic Category 1.The previously specified redundancy and physical separation is maintained.
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| The ability of the heating coils and fan to meet the revised load requirements have been verified.The revised flowrates for AH-12 ductwork do not exceed the current ductwork design limits.In addition, review of the additional return register confirms structural integrity in event of a seismic event.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 9.5A Table 9.4.0-2 MEM/HO-930100/5/OS1 Change to Facility as Described in the FSAR Title: PCR-02483 Rev.1, Downgrade the Boron Recycle System Evaporator Feed Pumps Functional Summar This is a design change to downgrade a portion of the Boron Recycle System (BRS)from ASME Section III (Q-List)to Radwaste Q-List.This change allows substitution of Radwaste Q-List Recycle Evaporator Feed Pumps for the existing ASME Section III pumps which require repl'acement.
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| Safet Summar The portion of the BRS that is downgraded will remain seismicly supported since the downgraded piping and Recycle Evaporator Feed Pumps are connected to seismic piping and must remain seismicly designed to protect the adjacent ASME designed pressure boundary.The BRS does not satisfy the functional requirements of Safety Class 3 components as defined by ANSI N18.2-1973.
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| This is not considered a major radwaste change as defined in Section 6.15 of the Technical Specification.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Table 3.2.1-1 Figure 9.3.4-6 MEM/HO-930100/6/OS1 Change to Facility as Described in the FSAR/Title: PCR-03170, Vent Stack Flow Rate Monitoring Functional Summar This project modifies the Effluent Radiation Monitoring Systems for the RAB,Vent Stack No.1 and the Waste Processing Vent Stacks No.5 and No.5A.The existing design utilizes Kurz velocity probes of the hot wire anemometer type mounted in each vent stack.The velocity of the effluent discharge is measured in each stack and a signal is sent to a microprocessor which controls the flow extraction rate of the sample by adjusting a valve located on the sample conditioning skid, thereby maintaining a near isokinetic sample.This modification will result in the deletion of the Kurz systems for flow determination in each of the subject vent stacks.The new scheme will rely on contact closure in the auxiliary relay panels on relays that are energized when their associated fans are running.The contact closure will signal a microprocessor a specific fan is running and thereby exhausting a predetermined volume of air to the vent stack.The microprocessor will be programmed with the fan flow information for each fan that has the capability of exhausting to that specific vent stack.The microprocessor has the capability of summing the total flow from all fans that are running at any specific time and deriving a flow velocity based on the configuration and unique characteristics of each stack.After the stack velocity is determined, the microprocessor will signal the existing conditioning skid to adjust the sample extraction rate assuring a representative sample is taken at near isokinetic conditions.
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| Safet Summar The Effluent Radiation Monitoring System is neither an initiating nor a mitigating system although release of radioactivity is a key safety consideration.
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| Implementation of this modification will reduce equipment down time, while improving the required effluent accountability effort and reducing the need for compensatory flow rate estimates.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Sections 8.3.1 and 11.5.2MEM/HO-930100/7/Osl
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| Change to Facility as Described in the FSARTitle: PCR-04220, Deletion of Locker Room Area Radiation Monitors RM-1WR-3652A and RM-1WR-3652C Functional Summar RM-1WR-3652A and RM-1WR-3652C are area radiation monitors located in the mens and womens locker rooms of the Waste Processing Building, elevation 261.As part of an ongoing effort to improve the Radiation Monitoring System (RMS)reliability, the RMS Task Force identified these monitors for deletion/removal.
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| This modification disconnects and removes all portions of radiation monitors RM-1WR-3652A and RM-1WR-3652C, and installs terminal boxes to complete the communications loop on which these monitors communicate.
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| This modification does not affect any control features of the Radiation Monitoring System.Safet Summar These radiation monitors are located in an area where during normal operations, refueling and design basis accident conditions, radiation exposures can not rapidly increase due to equipment malfunction, equipment failure or personnel error.Since radiation monitors RM-1WR-3652A and RM-1WR-3652C are not safety related and are not required for safe shutdown, or to operate during accidents, it can be concluded that removal of the monitors does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Table 12.3.4-1 MEM/HO-930100/8/OS1 Change to Facility as Described in the FSAR Title: PCR-04418, Chemical and Volume Control System (CVCS)Mixing Tank Level Indication.
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| Functional Summar This modification adds a local sightglass level gauge for the Chemical Mixing Tank being part of the CVCS.The Chemical Mixing Tank is used to introduce chemicals into the Reactor Coolant System (RCS)in order to maintain Reactor Coolant Water Chemistry.
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| Operations must add the chemicals and then fill the mixing tank with water before sending the solution to the RCS.At this time, operations monitors the tank level during filling by peering down the funnel atop the mixing tank.This situation is unacceptable due to the possibility of overflowing the mixing tank.In addition, this situation could expose plant personnel to the chemical solutions by overfilling the tank due to the lack of tank level indication.
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| Safet Summa The CVCS is an initiating system.However, the Chemical Mixing Tank and associated piping including the vent, drain and fill valves are non safety components.
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| Only the Chemical Mixing Tank inlet isolation valve and the Chemical Mixing Tank outlet isolation valve are safety related valves.These root valves are not affected by the installation of the level gauge.Since the Mixing Tank is used only to add chemicals to the Reactor Coolant System (RCS), the tank is normally valved out of service.All valves associated with the mixing tank are normally closed valves, this includes the fill valves, the tank vent valves, and drain valves.The safety related inlet and outlet isolation valves are also normally closed valves.The new level gauge is a heavy wall pyrex glass tubular sightglass.
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| Included with the sightglass are two 316 stainless steel isolation valves and guard rod protectors.
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| The gauge is mounted such that full range indication on the Chemical Mixing Tank is provided.The level gauge upper connection ties into the fill line below the two normally closed fill isolation valves.The lower connection is below the outlet of the mixing tank on a normally isolated section of piping.In addition, a normally closed instrument root valve off of each process line connection is provided.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.3.4-03MEM/HO-930100/9/OS1 Change to Facility as Described in the FSAR PCR-04461, Waste Gas Recombiners Cooler Condenser Drains Functional Summar This modification removes, one existing excess drain line along with its.two isolation valves from the closed cooling water side on the two Waste Gas Recombiners Cooler Condensers.
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| These drains are piped to the equipment drain system.Since Component Cooling Water (CCW)contains corrosion inhibitors and is not contaminated, it is not desirable for it to be drained to the equipment drains.These same cooler condensers have drains on the inlet and outlets that are piped to the proper CCW drain.Safet Summa The only analyzed accident involving the waste gas system is the rupture of one waste gas decay tank.If the CCW side of the recombiner cooler condenser were to fail, it would not cause any safety problem.Also removing these two redundant drain lines cannot in any way decrease the reliability of the recombiner cooler condensers.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 11.3.2-02 MEM/HO-930100/10/OS1
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| Change to Facility as Described in the FSAR Title: PCR-04603, Condensate Polisher Service Air to Sluice Water Header Inlet Valve 1CE-691.Functional Summar Due to pressure build up in the Condensate Polisher Demineralizing (CPD)System, CPD Sluice Header Inlet Valve 1CE-691 could lift off the seat and allow sluice water to enter the Service Air System.Operating procedures require radwaste operators to isolate the manual valve 1SA-237 to prevent sluice water from entering the Service Air System.Isolating 1CE-691 by 1SA-237 is not the desired operating mode.This design change installs a second spring in the air actuator for valve 1CE-691, and it installs a check valve in line 7SA1-327-1 upstream of 1CE-691.The addition of the second spring to the air actuator does not alter the air actuator design.It is an optional arrangement offered by the vendor to increase the seating force of the air actuator.The addition of the 1" check valve is a good practice when two systems are cross-tied.
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| This provides passive protection of the Service Air System from introduction of sluice water from the CPD System.The implementation of these changes will permit system operation without the requirement for manual isolation between the two systems.Safet Summar The addition of the concentric spring to the air actuator for 1CE-691 and the addition of the check valve upstream of 1CE-691 do not adversely impact any accident mitigating or initiating systems nor are the basis to any technical specifications affected.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.3.1-02 MEM/HO-930100/11/OS1
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| Change to Facility as Described in the FSAR Title: PCR-04628, Feedwater Preheater Bypass Valve Closure (Stroke)Time Functional Summa This change involves the required response time for Feedwater Preheater Bypass Valves (FPBV)in the containment accident analysis.This change did not result in any physical change to plant equipment.
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| The FPBV response time has been changed from 10 seconds to 8 seconds to be consistent with the assumptions of the existing containment safety analysis.Saf et Summar The FPBV are normally open during plant startup, power operation, and plant cooldown to allow main feedwater flow to be directed toward the steam generator upper feedwater nozzle (AFW inlet nozzle).The FPBVs are containment isolation valves as noted in FSAR Table 6.2.4-1 and are the second radiological barrier.(The feedwater lines are a closed system and represent the primary barrier.)The FPBVs are also safeguards system isolation valves in that they receive a main feedwater isolation signal (MFIS)and prevent feedwater to a faulted steam generator when they are fully closed.Prevention of excessive amounts of feedwater to a faulted steam generator minimizes mass and energy releases to Containment during a secondary side pipe break, either main steamline break (MSLB)or main feedline break (MFLB)and therefore, minimizes the Containment pressure and temperature increases.
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| This in turn will minimize challenges to the Containment structure (a radiological barrier), minimizes containment leakage and ensures safety components inside containment are operating within their qualified pressure and temperature limits.For each steam generator, a MFIS terminates main feedwater (MFW)by initiating closure, (if not already closed)of the feedwater isolation valves (FIV), feedwater isolation bypass valves (FIBV), feedwater control valves (FCV), feedwater control bypass valves (FCBV)and the FPBVs.The existing containment analyses utilize design input regarding main feedwater (MFW), auxiliary feedwater (AFW)flow rates and valve closure times.The existing MFW input data assumes termination of all MFW 8 seconds after receipt of a MFIS when the plant is operating at or above 30 percent power, and 10 seconds when the plant is in a shutdown condition.
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| Therefore, the more stringent of the valve closure times is the criteria by which the valve must operate to remain within the=existing analytical limits.Since the peak Containment pressure and temperature have not changed, there is no impact to the Containment radiological barriers or equipment qualification.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists..FSAR
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| ==Reference:==
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| Table 6.2.4-1 MEM/HO-930100/12/OS1
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| Change to Facility as Described in the FSAR Title: PCR-04819, Normal Service Water (NSW)Sample Point in Secondary Sample Room Functional Summar This modification provides a permanent NSW chemistry monitoring location in the 240'levation of the Turbine Building.Long-term sampling of service water is required in order to monitor the effectiveness of cooling water chemical additions.
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| The permanent NSW monitoring point is relocated from the temporary location on the 261'levation of the Turbine Building to the 240'levation to minimize chemistry personnel travels The NSW monitoring point consists of a supply line from NSW through an isolation valve to the monitoring equipment which is provided by plant chemistry personnel and chemical vendors.This monitoring equipment can be changed out, added to, or removed as required upon E&RC Manager approval without any plant modifications.
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| The NSW exits the monitoring equipment through an isolation valve and returns back into the NSW System.Safet Summar The subject portion of the Service Water System is non-seismic Category I, non-safety class and is not considered available during accident and emergency conditions.
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| No credit is taken in the safety evaluation for this portion of the system.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.2.1-2 MEM/HO-930100/13/OS1 II Change to Facility as Described in the FSAR i Title: PCR-04987, Essential Services Chilled Water System Refrigerant Transfer System Functional Summar This plant modification made three changes associated with the Essential Services Chilled Water System.First, the modification provides information on the quality classification of the components of the essential services chillers.The change clarifies that some components that are not ASME Section III are still quality classification A.This modification states that the Refrigerant Transfer System is quality classification B.The quality classifications as described in this modification are consistent with the function of the equipment components.
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| Secondly, this modification provides a design for the addition of compression type fittings for the service water and refrigerant line connections to the heat exchanger for the refrigerant storage tank.This tank as well as the service water and refrigerant lines connecting to it are quality classification B.The change does not affect the operation of the essential services chillers.It merely provides a convenient means of disconnect for the heat exchanger that is less likely to leak than the present method.Thirdly, this modification changes the filler material that is specified for brazed joints.The change is made because the brazing procedure used on site requires filler material different from that specified by the vendor.The two filler materials allowed by the site brazing procedure make a joint as strong as the filler material specified by the vendor.Safet Summar The changes made by this modification can not create an accident or affect the mitigating ability of the chillers to respond to an accident or affect the operation of any equipment.
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| Defining the quality classification of the chiller components has been done prior to this modification.
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| This modification provides an explanation of the quality classifications.
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| It does provide a quality classification boundary between the operating chiller and the refrigerant transfer system not clearly defined previously.
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| This boundary provides a separation between the operating chiller and the refrigerant transfer system, which only provides a maintenance function.The addition of the compression fittings only affects the refrigerant transfer system.Therefore, the addition does not affect the operating chiller.The filler materials specified will produce a joint as strong as the filler material specified by the vendor.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR Reference Figure 9.2.8-03 Table 3.2.1-1 MEM/HO-930100/14/OS1 Change to Facility as Described in the FSAR PCR-05371, Refueling Water Storage Tank (RWST)Boron Concentration Functional Summar The purpose of this plant modification is to satisfy the requirements for a substantial increase in Reactor Coolant System (RCS)boron concentration starting in Cycle-5.RCS boron concentration was increased to permit a reduction and eventual elimination of all the Wet Annular Burnable Absorbers (WABAs)and many of the Integral Fuel Burnable Absorbers (IFBAs)and thereby reduce fuel costs.The impact on the plant is as follows: The boron concentration in the RWST and Safety Injection System (SIS)Accumulators has been increased from 2000-2200 to 2400-2600 ppmB to a)satisfy shutdown margin requirements and b)ensure post LOCA subcriticality (since RCCAs are assumed to be disabled by the LOCA)for Cycle-5 and beyond.The volume of NaOH in the Spray Additive Tank (SAT)has been changed from 2736-2912 gal to 3268-3964 gal to maintain the long-term pH in the containment sump between 8.5 and 11.0 after mixing with the RWST, SIS, RCS and other borated/unborated water volumes after a LOCA.The minimum pH during the injection phase is reduced from 8.6 to 8.2.The minimum volume of boric acid in the Boric Acid Tanks is changed from 21,400 to 24,150 gal in Modes 1-4 and from 7,100 to 6,650 gal.in Modes 5-6 to satisfy shutdown margin requirements and, together with makeup water, to accommodate RCS shrinkage.
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| The requirement that K,~~8 0.95 during refueling operations is changed to refer to the Core Operating Limits Report (COLR)where a boron concentration is specified to ensure that K,zz 5 0.95 (Tech Spec 3.9.1.a, 6.9.1.6.l.g and 6.9.1.6.2.a).
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| The existing RCS pH control program is to specify a slightly higher LiOH for a short period early in the cycle when RCS boron concentration will be significantly higher than in previous cycles.The higher LiOH concentration will prevent additional corrosion and activation of corrosion products that would occur as a result of the higher RCS boron concentration.
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| To accommodate the capabili,ty to provide a higher RCS boron concentration, the boric acid flow controller is rescaled and recalibrated to allow a higher flowrate of boric acid to the blender thus providing the capability to produce a higher boron concentration.
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| The switchover time from cold leg recirculation to hot leg circulation (and vice versa)has been revised from 24 (or 18 hours)to 6.5 hours.MEM/HO-930100/15/OS1 Change to Facility as Described in the FSAR 1 Functional Summar: (continued)
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| The earliest time after a LOCA at which post-LOCA hydrogen generation is expected to reach the 3%concentration level is revised from 7.33 days to 7 days.Tritium production is increased.
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| 10.The RWST level setpoints are not affected, however, the reset value for the HI and LO level alarm is revised to provide a greater operating band.Safet Summar RWST and SIS Accumulator Boron Concentration Increase The boron concentration in the RWST and SIS Accumulators must be increased to satisfy shutdown margin requirements for Cycle-5 and beyond, and to ensure post LOCA subcriticality assuming a)RCCAs are disabled b)all WABAs and some IFBAs are removed from the fuel, and c)the RWST/SIS accumulator solution becomes diluted with RCS and other water sources in the containment sump.The proposed RWST/SIS accumulator concentration range was determined by Westinghouse and is believed satisfactory for all'uture cycles.Westinghouse has determined a)that the higher concentration will have no adverse effect on the stainless steel container materials despite the slightly lower pH at 2600 ppmB than at 2200 ppmB;b)there is no danger of boron precipitation; c)corrosion of carbon steel by leakage of the higher ppmB water will not be increased significantly because the pH change is small and the pH is still in the range where corrosion rates are nearly independent of pH (carbon steel must be protected from corrosion by boric acid in any case);and d)none of the accident scenarios are adversely impacted by the higher boron concentration in the RWST and SIS Accumulators.
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| The boron concentration increase was reviewed for impact to material issues in affected systems by CP&L.It was concluded that there is no adverse impact by this change.In addition, it was determined that the EQ, Mechanical, HVAC, Electrical, I&C were impacted but that the proposed changes were acceptable.
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| The Civil Discipline was determined as not impacted.The mechanical issues covered include component operability, and process impact, fluid systems performance, seismic qualification, functional and regulatory criteria, accident analyses.The I&C issues covered included setpoint changes, fluid density effects, level changes, etc.MEM/HO-930100/16/OS1 Change to Facility as Described in the FSAR Safet Summar: (continued)
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| As noted in FSAR Section 4.6.4, the only FSAR Chapter 15 events crediting boron concentration in the RWST are Main Steam Line Break (MSLB), Main Feed Line Break (MFLB)and Loss of Coolant Accident (LOCA).The MSLB (15.1.4 and 15.1.5)and MFLB (15.2.8)have already.been evaluated at the existing lower bound concentration of 2000 ppm.Thus an increased boron concentration represents a more conservative situation since any return to power effects are further minimized.
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| Also, since"return to power effects" are minimized, the heat release from the core to the RCS and subsequently to the steam generator secondary side are also minimized.
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| Thus, the energy released to containment or outside containment during a MSLB or MFLB should not increase and therefore, the existing analyses (Chapters 6.2.1.4, 3.6A.3.2, and Appendix 3.11E)remain bounding.The LOCA (15.6.5)short-term transient is not impacted by the boron concentration increase as boron is not credited for mitigation purposes to meet 10CFR50.46 and 10CFR50 Appendix K criteria.However, the long-term post-LOCA boron concentration needed to keep the reactor core subcritical is impacted.As noted in Westinghouse's Evaluation, the post-LOCA sump boron concentration to keep the core subcritical long term (no credit for control rods)assumes specific inventories and concentrations.
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| Specifically, the Boron Injection Tank was assumed to be at the same concentration as the RWST.However, Calculation NSSS-0056 determined that without crediting the boron concentration in the BIT, the reduction in the Containment sump mass average boron concentration would be relatively small.Nuclear Fuels Section response to IRR-HM-1124 indicates that this was acceptable.
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| The FSAR (15.6.5 and 6.3.3.3)has been updated though to reflect that the long-term LOCA transient does credit boron concentration (namely, RWST and SIS accumulators).
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| The LOCA mass and energy FSAR 6.2.1.3 releases to containment for peak pressure and temperature are not impacted by the boron concentration increase since WCAP 8312/8264 (source input)maximize heat release with no credit for boron.Also, by inspection of FSAR Tables 6.2.1-4 and 6.2.1-9, Containment peak pressure and temperature generally occur prior to actuation of safety injection, thus, the effect of boric acid from the RWST would not impact the results.The LOCA mass and energy releases 6.2.1.5.1 to containment for the Appendix K (to 10CFR50)analysis are designed to produce a conservatively low containment pressure and do not credit boric acid addition.MEM/HO-930100/17/OS1 Change to Facility as Described in the FSAR Safet Summar: (continued)
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| The natural circulation cooldown evaluation (1364-096805) was impacted only in that it will take longer to achieve the shutdown margin required for cold shutdown.This is due mainly to the higher BOC initial RCS boron concentration.
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| However, Calculation NSSS-0055 determined that this was acceptable.
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| The boron dilution accidents (FSAR 15.4.6)are adversely impacted by a higher initial RCS boron concentration.
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| These accidents were reevaluated by Westinghouse which concluded that Tech Spec Figure 3.1-1 required modification for Modes 3, 4, and 5: The new figure is based on conservative assumptions regarding future fuel designs and core loading patterns, and is expected to satisfy all shutdown margin criteria.However, these shutdown margin criteria are reevaluated each cycle in the Reload Safety Evaluation based on the final fuel design and the final core loading pattern to verify compliance with all applicable shutdown margin criteria.Westinghouse reanalyzed the Inadvertent Actuation of ECCS event (FSAR Section 15.5.1)assuming 2700 ppmB.The results changed substantially but not because of the increased boron concentration.
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| FSAR Figures 15.5.1-1 and-2 changed because decay heat and automatic pressurizer spray operation, previously neglected, are now modeled in the analysis.Spray Additive Tank (SAT)NaOH Volume Increase The higher RWST/SIS accumulator boron concentration will have no adverse impact to materials exposed to containment spray or the sump solution provided the long-term pH is maintained between 8.5 and 11.0 as is currently required.Although the short-term injection phase minimum spray pH will decrease to about 8.2, this is well above the minimum acceptable value of 7.0 identified in the NRC SRP 6.1.1 Revision 2 and MTEB 6-1.The pH range and potential impact on materials has been reviewed as documented in the PCR 5371 design impact mechanical discipline evaluation and found to be acceptable.
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| In addition, the EQ review was also performed concluding no adverse impact.Iodine absorption by containment spray, and iodine retention by the sump solution is enhanced when sodium hydroxide is used as a spray additive.Although the minimum spray pH will decrease to about 8.2, this will not significantly degrade its enhancement capability.
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| The calculated spray removal coefficient (2)for elemental iodine as documented in FSAR Section 6.5.2.3.2, is 18'hr.and is independent of the pH value.This is the same method recently adopted by the NRC in Revision 2 of SRP 6.5.2 (December, 1988).The value used in the radiological offsite dose calculations presented in FSAR Section 15.6.5 has been conservatively reduced by approximately 50 percent to about 10 hr.~and was accepted by the NRC.MEM/HO-930100/18/OSl 0
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| ~Change to Facility as Described in the FSAR J4Saf et Summa (continued)
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| The long-term capability for the retention of iodine will not change either.This is because the minimum sump pH reached at the onset of the recirculation mode will be 7.0, as is required by SRP 6.5.2 Revision 2 and at the end of sodium hydroxide addition, the sump pH will reach 8.5, unchanged from the original criteria.At this time the minimum spray pH will be about 9.1 thereby, enhancing the spray solutions capability to retain iodine.Thus, there is no adverse impact to the CSS capability to absorb or retain iodine and no impact to the radiological offsite dose (or Control Room personnel dose)calculations.
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| The calculated maximum pH has increased from 10.8 to about 11.0.Due to previous NRC concerns regarding toxicity level, after an inadvertent Containment Spray actuation (Ref.NRC SRQ No.450.3), CP6L had previously informed the NRC that a time of 50 minutes would be needed to reach a pH of 10.5.Although the FSAR was later revised from 50 mins.to 100 mins., the NRC accepted 50 mins.as an adequate time to terminate NaOH addition.With the proposed increase in boron concentration and SAT volume, a new time of=80 mins.has been estimated to reach a pH of 105.This longer time is more conservative, since it gives the operators additional time to manually terminate NaOH addition before reaching high toxicity levels.The proposed NaOH volume range is based on the proposed boron concentration in the RWST, SIS accumulators and an assumed range of boron concentration in the RCS and other water volumes expected to enter the sump after a LOCA.The NaOH concentration will not change;thus there is no concern about the SAT and associated piping/valve materials compatibility or solubility.
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| The flowrate through the eductors will be unchanged, but the time to empty the SAT contents into the sump via the containment spray will increase.PATH-1 requires TSC to sample the sump and determine pH.Containment spray may be reactivated if necessary to achieve a sump pH between 8.5 and 11.0.This change has no impact on any other accident scenario or anticipated transient because this system is not actuated except on detection of high containment pressure conditions.
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| The mechanical design impact evaluation included in PCR-5371 discusses fluid system performance of the SAT, level setpoint changes, seismic qualification, etc, and concluded that there is no adverse performance effects by these changes.MB 1/HO-930100/19/OS 1
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| Change to Facility as Described in the PSAR Safet Summar: (continued)
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| Minimum Boric Acid Volume in the BAT Westinghouse determined the proposed minimum boric acid usable volumes based on shutdown margin requirements assuming an extended 18 month Cycle-5 fuel design and core loading pattern.These volumes together with reactor makeup water must also accommodate RCS shrinkage during cooldown.The changed minimum volumes have no adverse impact to any accident scenario or anticipated transient because the tanks normally operate near full capacity.The minimum volume requirements are reevaluated each RSE and verified to satisfy all criteria.The mechanical design impact evaluation included in PCR-5371 discusses the BAT seismic qualification, level setpoint changes, fluid performance, etc., and concluded there is no adverse impact by these changes.RCS and Refueling Canal Boron Concentration Westinghouse advised CPGL that the current requirement for boron concentration during refueling yielded K,qi near but less than 0.95 during the Cycle-4 refueling operation.
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| To ensure that K,zq 6 0.95 during Cycle-5 and future refuelings, boron concentration may need to be greater than 2000 ppm.Therefore, it is proposed that the boron concentration be provided in the Core Operating Limits Report (COLR)each cycle.The COLR is available before each refueling outage and the necessary procedures will be revised to require the higher boron concentration before the refueling outage.None of the accident scenarios or anticipated operational transients are adversely affected by the proposed'change, and no safety concerns have been identified.
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| Modified Reactor Coolant pH for Cycle-5 and Beyond Based on Westinghouse studies sponsored by EPRI, Westinghouse concluded that the slightly higher LiOH concentration specified does not constitute an unreviewed safety question as defined by 10CFR50.59(a)(2).
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| The phenomenon of interest for safety effects is potential stress corrosion cracking of the steam generator tubes caused by the change in reactor coolant chemistry.
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| It is Westinghouse's judgement that the modified Li/B coolant chemistry program will not adversely impact the integrity of the steam generator tubing material.None of the accident scenarios or anticipated operational transients are adversely affected by the proposed change, and no safety concerns, have been identified.
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| The impact of the maximum LiOH concentration (3.77 ppm)was reviewed for materials impact and determined to be acceptable.
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| MEM/HO-930100/20/Osl
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| Change to Facility as Described in the FSAR Safet Summar: (continued) 6.Boric Acid Flow to Blender Although the boric acid flow controller to the blender is being rescaled and recalibrated to provide the capability to borate up to.2800 ppm, there is no adverse safety impact by these changes.The flow controller is safety Class 3 for its pressure boundary only since it is included in the emergency boration path boundary too.The valve fails open but its failure to open or close has already been evaluated in FSAR Section 7.4.2.2.1 and Table 9.3.4-4 since it is controlled by the NNS Instrument Air System.7.Switchover Time Post-LOCA boron precipitation in the core may occur sooner during the recirculation phase because of the higher boron concentration.
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| Therefore Westinghouse recalculated the hot leg switch-over time and obtained 6.5 hours.The current value, 24 or 18 hours, is a generic value originally provided to all Westinghouse plants.The decrease to 6.5 hours is partly a result of the higher boron concentration, but primarily a result of the HNP specific analysis which modeled the relatively low RCS water inventory and high core thermal power.The switchover time will impact emergency operation procedures but not component operation or performance.
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| Post-LOCA Hydrogen Generation The increased boron concentration creates a slightly more acidic solution for the injection phase spray.Since the NaOH concentration and Containment Spray System (CSS)flowrates remain essentially the same, a slightly lower pH will result, as determined in Calculation 14.06.000-21.
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| There is no impact to aluminum corrosion rate of aluminum decreases with decreasing solution pH.However, the post-LOCA hydrogen production rate, due to boric acid reacting with zinc is pH dependent.
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| NUREG/CR-2812,"The Relative Importance of Temperature, pH and Boric Acid Concentration on Rates of Hq Production from Galvanized Steel Corrosion", Sandia Laboratories, November 1983, indicates an increased corrosion rate of about 7%for 2600 ppmB over 2000 ppmB for high temperature (h 260'F).High containment temperature exists for a relatively short time (15-30 min per FSAR Figure 6.F 1-5)and the total calculated increase in Hz production, about 3.5S, is considered very small.MEM/HO-930100/21/OS1 Change to Facility as Described in the FSAR Safet Summar: (continued)
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| Nevertheless, using this input from Westinghouse the earliest time to a 3 percent hydrogen concentration will be reduced from 7.33 days to approx.7 days (Ref.Gale.3-A-8¹03 and 3-A-8¹05).This does not impact equipment operation or operator procedures since their actions are based on hydrogen concentration level, not time in to a LOCA event.'uring the recirculation phase of a LOCA, the maximum spray pH could increase from 10.8 to 11.0.This has a potential to increase the hydrogen generation rate from aluminum corrosion.
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| However, the calculated hydrogen generation as reported in FSAR Section 6.2.5.3.1 used a fixed but conservatively high corrosion rate provided by Regulatory Guide 1.7 and is thus independent of pH.Therefore, there will be no change to the calculated hydrogen production rate from aluminum.In any case, the increase in calculated pH (from 10.8 to 11.0)will only occur for about 10 minutes since at the completion of NaOH, the spray pH will be the same as that of the sump, i.e., 9.7.This very small amount of time at a slightly higher pH will have negligible impact on the actual hydrogen production rate.Tritium Production The purpose of the higher RWST/SIS accumulator boron concentration is to permit a higher RCS boron concentration.
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| A higher RCS boron concentration requires a slightly higher RCS LiOH concentration early in each cycle to maintain the desired RCS coolant chemistry.
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| The higher boron concentration will cause more tritium production.
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| Although LiOH concentration will be slightly larger for a brief period at BOL, it will be slightly lower for much of the cycle, and contributes no net change to tritium production.
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| Also, tritium production from lithium is minimized by using Li enriched to 99.9%Li~.The maximum tritium produced during any future cycle will be about 14%over that produced during Cycle-4.An increase in Tritium production up to 25%was previously evaluated and approved.FSAR Table 11.1.1-2 presents Design Basis Reactor Coolant Fission and Corrosion product Specific Activity which includes the maximum permissible tritium concentration of 3.5 pCi/gm.Although Chapter ll indicates that these activities are the source terms used for calculating the radiological consequences for Chapter 15 events, and although Chapter 15 references Chapter 11 as the source of the source terms, tritium is not currently modeled in the calculation of offsite doses for Chapter 15 events;thus the calculated offsite doses do not change as a result of increased tritium production.
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| The omission of tritium is not significant because tritium emits a low energy beta (19 kev (maximum), 6 kev (average))
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| and, although it has a 12.3 year decay half-life, its biological half-life is only about 12 days.Thus, tritium is one of the least hazardous radionuclides and its inclusion in the radiological consequences in Chapter 15 would not increase the calculated doses significantly.
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| MEM/HO-930100/22/OS1 Change to Facility as Described in the FSAR Safet Summar: (continued)
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| The other nuclides listed in Table 11.1.1-2 are also affected by initial boron concentration.
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| The differential equations used to calculate those activities (Section 11.1.1.1, Equations 1-4)use initial boron concentration to model the feed and bleed process.An increase in initial RCS boron concentration requires additional feed and bleed to reduce the concentration from the initial (BOC)value to the final (EOC)value.The larger amount of feed and bleed will flush more material (fission products)out of the RCS and therefore reduce the design basis activity concentrations.
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| Thus, Table 11.1.1-2 continues to be valid for its intended purpose.10.RWST HI/LO Level Reset Values The RWST HI and LO level alarm reset values provided in temporary Modification PCR-5421 have been"rolled over" into this PCR 5371 and made permanent.
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| Essentially, the reset values will be closer to the setpoint values for both HI and LO alarms.This will provide a greater operating band for the RWST tank level during normal plant operation.
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| No changes are being made to actual HI and LO level alarm setpoints.
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| Since no changes to the alarm setpoints are being made, there will be no changes to the minimum water inventory required in the RWST for its normal and safety functions.
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| The alarm reset values do not have any safety functions.
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| Hence, there is no impact to any safety analysis described in the FSAR.Since the operating band will be increased, this actually enhances safe plant operation by reducing the amount of operator time or distraction in trying to refill the RWST to a level that clears the alarms.These changes do not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Sections 3, 4, 6, 9, ll, 15 MEM/HO-930100/23/OS1 Change to Facility as Described in the FSAR Title: PCR-05393, Waste Gas Traps Functional Summar This plant modification to the Waste Gas Processing System (WGPS)installed new gas traps which have sight level gages.The Waste Gas traps were initially purchased without level gages.Thus there was no.indication for the operator to know the separation of condensed water from the gas.Safet Summar The new Waste Gas traps do not affect normal operation of the WGPS.The WGPS is a non-safety class system and is not required for safe shutdown of the plant.Xn case of a level gage glass break, isolation valves are provided to isolate the particular gage.FSAR Section 15.7.1 has analyzed the failure of the WGPS and associated piping and has concluded that the radioactivity released to the environment is less than the 10CFR100 dose guidelines.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 11.3.2-1 MEN/HO-930100/24/OS1 Change to Facility as Described in the FSAR Title: PCR-05403, Radiation Control Area/High Temperature Ionization Detectors Functional Summar This modification allows for the replacement of Nittan Model NID-48F ionization detectors with Base Model 48RBL-CP as part of its UL approved FC-72 Fire Detection System.Ionization type smoke detectors contain electronic components with Americium 241 (a radioactive source)and as such have a finite life expectancy.
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| The typical ionization detector is designed to last a minimum of 10 years depending on its operational environment.
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| The Nittan detectors installed at Harris are reaching the end of their operational life cycle.Nittan Corporation closed its manufacturing facility in Chicago after the peak of nuclear construction in the US and relocated in Surry, England.Since Nittan no longer manufactures the detector and base in the United States, the Underwriter's Laboratory (UL)listing was not renewed.The UL listing was simply not renewed due to the annual cost involved and not based on any noncompliance issue.Nittan indicated that no change had been made in the devices which would cause them to deviate from the original UL listing requirements.
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| The manufacturing process in the United Kingdom (UK)is controlled by British standards and the Fire Offices Committee of the British Standards Institution (FOC)approval in lieu of the UL requirements.
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| This ensures adequate monitoring and quality control over the manufacturing process in the UK.The FOC certification is considered an acceptable replacement for the U.L.listing.This modification allows the use of the Nittan Model NID-48F ionization detector without a Underwriter's Laboratory Listing.Safet Summar Nittan manufacturers the Model NID-48F ionization detector and associated Base Model 48RBL-CP in the United Kingdom to British standards.
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| Nittan has indicated that no changes have been made in the devices which would cause them to deviate from the original UL listing requirements which was discontinued due to the annual renewal cost involved and not due to any noncompliance issues.A sample detector from the United Kingdom was examined by CP&L Engineering and I&C Maintenance.
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| The detector was identical with the exception of the mesh insect screen which had a smaller cross-section to better prevent intrusion.
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| The sample base was identical with exception of Terminal Point P+.Terminal Point P+did not have the factory installed jumper bar between the two P+terminals.
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| The P+terminal is used for input to the local graphic panels and is electrically equivalent.
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| The detectors and bases are considered a direct replacement compatible with the FC-72 series Fire Detection Control System installed at the Harris Plant.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 9.5.1 MEM/HO-930100/25/OS1 Change to Facility as Described in the FSAR PCR-05495, Heater Drain Pump Seal Flow Switch Isolation Valves and Piping Functional Summa This plant modification to the Heater Drain Pump Seal Water System adds isolation valves and bypass lines around the heater drain pumps seal flow swi.tches and flow orifices.This modification provides a means of taking the flow switches or orifices out of service without reducing plant power.Safet Summa The piping and isolation valves added at the flow switches are the same material type, grade and rating as the existing components therefore mechanical reliability is maintained.
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| The addition of the isolation valves and material change (stainless steel)on the piping with the seal water flow orifice increases the reliability since the previous piping was experiencing erosion.The added bypass at the flow orifice provides the same approximate pressure reduction as the original flow orifices.The Heater Drain Pump Seal Water System is classified as non-nuclear safety.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 10.2.2.-06 MEN/HO-930100/26/OS1
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| Change to Facility as Described in the FSAR Title: PCR-05676, Cooling Tower Basin Low Level Setpoint Functional Summar This modification to the Cooling Tower Basin has two phases.The first is the lowering of the low level setpoint on Level Switch LS-01-CW-1931.
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| This is to allow the basin to be operated at a lower level (to reduce blowdown from the basin)without alarming.The second phase is the raising of the cooling tower weir to allow the water level to be raised in the basin while eliminating blowdown from the basin.With the raising of the weir, the level control setpoints were also revised to control operation of the water level above the weir plate.Safet Summa The following key safety areas were reviewed.Structural analysis, materials compatibility, and mechanical failure.The cooling tower basin design calculation was reviewed.The elevation change on the weir plate and the level control setpoint changes will allow the water level in the basin to be raised.This elevation change is within the allowable for the basin per the existing calculation.
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| The replacement of the weir plate is with the same material type and grade as the original design therefore material compatibility is satisfied.
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| The operation of the circulating water pumps was reviewed due to the changing of the water level.The short term lowering of the basin level is a negligible change to the circulating pump suction.The long term raising of the water level will have a more positive effect on the pump operation.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a'ifferent type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 10.4.5-2 MEM/HO-930100/27/OS1 Change to Facility as Described in the FSAR PCR-.06265, Leak Repair of RTD Bypass Manifold Isolation Valve 1RC-22 Functional Summa This modification removes the handwheel on the C Loop RTD Bypass Mani, fold Isolation Valve 1RC-22 and adds a seal cap over the valve stem to repair a leak on the valve.This leaves the valve inoperable in the full open position.Safet Summar The seal cap installed by this modification was supplied by Kerotest to the same specification and quality standards as the original valve components.
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| The body and bonnet will remain as the primary pressure retaining parts with the seal cap as a backup providing full pressure seal capability.
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| The seismic/stress analysis effects due to the weight of the cap offset by the elimination of the handwheel were evaluated by NED Stress Analysis (working through Westinghouse) and found to be acceptable.
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| The system is thus kept in the"as analyzed" condition.
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| This valve's operability is not required for plant operation or shutdown.This change does not increase the probabili.ty or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 5.1.2-1 MEM/HO-930100/28/OS1 Change to Facility as Described in the FSAR Title: PCR-06273, Leak Repair of RTD Bypass Manifold Isolation Valve 1RC-953 Functional Summar This modification removes the handwheel on the C Loop RTD Bypass Manifold Isolation Valve 1RC-953 and adds a seal cap over the valve stem to repair a leak on the valve.This leaves the valve inoperable in the full open position.Safet Summar The seal cap installed by this modification was supplied by Kerotest to the same specification and quality standards as the original valve components.
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| The valve body and bonnet will remain as the primary pressure retaining parts with the seal cap as a backup providing full pressure seal capability.
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| The seismic/stress analysis effects due to the weight of the cap offset by the elimination of the handwheel were evaluated and found to be acceptable.
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| The system is thus kept in the"as analyzed" condition.
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| This valve's operability is not required for plant operation or shutdown.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 5.1.2-3 MEM/HO-930100/29/OS1 Change to Facility as Described in the FSAR Chemical and Volume Control System (CVCS)Mixed Bed Demineralizer Resin Usage Functional Summer Two flushable mixed bed demineralizers in the CVCS assist in maintaining reactor coolant purity.Normally, lithium-form cation resin and hydroxyl-form anion resin are loaded into the demineralizers.
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| However, this change allows that during reactor shutdown prior to refueling, a fresh charge of mixed resin containing hydronium-form cation resin may be used instead of the lithium-form cation resin.In both cases, the anion resin is converted to the borate form during operation.
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| Both types of mixed bed formulations remove fission and corrosion products.The resin beds are designed to reduce the concentration of ionic isotopes in the purification stream, except for cesium, yttrium and molybdenum, by a minimum factor of 10.Safet Summar The CVCS mixed bed demineralizers are used to maintain RCS purity.The type of resin used will not affect its operation.
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| Operation of the demineralizers does not effect the safe shutdown of the plant and provides no safety function.Any malfunctions will not affect safe operation of the plant since, the system can be isolated without affect on RCS chemistry.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 9.3.4 MEM/HO-930100/30/OS1
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| Change to Facility as Described in the FSAR Title: Meteorological Program Change Functional Summar FSAR Section 2.3.3 was r'evised to incorporate changes and improvements to.the on-site meteorological data systems.These changes include: 1)replacement of the data logger sensor system with a microcomputer based sensor system;2)Deletion of the Cambridge Dewpoint measurement system;3)Changing of the verb tense for data obtained from obsolete systems and the corresponding system descriptions.
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| The data logger sensor system employed a Monitor Labs 9300 data logger and a Techtran microdisc.
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| The data logger system was replaced by an ADAC microcomputer based data acquisition and storage system to permit access to immediate past historical records by ordinary microcomputers and to permit simultaneous, multiple, remote data access.The Cambridge cooled mirror dewpoint sensors were removed from service because the reliability of these systems were low in comparison to the lithium chloride system and no significant differences were found between dewpoint temperatures measured by either system.Safet Summar The Meteorological Program is used to aid in determining the appropriate protective action recommendations to be made for the public.The addition of a microcomputer based data acquisition which allows simultaneous, multiple, remote access to the data improves the ability to make such recommendations.
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| The consequences of any accident, if affected at all, are reduced.The dewpoint measurement systems are not used in determining accident consequences or protective action recommendations and therefore removal of the Cambridge dewpoint systems has no effect on safety.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 2.3.3.1 MEM/HO-930100/31/OS 1
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| Change to Facility as Described in the FSARTitle: PCR-06133, Essential Services Chilled Water System Chilled Water Setpoints Functional Summar This modification reduces the essential chiller low flow trip from approximately 900 gpm to 700 gpm.Essential chiller trips have occurred as a result of chilled water flow oscillations when the nonessential chilled water loop is aligned to the safety related chilled water loop.The low flow trip setpoints for FS-9429Al/Bl previously corresponded to approximately 900 gpm.Normal flow is approximately 1200-1300 gpm.This modification reduced the setpoint for the low flow trip to 700 gpm for each instrument.
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| This modification also changes some wording in the FSAR.The FSAR stated that the chilled water pump is monitored for flow failure by a chiller trip on low-low chilled water flow.There is only one chilled water low flow switch, so the low-low is changed to just low by this modification.
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| the FSAR stated that the flow switch in the chilled water line is interlocked with the chiller to prevent the chiller start up before the flow of chilled water is established.
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| This is correct but the flow switch will also trip the chiller in the event of chilled water flow failure.This description was added to the FSAR.The FSAR stated that the flow switch in the chilled water line is interlocked with the chiller to prevent the chiller start up before the condenser water is established.
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| This is not correct and the statement is eliminated.
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| Safet Summar Flow is determined by measuring the pressure drop across the chiller and converting this value to the flow that would produce that pressure.When the nonessential loop is aligned to the safety related loop, pressure fluctuations occur until the flow reaches steady state conditions.
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| According to the vendor manual, a chilled water flow permissive is required.This permissive may be a low flow trip or a chilled water pump relay energized indication.
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| Either of these type permissives is satisfactory, because both will indicate that the chilled water pump is running.Xt is also important that the water in the chiller not freeze.There are two other permissives, low refrigerant pressure and low chilled water temperature trip, that will do this.A low flow trip setpoint at 700 gpm is sufficiently high to positively indicate that the chilled water pump is running and sufficiently low to avoid nuisance trips.Avoiding nuisance trips is necessary to insure pump reliability.
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| This modification contributes to increased safety.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 7.3.1 and 9.2.8.5 MEM/HO-930100/32/OS1 Change to Facility as Described in the FSAR PCR-03950, Reactor Auxiliary Building Normal Ventilation (RABNV)Supply System S-3 Fan Modulating Dampers Functional Summa This modification to the RABNV System converted the previous inlet modulating dampers AV-D104-1 and AV-D105-1 to isolation dampers and the outlet dampers AU-D106-1 and AV-D107-1 to modulating dampers.The RABNV supply and exhaust system provides normal ventilation for areas of the RAB and maintains airflow from areas of low potential radioactivity to areas of progressively higher potential radioactivity.
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| The system maintains the RAB at a slight negative pressure below outside pressure to prevent an uncontrolled and unmonitored release of radiation during normal plant operation.
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| Differential pressure transmitters are installed at different RABNV elevations and at the outside atmosphere.
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| The supply fan flow control dampers are modulated by a pressure differential controller which receives a signal from these transmitters to maintain a slight RAB negative pressure below outside pressure.When an RABNV exhaust fan is placed out of service (3 fans operating) and/or RAB hatches out, doors open, etc., the RAB pressure becomes less negative than the design requirement and the supply fan modulating damper throttles back to reduce supply flow and accommodate the negative pressure requirement.
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| This modification was implemented because the damper was located so close to fan intake, an uneven velocity profile resulted in an unstable fan balance, resulting in vibration problems and ultimately fan blade failures.Safet Summar This modification does not change the design control features nor operating parameters of the RABNV System.Only the damper functions are switched.The modification increases the reliability of the RABNV System.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.This change does not increase the probability or consequences of analyzed accidents, nor introduce a di.fferent type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.4.3-02 MEM/HO-930100/33/OS1 8
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| Change to Facility as Described in the FSAR Tit1e: PCR-03716, Boron Recycle Holdup Tank Bladder Functional Summa This modification reworked the diaphragm seal ring in the Boron Recycle Holdup Tank (BRHT), replaced the diaphragm, added a water source to the bottom of the BRHT Loop Seal and added a Condensate Trap to prevent blowing a loop seal charge on line 7BR3/4-406-172.
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| The previous design of the Recycle Holdup Tank Diaphragm Seal Ring could have allowed hydrogen and fission gases to escape into the room atmosphere before being processed by the waste gas system.The bladder mounting/sealing detail of the diaphragm seal ring was enhanced to allow future work on the seal ring to be performed on top of the bladder (clean side).Safet Summa The seal ring and diaphragm work was on the non-Q/non-seismic portion of the tank.The diaphragm was replaced with material of superior quality.The piping line from the primary makeup water source was routed to the BRHT Overflow Loop Seal, ensuring a water solid loop seal which will also prevent hydrogen/fission gas egress from the BRHT.This non-Q water source piping breeches safety class 3 tank piping (atmospheric pressure), and is treated as such (Eg.seismically supported&built to ASME standards).
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| A drain line (7BR3/4-406-1&2) with a loop seal on the eductor line from the recycle holdup tank occasionally loss its charge (water column)due to evaporation, line pressurization, etc.This modification installs a ball float condensate trap upstream of the loop seal.A check valve downstream of the loop seal was removed.The loop seal water column and drain trap will prevent backflow from equipment drains.The pipe supports added to the new piping have been located and designed to assure that the pipe and support structural members adhere to acceptable standards for stresses and deflection.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 9.3.4-06 and 9.2.3-2 MEM/HO-930100/34/Osl
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| Change to Facility as Described in the FSAR PCR-00745, Condensate Storage Tank Diaphragm Removal and Nitrogen Piping Reroute.Functional Summar This modification remove's the diaphragm from the Condensate Storage Tank (CST)and relocates the nitrogen sparging flow indicators on the CST and the Reactor Makeup Water Storage Tank (RMWST).The diaphragm was installed in the CST for dissolved oxygen control.An inspection revealed water on top of the diaphragm, indicative of a leak.The diaphragm was twisted in the center of the tank, preventing plant personnel from making the repair.The diaphragm was removed during RFO¹4 and may be replaced during RFO¹5.The CST will be continuously sparged with nitrogen to keep the oxygen content at an acceptable level.The nitrogen sparging flow indicators on the CST and RMWST were relocated farther upstream to provide better operator access and eliminate a potential safety hazard.The nitrogen flow indicators are also replaced with ones with a narrower scale.Safet Summar Although the modification is being performed on non-safety related components (ie, diaphragm, non-Q portion of nitrogen piping)it was implemented as a safety related change.The nitrogen piping and pipe supports used for this purpose have been located and designed to assure that the pipe and support structural members all adhere to acceptable standards for stresses and deflection.
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| The nitrogen piping in the bottom of the tank is configured in such a way (located away from the tank outlet)that the nitrogen will not be drawn into the auxiliary feedwater suction.The removal of the CST Diaphragm and relocation of the nitrogen flow meters do not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 9.2.3-02, 9.2.6-01, 10.1.0-04 MEM/HO-930100/35/OS1 Change to Facility as Described in the FSAR Title: PCR-04961, Emergency Core Cooling System Atmospheric Release Scenario via Refueling Water Storage Tank Functional Summar A scenario was discussed in Westinghouse Letter CQL-89-556 involving an unidentified potential leakage path to the environment during post-LOCA recirculation.
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| The leakage path would occur after resetting the Safety Injection (SI)signal, while the Emergency Core Cooling System (ECCS)is in recirculation phase, with a Volume Control Tank (VCT)low-low level signal, and the single failure of Refueling Water Storage Tank (RWST)supply line check valve 1CS-294.PCR-4961 addresses this problem and eliminates it through the use of different control switches.The valves in question are 1CS-291 (LCV-115B) and 1CS-292 (LCV-115D)
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| RWST to charging pump valves.These valves had control switches which are spring return to normal (center)position.The replacement control switches have a pull-to-lock position.As this switch is placed into pull-to-lock, it will first initiate a close signal to its associated valve.Then after pull-to-lock is achieved, any valve motion signal is blocked and the valve will remain closed.Safet Summar Replacement switches are physically identical to the original ones in appearance and operation.
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| The pull-to-lock feature is noted on the faceplate engraving and will function the same way as other pull-to-lock switches on the Harris Main Control Board.Control circuitry for valves 1CS-291 and 1CS-292 will not be affected, hence valve operation is the same as before with exception of the pull-to-lock feature.The pull-to-lock feature will play an important role in preventing the potential leakage path identified by Westinghouse Letter CQL-89-556.
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| Under special conditions, pull-to-lock can be utilized to maintain 1CS-291 and 1CS-292 closed.Pull-to-lock for these valves should only be used in that special case to avoid the release scenario mentioned earlier since its use will also prevent an SI signal from opening these same valves.The original control switches, being spring return to normal position, could only block the SI open signal when physically held in the closed position.On the other hand, pull-to-lock on the new switches could be left unattended defeating the SI open signal until released.To minimize the possibility of switch misoposition, annunciation is added to the MCB to provide the operator with visual and audible indication when either control switch is placed in the pull-to-lock position.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Table 6.3.1-1 and 6.3.2-6 MEM/HO-930100/36/OS1 Change to Facility as Described in the FSAR Title: PCR-02512, Service Air Regulators to Chiller Expansion Tank Functional Summar This modification insta11ed additional air regulators on the Service Air Makeup lines to the Essential Services Chilled Water System (ESCWS)Expansion Tanks (Train A&B)to provide improved tank pressure control.Safet Summar The added air regulators are non-safety related and are installed in the non-safety related portion of the system.This modification improves the safe operation of the ESCWS by reducing tank pressure perturbations, allowing the ESCWS pumps to operate more efficiently.
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| Although this modification increases the number of mechanical joints in the Service Air System the Service Air System is not required for safe shutdown of the plant.In addition, failure of service air make up to the ESCWS will not affect the safe shutdown of the plant during a design basis accident situation because service air is assumed to fail and service air to the ESCWS expansion tanks is automatically isolated in an accident situation.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.2.8-1 and 9.2.8-3 MEM/HO-930100/37/OSl Change to Facility as Described in the FSAR PCR-05899, Reactor Cavity Seal Ring Instrument Air Piping Functional Summar This modification removed Reactor Cavity Seal Ring piping regulators lIA-979 and 1IA-983 and their associated check valves.These components are replaced with cart mounted units to be utilized when needed.Provisions are made to permit connecting the cart into the permanent piping by flexible hose.The use of the cart mounted regulators and check valves will permit calibration of these items prior to their connection into the system and repairs will be more convenient to implement outside the containment if necessary.
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| Originally, two flexible hoses were required in the system.One from the compressed gas bottles to the regulator and one from the piping to the bottom of the cavity.The cart concept introduces two new flexible hoses with quick disconnects.
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| These segments of the piping system were previously rigid piping.However, steel reinforced high pressure hydraulic hoses are used to prevent/minimize damage to the hoses.Safet Summar The normal pressurization for the Reactor Cavity Seal Ring is with the plant's Instrument Air System.Once inflated, the seal ring does not consume air since it is a leaktight, fixed volume component.
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| In the event an air supply line is ruptured, check valves are provided to prevent bleed down of the inflatable cavity seals.In the event the instrument air is lost, the bottled gas system is the back-up.The b'ack-up system may use either compressed air or compressed nitrogen.The cart is not permitted to be installed in the system with the Unit at power.The cart mount components are of the same types of components used previously in the system.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.3.1-3 MEM/HO-930100/38/OS1 Change to Facility as Described in the FSARTitle: PCR 05527, Spent Fuel Bridge Crane Drive System Replacement Functional Summar This modification upgrad'es various control and operational features of the Spent Fuel Bridge Crane which improve the safety and reliability of fuel handling operations.
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| The modification includes a Mitsubishi motor drive and a GE/FANUC model 90-30 programmable controller mounted in a new cabinet and located on the bridge platform.A new sliding control pendant mount is added.Software and a laptop computer will be used in making any changes in the program for the controller in the future.Safet Summa The Spent Fuel Bridge Crane is non-nuclear safety related seismic category I.The proposed spent fuel handling crane upgrades will increase the safety and reliability of the refueling operation as well as the receipt of spent fuel from other plants.The crane safety requirements and interlocks as outlined in FSAR Section 9.1.4.2.2.4 will not be affected by this modification.
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| The modification to the control pendant will improve the operator's ability to safely transport and store spent fuel as the proposed pendant mount will allow the operator to control the crane with one hand and leave one hand free to guide the fuel bundle.The crane is supplied with a non-safety related power source.If there is an electrical failure with a load on the crane the load will not drop.The brakes on the crane motor are equipped with a fail safe solenoid that is backed up by a dog-and-rachet mechanical brake which will safely hold the load.As a result of this modification there will be no additional electrical load on the crane.All electrical parts are Class"E" and are not seismically supported, due to their light weight and potential impact on spent fuel being bounded by the Spent Fuel Bundle Load-Drop Analysis.This preserves the integrity and safety of any equipment that may be located under the crane travel.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Table 3.2.1-1 and Figure 9.1.4-3 MEM/HO-930100/39/OS1 Change to Facility as Described in the FSAR Title: pCR-05307, Residual Heat Removal/Low Head Safety Injection Isolation to the Refueling Water Storage Tank Functional Summar This modification installs an additional manually operated isolation valve between the Reactor Coolant System (RCS)hotlegs and the Refueling Water Storage Tank (RWST).The change is being made to prevent the potential for an unmonitored release via the RWST during Residual Heat Removal (RHR)/Low Head Safety Injection (LHSI)Recirculation.
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| Safet Summar The valve being added is part of the Safety Injection System and is safety class 2.The valve will increase system isolation reliability.
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| The new valve meets the original design and material specification requirements of the Safety Injection System.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 6.3.2-3 MEM/HO-930100/40/OS1 Change to Facility as Described in the FSAR Title: PCR-05519, Provide Level Indication For 1A-SA and 1B-SB Emergency Diesel Generator Jacket Water Standpipe Functional Summar This plant modification adds a more effective and reliable means of monitoring 1A-SA and 1B-SB Emergency Diesel Generator (EDG)Jacketwater level.A local level indication instrument is added to the EDG Jacketwater Standpipe.
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| The standpipe fill/drain line located at the bottom of the standpipe is utilized as the instrument tap.A root valve is added on the fill/drain line to allow instrumentation sensing line attachment.
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| The previous instruments had a limited span which could lead a misidentification of the actual standpipe level.Safet Summar This change incorporates a local level indicator for the EDG Jacketwater Standpipe to assist plant operators in monitoring standpipe level.The addition of the local level indicator on both 1A-SA and 1B-SB EDG Jacketwater Standpipe affects the ASME Section III jacketwater piping associated with each diesel.The mechanical changes incorporate an ASME piping root valve to the standpipe fill/drain line and maintains the level of design for this section of the Jacketwater Cooling System piping.Instrumentation requirements and design incorporates a seismically qualified instrument and seismic Category I instrument support for mounting.The stainless steel construction of the process tubing and instrument bellows and housing maintain material compatibility with the jacketwater process fluid.Since this application adds only a local level indicator, no diesel control feature or operating parameter is affected.The operability and design intent of the EDG or Jacketwater Cooling System is unchanged.
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| By maintaining the ASME Section III piping design for the jacketwater piping change and the seismic integrity of the process sensing line and instrument, no new type of failure is introduced by this modification.
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| The independent alarm associated with each standpipe level is not altered or affected by this modification.
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| This level alarm ensures operations is alerted in the event of reduced standpipe inventory.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 9.5.5 and Figure 9.5.5-01MEM/HO-930100/41/OS1
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| Change to Facility as Described in the FSAR PCR-02982, Containment Isolation Valve Access Functional Summar This modification provides access to the Containment Isolation System (CIS).HVAC valves.Quarterly LLRT necessitates inspection and possible adjustment or replacement of seals for 42" and 8" butterfly valves to achieve a successful leakage rate.The valves are located in Containment Pre-entry Purge (CPP)and Normal Containment Purge (NCP)Systems.This modification involves the addition of four access doors to the ductwork to gain access to the 42" CIS valves and a removable duct spool piece to gain access to one 8" CIS valve to facilitate valve seal adjustment or replacement.
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| Safet Summar The CIS provides isolation of lines penetrating containment (which are not required to be open for operation of the ESF System)to limit release of radioactive materials to the atmosphere during a LOCA.The HVAC isolation valves will not be modified and will not be adversely impacted by this modification.
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| The valves will remain functional to isolate the containment building within the required closure time(s)upon a fuel handling accident inside containment and following a LOCA or steam line break.The CPP and NCP Systems are not safety related and are not required to operate under accident conditions.
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| Upon loss of power, the systems will shutdown.A small portion of the ductwork beyond the isolation valves was designed to Safety Class 3 requirements.
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| This portion of, the ductwork and piping beyond the isolation valves have been voluntarily upgraded, but for operational considerations, this ductwork and piping is classified as non-nuclear safety.Access ways installed in either safety related or NNS-seismic ductwork will maintain the same Q-class integrity as the ductwork that they are installed in.Installation of access ways will not impair the ability of the CIS to achieve its safety function.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 6.2.2-11 and 6.2.2-15MEM/HO-930100/42/OS1 Change to Facility as Described in the FSAR Title: PCR-01982, Vacuum Pump Auto Start on Low Vacuum.Functional Summer The taps on the Main Condenser for the pressure switches that provide an automatic start for the Standby Condenser Vacuum Pump were located at a higher elevation than the pressure switches.The steam in the instrument tubing running to the instruments sometimes condensed.
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| This created a water column in the lines giving an indication of higher condenser pressure than actually existed.PCR 01982 was originally a temporary modification to disable the low vacuum auto start signal until a permanent solution could be designed.PCR 01474 is a temporary modification that added purge meters to the instrument lines to keep them purged of condensation.
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| This modification provides for taps on the condenser that are lower than the instruments.
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| It routes the pipe and tubing such that the flow of condensate will be back to the condenser.
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| It eliminates the need for heat tracing.It removes the piping, tubing, hangers, and heat tracing for the taps that are currently used for the instruments.
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| It eliminates the condenser vacuum pump auto start auto trip feature.This modification makes temporary modifications PCR 01982 and PCR 01474 permanent.
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| One fitting on the purge lines was changed to allow an instrument to be connected to the purge lines while'he flow meter is also connected.
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| Safet Summar This modification is an improvement over the previous design.It eliminates the condensation problem by the simple method of sloping the lines back to the condenser.
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| It also eliminates the need for heat tracing.When water cannot stand in the piping and tubing it will not freeze.This is because the heat from the latent heat of vaporization will prevent the temperature of the condensed water in the tubes from dropping significantly below the saturation temperature.
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| The new lines will still serve the same function but will do so more reliably.Eliminating the condenser vacuum pump auto start auto trip feature was done by temporary modification PCR 01982 and has proven successful.
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| Operating experience has demonstrated that averting a low vacuum trip by starting a second vacuum pump would be a rare occurrence.
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| These changes do not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 10.1.0-02 MEM/HO-930100/43/OS1
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| Change to Facility as Described in the FSAR Title: Radiation Monitoring System Data Base Changes Functional Summar II This change to the Radiation Monitoring System Data Base resulted from review of an industry event which identified where a non-safety related (RM-11)processor malfunctioned and downloaded erroneous data to all radiation monitors (safety and non-safety) resulting in the monitors being inoperable.
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| The Harris Plant has the same design, with three slide switches available to inhibit RM-11 control of the safety-related (RM-80)monitors.These switches include a database load switch, a database alter switch, and a control switch.It was determined that these switches at Harris were all placed to the off (no inhibit)position, resulting in the RM-11 being able to download incorrect operational and database data to both trains of safety-related radiation monitors.These radiation monitors have been operating correctly and corrective actions to place the appropriate inhibit switches to the inhibit position have been implemented.
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| The FSAR has been changed to agree with the different configuration of the RM-80 switches to prohibit the RM-11 downloading bad data to safety-related monitors.Safet Summar This change will prevent non-safety related RM-ll malfunction from inadvertently downloading incorrect data or initiating incorrect control function which would compromise the function of the safety related RM-80 monitors.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 11.5.2 MEM/HO-930100/44/OS1
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| Change to Facility as Described in the FSAR Title: PCR-06682, Lead Shielding Storage-Containment Functional Summar This modification allows storage of lead shielding on the 286'levation and 221'levation of Containment.
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| On elevation 286', 24,000 pounds of shielding is approved to be stored inside the Reactor Vessel Head Laydown Area during normal plant operation.
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| On elevation 221', 40,000 pounds of shielding is approved to be stored on the slab near azimuth 0'nd azimuth 145'40,000 each location).
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| Safet Summar Review of applicable calculations confirm that the additional loading to the 286'levation and 221'levation slabs in Containment is acceptable.
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| Potential interaction of shielding with safety related components during a seismic event was considered since the shielding is not secured to the slab.Shielding height is limited to 2'-0.On the 286'levation there is a concrete ring shield 2'-0 in height and 1'-0" in thickness surrounding the lead shielding and thereby assuring that it will be contained and will not create any interaction.
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| For the 221'levation, both interaction of shielding with safety related components'and the possibility of the shielding blocking recirculation sump screens were evaluated.
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| There is no safety related equipment in these two areas which the shielding could impact.The azimuth 0'rea is 15-20 feet from the sumps.The sump screens have an 18" high baffle wall protecting them from heavy debris.For the lead shielding to be moved 15-20 feet and lifted 18" during the recirculation phase of accident operations is not considered credible.Allowing storage of lead shielding in the aforementioned areas will not affect the safety, operability or reliability of the containment structure or any containment building system.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 1.2.2-07 and 1.2.2-03 MEM/HO-930100/45/OS1 Change to Facility as Described in the FSAR Title: PCR-06238, Generator Hydrogen Dryer Functional Summar This modification replaces the Lectrodryer Hydrogen Dryer Unit with a.new Hydrogen Dryer built by Pnenmatic Products Corporation.
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| The Model 25HCP is a dual tower dryer and it offers dehumidification of the generator hydrogen as well as oil and particulate removal.The operating principle i's the same;however, the desiccant regeneration is different.
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| The new dryer is located west and north of the old unit on the 261'levation of the Turbine Building.Safet Summer The Generator Hydrogen Gas System is a non-safety related system.It is not an accident mitigating system nor is it an initiating system.The dryer is used to maintain low moisture in the generator.
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| This avoids failure of the generator retaining rings due to a damp environment since the present ring material is sensitive to stress corrosion cracking.The dryer is for improved performance of the generator.
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| No safety related equipment is located around the hydrogen dryer.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction'han already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 1.2.2-64, 1.2.2-76, 9.2.1-02 and 10.2.2-06 Table 9.5.1-2 MEM/HO-930100/46/Osl Change to Facility as Described in the FSAR Title: PCR-06607, Turbine Generator Electrohydralic Control (EHC)System Accumulator Filters Functional Summar This modification installs cartridge type filters on the Turbine Generator EHC System accumulator surge lines.0-ring deterioration and particulate fouling of the Turbine Generator Governor Valves has been detected and is believed to have originated in the EHC System accumulators.
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| Safet Summa i The cartridge type filters installed on the accumulator surge lines will capture any future particulate fouling of the Turbine Generator Governor Valve servo assemblies and prevent valve failure.Components are sized for minimum pressure drop to limit the affect of installation on system fluid volume requirements.
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| Material compatibility with system conditions has been addressed for proper application of components.
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| System operability is enhanced by decreasing the probability of valve failure by fouling.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 10.2.2-07 and 10.2.2-08 MEN/HO-930100/47/OS1
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| Change to Facility as Described in the FSAR Title: PCR-06111, Condenser Nitrogen Sparging Functional Summar This modification introduces a nitrogen sparging layout which injects nitrogen gas into the condenser to reduce the content of dissolved oxygen in the condenser condensate.
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| At times dissolved oxygen has been measured significantly higher than the 2 ppb that is desired as a targeted level of performance.
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| The vendor nitrogen supply storage tank located south of the turbine building is used for the nitrogen sparging supply.The storage tank has sufficient capacity to supply existing plant uses as well as the new sparging layout.A tieing to the storage tank header supplies the sparging lines which runs on the east and west sides of the condenser.
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| The sparging header branches to supply nitrogen to eight flow meters which allow 0-3.9 SCFM each into the condenser.
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| Safet Summa The nitrogen system and main condenser system are affected by this modification.
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| The nitrogen supply system converts liquid nitrogen to nitrogen gas for uses such as pressurizing various plant accumulators, to purge tanks, and to support other plant components and systems.The nitrogen storage tank located south of the turbine building has sufficient volume and capacity for existing plant use and the nitrogen sparging layout.The main condenser system accepts exhaust steam from the turbines and condenses the steam by heat exchange with cooling water from the Circulating Water System.The condenser is located directly below the two low pressure turbines.Operability of the condenser vacuum pumps have not been affected by the added injection of the nitrogen gas.The nitrogen supply system and main condenser system are neither an initiating or mitigating system.All components and systems affected by the nitrogen sparging layout are non-safety
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| -Quality Class E.Neither the condenser or nitrogen supply system are required for safe shut down of the plant.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figures 9.3.1-2, 10.1.0-4, 10.2.2-5 and 10.2.2-6 MEM/HO-930100/48/OS1 Change to Facility as Described in the FSAR PCR-06442, Service Water Isolation Valve For Turbine Digital Electrohydralic (DEH)Unit Coolers Functional Summa This modification adds a lh" manual globe valve to the Normal Service Water (NSW)System to allow isolation of service water from the Turbine DEH Unit Coolers.The valve is normally open and is closed only for maintenance purposes.Safet Summar NSW is neither an accident initiating nor mitigating system.The valve and line affected are non-safety and non-seismic.
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| This NSW line provides the heat sink for the Turbine DEH Unit Coolers.The Turbine DEH is also non-safety, non-seismic.
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| This valve model added is used throughout the plant, including NSW, and is compatible with the system.It is a manual valve, so no control functions are impacted.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.2.1-02 MEM/HO-930100/49/OS 1
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| Change to Facility as Described in the FSAR PCR-05446, Turbine Vapor Extractors Drains Functional Summar This modification to the Turbine Lube Oil Vapor Extraction System and the Main Generator Bearing Oil Ventilation System adds a loop seal from their drains and reroutes them along with the hydrogen loop seal drains to newly installed holding tanks.Safet Summar The Lube Oil System and the Generator Bearing Oil Ventilation System (Hydrogen Loop Seal)serve no safety related purpose nor do they support any initiating or mitigating system required for safe shutdown.The rerouting of the drain lines and the addition of holding tanks for the drain oil will not change the performance of either system.This modification will allow oily water from the sources above to be collected in a holding tank instead of being deposited in the Industrial Waste System and possibly directed to the Secondary Waste Treatment System.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists FSAR
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| ==Reference:==
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| Figure 10.2.2-9 MEM/HO-930100/50/OS1 Change to Facility as Described in the FSAR Title: PCR-06584, Emergency Diesel Generator (EDG)Starting Air/Control Air Filters Functional Summa This modification installed instrument style air filters on the EDG engine control/starting air lines to protect engine components from particulate ingress.This action was initiated in response to the EDG system engineer finding particulate matter in the oil governor booster shuttle valve.This debris was believed to have contributed to the diesel not reaching the desired rated speed in the Tech Spec time requirement of 10 seconds.This finding was discussed with the diesel manufacturer's representatives.
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| It was acknowledged that any type of particulate matter in the Diesel Starting Air System could possibly hamper pneumatic components from operating at optimum performance.
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| During review of this matter, it was noted that the diesel right and left bank starting air headers did not have the blowdown/drain orifice that the vendor shows on the engine drawings.This matter was reviewed, and was concluded that the headers should have this orifice to allow combustible gases to escape from the header during diesel operation.
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| This orifice would allow particulates to also be blown out the header upon a diesel starting air admission.
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| The addition of the blowdown orifice was also accomplished.
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| Safet Summar The Diesel Starting Air System is designed to supply sufficient compressed air by two physically separate air start systems.Each system has a air receiver that is capable of providing five cold engine start attempts.Each system has a air dryer that has a pre-filter and after filter to prevent contamination of the desiccant and ensure that the starting air is free of debris.The design comprises a degree of redundancy, since each header incorporates a filter for the overspeed trip components, combustion air shut-off cylinder and the governor oil booster cylinder.Each header is capable of starting the engine and providing control air requirements to operate the diesel.Failure of one of the air filters would not impact the ability of the diesel to start, reach rated speed, or shutdown.The addition of the 1/4" drain/blowdown orifice in the right and left bank starting air header on each diesel is recommended by the diesel manufacturer.
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| The existing header did not have the header nub attachment as detailed in the engine manual.The vender acknowledged that the orifice was acceptable and did not impact the operation of the diesel or seismic qualification.
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| CP&L evaluated this application and concurred with the design.MEM/HO-930100/51/OS1
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| Change to Facility as Described in the FSAR PCR-06584, Emergency Diesel Generator Starting Air/Control Air Filters Safet Summar: (continued)
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| CP6L and the manufacturer (Enterprise Engines)has reviewed the design details of the modification and agrees that the modification would not impact the operability of the diesel.The review concluded that the diesel's starting, operability, shutdown capability or emergency start ability is not adversely impacted by the addition of the filters or blowdown orifice.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.5.6-1 MEM/HO-930100/52/OS1 Change to Facility as Described in the FSAR Title: PCR-06211, Auxiliary Feedwater (AFW)Recirculation Isolation Valve Auto Closure Deletion Functional Summar This modification removes the automatic closure capability of the motor driven AFW pumps recirculation isolation valves (1AF-5 and 1AF-24).The automatic closure feature shut the recirculation isolation valve when a motor driven AFW pump was running and the other train's safety bus was deenergized.
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| Safet Summar The automatic closure is not required by the Technical Specifications and is no longer required by the FSAR commitments due to a reanalysis of the Loss of Normal Feedwater (LONF)with offsite power available.
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| The automatic closure feature was added to ensure that an AFW flow rate of 475 gpm for the LONF event could be achieved as'pecified in the Westinghouse analysis for FSAR Chapter 15 (reference Westinghouse Letter FCQL-193 dated 7/15/83).In 1987, Westinghouse reanalyzed the LONF event (along with several other events)and concluded that 430 gpm was a satisfactory flow rate (reference Westinghouse Letter CQL-87-518 dated 9/22/87).This new analysis has been incorporated into FSAR Chapter 15, and has deleted the requirement for a flow rate of 475 gpm.Actual tests on the pumps and the mechanical design calculation AF-13 show that 430 gpm to two intact steam generators can be met from one motor driven AFW pump with the recirculation line left unisolated.
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| Therefore, it is permissible to remove the automatic closure of the recirculation isolation valves.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 10.4.9 Figures 7.3.1-9A and 7.3.1-9B MEM/HO-930100/53/OS1 Cl Change to Facility as Described in the FSAR Title: PCR-05309, Turbine Vacuum Trip Relay Functional Summar This modification installs a reliable source of continuous power for the.low condenser vacuum setpoint transfer relay Ll/1355, reterminates cables 1355C&D in the Turbine Emergency Trip Cabinet to provide correct annunciation and updates the FSAR to delete reference to the low condenser vacuum trip setpoint (6 inhg absolute).
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| Relay Ll/1355, We'stinghouse Model AR880SR 120VAC relay was fed from a non-safety power panel.Any AC System transient associated with this circuit during plant operation could result in a Turbine Trip thus creating a challenge to the Reactor Protection Systems.This change to the facility replaces the 120VAC relay and power supply with a 125 VDC Westinghouse relay ARD880SR and installs a fused coordinated 125 VDC reliable and continuous power source.By changing Ll/1355 power from AC to a continuous (battery backup)DC power source the possibility of a Turbine Trip and subsequent challenges to the Reactor Protection Systems have been reduced.PCR-5309 also provided for the termination of cables 1355C and D in the Turbine Emergency Trip Cabinet which will allow indication per the original design intent.Additionally, FSAR Section 10.4.1.3 was revised to delete reference to the low condenser vacuum trip setpoint.Safet Summar The main condenser is not required for the safe shutdown of the reactor or to perform in the operation of reactor safety features.The relay power enhancement will reduce the probability of loss of condenser vacuum control which would introduce a Turbine Trip and subsequently a Reactor Trip.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 10.4.1.3MEM/HO-930100/54/OS 1
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| Change to Facility as Described in the FSAR Title: PCR-04034, Reactor Coolant System (RCS)Shim Gap Functional Summar This engineering evaluation addresses gaps between the reactor coolant loop (RCC)crossover leg piping and pipe support bumpers.Field measurements show that at 450'F, contact exist between the crossover leg piping and the pipe support bumpers for all three loops.Original plant design by Westinghouse called for unrestrained thermal growth of the RCS.The engineering evaluation contained PCR-0434 documents acceptability of this condition for plant operation.
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| No changes to the existing plant configuration are provided by this PCR.Safet Summa The reactor coolant loop crossover legs carry reactor coolant fluid from the steam generators to the reactor coolant pumps.Field measurement taken on 10/10/88 at 450'F identified contact between the crossover leg piping and the pipe support bumpers.The design philosophy for the reactor coolant piping is to provide for free thermal expansion of the system.The piping/bumper interference is inconsistent with the original RCL design, and has a definite impact on the qualification of the RCL system.Since the reactor coolant system is Westinghouse scope, Westinghouse Electric Corporation was contacted to provide an evaluation of the condition.
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| The condition was evaluated by Westinghouse, who performed a set of RCL thermal analyses which included appropriate bumper restraints.
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| The loadings from these thermal analyses were factored into the RCL piping fatigue evaluation, the primary equipment nozzle load evaluation, the primary equipment support load evaluation, and the primary loop piping leak-before-break evaluation.
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| Westinghouse concluded that the condition of the primary loop system with the known interference in the cross over leg piping restraints is acceptable for the life of the plant from an analysis standpoint.
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| Results of the Westinghouse thermal analyses provided new thermal loadings of 1030 kips on the pipe rupture bumpers.The evaluation of the bumpers for the revised loadings was completed by CP&L.The results of the evaluation show that the bumper structural steel and concrete are acceptable for the revised thermal loadings.The analyses by Westinghouse and CP&L demonstrate that with the bumper interference condition, the RCS piping, equipment, and component supports remain capable of performing their safety-related functions within the original plant design criteria.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 5.4.14 MEN/HO-930100/55/OS1
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| Change to Facility as Described in the FSAR Title: PCR-05833, Emergency Sequencer Panel Containment Spray Pump Indication Functional Summar The Emergency Sequencer System (ESS)is composed of two separate systems, one for each safety train.The'rimary function of the sequencer is to activate.the large ESF loads in response to emergency system features activation signals.The sequencer will start the emergency safety equipment during loss of coolant (LOCA)or loss of off-site power (LOOP)or both, so as to minimize the effects of excessive voltage drops on the safety buses due to simultaneous large motor starting.The sequencer is designed for testing during operation to verify that the ESS panels 1A-SA and 1B-SB operate per design.This testing is accomplished through logic, which generates simulated LOCA and/or LOOP signals and processes them into the appropriate logic program.The test is performed by EPT-033 (not a tech spec requirement but a NRC commitment per SDD-155.02).
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| During the EPT-033 test it was found that containment spray monitoring lights L17 and Load Block-2 BC energized without containment spray SOS/PB pressed.These lights should energize only when SOS/PB is pressed.In addition, it was found that containment spray circuit was only tested manually by SOS/PB, while having the capability to test Load Block-2 BC automatically.
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| This modification will reconfigure the spray circuit L17 and BL-2 light, so that they will energize only when SOS/PB is pressed or automatically in Load Block-2 BC by increasing the TTIX relay setpoint from 0.6 sec to 10 sec.Safet Summa The FSAR discusses the sequencer and the timing and philosophy of the sequencing programs.This modification does not change the logic associated with the sequencer.
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| The technical specifications require that a test be performed to ensure that large loads are activated in a manner which limits the demands placed on the emergency buses and diesel generator.
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| This modification does not change the Load Block timing nor the operating philosophy of the sequencer testing.This modification does not require a change to the margin of safety as detailed in the technical specifications nor the operating license.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 7.3.1.51MEM/HO-930100/56/OS1 Change to Facility as Described in the FSAR PCR-05534, Essential Services Chilled Water System (ESCWS)Expansion Tank Reliability Improvements Functional Summar This modification replaces the ESCWS fire protection water makeup source which is required during normal operation with demineralized water.The demineralized water will prevent the introduction of sediment into the ESCWS which will improve system operation.
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| Additionally, the setpoints of two bistables, which provide alarms on hi-hi expansion tank level, are being revised.Safet Summar 0 Replacement of fire protection water with demineralized water for the makeup source to ESCWS will improve the normal operation of this safety-related system by minimizing ESCWS corrosion and deposition.
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| This modification does not affect the ESCWS operation during an accident condition because the Demineralized Water System is assumed to fail as was the Fire Protection Water System at which time makeup water to the ESCWS is provided by the Emergency Service Water System.The emergency service water makeup to ESCWS function during an accident condition is not affected by this modification.
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| Setpoint changes to expansion tank level instrumentation are necessary for chilled water volumetric expansion considerations.
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| These setpoints are associated with expansion tank level annunciation which is deemed nonsafety, changes to level alarm bistables does not invalidate the design basis of the ESCWS.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Sections 7.3.1.5.4 and 9.2.8 Figures 7.3.1-16, 9.2.3-1, 9.2.8-3, and 9.5.1-2MEM/HO-930100/57/OS1 Change to Facility as Described in the FSAR Title: PCR-06346, Leak Repair of RTD Bypass Manifold Isolation Valves 1RC-2 and 1RC-53 Functional Summar This modification removed the handwheels on the A Loop RTD Bypass Manifold Isolation Valves 1RC-2 and 1RC-53 and added a seal cap over the valve stems to eliminate the potential for a leak from the stem area.This leaves the valves inoperable in the full open position.Safet Summar The seal caps installed by this modification were supplied by Kerotest to the same specification and quality standards as the original valve components.
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| The valve body and bonnet will remain as the primary pressure retaining parts with the seal cap as a backup providing full pressure seal capability.
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| The seismic/stress analysis effects due to the weight of the caps offset by the elimination of the handwheels were evaluated and found to be acceptable.
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| The system is thus kept in the"as analyzed" condition.
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| These valves are not required for plant operation or shutdown.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than air'eady evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 5.1.2-1 and 5.1.2-3 MEM/HO-930100/58/OS1 Change to Facility as Described in the FSAR Title: PCR-06547, Alternate Miniflow Redesign Functional Summar This modification replaces the safety injection (SI)alternate miniflow relief valves (1CS-744 and 1CS-755)with orifices and enhanced open/close logic on the motor operated alternate miniflow isolation valves (1CS-746 and 1-CS-752).
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| The isolation valves will receive an open signal on high RCS pressure to ensure minimum flow for protection of the charging/SI pumps (CSIP's)via the new orifices.They will also receive a close signal on low RCS pressure to ensure adequate SI flow is delivered to the core.The purpose of the alternate miniflow sub-system is to provide an alternate flow path for protection of the CSIP's for those postulated accidents in which RCS pressure can increase above CSIP shutoff head following SI actuation.
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| Safet Summar The motor operated alternate miniflow isolation valves will close only during those accidents where SI is actuated and RCS pressure is expected to increase above the motor operated valve (MOV)opening setpoint.In these accidents however (inadvertent SI and feedline break)SI flow provides little benefit.Consequently, MOV closure is not critical to ensure satisfactory performance of the SI system.Automatic closure does, however, increase plant safety by providing an additional backup.To ensure their integrity and long-term availability for accident mitigation, each operating CSIP must pass at least 60 gpm.The new orifices are sized to ensure at least 60 gpm will be passed by each pump in the condition of maximum degradation that will satisfy the ECCS analyses assumptions.
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| The combination of MOV opening setpoint and MOV stroke time provide adequate pump protection for the highest expected rate of RCS pressure increase.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 6.3.2 Table 6.3.2-3 and 7.3.1-5 MEM/HO-930100/59/Osl Change to Facility as Described in the FSAR PCR-06091, Steam Generator Chemical Addition System Functional Summar This modification installs a Boric Acid Chemical Addition System for.the Secondary plant.The system is installed in the Turbine Building elev.261'.The new system which includes pumps and a tank will allow a 50 ppm boron steam generator soak at start-up and 5-10 ppm boron to be maintained during normal power operation.
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| t Safet Summar This modification provides for the addition of boric acid to the secondary water in order to minimize the denting and intergranular stress corrosion cracking that may occur in steam generators.
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| This is a nonsafety related modification.
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| The system is designed to meet the pressure ratings of the feed water piping it connects to.A failure of the system will not adversely impact the operation of the plant.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 10.3.5 MEM/HO-930100/60/OS1 Change to Facility as Described in the FSAR Title: PCR-6334, Revised Dose Calculation For Refueling Water Storage Tank (RWST)Level Transmitters.
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| Functional Summar I This modification removes the RWST level transmitters LT-990, LT-991, LT'-992, and LT-993 from the Environmental Qualification (EQ)Program.Safet Summer The subject level transmitters were removed from the EQ program based on a recalculation of the expected radiation dose and temperature in the area in which the devices are installed.
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| The determining EQ factor for temperature is a rise of 18'F above the normal ambient temperature in the event of a Design Basis Accident.FSAR Figure 3.'11B-15 shows that for Zone P, the RWST area, the maximum temperature is 122'F for both the normal and the accident conditions.
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| Therefore, based on the EQ criteria, temperature alone does not require designating the area as EQ harsh.The RWST area radiation dose was re-examined to determine if the dose was excessively conservative, general to the entire tank area, or could otherwise be reduced.The results indicate that the radiation dose adjacent to the RWST is significantly lower than the values given on FSAR Figure 3.11B-29.The worst case radiation level determined by the review assigns a 40 year normal radiation dose of 3.1 E+2.When added to the existing accident dose of 2.0 E+2 a TID of 5.1 E+2 is obtained.This is below the EQ threshold of 1.0 E+4 required to designate an area as EQ harsh and therefore establishes the basis for removal of the RWST level transmitters from the EQ program.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 3.11B-29 MEM/HO-930100/61/OS1
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| Change to Facility as Described in the FSAR 0 Title: PCR-06551, Emergency Diesel Generator (EDG)1A-SA Cylinder Block Cavity Drains Functional Summer This modification remov'es EDG cylinder block cavity drain valves 1DJ0.-21, 1DJ0-22, 1DJ0-23, and 1DJO-24 and associated piping and installs 1 1/4" hex head steel plugs in their place.This action was taken due to leakage that had occurred at these locations.
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| Safet Summar During construction, it was anticipated that the cylinder block cavity would require drain lines and valves.A field change was implemented to replace plugged connections with drain lines and valves.On at least two occasions leakage has occurred at threaded pipe joints.Loosening of the joints occurred on valve operation.
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| Retightening of the connections did not provide a tight seal.This modification returns the equipment to its original design.Materials used in the design are compatible with existing equipment.
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| The plugs that replace the valves are considered passive components that are no more likely to fail than the piping it replaces.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Figure 9.5.5-01MEM/HO-930100/62/Osl Change to Facility as Described in the FSARTitle: Meteorological Program Change Functional Summar The Harris, Meteorological Program has been upgraded by removal of the Westinghouse sensor system and Esterline Angus strip chart recorders.and replacement with an Yokogawa Hybrid Recorder.A change has been made to the historical data base.input and primary data acquisition system from the Westinghouse sensor system to the ADAC sensor system.The Westinghouse sensor system and the Esterline Angus strip chart recorders represent twenty-year old technology in which manufacturer support had been discontinued.
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| In planning for the phaseout of these two systems, the ADAC sensor system was installed and became operational in 1987.The ADAC se'nsor system is now the primary data acquisition system and serves as the input to form'the historical data base.The new hybrid recorder will serve as the backup sensor system to the ADAC sensor system.The hybrid recorder will produce hard copy analog trend traces and digital fifteen minute averaged data in hard copy format and stored on a memory card for both differential temperatures, upper and lower level wind directions and wind speeds, ambient temperature, and dew point temperature.
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| An RS232 communication port on the hybrid recorder permits us to poll and retrieve the fifteen minute averaged data via a telephone line modem connection to a host computer in the CP&L General Office.Safet Summar The purpose of the Meteorological Program is to measure and display meteorological data during normal and accident conditions.
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| The system has no effect on the design, operation, and maintenance of systems important to safety.The Meteorological Program is used to aid in determining the appropriate protective action recommendations to be made for the public.Variables measured include those required to comply with R.G.1.97 Rev.2.In addition, the requirements of R.G.1.23 Rev.0 are also met.The meteorological instrumentation bases are described in Section 3/4.3.3.4 of the Technical Specifications.
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| The bases as stated are: "...the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere".
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| No changes have been made to the Meteorological Program that would decrease data recovery rates or data availability.
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| If anything, the changes to the program should increase data availabilities and recovery rates because of the increased serviceability of the newer technology used in the hybrid recorder.Also, additional analog trend traces of both differential temperatures, ambient temperature, and dew point temperature are now available with the hybrid recorder.The Esterline Angus strip chart recorders provided trend traces for the wind direction and speed parameters only.Additional remote access of digital fifteen-minute averaged data is provided by the hybrid recorder.The Westinghouse sensor system only provided for remote access of the latest fi.fteen-minute averaged data.The hybrid recorder can provide remote access for up to thirty-seven days of fifteen-minute averaged data history.Since all of the required parameters continue MEM/HO-930100/63/OS1
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| Change to Facility as Described in the FSAR Safet Summar: (continued) to be measured and the added hybrid recorder increases the access to meteorological data used in estimating potential doses to the public, the safety margin is increased, if affected at all.The ADAC sensor system replaced the Westinghouse sensor system as the input to the historical meteorological data based in September of 1992.An eighteen-month study that compared Westinghouse sensor system values with ADAC sensor system values was completed before the data base change was made.No significant differences between the values were discovered.
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| Data recovery rates are also comparable with both systems well above the R.G.1.23 required 90%.Since there are no significant differences in the values produced by the two systems, and the data recovery rates are comparable and meet R.G.1.23 guidance, there is no change in the safety margin associated with this change.The ADAC sensor system interfaces with the control room through the ERFIS via telephone lines.The hybrid recorder will not directly interface with the ERFIS.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 2.3.3 MEM/HO-930100/64/OS1 Change to Procedure as Described in the FSARTitle: EOP-EPP-011, Rev.5/1, Transfer Between Cold Leg and Hot Leg Recirculation Functional Summar In January, 1992 Westinghouse notified the NRC of concerns regarding.
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| the methodology used in evaluating the time for Hot Leg Switchover (Westinghouse Letter ET-NRC-091-3659 and Interim Report 91-046).The concerns resulted from increased RWST boron concentrations used in many Westinghouse plants to accommodate higher fuel enrichments and longer fuels cycles.The increased RWST boron concentrations significantly reduced the time for switchover from cold le to hot leg recirculation.
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| eg In June, 1992 Westinghouse notified the NRC of an additional concern related to the hot leg recirculation alignment-failure of the low head Safety Inje t SI njec ion ()hot leg injection valve to open (Westinghouse Letter ET-NRC-092-3712 and Interim Report 91-046, Rev.1).For plants with low head SI hot leg injection (including SHNPP), this failure reduced the amount of SI flow delivered to the core via the RCS hot legs during recirculation.
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| This failure had not been previously recognized.
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| The reduction in SI flow resulted in two potential problems;1)further reduction in the time for Hot Leg Switchover, and 2)reduction in heat removal from the core.Evaluations were performed by Westinghouse to determine the impact of the reduced injection flow on boron precipitation and core cooling.The results of this evaluation, included in an attachment to Westinghouse Letter ET-NRC-092-3782 were;1)a minimum SI flow to the hot legs of 1.3 times the core boil-off rate is required to prevent boron precipitation.
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| The flow must be supplied f th rom e ig ea I pumps since the low head SI hot leg injection valve is assumed to be shut.The boil-off rate is calculated at the time of entry into hot leg recirculation.
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| Credit can be taken for all high head SI flow reaching the core since the LOCA of concern in the boron precipitation analysis is a cold leg break, and 2)a minimum SI flow to the cold legs of 1.5 times the core boil-off rate is required to maintain adequate core cooling for a hot leg break.(Boron precipitation is not a concern for a hot leg break.)The SI flow could come from either the low head or high head SI pumps.Since the break is in the hot legs, credit can be taken for all SI flow reaching the core.For SHNPP, one Charging Safety Injection Pump (CSIP)can deliver the required igh head SI flow to either the hot or cold legs.However, since credit can not be taken for identification of the break location, both hot leg and cold leg injection flows must be delivered simultaneously.
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| To satisfy the cold le injection requirements for the assumed failure, Westinghouse recommends that low head SI be re-established.
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| Westinghouse recommends that EOPs contain instructions to align the low head SI for cold leg injection if the hot injection valve can not be opened.Westinghouse also emphasizes that high head SI hot leg injection be established without delay.The contingency instructions added to EOP-EPP-ll 1, Steps 3a and 3b incorporate these recommendations.
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| The contingency for Step 3a ensures the operator proceeds to Step 4 to establish high head hot leg injection to preclude, boron precipitation.
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| The contingency instructions for Step 3b specify that one MEM/HO-930100/65/OS1 Change to Procedure as Described in the FSAR Functional Summar: (continued) of the two low head SI cold leg injection valves be opened.(Only one is opened to ensure a single operation RHR pump can not experience runout).To ensure minimal delays re-establishing low head SI cold leg injection and establishing high head SI hot leg injection, local operation of 1SI-359 is not included as a contingency.
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| The contingency instructions for Step 3b also require the plant operations staff be consulted for recommendations concerning subsequent transfers between hot and cold leg recirculation.
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| The procedural steps as written should be adequate, assuming no additional equipment failures.The instructions ensure the plant operations continues to evaluate plant conditions and is prepared to recommend additional actions if required to maintain core cooling and minimize the potential for boron precipitation.
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| Evaluations performed as part of PCR-5371 demonstrated that for this failure, boron precipitation would not occur if high head SI hot leg injection was initiated within 6.5 hours following a LOCA.Safet Summar The contingency to open one of the low head SI to cold leg injection valves ensures adequate SI flow is delivered to the RCS and is only implemented following failure of the low head SI hot leg injection.
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| The contingency was added in response to a Westinghouse 10CFR21 report (Westinghouse Letter ET-NRC-092-3782) concerning the failure of the low head SI hot leg injection valve to open when aligning for hot leg recirculation.
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| Further discussion on the safety aspects of this changes is described above in the Functional Summary.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 6.1.2 Table 6.3.1-1 MEM/HO-930100/66/OS1 Change to Procedure as Described in the FSARTitle: OP-120.07, Rev.7, Waste Gas Processing Functional Summar This change allows two Waste Gas Decay Tanks (WGDTs)to be cross-connected for better management of the Waste Gas System.This cross-connection allows pressure to be equalized so that the low pressure mode of operation can be bypassed without initially pressurizing the WGDT with nitrogen.Safet Summar No physical changes are being made to the Waste Gas System.With the two WGDTs interconnected the potential accident analyzed in FSAR Section 15.7.1 remains the same, i.e.tank rupture, but the accident now would involve the release of both tanks contents.Even with two WGDTs interconnected the margin of safety is not reduced in that the maximum curie content of WGDTs is limited to 105,000 curies in Technical Specification 3/4.11.6 and the maximum curie content of the interconnected tanks is limited to 59,000 curies by the FSAR Section 15.7.1 accident analysis.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Sections 11.3.1, 11.3.2 and 15.7.1.MEM/HO-930100/67/OS1 Change to Procedure as Described in the FSAR PLP-626 Rev.1, Power Ascension Testing Program After a Refueling Outage Functional Summar The program for power ascension testing has remained basically the same as.the initial plant startup program.The intent of this change to PLP-626 is to more align the Harris test program with current vendor and industry adopted standards.
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| The major changes are the elimination of the 50%flux map and the allowance for a greater than 90%power level flux map as the"high power flux map".The revised power ascension program was implemented for Cycle 5.Safet Summar The elimination of the 50%flux map does not reduce the margin to safety.Adequate margin exists at this power level to ensure safe operation.
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| Fq(z)and Fah are measured at 30%power to ensure Technical Specification compliance prior to exceeding 30%.The previous need for the 50%flux map was to provide new full power total current'if QPTR (OST-1039) at less than 50%failed.If OST-1039 fails at 50%power, the flux map would still be performed and a new NIS full power total current developed.
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| The change for a greater than 90%power level map" is a fall out of previous cycle flux maps map typically allow for power operation up appropriate power level, 100%allowable power the high power flux map.flux map as the"high power flux at 75%.The results of this flux to 97%thru 99%.Once at the operation is always confirmed by This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 4.3.2.2.8 MEM/HO-930100/68/OS1 Change to Procedure as Described in the FSAR Title: AOP-017, Revision 6, Loss of Instrument Air Functional Summar AOP-017 was revised to allow control of Residual Heat Removal (RHR)cooldown rate by either throttling RHR to Reactor Coolant System (RCS)loop discharge valves or cycling the pump on and off.The previous method of controlling RCS cooldown (on-off operation of the RHR pump)was specified in FSAR Section 7.4.2.2.1.
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| This method could easily result in exceeding the starting duty cycles of the RHR pump.It also places more of a burden on the operators since one must be present at the pump (in a contaminated area with elevated radiation level)prior to each start.An equally effective method was added.This method is to locally throttle or open-shut the RHR to RCS discharge valves (1SI-340 and 1SI-341).When the valve(s)is shut (or throttled), the pump will run on r'ecirc flow, with the cooling effect of RHR isolated (or minimized) from the RCS.This method is used (and therefore has been reviewed)in AOP-004"Remote Shutdown".
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| Safet Summar e This method of controlling RCS cooldown is more effective, better for equipment, and easier for the operator to control.Failure of the discharge valve (to open or close)has been analyzed in the FSAR.If the valve were to fail in the open or throttled position, the existing method of cycling the RHR Pump could be used.If the valve failed in the closed position, the other train of RHR would be used.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Section 7.4.2.2.1MEM/HO-930100/69/OS1 Cl Change to Procedure as Described in the FSAR Title: OP-108 Revision 4, Boron Thermal Regeneration System (BTRS)Functional Summar As stated in FSAR Section 9.3.4.1.2.4, a thermal regeneration demineralizer.
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| can be used as a deborating demineralizer, which would be used to dilute the RCS down to very low boron concentrations towards the end of a core cycle.This revision to OP-108 will allow dilution operation without the use of the BTRS Chillers.Safet Summar Allowing dilution without using the chillers is safe due to the fact that it is a more conservative way to operate the system than was originally designed.The'bility of a deminineralizer to remove boron is directly proportional to the boron concentration and inversely proportional to the water temperature.
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| The new section will not use the chillers to cool the water to 50'so the temperature of the fluid entering the demineralizer will be higher.Therefore the dilution rate will be slower.A slower dilution rate implies a slower reactivity rate which will give the operators more time to respond should the system fail to the dilution mode.This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction than already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR Reference Section 9.3.4.12.4 MEM/HO-930100/70/OS1 Change to Procedure as Described in the FSAR Title: CRC-001 Rev.8, SHNPP Environmental and Chemistry Sampling and Analysis Program Functional Summar Chemical control of the Waste Processing Building Cooling Water and Component Cooling Water Systems are maintained in accordance with CRC-001.Corrosion coupon monitors are installed on each system to determine the effectiveness of the corrosion control program.These results serve as the basis for making any needed changes in the corrosion control program and may indicate a need to vary the inhibitor concentrations from time to time.Qualitative analyses have been performed to determine the effects of various chemical concentrations on the materials contained within these systems.Based upon the findings, it was determined to be more prudent to control the program based upon actual results from the installed corrosion coupons than to set limits on the individual inhibitors being used.Therefore, the FSAR is being revised to recognize this change and the chemistry specifications contained in Tables 9.2.2-2 and 9.2.10-2 are being removed.Safet Summar This change will not relax requirements to monitor ph, chloride, and fluoride but will use actual analysis results to monitor the effectiveness of the corrosion inhibitors.
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| This change is consistent with requirements of the Standard Review Plan.The enhanced method of monitoring corrosion inhibitor effectiveness allows earlier detection and corrective actions to assure adequate corrosion protection and structural integrity of the system is maintained.
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| This change does not increase the probability or consequences of analyzed accidents, nor introduce a different type of accident or equipment malfunction that already evaluated in the FSAR.Thus, no unreviewed safety question exists.FSAR
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| ==Reference:==
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| Sections 9.2.2.2, 9.2.10.2, Tables 9.2.2-2 and 9.2.10-2 MEM/HO-930100/71/OS1 P~}}
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