ML20202F377
| ML20202F377 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/31/1998 |
| From: | Zinke G Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GAZ-98-08, GAZ-98-8, MN-98-10, NUDOCS 9802190188 | |
| Download: ML20202F377 (15) | |
Text
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.e MaineVankee P.O. BOX 408
- WISCASSET, MAINE 04578 * (207) 882-6321 Febmary 12,1998 MN-98-10 GAZ-98-08
- UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk
- Washington, D.C. 20555
References:
- 1) License No. DPR 36 (Docket No. 50-309)
(b) MYAPCo Letter to USNRC dated March 11,1981 (FMY-81-33)
Subject:
Annual Report of Facility Changes and Relief and Safety Valve Failures and Challenges Gentlemen:
In accordance with to CFR 50.59(b), the Attachment to this letter contains brief descriptions of the facility changes completed at the Maine Yankee Atomic Power Station from January 1,1996 to January 31,1998 and a summa y of the safety evaluations for each change, in Reference (b), Maine Yankee committed to reporting any challenges and, or failures of PORV and pressurizer safety valves. During 1996 and 1997 there were no such events.
Verv truly yours, (M
George A. Zinke, hbaker Regulatory Affairs Department Attachments c: Mr. Hubert J. Miller Mr. Ron Bellamy M , Richard J. Rasmussen _'
I Mr. Michael K. Webb
/M {-
Mr. P. J. Dostic
- h.r. Michael T. Masnik l Mr. Clough Toppan Mr. Uldis Vanags 9802190188 980131'#
PD81 ADOCK 05000309 E. I.t ..
1.IllllIl.l 1.1I.ll L
4 Attachment A dn August 7,1997, hiaine Yankee certified per 10CFR50.82 that the company had nermanently ceased power operation and that all fuel was in the Spent Fuel Pool (Reference h1N-97-89). This is a permanent, non-revocable certification that changed hiaine Yankee's licensing basis by no longer allowing fuel in the reactor vessel and no longer allowing operation of the reactor.
The most significant effect of this license basis change was to eliminate nuclear safety functions for the majority ofstructures, systems and components (SSCs). Thos 3SCs which had enh performed a reactor safety function (i.e., SSCs which do not support a spent . ! or radiation protection safety function) need no longer be maintained under nuclear grade contro ..
Part I of this Attachment cummarizes changes made pursuant to 10CFR50.59 that are still pertinent to hiaine Yankee's current licensing basis in light of these ecrtifications. Those listed in Part 11 were changes made pursuant to 10CFR50.59 that are no longer pertinent (i.e. the safety evaluations subsequent to the 10CFR50.82 certifications superceded those listed in Part 11 due to the immediate climination of various safety functions).
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, , Part I The following changer were made, and are still pertinent to Maine Yankee's current licensing basis in accordance with the cessation of power operation:
Yellow Tae 97-22. SFP IIcal Exchancer Alternate Cooline Temporary valves were installed on the SFP heat exchanger primary component cooling (PCC) inlet r.nd outlet, in support of the alternate SFP cooling project. Altemate SFP cooling was provided by SCC (Secondary Component Cooling) via hoses. This yellow tag and 50.59 only covered the valves on the SFP heit exchanger, not the rest of the attemate SFP cooling equipment. Altemate cooling
- was required while the normal cooling, PCC, was out of service for maintenance and testing. The fire protection system remained as the backup cooling. This alternate cooling was tied into SCC in such a way that a failure of the alternate cooling hose system would trip closed the SCC non-safeguards iso'ation valves. This assured that SCC flow to the RHR heat exchanger and other safety related components would not be interrupted. The altemate cooling system was analyzed to be capable of removing the maximum expected heat load from the SFP, while maintaining the bulk pool temperature below the assumed value in the safety analysis. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59 TE 090-97. Movement of Fuel Assemblies A053 and C206 Fuel assemblics A053 and C206 had one end of the upper tie plate removed (and, subsequently, replaced). This Technical Evaluation (T E) demonstrates that it is acceptable to move these assemblies utilizing nomial operations procedures for fuel handling in the SFP. The two assemblics ,
being evaluated were last used in the reactor core in the mid 1970s, so the postulated drop accident is bounded by the nomial fuel handling accident, which' assumes the dropped assembly was discharged 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> aller operation. This TE did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 153-97. Increase Service Water Pump Motor Runnine Current This TE evaluated the effect of higher running currents on the four service water (SW) pumps.
These SW pumps provide cooling water to the Primary and Secondary Component Cooling Water heat exchangers. With the plant now in decommissioning, the primary heat load of concern is SFP cooling. Following an overhaul ofSW pump P-29A, its running current increased above its previous value to the point of challenging the circuit breaker's long time trip setting. This TE concluded that the higher running current required t.nd the change in circuit breaker setting do not adversely effect the ability of the SW pumps to perform their function. This TE did not involve an unreviewed safety question as defined in 10 CFR 50.59.
IEJ05-97. Modification to Fuel Transfer Tnbe (FU-7)
The modification removes valve FP-21 fm - a west end of the fuel transfer tube (FU-7) and replaces it with a blind flange. This valve nolates the SFP from the fuel handling cavity in containment, and must be manually opened to allow fuel to pass through. The new blind flange will climinate possible losses of Spent Fuel Pool coolant inventory due to inadvertent operator actions.
The blind flange has double o-rings and a test connection to ensure that the o-rings are performing their intended ftmetion. The replacement blind flange has been analyzed and meets design basis requirements. This evaluation did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc 13-30-1. Core Component Trar-kr This procedure revision incorporates the limitations imposed by installation of the new SFP racks.
Additions to the procedure include the recognition of two regions of racks, initial enrichment and bumup criteria for placement of fuel in particular regions and placement of fuel to prevent radiation
, heating effects on the SFP wall. This procedure provides administrative controls to meet the conditions provided in Technical Specification Amendment 144 induding concems of photon
- heating of the SFP walls, therefore no separate safety evaluation was required. This procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc 1-17-1. Fuel Pool Makeup. Cooline & Purification This procedure specifies operator actions to makeup to the SFP, purify the fuel pool water, control the temperature and lower the water level. The procedure specifies methods, such as requiring makeup from a borated source and ensuring that water temperature is appropriate, that ensure that no adverse impact on any of the accidents analyzed in the FSAR is created. This procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
il Proc 4-1-49. Alternate Cooline to SFP IIcat Exchanger Alternate cooling during the Primary Component Cooling (PCC) outage was supplied to E-25 (SFP heat exchanger) from secondary component cooling (SCC). The Yellow Tag process (see Yellow Tag 97-22) and this procedure controlled the temporary modification. This alternate cooling was only used below 210 F and below transthermal mode. The SCC for attemate cooling is on the shell side of the SFP heat exchanger, so a hose leak will not effect the water level in the SFP, Backup cooling, via the fire protection system, was still available. This attemate cooling was tied into SCC
- in such a way that a failure of the aMemate cooling hose system would trip closed the SCC non-safeguards isolation valves. This assured that SCC flow to the RHR heat exchanger and other safety related components would not be interrupted. The alternate cooling system was' analyzed to be capable ofremoving the maximum expected heat load from the SFP, while maintaining the bulk pool temperature below the assumed value in the safety analysis. This temporary procedure did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc 4-1-71. Spent Fuel Pool IIcat-Up Rate Test This temporary procedure provided for the performance of a spent fuel pool heat up rate test in order to determine actual decay heat load stored in the SFP. The procedure also provided for an evaporation rate test to benchmark the actual evaporation rate at various SFP temperature conditions.
This test was not described in the FSAR and was evaluated to ensure that it did not have a negative effect on nuclear safety, since SFP temperature was allowed to rise above normal. This procedure required that the pool temperature remain below its maximum design temperature rating of 154 F.
Specific restrictions were also provided for pool level. This test did not have an effect on any of the analyzed accidents in the FSAR, because fuel handling was prohibited during the performance of-the test. Since the SFP cooling system and its support systems remained available for operation throughout the test, and normal makeup systems including fire water were available, the safety function of this equipment was not impacted. This procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
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Proc 4-17 23. Spent Fuel Pool Heat-Up Rate Test
, This temporary procedure provided for the perfonnance of a spent fuel pool heat up rate test in order i
to determine actual decay heat load stored in the SFP The procedure also provided for an i evaporation rate test to benchmark the actua! evaporation rate at various SFP temperature conditions.
1 This test was not described in the FSAR and was evaluated to ensure that it did not have a negative i effect on nuclear safety, since SFP temperature was allowed to rise above nonnal. This procedure required that the pool temperature remain below its maximum design temperature rating of 154 F.
- Specific res.rictions were also provided for pool level. This test did not have an effect on any of the analyzed accidents in the FSAR, because fuel handling was prohibited during the perfom1ance of i- the test. Sinci. the SFP cooling system and its support systems remained available for operation
! t! :ghout the test, and normal makeup systems including fire water were available, the safety 1 fe .mn of this equipmer.t was not impacted. This 50.59 also evaluated the NPSH requirements of i- the SFP cooling pumps at the anticipated elevated temperatures, it concluded that throttling of the 4
SRP heat exchanger d:., charge valve ensured that the pumps remained operable at the higher temperatures. _This procedure change did not involve at unreviewed safety question as defined in
!' 10 CFR 50.59 Proc 0-06-2. Administrative Controls for Procedures and Procedure Channes i
!' This nrocedure change involves a change to the procedure cancellation process for the plant in the ,
defueled and decommissioning mode. This procedure now incorporates screening criteria for
- procedure elimination that cover those elements of nuclear safety and compliance necessary for
, continued safe operation of Maine Yankee during the decommissioning phase. This screening will take the place of an individual 50,59 evaluation, if all responses are negative. If all screening a responses are negative, for a particular procedure cancellation, that cancellation will be bounded by .
l this 50.59 cvaluation. Assessment of the criteria may only be performed by knowledgeable i personnel. Use of the screening criteria ensures that procedures that could have an effect on the probability or consequences oflicense basis accidents during decommissioning are retained, and only l'
allows cancellation of procedures unrelated to those accidents. This procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59, i
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FSAR CN 97-33. Chapter VIII Final Safety Analysis Report (Rev 13)
The purpose of this change was to reclassify all electrical systems previously described in the FSAR to non-nuclear safety (NNS). This revision to Chapter Vlli of FSAR was made aller Maine Yankee certified per 10CFR50.82 that the company had permanently ceased power operation and that all fuel was in the Spent Furl Pool (Reference MN 97 89). This evaluation concluded that no electrical systems meet 10CFk50.2 definition of" Safety Related". In the current plant condition there is no requirement to maintain the integrity of the reactor coolant pressure boundary or to maintain SSCs necessary to bring the plant to cold shutdown. There is no accident scennio, in the current plant condition, which requires electrical power to mitigate the consequences of the accident or to limit the dose to the public. Consequently, discussion of these systems in the FSAR is no longer pertinent in accordance with 10CFR50.34. This FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Initial FSAR Channes Effecting the MY Defueled Condition (Rev 13)
The purpose of this change to the FSAR is to delete references to SSC's no longer performing a safety function in the defueled condition, and to add new conservative quality assurance criteria applied to some non nuclear safety (NNS) SSCs into the licensing basis, On August 7,1997, Maina Yankee (MY) certified per 10CFR50.82 that the company had permanently ceased power operation and that all irradiated fuel had been permanently removed from the reactor vessel. This is a permanent, non-revocable certification that changed Maine Yankee's licensing basis by no longer allowing fuel in the reactor vessel and no longer allowing reactor operation.
The most significant elTect of this license basis change was to eliminate nuclear safety function for the majority of Structures, Systems and Components (SSCs). Those SSCs which only performed -
a reactor safety function (i.e., SSCs which do not support a spent fuel or radiation protection safety function) need no longer be maintained under nuclear grade controls and are no longer required to be discussed in the FSAR. Additionally, a set of criteria is added into the licensing basis imposing
- new conservative quality assurance criteia applied to some non nuclear safety (NNS) SSCs. This criteria is applied to some NNS SSCs which meet importent to the Defueled Condition (ITDC) criteria. ITDC criteria is added as a result of Maine Yankee's intent to preserve a level of graded quality assurance for certain NNS SSCs which meet the ITDC criteria.
The attribw af the ITDC criteria are generally related to spent fuel storage and handling, protection of worken; . ^e public from radiological consequences, and monitoring of radiological release paths.
t The evalu tion of these two basic changet, follows:
Deletion of references to SSC's no longer perfomiing a safety function in the defueled condition:
With the cessation of any activities associated with fuel in the reactor vessel, or in transit between the containment and the spent fuel pool, the possibility of any accident associated with the ;
associated activities is precluded, therefore, the probability o consequences of these accidents could l not be increased. By dermition, the deleted SSC's are no longer perfomiing a safety function, therefore, there can be no increase in the probability or consequences due to or as the result of a failure of equipment important to safety. This evaluation did not authorize the physical removal of any SSCs, therefore no possibility of any new accidents / malfunctions is cr ated. (The physical removal of SSCs will be evaluated separately), i 1
Addition ofITDC criteria:
Tia addition of the ITDC criteria exceeds regulatory requirem :nts by providing augmented quality assurance for certain NNS related SSCs. The result of this criteria will be to increase nuclear safety margin, therefore no unreviewed safety question is created.
FSAR Channes Effectine the MY Defueled Condition (Rev 14): Defueled Safety Analysis Report (DS AR)
The purpose of this change was to formally reclassify SSCs that no longer need to be maintained under nuclear grade controls, and to clarify the safety classification of the remainder of SSCs. This was done by formally defming the classification of SSCs that are important To the Defueled Condition (ITDC), as discussed in the previous section " Initial FS AR Channes Effectine the MY JMurled Condition (Rev 13)". Certain engineered requirements are maintained for equipment classified as ITDC and are specified in the DSAR. The DSAR also includes a general discussion of decommissioning activities and potential accidents based on NUREG/CR-0130.
The accident analyses included in the DSAR is changed. This change falls into three categories, applicable previous analyses, analyses for accidents which are precluded, and new or modified analyses. The only analyses previously included that are still applicable are the Fuel Handling Incident, the Spent Fuel Cask Drop Accident, and Radioactive Liquid Waste System Leaks and Failures. All other analyses previously listed are precluded since no fuel may be loaded into the 4
reactor vessel. Analyses for previous incidents which were of lesser significance now represent tl.e bounding conditions. The following new analyses have been added to the accident analysis section of the DSAR:
Spent Fuel Criticality analysis (Misplaced Assembly, Dropped Assembly, Assembly adjacent to the racks, Assembly in the corner of the racks, and Boron Dilution)
Spent Fuel pool Accidents (Loss of SFP Cooling, Blocked / Improper Cell Flow, Loss of Forced Flow, Loss of Heat Sink, and Loss of SFP Inventory)
Low Level Waste Release Incident (Radioactive Waste Gas System Leaks and Failures, and Low Level Waste Storage Building Accident)
Except for the new/ modi 6ed analyses, the changes are the result of previously approved changes, either pursuant to 10CFR50.59, previously reviewed and approved by the NRC or as a result of the 50.82 certifications (specifically refer to " Initial FSAR Chances Effectine the MY Defueled Condition (Rev 13)"previously discussed in this report). These excepted changes did not require additional safety evaluation.
The evaluation of the full spe. um of analyses, including new/ modified and previous, indicates that the consequences of accidents previously described in the FSAR have decreased. There has been no increase in probability of any accidents previously analyzed (none of the changes have any impact on this). No equipment important to safety is impacted by this change. md no new accidents, different than previously analyzed, have been introduced. Based on this evalua: ion, no unreviewed safety question is ercated.
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Part if The following changes were made, but are no longer neninent to Maine Yankee's current licensing basis in accordance with the cessation of power operation:
EDCR 91-051, Revenue Metering Upgrade /SCADA This EDCR upgraded the existing printing demand meters mounted on the electric control board 4
with more accurate, state of the art equipment. The new meters allow for remote reading vin phone lines for CMP to retrieve data. The SCADA system was installed on the ll5kV system to allow CMP to monitor the status of switching and metering devices remotely. This design change did not involve an unreviewed safety question as defined in 10CFR50.59.
EDCR 95-048, Condensate Chemical Addition Modification A new chemical addition system was installed. The old chemical feed for the condensate system was replaced with new TOTES chemical tanks and new pumping skids. This design change did not involve an unreviewed safety question as defmed in 10CFR50.59.
EDCR 96-028, EFW Pump Room Ventilation Upgrade - FN-35A, FN-35B Design Basis Screen 95-017 determined that the existing single train NNS ventilation system for the Emergency Feedwater Pump Room was inadequate. This design change upgraded the system to two safety related trains. This design change did not involve an unreviewed safety question as defined in 10CFR50.59.
EDCR 96-033, HV-7 Replacement The Containment Spray Building heating and ventilation unit, HV-7, was replaced to make it more reliable during wintet conditions. This design change did not involve an unreviewed safety question as defined in 10CFR50.59.
EDCR 96-036, PCC Relief Valves in Containment This design change installed liquid relief valves on the containment recirculation air cooler primary component cooling (PCC) piping. The relief valves were installed outside containment, inside the containment isolation boundary. This design change did not involve an unreviewed safety question as defined in 10CFR50.59.
EDCR 96-041, Turbine Building Flood Relief Panel
' A panel was installed in the north wall of the turbine hall to allow flood waters from potential turbine hall pipe breaks to exit the building. This will prevent flooding of safety related equipment. This design change did not involve an unreviewed safety question as defined in 10CFR50.59.
Yellow Tag 87-203, Safety Parameter Display System (SPDS)
Lifled the leads in the Safety Parameter Display System cabinet removing a redundant ground wire from Reactor Protection System Channel A, Temperature Cold (Tc) and Temperature Hot (Th) eliminating a ground loop. This temporary change did not involve an unreviewed safety question as defmed in 10 CFR 50.59.
Yellow Tag 87-205, Safety Parameter Display System (SPDS)
- Lifled the leads in the SPDS cabinet removing a redundant ground wire eliminating a ground loop from steam generator pressure input to the Reactor Protection System, Channel A. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 92-25, Safety injection Tank Pressure Switch A temporary eJilication was provided to bypass the Safety injection Tanks' high and low pre-alarms, associated with the main control board alann annunciator windows. These pre-alanns were unreliable. The Safety Injection Tanks' high and low alarms still function for the "Alann condition."
This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 92-49, Reactor Protection System (RPS)
The temporary modification disconnects the non-nuclear system computer points from the Thermal Margin Low Pressure Calculator of the RPS to eliminate a common ground problem. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 92-54, Subcooled Margin Monitor Lilled certain thermocouple input leads and shorted them to corresponding core exit thermocouple transmitters to protect the Subcooled Margin Moni tor calculator input channels from cxcessive voltage transmitter outputs that have no core exit thermocouple inputs. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 93-04, Pressurizer Proportional IIcaters Provide forced air cooling to the existing convection coolcd cabinet to aid in the dissipation of heat within the cabinet and increase the reliability /perfonnance of the B pressurizer proportional heater.
This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 96-01, Auxiliary Feed Water (AFW) System A temporary modification to the auxiliary feed water Appendix R flange to allow draining of the first point heater feed line during smtdowns. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 96-17, Feedwater Suction flow readings for P-2C (turbine driven main feedwater pump) are required and the existing flow venturi failed, so an annubar was installed to replace the venturi in the suction line of P-2C.
In addition, the pressure taps where CD-249 & 250 were instolled were capped. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59 l
l Yellow Tag 96-28, Spray Hullding Exhaust Ventilation (VP-A-56,57) i A mechanical blockmg device was installed to hold VP-A-56 & 57 (fan inlet vanes) in full open position, so that they wouldn't fail shut on a loss ofinstrument Air to their actuators. This temporary change did not involve an unreviewed sa rety question as defined in 10 CFR 50.59 Yellow Tag 96-34, Main Steam (MS-P-168)
A bypass valve on the positioner for MS-P-168 was installed to stop its erratic operation during Aux Feedwater Pump (P-258) runs. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59 1
i Yellow Tag 96-43,'I urbine Buildir.,, dall Turbine 11all liELB issues required the installation of a temporary vestibule inside the security br located in the north wall of the Turbine 11alljust west of the roll up door. Ilinged sections that can be easily and quickly opened from the outside, were installed on the Turbine liall cast wall louvers, This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59 Yellow Tag 97-12, Servlee Air (SA 180) and Containment Integrity SA-180 and associated piping were removed to allow temporary lines to be routed into containment.
This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Yellow Tag 97-33, Temporary Set Point Chenge for TIS-6088 The Protected Switchgear Room high tempenture alami set point was previously lowered due to Turbine IIall HELB concems. These concerns no longer exist, so the high temperature set point was changed back to its previous higher value to eliminate unnecessary alarms. This 50.59 was also used for Technical Evaluation 188-97. This temporary change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 054-96, Component Substitution of PIT-2021 Replacement of PIT-2021 from a Mensor Model 10100-001 to Volumetrics Model PM-05. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 069 96, RWST Level Setpoint Changes As part of the MY Setpoint Program, this TE incorporates the necessary setpoint changes as a result ofincorporating instrumentation uncertainties as calculated by CALC-035 93 Rev 0. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE I14-96, /-29H and P-29C Low Raw Water Pressure Switch Setpoint Evaluation ofchanging the setpoints of pressure switches PS-2102B and PS-2102C from 2 psiG to 10 psiG. These pressure switches actuate on low Raw Water pressure to P-29B/C (Service Water pump) bearings to establish backup (Safety Class) bearing lubrication from the respective SW pump's discharge. This tech eval did not involve an unreviewed safety question as defmed in 10 CFR 50.59.
TE 121-96, Evaluation of PCC/ SCC Flow Tests This TE provides evaluation ofprimary/ secondary component cooling flow test results and discusses the affect these resulta have on maximum pemiissible Service Water inlet temperatures and addresses the as-found mispositioning of PCC-T-20, PCC bypass. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 122 96, Permanent Installation of Power to Guard Post Unit No. 6 Evaluation of the power supplied to a Guard Shack near the personnel hatch. The power supply is a spare breaker offlighting panel LP-R4 in the upper Spray Building. LP-R4 is fed from motor control center MCC-8Bl. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 128-96, Disabling the Protected SWGR and Cable Tray Room Ventilation Dampers Maine Yankee's past practice has been io open the Protected Switchgear Room and Cable Tray Room doors and provide forced Dow, when required, to maintain acceptable temperatures and air Gow through the battery rooms. With current ilELB issues in the Turbine Hall, the Protected Switchgear Room door cannot be opened. Therefore, the ventilation dampers were yellow tagged in the summer mode to ensure that a fresh air supply is available. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 226-96, Cable and Wire Separation Criteria This Tech Eval documents Maine Yankee's original cable and wire electrical separation criteria. i This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 005-97, Siltemp Barrier Material Compare Siltemp Material to Asbestos Wrap which is an approved separation barrier material. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59 TE 013-97, inspection of Fuel Assemblics by UT and Repair of CE Fuel Assemblies This 50.59 addresses activities related to fuel inspection and repair in the SFP. Fuel inspecttan includes detemiination oflocations ofleaking rods by ultrasonic testing, visual inspection and eddy current testing. Repair includes replacement of fuel rods with stainless steel rods, rod exchanges and recaging. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 014-97, Material and Configuration Change for a Mueller Strainer and Screen Evaluate changing Service Water pump back up seal water strainer screens from brass to Monel.
Strainer conGguration is also changing due to obsolescence. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 018-97, Control and Safety Related Instrument Cables in Common Raceways Low level control cables were installed into safety related instrument raceways during and aller the original plant construction. This TEjustifies these installations. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 019-97, Inspection of Fuel Assemblies by UT and Repair of W Fuel Assemblies This 50.59 addresses activities related to fuel inspection and repair in the SFP. Fuel inspection includes detemiination oflocations ofleaking rods by ultrasonic testing, visual inspection and eddy current testing. Repair includes replacement of fuel rods with stainless steel rods and rod exchanges.
This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE I11-97, ABB/CE Steam Generator Tube Plugs.Tubesheet Plugs and Tube Stabilizers Determine and document the adequacy of the pressure boundary ftmetion of various tube plugs and intemal tube stabilizers supplied by ABB/CENO. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
TE 129-97, Removal of Ctmt Spray Pumps to Support Alodifications per EDCR 97 35 Tech Eval for EDCR 97-35 which will disconnect the containment spray pumps, lin them from their barrels and transfer them to a temporary enclosure crected specifically for the purpose of housing them during the modification. This tech eval did not involve an unreviev'ed safety question as dermed in 10 CFR 50.59.
TE 136-97 Steam Generator Tube Removal and installation of Tubesheet Plugs Evaluation of the ABB/CE steam generator primary side tube removal procedure and installation of welded tubesheet plags. This tech eval did not involve an unreviewed safety question as dermed in 10 CFR 50.59.
TE 146-97, Sialn Steam Valve llouse llELH Control cable 114PLi97 is currently routed through the Main Steam Valve llouse, wherchy a postulated HELB event in the MSVil is not accounted for in the original IPEEE Analyses for the emergency core cooling system. The re-routing of a replacement cable will restore the safety integrity level, as a result of the latest interpretation of the separation criteria. This tech eval did not involve an unreviewed safety question as defmed in 10 CFR 50.59.
TE 172-97, Cable Separation Safety Assessment Safety assessment to support FSAR ch nges necessary to resolve cable separation issues. This tech eval did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc AOP 2-40, Severe Weather r
This procedure provides operators with guidance on dealing with severe weather. Actions taken help to minimize damage to the site and protect equipment required for safe operation of the plant. This procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc 80268-21, Load Test Procedure, Rack Vertical Lifting Rig Defines the method for " testing to verify continuing compliance" as required by ANSI 14.6, as applied to the MY fuel rack safety related lin rig. This procedure change did not involve an unreviewed safety question as dermed in 10 CFR 50.59.
Proc 3.1.14A/B,'A/H' Train EDG/ECCS Cold Shutdown Test This procedure was changed to require diesel start times of 15 seconds, rather than 10 seconds. This procedure tests the auto start featurcs of the emergency diesel generators as well as safeguards relays.
This also required a change to the FSAR. This procedure change did not involve an unreviewed safety question as defmed in 10 CFR 50.59.
Proc 4-26-1, Fuel Rod / Pellet Retrieval This temporary procedure was developed to allow the retrieval of broken fuel rods and fuel pellets thM have gone adrin in the SFP during past fuel reconstitution campaigns. This procedure change
- did not involve an unreviewed safety question as dedned in 10 CFR 50.59.
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.4 Proc 10-1, Core Reloading This procedure contains the administrative controls for performing core reloads. It also contains .
requirements to ensure all applicable Tech Specs are observed during the reloading process. This l procedure change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Proc AOP 2- 84, Control Room Evacuation This procedure directs the actions of operating personnel to maintain contral of SFP cooling equipment in the event the Control Room has been evacuated for reasons other than a fire or station blackout. This procedure change did not involve an unreviewed safety question as defined in 10
- - CFR 50.59.
FSAR CN 96-115, FSAR Change Deleted the statement in FSAR Section 8 regarding the minimum hours of operation of the emergency diesel generators as supplied by the integral and day tanks. This FSAR change did not involve r.n unreviewed safety question as defined in 10 CFR 50.59.
FSAR CN 96-120, FSAR Section 8.3.1.2 Modified discussion related to the impact of starting 6.9 kV large motors. Analysis indicates that starting loads are not negligible, but are acceptable. This FSAR change did not involve an
. unreviewed safety question as defined in 10 CFR 50.59.
MFU 96-022, FSAR Section 8.2.2 Clarifications to FSAR Section 8.2.2 to reflect that the original grid stability study was performed as part of ariginal plant design, that the Maine Yankee generator is no longer the largest on the grid
, and thm subsequent studies (consistent with the original studies) indicate that the utility grid system will remain stable aller the tripping of Maine Yankee or any other unit on the system. This FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
MFU 96-023, FSAR 8.3 Station Service ClariFcations made to information about the Emergency Diesel Generators, Heater Drain Tank motor sizing, and communications systems. Tlis FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
MFU 96-168, FSAR Section 2.3.3.3 Extreme Low Lesel Changes to FSAR Section 2.3.3.3 to clarify wording about low tidal elevation and minimum
. 'sebmergence of service water pumps. This FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
FSAR CN 96-024, FSAR Section 9.1 CVCS Multiple clarifications of text per Operations' comments. This FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
T' l
' FSAR CN 97 020, Change to FSAR Section 9.10.5.1 (Refueling Machine)
) FSAR Section 9.10.5.1 states that during withdraw:il or insedion of a fuel assembly, variations from j- normal loads in excess of 10 percent automatically stop the motion of the hoist winch mechanism. l Refueling System Interlock Test procedure 3.1.10, Rev 18, violates this FS AR requiremem.- This
!- FSAR change did not involve an unreviewed safety question as defined in 10 CFR 50.59.
Installation of New Parking Lot Below 345kV Transmission Lines This activity involved the construction of a r:w parking lot or, the west access road to be located undemeath the 345kV transmission lines. This project has been reviewed and approved by the Dept.
i- of Environmental Protection as recorded per docket # L-17973-26-A-N. The safety evaluation l
, associated with this 50.59 only addressed the acceptability of the construction phase of the lot and 1 i- its use during the refueling outage. The plant is currently relying on power supplied from the 115kV
, system. There exists no actions during the construction and usage of this lot that could affect the
- reliability of this supply. No USQ evaluation was required.
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